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完整後設資料紀錄
DC 欄位 | 值 | 語言 |
---|---|---|
dc.contributor.advisor | 吳文方 | |
dc.contributor.author | Hong-Lin Wei | en |
dc.contributor.author | 魏宏霖 | zh_TW |
dc.date.accessioned | 2021-06-15T06:47:12Z | - |
dc.date.available | 2013-07-06 | |
dc.date.copyright | 2011-07-06 | |
dc.date.issued | 2011 | |
dc.date.submitted | 2011-06-11 | |
dc.identifier.citation | 1. International Atomic Energy Agency, Assessment and Management of Ageing of Major Nuclear Power Plant Components Import to Safety: PWR Pressure Vessels, IAEA-TECDOC-1120. 1999.
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R., “Analysis of Stresses and Strains Near the End of a Crack Traversing a Plate,” ASME journal of Applied Mechanics, Vol. 24, pp. 361-364, 1957. 20. Cindy, L. R., et al., Atomistic Aspects of Crack Propagation in Brittle Materials: Multimillion Atom Molecular Dynamics Simulations, Annual Review of Materials Research, Vol. 32, pp.377-400, 2002. 21. Kanninen, M. F. and Popelar, C. H., Advanced Fracture Mechanics, Oxford University Press, 1985. 22. 楊銘堯,符合實驗結果之金屬疲勞裂縫延伸電腦模擬,國立台灣大學機械工程學研究所碩士論文,2001年6月。 23. Paris, P. C. and Erdogan, F., “A Critical Analysis of Propagation Laws,” Journal of Basic Engineering, Vol. 85, pp. 528-534, 1960. 24. Sundararajan, C., “Probabistic Assessment of Pressure Vessel and Piping Reliability,” ASME Journal of Pressure Vessel Technology, Vol. 108, pp. 1-13, 1986. 25. Wilson, S. A., Estimating the Relative Probability of Pipe Severance by Fault Cause, Report GEAP-20616, General Electric, 1974. 26. Harris, D. O., Lim, E. Y. and Dedhia, D. Probability of Pipe Fracture in the Primary Coolant Loop of a PWR plant, U.S. Nuclear Regulatory Commission Report NUREG/CR-2189, Vol. 5, 1981. 27. Harris, D. O., Lim, E. Y., Dedhia, D., Woo, H. H. and Chou, C. K., Fracture Mechanics Models Developed for Piping Reliability Assessment in Light Water Reactors, Report No. NUREG/GE-2301, U.S. Nuclear Regulatory Commission, Washington, D.C., 1982. 28. Harris, D. O., Analysis of the Probability of Pipe Rupture at Various Locations in the Primary Cooling Loop of a Babcock and Wilcox 177 Fuel Assembly Pressurized Water Reactor-Including the Effects of a Periodic Inspection, Report SAI-050-77-PA, Science Applications inc., 1977. 29. Proven, J. W., Probability Fracture Mechanics and Reliability, Martinus Nijhoff Publishers, Dordrecht, 1987. 30. Marshall, W., An Assessment of the Integrity of PWR Vessels. HM Stationary Office, London Report of a Study Group Chaired by W. Marshall. 31. Khaleel, M. A. and Simonen, F. A. “A Model for Predicting Vessel Failure Probabilities Including the Effects of Service Inspection and Flaw Sizing Errors,” Nuclear Engineering and Design, Vol. 200, pp. 353-369, 2000. 32. American Society for Testing and Materials International, Standard Practice for Light-Water Cooled Nuclear Power Reactor Vessels, ASTM E185-82, E7.6 (IF), 1982 33. United States Nuclear Regulatory Commission, Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report, Report TAC No M93925, July 1998. 34. Sundararajan, C., “Probabilistic Assessment of Pressure Vessel and Piping Reliability,” ASME Journal of Pressure Vessel Technology, Vol. 108, pp. 1-13, 1986. 35. Eealds, H. L. and Wanhill, R. J. H., Fracture Mechanics, ASTM Committee E-24 on Fracture Test, 1985. 36. Ang, A. H-S. and Tang, W. H., Probability Concepts in Engineering Planning and Design, Vol. Ⅱ, John Wiley and Sons, Inc., New York, 1988. 37. Harris, F. O. and Dedhia, D., Theoretical and User’s Manual for pc-PRAISW, A Probabilistic Fracture Mechanics Computer Code for Piping Reliability Analysis. U.S. Nuclear Regulatory Commission, Washington, DC NUREG/CR-5864, 1991. 38. Rybicki, E. F., Shadley, J. R. and Peterson, A. G., Experimental Residual Stress Evaluation of a Section of Clad Pressure Vessel Steel, Electric Power Research Institute, TR-101989, 1993. 39. Dirk, J. W. Policy Issue to NRC Commissioners, Enclisure A: NRC Staff Evaluation of Pressurized Thermal Shock, SECY-82-465, Division of Nuclear Reactor Regulations, U.S. Nuclear Regulatory Commission, Washington, D. C., 1982. 40. 林獻洲,核能電廠反應器壓力槽焊道之可靠度評估,國立台灣大學機械工程學研究所碩士論文,2009年7月。 | |
dc.identifier.uri | http://tdr.lib.ntu.edu.tw/jspui/handle/123456789/48141 | - |
dc.description.abstract | 核能電廠具有發電成本低與二氧化碳排放量少等優點,在面臨全球能源短缺與環保問題下,其為乾淨能源選項之一,然而核能電廠運轉首重結構與機械設備的安全,避免它們因老化而引起風險。本論文以反應器壓力槽為安全評估的對象,其為核能電廠蒸汽供應系統中極重要的壓力邊界組件,係由數片鋼板焊接而成,其中的焊道最有可能產生裂縫。輻射脆化為反應器壓力槽最主要的老化機制,當裂縫隨運轉時間成長至某一程度時,反應器壓力槽焊道即有可能產生脆裂。先前研究指出,包覆層厚度與裂縫尺寸為影響焊道脆裂的主要因素,因此本論文針對沸水式反應器壓力槽之垂直與水平焊道,利用破裂力學理論、並考慮輻射脆化現象,依隨機變數觀念進行模擬分析。分析結果顯示,當包覆層厚度超過0.35 in時,垂直焊道失效機率有上昇的趨勢;而當考慮檢測與修復時,使用適當的檢測能力相較於增加檢測週期或減少檢測範圍更能有效的降低失效機率;此外,在相同的環境負載下,水平焊道失效機率較垂直焊道來的低,就此,國外管制單位對於水平焊道檢測豁免的作法不失為可行之道。 | zh_TW |
dc.description.abstract | Nowadays, we are facing problems of global energy shortage as well as the need of environmental protection. The advantage of low cost and small amount of CO2 discharge makes nuclear power an important choice for energy. However, the safety of structures, systems and mechanical components employed in a nuclear power plant has to be assured before a plant can be constructed. One of the most important pressure boundary components in the steam supply system of a nuclear power plant is the reactor pressure vessel (RPV). It is welded together by several steel plates. Cracks occur more frequently in welds rather than in base plates of a RPV. When a predominant crack grows along with operating time to a certain size, it may result in brittle fracture in the weld of a RPV. It has been pointed out that clad thickness and crack size affect the embrittlement and fracture of the weld. The present study employs a probabilistic fracture mechanics approach by taking into account radiation embrittlement to find fracture-failure probabilities of RPV welds. The result shows that, when the clad is thicker than 0.35 inch, the failure probability at axial weld should be paid more attention to. As for the effect of inspection and repair, it is found that adopting a more advanced inspection instrument reduces failure probability more than increasing inspection cycles or covering more inspection areas. It is also found the probability of failure at circumferential welds is smaller than that at axial welds. The finding reassures the proposition made by the United States Nuclear Regulatory Commission (USNRC) that inspection of circumferential welds of a RPV can be exempted. | en |
dc.description.provenance | Made available in DSpace on 2021-06-15T06:47:12Z (GMT). No. of bitstreams: 1 ntu-100-R98522527-1.pdf: 3201057 bytes, checksum: 41febb559e8dbebc9770d521f2a9b2dc (MD5) Previous issue date: 2011 | en |
dc.description.tableofcontents | 口試委員會審定書 I
致謝 II 摘要 III ABSTRACT IV 目錄 V 表目錄 VIII 圖目錄 IX 符號說明 XI 第一章 緒論 1 1.1 研究背景與動機 1 1.2 文獻回顧 2 1.3 論文架構 4 第二章 反應器壓力槽之老化機制 6 2.1 反應器壓力槽簡介 6 2.2 沸水式反應器壓力槽 6 2.3 老化機制 7 2.3.1 輻射脆化 7 2.3.2 熱老化 9 2.3.3 回火脆化 9 2.3.4 疲勞 10 2.3.5 腐蝕 11 2.3.6 磨耗 11 第三章 基本理論概述 15 3.1 可靠度基本理論 15 3.2 機率分佈 17 3.2.1 離散性隨機變數分佈函數 17 3.2.2 連續性隨機變數分佈函數 17 3.3 機率點圖法 20 3.4 適配度檢定法 22 3.5 破裂力學 23 3.5.1 破裂準則 23 3.5.2 裂縫成長模式 24 3.6 機率破裂力學 26 第四章 反應器壓力槽焊道之機率破裂力學分析模式 34 4.1 初始裂縫深度分佈 34 4.2 應力強度因子 34 4.2.1 耦合場分析 35 4.2.2 焊道殘留應力 36 4.2.3 熔覆應力 36 4.3 反應器壓力槽材料監測 36 4.3.1 夏比衝擊試驗 37 4.3.2 快中子照射通量之估算 38 4.4 輻射脆化評估模式 38 4.5 低溫超壓事件 40 4.6 蒙地卡羅模擬 42 4.7 應力強度干涉理論 43 4.8 檢測機率 44 4.9 檢測範圍 44 第五章 分析結果與討論 52 5.1 應力分析 52 5.2 裂縫成長分析 53 5.3 低溫超壓事件之應力負載 55 5.4 失效機率分析 55 5.5 檢測與修復 58 5.6 分析結果與討論 59 第六章 結論 84 參考文獻 85 | |
dc.language.iso | zh-TW | |
dc.title | 包覆層厚度對核能電廠反應器壓力槽可靠度之影響 | zh_TW |
dc.title | Effect of Clad Thickness on Reliability of Reactor Pressure Vessels in Nuclear Power Plants | en |
dc.type | Thesis | |
dc.date.schoolyear | 99-2 | |
dc.description.degree | 碩士 | |
dc.contributor.oralexamcommittee | 黃金城,周雄偉,游章雄 | |
dc.subject.keyword | 反應器壓力槽,裂縫成長,包覆層厚度,輻射脆化, | zh_TW |
dc.subject.keyword | Reactor Pressure Vessel,Crack Growth,Clad Thickness,Radiation Embrittlement, | en |
dc.relation.page | 88 | |
dc.rights.note | 有償授權 | |
dc.date.accepted | 2011-06-12 | |
dc.contributor.author-college | 工學院 | zh_TW |
dc.contributor.author-dept | 機械工程學研究所 | zh_TW |
顯示於系所單位: | 機械工程學系 |
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