核能電廠餘熱移除系統振動台試驗之模擬與分析

Analytical Studies for Seismic Behavior of Residual Heat Removal System in Nuclear Power Plants Experiment

梁瑋洛(Wai-Lok Leong)

指導教授 ： 黃尹男

國立臺灣大學/工學院/土木工程學研究所/碩士(2018年)

https://doi.org/10.6342/NTU201803008

摘要

Huang, Whittaker 與 Luco等人綜合核電廠既存的機率式地震風險評估 (Seismic Probabilistic Risk Assessment, SPRA) 方法與地震工程學界在耐震性能評估上的做法，發展出一SPRA方法(簡稱為HWL-typed SPRA)。HWL-typed SPRA之易損性曲線為結構元件反應參數之函數，有別於傳統SPRA之易損性曲線為地表運動參數之函數。前者配合結構非線性動力歷時反應分析，藉由分析直接將原本傳統易損性分析時須要考慮之結構反應參數之保守性及變異性納入考量。HWL-typed SPRA包含五個步驟，分別為：(一)、系統分析及元件易損性分析，(二)、危害度分析，(三)、結構動力歷時反應分析，(四)、元件損傷評估，(五)、地震風險量化計算。
本研究針對元件易損性分析方法進行探討，在建立元件易損性曲線之程序上，以實驗方式評估數值模型與數值分析之合理性，再以數值分析方式來建立以元件反應為參數之易損性曲線。以案例電廠之餘熱移除(Residual Heat Removal, RHR)C串管線系統為例，除了以設計分析資料建立其傳統地震易損性曲線，亦依本研究發展之易損性分析方法，對RHR-C管線系統進行振動台耐震性能試驗，進行系統識別得知管線系統之模態行為，並以實驗資料建立可靠之數值模型。藉由多筆管線數值模型之非線性動力歷時分析資料，建立RHR-C管線系統之以元件反應參數為函數之地震易損性曲線，並討論以地表運動及結構反應參數為函數之易損性曲線之差異。本研究結果顯示與傳統地震易損性分析方法之結果相比，以元件反應為參數之地震易損性分析結果在中位數推估上較為保守，且對數標準差較小，故建議使用HWL-typed SPRA之以元件反應為參數之地震易損性分析方法進行核電廠元件耐震容量之評估。

關鍵字

餘熱移除系統 ； 元件易損性分析 ； 非線性動力分析 ； 振動台試驗 ； 機率式地震風險評估

並列摘要

Seismic probabilistic risk assessment (SPRA) has been widely used to compute the frequencies of core damage and release of radiation of nuclear power plant (NNP). In 2011, Huang, Whittaker, and Luco published a new SPRA methodology, different from existing SPRA, including nonlinear response-history analysis of structure to estimate seismic demands for component and the component fragility curves are functions of component response parameters
In the study presented herein, the methodologies of developing ground-motion-based and response-based component fragility curves are presented. A residual heat removal (RHR) piping system which contributes significantly to the seismic risk of a sample NPP has been analyzed experimentally and numerically. System identification of the tested piping system was conducted using Enhanced Frequency Domain Decomposition (EFDD), Multiple Singular Spectrum Analysis (MSSA) and Singular Value Decomposition (SVD). The numerical models validated using the results of system identification analysis and experiments were used to develop response-based fragility curves for the system. Comparing to the results of ground-motion-based component fragility analysis, those of response-based fragility analysis are more conservative and reliable, so the former is recommended for seismic probabilistic risk assessment.

並列關鍵字

seismic probabilistic risk assessment ； nonlinear response-history analysis ； residual heat removal piping system ； component fragility analysis ； system identification

參考文獻

American Society of Civil Engineers (ASCE). (2014). “Seismic Analysis of Safety-Related Nuclear Structures” ASCE 4-14, American Society of Civil Engineers, Reston, Virginia.

ASME Boiler and Pressure Vessel Committee Subcommittee on Nuclear Power (ASME BPVC) (2007). “2007 ASME Boiler & Pressure Vessel Code Section II Part D Properties (Customary) Materials” American Society of Mechanical Engineers, Two Park Avenue, New York

ASME Boiler and Pressure Vessel Committee Subcommittee on Nuclear Power (ASME BPVC) (2007). “2007 ASME Boiler & Pressure Vessel Code Section III Division 1-Subsection NB-Class 1 Components Rules for Construction of Nuclear Facility Components” American Society of Mechanical Engineers, Two Park Avenue, New York

ASME Boiler and Pressure Vessel Committee Subcommittee on Nuclear Power (ASME BPVC) (2007). “2007 ASME Boiler & Pressure Vessel Code Section III Division 1-Subsection NC-Class 2 Components Rules for Construction of Nuclear Facility Components” American Society of Mechanical Engineers, Two Park Avenue, New York

ASME Boiler and Pressure Vessel Committee Subcommittee on Nuclear Power (ASME BPVC) (2007). “2007 ASME Boiler & Pressure Vessel Code Section III Division 1-Appendices Rules for Construction of Nuclear Facility Components” American Society of Mechanical Engineers, Two Park Avenue, New York