Seismic fragility analysis of nuclear power plants considering structural parameter uncertainty

The Korea Atomic Energy Research Institute (KAERI) and Brookhaven National Laboratory (BNL) are continuing a collaborative effort to achieve a better understanding of the effects of aging on the performance of structures and passive components (SPCs) in nuclear power plants (NPPs). This paper presents a seismic fragility analysis of a condensate storage tank (CST) with multiple degradation scenarios that are treated in a non-perfectly correlated manner. The analysis utilizes a set of optimum Latin Hypercube samples to characterize the deterioration behavior of the fragility capacity as a function of age-related degradations. This study is an addition to the previous study summarized in an ICONE19 paper entitled “Seismic Fragility Analysis of a Degraded Condensate Storage Tank” [1], which considered individual degradation scenarios and multiple degradations occurring in a perfectly correlated manner.

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The numerical computation of seismic fragility of base-isolated Nuclear Power Plants buildings

Nuclear Engineering and Design ◽

10.1016/j.nucengdes.2013.04.029 ◽

2013 ◽

Vol 262 ◽

pp. 189-200 ◽

Cited By ~ 45

Author(s):

Federico Perotti ◽

Marco Domaneschi ◽

Silvia De Grandis

Keyword(s):

Numerical Computation ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Seismic Fragility

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Seismic Performance of Piping Systems of Isolated Nuclear Power Plants Determined by Numerical Considerations

Energies ◽

10.3390/en14134028 ◽

2021 ◽

Vol 14 (13) ◽

pp. 4028

Author(s):

Sungjin Chang ◽

Bubgyu Jeon ◽

Shinyoung Kwag ◽

Daegi Hahm ◽

Seunghyun Eem

Keyword(s):

Seismic Response ◽

Nuclear Power ◽

Seismic Isolation ◽

Power Plants ◽

Nuclear Power Plants ◽

Mesh Size ◽

Seismic Fragility ◽

Response Analysis ◽

Seismic Safety ◽

Seismic Isolation Systems

The interest in the seismic performance of nuclear power plants has increased worldwide since the Fukushima Daiichi Nuclear Power Plant incident. In Korea, interest in the seismic safety of nuclear power plants has increased since the earthquake events in Gyeongju (2016) and Pohang (2017). In Korea, studies have been conducted to apply seismic isolation systems to ensure seismic safety while minimizing the design changes to nuclear power plants. Nuclear power plants with seismic isolation systems may have a higher seismic risk due to the failure of the piping system in the structure after a relatively large displacement. Therefore, it is essential to secure the seismic safety of pipes for the safe operation of nuclear power plants. The seismic safety of pipes is determined by seismic fragility analysis. Seismic fragility analysis requires many seismic response analyses because it is a statistical approach to various random variables. Typical numerical conditions affecting the seismic response analysis of pipes are the convergence conditions and mesh size in numerical analysis. This study examined the change in the seismic safety of piping according to the numerical conditions. The difference in the seismic response analysis results of the piping according to the mesh size was analyzed comparatively. In addition, the change in the seismic fragility curve of the piping according to the convergence conditions was investigated.

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Development of Guideline on Seismic Fragility Evaluation for Aged Piping

10.1115/pvp2021-61509 ◽

2021 ◽

Author(s):

Yoshihito Yamaguchi ◽

Jinya Katsuyama ◽

Koichi Masaki ◽

Yinsheng Li

Keyword(s):

Risk Assessment ◽

Fracture Mechanics ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Seismic Fragility ◽

Degradation Mechanisms ◽

Probabilistic Fracture Mechanics ◽

Age Related ◽

Piping Systems

Abstract The seismic probabilistic risk assessment is an important methodology to evaluate the seismic safety of nuclear power plants. In this assessment, the core damage frequency is evaluated from the seismic hazard, seismic fragilities, and accident sequence. Regarding the seismic fragility evaluation, the probabilistic fracture mechanics can be applied as a useful evaluation technique for aged piping systems with crack or wall thinning due to the age-related degradation mechanisms. In this study, to advance seismic probabilistic risk assessment methodology of nuclear power plants that have been in operation for a long time, a guideline on the seismic fragility evaluation of the typical aged piping systems of nuclear power plants has been developed considering the age-related degradation mechanisms. This paper provides an outline of the guideline and several examples of seismic fragility evaluation based on the guideline and utilizing the probabilistic fracture mechanics analysis code.

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