• Corrosion Science and Technology
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• 제14권1호
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• pp.1-11
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• 2015

Local wall thinning and integrity degradation caused by several mechanisms, such as flow accelerated corrosion (FAC), cavitation, flashing and/or liquid drop impingements, are a main concern in carbon steel piping systems of nuclear power plant in terms of safety and operability. Thinned pipe management program (TPMP) had been developed and optimized to reduce the possibility of unplanned shutdown and/or power reduction due to pipe failure caused by wall thinning in the secondary side piping system. This program also consists of several technical elements such as prediction of wear rate for each component, prioritization of components for inspection, thickness measurement, calculation of actual wear and wear rate for each component. Decision making is associated with replacement or continuous service for thinned pipe components. Establishment of long-term strategy based on diagnosis of plant condition regarding overall wall thinning is also essential part of the program. Prediction models of wall thinning caused by FAC had been established for 24 operating nuclear plants. Long term strategies to manage the thinned pipe component were prepared and applied to each unit, which was reflecting plant specific design, operation, and inspection history, so that the structural integrity of piping system can be maintained. An alternative integrity assessment criterion and a computer program for thinned piping items were developed for the first time in the world, which was directly applicable to the secondary piping system of nuclear power plant. The thinned pipe management program is applied to all domestic nuclear power plants as a standard procedure form so that it contributes to preventing an accident caused by FAC.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2004년도 봄 학술발표회 논문집
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• pp.309-316
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• 2004

A probabilistic structural integrity assessment is performed for a reactor pressure vessel under PTS(Pressurized Thermal Shock). A semi-elliptical finite axial crack is assumed to he in the beltline region(either base metal or weld meta)1 of the reactor vessel inside surface. The selected random variables are initial crack depth, neutron fluence on the vessel inside surface, copper, nickel, and phosphorus content of the vessel material, and RT/sub NDT/. The probabilities of crack initiation or vessel failure where the crack is propagated through vessel wall are calculated. The probabilities obtained with random crack size are compared to these obtained with deterministic us. Since the failure function cannot to explicitly by selected by selected random variables, Monte Carlo Simulation is applied to perform probabilistic analysis The influence of the amount of neutron fluence is also examined to assess the structural reliability for vessel life time.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회A
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• pp.309-314
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• 2007

Recently, the LRFD and the PSF based on structural reliability assessment have been applied to NPP designs in behalf of the conventional deterministic design methods. In the risk-informed structural integrity, it is especially possible to optimize design procedures considering cost, manufacturing and maintenance because the structural reliability concepts have confirmed the reliability for which a designer aims. Generally, in order to evaluate the PSF, the LRFD which is the design concept for evaluating safety factors respectively on the limit state function including load and resistance. This study certifies the concept and its applications of the PSF using the LRFD based on the structural reliability engineering.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2011년도 추계학술대회 논문집
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• pp.221-226
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• 2011

The structural integrity of APR1400 reactor vessel internals has been being assessed referring the US Nuclear Regulatory Commission regulatory guide 1.20 comprehensive vibration assessment program. The program is composed of a vibration and stress analysis, a limited vibration measurement, and an inspection. This paper covers the vibration and stress analysis on the reactor vessel internals due to the pulsation of reactor coolant pump. 3-dimensional models to calculate the hydraulic loads and structural responses were built and the pressure distributions and the structural responses were predicted using ANSYS. The peak stress of the reactor vessel internals is much lower than the acceptance limit.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.963-970
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• 2018

The U.S. Nuclear Regulatory Commission, Regulatory Guide 1.20 needs to perform a comprehensive vibration assessment program for reactor internals during preoperational and startup testing for nuclear power plants and extended power uprate. Although the measurement program is one of the core programs, it is rarely carried out except for a first-of-a-kind or a unique design. This article describes measurement results of vibration and stress for the comprehensive vibration assessment program for an APR-1400 reactor internals. The measurement was performed at an upper guide structure during the pre-core hot functional test of Shin Kori unit 4 reactor internals because the Shin Kori unit 3 and 4 are the first construction project for the APR-1400, and the upper guide structure assembly was to design change compared with the valid prototype. We confirmed that all measured results are within the test acceptance criteria. It means that the structural integrity of the APR-1400 reactor internals was secured for the flow-induced vibration.

• 기계저널
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• 제35권6호
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• pp.504-516
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• 1995

작년에 발생한 성수대교의 파손사고는 인적손실은 물론 경제적 그리고 사회적으로 심각한 영향을 미쳤다. 현재 수많은 공장에서 가동되고 있는 크레인 관련 시설들은 변동응력의 작용과 용접구 조물이라는 점에서 성수대교와 유사한 조건을 가지고 있다. 또한 그중에서 많은 부분은 설계시 에는 고려치 못했던 부하조건의 가혹화와 노후화 때문에 손상 가능성이 항상 존재하고 있다. 이러한 점을 고려하여 필자들은 설비의 예방정비 차원에서 크레인 거더 및 런웨이 거더의 안정성 평가에 주목하게 되었고 나름대로 얻은 경험을 지면에 정리하여 보았다.

• Wind and Structures
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• 제13권1호
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• pp.1-20
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• 2010

The recommended factored design wind load effects for overhead lattice transmission line towers by codes and standards are evaluated based on the applicable wind load factor, gust response factor and design wind speed. The current factors and design wind speed were developed considering linear elastic responses and selected notional target safety levels. However, information on the nonlinear inelastic responses of such towers under extreme dynamic wind loading, and on the structural capacity curves of the towers in relation to the design capacities, is lacking. The knowledge and assessment of the capacity curve, and its relation to the design strength, is important to evaluate the integrity and reliability of these towers. Such an assessment was performed in the present study, using a nonlinear static pushover (NSP) analysis and incremental dynamic analysis (IDA), both of which are commonly used in earthquake engineering. For the IDA, temporal and spatially varying wind speeds are simulated based on power spectral density and coherence functions. Numerical results show that the structural capacity curves of the tower determined from the NSP analysis depend on the load pattern, and that the curves determined from the nonlinear static pushover analysis are similar to those obtained from IDA.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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• pp.85-89
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• 2004

The inner side between HP stationary blades in #1 turbine of Nuclear Power Plant A is damaged by the FAC(flow assisted corrosion) which is exposed to moisture. For many years the inner side is repaired by welding the damaged part, however, FAC continues to deteriorate the original material of the welded blade ring. In this study, we have two stages to verify the integrity of stationary blade ring in nuclear power plant A. In the stage I, replication of blade ring is performed to survey the microstructure of blade ring. In the stage II, the stress analysis of blade ring is performed to verify the structural safety of blade ring. Throughout the two stages analysis of blade ring, the stationary blade ring had remained undamaged.

• 한국안전학회지
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• 제17권3호
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• pp.66-72
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• 2002

During the last ten years, effort has been made for reducing maintenance cost for aged equipments and ensuring safety, efficiency and profitability of petrochemical and refinery plants. Hence, it was required to develop advanced methods which meet this need. RBI(Risk Based Inspection) methodology is one of the most promising technology satisfying the requirements in the field of integrity management. In this study, a qualitative assessment algorithm for RBI based on the API 581 code was reconstructed for developing an RBI software. The user-friendly realRBI software is developed with a module for evaluating qualitative risk category using the potential consequence factor and the likelihood factor.

• 국제초고층학회논문집
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• 제2권2호
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• pp.141-152
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• 2013

This paper presents a brief overview of the recently developed performance-control method of moment frame design subjected to monotonously increasing lateral loading. The final product of any elastic-plastic analysis is a nonlinear loaddisplacement diagram associated with a progressive failure mechanism, which may or may not be as desirable as expected. Analytically derived failure mechanisms may include such undesirable features as soft story failure, partial failure modes, overcollapse, etc. The problem is compounded if any kind of performance control, e.g., drift optimization, material savings or integrity assessment is also involved. However, there is no reason why the process can not be reversed by first selecting a desirable collapse mechanism, then working backwards to select members that would lead to the desired outcome. This article provides an overview of the newly developed Performance control methodology of design for lateral resisting frameworks with a view towards integrity control and prevention of premature failure due to propagation of plasticity and progressive P-delta effects.