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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 | Li-Chuan Lee | en |
dc.contributor.author | 李禮全 | zh_TW |
dc.date.accessioned | 2021-06-15T04:51:52Z | - |
dc.date.available | 2013-08-04 | |
dc.date.copyright | 2010-08-04 | |
dc.date.issued | 2010 | |
dc.date.submitted | 2010-07-31 | |
dc.identifier.citation | 1. U.S. Nuclear Regulatory Commission, Radiation Embrittlement of Reactor Vessel Materials, Regulatory Guide 1.99, Revision 2,1988.
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Moskvichev, V. and Doronin, S., “Statistical Fracture Modeling of Weld Joint for Nuclear Reactor Components,” Theoretical and Applied Fracture Mechanics, Vol.29, pp. 103-107,1998. 8. Tang, S. S., et al., “Effect of Adjusted Reference Temperature on the Probability of Failure in Boiling Water Reactor Vessel Welds,” American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP, Vol.388, pp.121-126,1999. 9. George J. Schuster and Steven R., “Fabrication Flaws in Reactor Pressure Vessel Repair Welds” , Transactions, SMiRT 19, Toronto, August 2007. 10. Irwin, G.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, 1975. 11. Rubinstein, R. Y., Simulation and Monte Carlo Method, Wiley, New York, 1981. 12. Kanninen, M. F., and Popelar, C.H., Advanced Fracture Mechanics, Oxford University Press, 1985. 13. Ang, A. H-S and Tang, W. H., Probability and Concepts in Engineering Planning and Design, Vol. I, John Wiley & Sons, Inc., New York, 1975. 14. 張紘睿,非破壞評估可靠度之模型適合度探討,國立台灣大學機械工程學研究所碩士論文,民國94年 15. Ebeling, C. E., An Introduction to Reliability and Maintainability Engineering, McGraw-Hill. 16. Sundararajan, C., “Probabilistic Assessment of Pressure Vessel and Piping Reliability,” ASME Journal of Pressure Vessel Technology, Vol. 108, pp. 1-33, 1986. 17. Wilson, S. A., Estimating the Relative Probability of Pipe Severance by Frult Cause, Report GEAP-20616, General Electric Company, 1974. 18. Harris, D.O., et al., Probability of Pipe Fracture in the Primary Coolant Loop of PWR Plant, Probability Fracture Mechanics Analysis, U.S. Nuclear Regulatory Commission Report NUREG/CR-2189, Vol.5, Washington, D.C.,1981. 19. Harris, D.O., Lim, E.Y., Dedhia, D.D., Woo, H.H., and Chou, C.K., Fracture Mechanics Models Developed for Piping Reliability Assessment in Light Water Reactor, Report No. NUREG/GE-2301, U.S. Nuclear Regulatory Commission, Washington, D.C.,1982 20. 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. 21. The Nuclear Human Resource Development (HRD) Center of Korea Atomic Energy Research Institute, http://www.kntc.re.kr/openlec/policy/part1/part1_chapter6.html 22. Marshall, W., An Assessment of the Integrity of PWR Vessels. HM Stationary Office, London Report of a Study Group Chaired by W. Marshall. 23. American Society for Testing and Materials International, Standard Practice for Light-Water Cooled Nuclear Power Reactor Vessels, ASTM E185-82, E7.6 (IF), 1982. 24. International Atomic Energy, Assessment and Management of Ageing of Major Nuclear Power Plant Component Important to Safety: PWR Pressure Vessels, IAEA-TECDOC-1120, 1999. 25. Harris, D.O., Dedhia, D.D., Theoretical and User’s Manual for pc-PRAISE, A Probabilistic Fracture Mechanics Computer Code for Piping Reliability Analysis. U.S. Nclear Regulatory Commission, Washington, D.C. NUREG/CR-5864, 1991. 26. United States Nuclear Regulatory Commission, Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report, Report TAC No M93925, July 1998. 27. Sundararajan, C., “Probabilistic Assessment of Pressure Vessel and Piping Reliability,” ASME Journal of Pressure Vessel Technology, Vol. 108, pp. 1-33, 1986. 28. Eealds ,H.L. and Wanhill, R.J.H.,Fracture Mechanics, ASTM Committee E-24 on Fracture Test, 1985. 29. Ang, A. H-S and Tang, W. H., Probability and Concepts in Engineering Planning and Design, Vol. II, John Wiley & Sons, Inc., New York, 1988. 30. Rubinstein, R.Y., Simulation and Monte Carlo Method, Wiley, New York, 1981. 31. 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,民國98年 32. Ford, F.P., Environmentally Assisted Cracking of Low Alloy Steels, EPRI Report NP-7473-L, GE, San Jose, CA, January 1992. 33. GE Nuclear Energy Interim Technical Report on EPRI RPC 102-4,Stress Corrosion Cracking in Low Alloy Steels, February 1994. 34. Khaleel, M.A. 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. 35. Dirk, J.W., Policy Issue to NRC Commissioners, Enclosure 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. | |
dc.identifier.uri | http://tdr.lib.ntu.edu.tw/jspui/handle/123456789/46032 | - |
dc.description.abstract | 反應器壓力槽為核能電廠中最重要的壓力邊界組件,它若發生破壞將會引發嚴重災害。由於反應器壓力槽結構巨大,所以需以焊接方式製造而成,而依據經驗,其垂直焊道最容易產生破壞。反應器壓力槽焊道內的裂縫會隨時間生長而導致破壞的可能,且材料也會受到輻射脆化,增加壓力槽破壞的機率。本研究以蒙地卡羅模擬法配合機率破壞力學理論,評估焊道預存裂縫沸水式反應器壓力槽之破壞機率,其間我們考慮到裂縫成長機制、輻射脆化影響、低溫超壓之偶發事件、材料性質之不確定性、初始裂縫之隨機性、不同程度之非破壞檢測水準與更新等學理與現象。研究結果顯示,針對一商用沸水式反應器壓力槽,若每十年進行一次檢測與修復,則其運轉至40年之設計壽命時,破壞機率為4.32×10-5;若將模擬時間延長至60年,則破壞機率變為5.94×10-5。此外,相較於輻射脆化,初始裂縫成長為影響破壞機率的主要因素,因此若能透過檢測與更新有效控制裂縫長度,即可有效減低反應器壓力槽的破機率。 | zh_TW |
dc.description.abstract | The reactor pressure vessel (RPV) is the most important pressure boundary component in a nuclear power plant. The failure of RPV will cause serious hazard, and the vertical welds of RPV attribute very high probability of failure. The failure probability of RPV in general increases along with time owing to crack growth in welds of RPV. The increase is even more significant when radiation embrittlement is taken into consideration. In the present study, Monte Carlo Simulation (MCS) and Probabilistic Fracture Mechanics are employed to study the reliability of a Boiling Water Reactor Pressure Vessel (BWRPV). It is assumed that a predominant initial crack exists in the vertical welds of the RPV. When subjected to environmental conditions, the crack grows which, in turn, increases the failure probability of the vessel. In the modeling and analysis, the initial crack distribution, crack growth mechanism, low temperature over pressurization (LTOP) transients, radiation embrittlement, NDT examination, renew process and uncertainty of material properties are all considered. The result indicates that the probability of failure (POF) of the analyszed RPV is 4.32×10-5 after 40 years of operation if non-destructive inspection is employed and renewal action is taken every 10 years. Under the same inspection and repair condition, the POF decreases to 5.94×10-5 after 60 years of operation. It reflects that the POF of BWRRPV can be reduced effectively through periodic inspections and complete repair actions afterwards. | en |
dc.description.provenance | Made available in DSpace on 2021-06-15T04:51:52Z (GMT). No. of bitstreams: 1 ntu-99-R97522530-1.pdf: 2151971 bytes, checksum: 905cb6d7dbcdfe14bafaddea1d025695 (MD5) Previous issue date: 2010 | en |
dc.description.tableofcontents | 摘要 I
Abstract II Index III 表目錄 V 圖目錄 VI 符號說明 VIII 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研究架構 3 第二章 基本理論 5 2.1破壞力學 5 2.1.1破壞準則 5 2.1.2裂縫成長模式 6 2.2非破壞檢測 7 2.2.1檢測機率曲線 7 2.3可靠度理論 8 2.4機率分佈函數 9 2.4.1離散型隨機變數之機率分佈函數 10 2.4.2連續型隨機變數之機率分佈函數 11 2.5機率破壞力學 13 2.6統計檢定 14 2.6.1機率圖紙法 14 2.6.2卡方檢定 15 第三章 沸水式反應器壓力槽機率破壞力學評估模式 22 3.1沸水式反應器壓力槽 22 3.2初始裂縫分佈 22 3.3應力強度因子 23 3.4材料監測 23 3.4.1Charpy衝擊試驗 23 3.4.2快中子照射量估算 24 3.5材料弱化評估模式 24 3.5.1輻射脆化 25 3.6檢測機率 26 3.7修復 26 3.8低溫超壓事件 27 3.9蒙地卡羅模擬法 28 3.10 應力強度干涉理論 29 第四章 分析結果 33 4.1應力負載 33 4.2裂縫成長 34 4.3低溫超壓事件下應力負載 35 4.4輻射脆化評估 36 4.5檢測與修復 37 4.6分析結果討論 38 第五章 結論 52 References 53 | |
dc.language.iso | zh-TW | |
dc.title | 含馬歇爾預存裂縫之沸水式反應器壓力槽可靠度評估 | zh_TW |
dc.title | Reliability Assessment of Boiling Water Reactor Pressure Vessels Having Marshall-Type Initial Cracks | en |
dc.type | Thesis | |
dc.date.schoolyear | 98-2 | |
dc.description.degree | 碩士 | |
dc.contributor.oralexamcommittee | 黃金城,游章雄,周雄偉 | |
dc.subject.keyword | 反應器壓力槽,機率破壞力學,低溫超壓事件,裂縫成長,輻射脆化,可靠度, | zh_TW |
dc.subject.keyword | Reactor Pressure Vessel,Probabilistic Fracture Mechanics,Low Temperature Over Pressure,Crack Growth,Radiation Embrittlement,Reliability, | en |
dc.relation.page | 56 | |
dc.rights.note | 有償授權 | |
dc.date.accepted | 2010-08-02 | |
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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