• Nuclear Engineering and Technology
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• v.56 no.3
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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.

• Proceedings of the Earthquake Engineering Society of Korea Conference
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• 2000.04a
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• pp.97-104
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• 2000

Evaluating a Review-Level Ground Motion is a key to efficiently perform Seismic Margin Assessment of nuclear power plants whose purpose is to determine a ground motion level for which a plant has high-confidence-of-a-low-probability of seismic-induced core damage and to identify any weaker-link components. In this study a method to obtain RLGMs is reviewed which is recommended by Electric Power Research Institute and implemented to be applied to Limerick site in eastern and central U. S as a case study. This method provides reasonable and site-specific RLGMs as minimum required plant HCLPF for SMA that meet a target mean seismic core-damage frequency based on seismic hazard results and generic values of uncertainty and randomness parameters of the core-damage fragility curves. In addition high-frequency RLGM is justifiably modified to reflect the increased seismic capacity of high-frequency components and spatial variation and incoherence of input ground motion on a basemat of large structures by establishing a method to obtain high0-frequency reduction factors according to EPRI guidelines.

• Proceedings of the Earthquake Engineering Society of Korea Conference
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• 2005.03a
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• pp.453-460
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• 2005

In this study, the seismic fragility analysis of a base isolated condensate storage tank installed in the nuclear power plant. The condensate storage tank is safety related structure in a nuclear power plant. The failure of this tank affect significantly to the core damage frequency of the nuclear power plants. The seismic analysis of the liquid storage tank was performed by the simple calculation method and the dynamic time storage analysis method. The convective and impulsive fluid mass is modeled as added masses proposed by several researchers. To evaluate the effectiveness of the isolation system, the comparison of HCLPF and core damage frequencies in non-isolated and isolated cases are carried out. It can be found from the results that the seismic isolation system increases the seismic capacity of a condensate storage tank and decreases the core damage frequency significantly.

• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.313-326
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• 2017

The systematic management of plant risk is crucial for enhancing the safety of nuclear power plants and for designing new nuclear power plants. Accident sequence precursor (ASP) analysis may be able to provide risk significance of operational experience by using probabilistic risk assessment to evaluate an operational event quantitatively in terms of its impact on core damage. In this study, an ASP methodology for two operation mode, full power and low power/shutdown operation, has been developed and applied to significant accident precursors that may occur during the operation of nuclear power plants. Two operational events, loss of feedwater and steam generator tube rupture, are identified as ASPs. Therefore, the ASP methodology developed in this study may contribute to identifying plant risk significance as well as to enhancing the safety of nuclear power plants by applying this methodology systematically.

• Steel and Composite Structures
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• v.21 no.5
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• pp.1145-1156
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• 2016

To study the effects of foam core density and face-sheet thickness on the mechanical properties and failure modes of aluminum foam sandwich (AFS) beam, especially when the aluminum foam core is made in aluminum alloy and the face sheet thickness is less than 1.5 mm, three-point bending tests were investigated experimentally by using WDW-50E electronic universal tensile testing machine. Load-displacement curves were recorded to understand the mechanical response and photographs were taken to capture the deformation process of the composite structures. Results demonstrated that when foam core was combined with face-sheet thickness of 0.8 mm, its carrying capacity improved with the increase of core density. But when the thickness of face-sheet increased from 0.8 mm to 1.2 mm, result was opposite. For AFS with the same core density, their carrying capacity increased with the face-sheet thickness, but failure modes of thin face-sheet AFS were completely different from the thick face-sheet AFS. There were three failure modes in the present research: yield damage of both core and bottom face-sheet (Failure mode I), yield damage of foam core (Failure mode II), debonding between the adhesive interface (Failure mode III).

• Nuclear Engineering and Technology
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• v.38 no.6
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• pp.535-550
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• 2006

The objectives of this paper are twofold to summarize the new findings and confirmed results from the Phebus FPT-1 experimental data and to report useful information to MELCOR users regarding the better use of MELCOR. For the core damage behavior, the early stage of a melt progression was predicted well; however, the late phase models, concerned with fuel dissolution, oxide cladding failure, fuel slumping, rubble debris heat up, effects of burn-up fuel, and so on, still showed limitations in MELCOR. For the fission product behavior, the comparison showed unexpected phenomena, various limitations, unresolved issues, and even absence of models. The issues summarized in this study have revealed the main areas where our endeavors need to be intensified in order to improve our understanding of severe accident phenomena. From the analysis of the Phebus FPT-1 test results, not only new core damage features, such as foaming or core expansion, but also possible new fission product release patterns due to effects from a high burn-up fuel have raised alternative challenging phenomena that should be solved in the next severe accident research phase.

• Steel and Composite Structures
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• v.51 no.1
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• pp.53-61
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• 2024

Aluminum foams sandwich panel (AFSP) has been used in engineering field, where cyclic loading is used in most of the applications. In this paper, the fatigue life of AFSP prepared by the bonding method was investigated through a three-point bending test. The mathematical statistics method was used to analyze the influence of different plate thicknesses and core densities on the bending fatigue life. The macroscopic fatigue failure modes and damage mechanisms were observed by scanning electron microscopy (SEM). The results indicate that panel thickness and core layer density have a significant influence on the bending fatigue life of AFSP and their dispersion. The damage mechanism of fatigue failure to cells in aluminum foam is that the initial fatigue crack begins the cell wall, the thinnest position of the cell wall or the intersection of the cell wall and the cell ridge, where stress concentrations are more likely to occur. The fatigue failure of aluminum foam core usually starts from the semi-closed unit of the lower layer, and the fatigue crack propagates layer by layer along the direction of the maximum shear stress. The results can provide a reference for the practical engineering design and application of AFSP.

• Proceedings of the Earthquake Engineering Society of Korea Conference
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• 2003.03a
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• pp.93-100
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• 2003

Seismic isolation systems can improve the seismic safety of nuclear power plants by decreasing seismic force transmitted to structures and equipment. This study evaluates the effectiveness of equipment seismic isolation systems by the comparison of core damage frequencies in non-isolated and isolated cases. It can be found that the seismic isolation systems increase seismic capacity of nuclear equipment and decrease core damage frequencies significantly. The effect of equipment isolation is more significant in the PGA range of 0.3g to 0.5g.

• Proceedings of the Korean Society For Composite Materials Conference
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• 2001.05a
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• pp.74-77
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• 2001

The impact response and damage of CLAS panel was investigated experimentally. The facesheet material used was RO4003 woven-glass hydrocarbon/ceramic and the core material was Nomex honeycomb with a cell size of 3.2mm and a density of 96 kg/$\textrm{m}^{3}$. The shield plane used was RO4003 and 2024-T3 aluminum. Static indentation and impact test was conducted to characterize the type and extent of the damage observed in two CLAS panels, and the performance of antenna used in a wireless LAN system. Correlation of peak contact force, residual indentation and the delamination area shows impact damage of the panel with an aluminum shield plane is larger than that of the panel with RO4003 shield plane, although tile former is more penetration resistant. The damage was observed by naked eye, ultrasonic inspection and cross sectioning. The shape and size of delamination was estimated by ultrasonic inspection, and the area of delamination linearly increases as impact energy increases. The performance of impact damaged antenna was estimated by measuring return loss and radiation pattern.

• Journal of the Korean Society of Safety
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• v.35 no.1
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• pp.97-106
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• 2020

The objective of this paper is to suggest a new quantification method for multi-unit probabilistic safety assessment (PSA) that removes the overestimation error caused by the existing delete-term approximation (DTA) based quantification method. So far, for the actual plant PSA model quantification, a fault tree with negates have been solved by the DTA method. It is well known that the DTA method induces overestimated core damage frequency (CDF) of nuclear power plant (NPP). If a PSA fault tree has negates and non-rare events, the overestimation in CDF drastically increases. Since multi-unit seismic PSA model has plant level negates and many non-rare events in the fault tree, it should be very carefully quantified in order to avoid CDF overestimation. Multi-unit PSA fault tree has normal gates and negates that represent each NPP status. The NPP status means core damage or non-core damage state of individual NPPs. The non-core damage state of a NPP is modeled in the fault tree by using a negate (a NOT gate). Authors reviewed and compared (1) quantification methods that generate exact or approximate Boolean solutions from a fault tree, (2) DTA method generating approximate Boolean solution by solving negates in a fault tree, and (3) probability calculation methods from the Boolean solutions generated by exact quantification methods or DTA method. Based on the review and comparison, a new intersection removal by probability (IRBP) method is suggested in this study for the multi-unit PSA. If the IRBP method is adopted, multi-unit PSA fault tree can be quantified without the overestimation error that is caused by the direct application of DTA method. That is, the extremely overestimated CDF can be avoided and accurate CDF can be calculated by using the IRBP method. The accuracy of the IRBP method was validated by simple multi-unit PSA models. The necessity of the IRBP method was demonstrated by the actual plant multi-unit seismic PSA models.