博碩士論文 993203601 詳細資訊




以作者查詢圖書館館藏 以作者查詢臺灣博碩士 以作者查詢全國書目 勘誤回報 、線上人數:25 、訪客IP:18.207.238.169
姓名 達曼多(Darmanto)  查詢紙本館藏   畢業系所 機械工程學系
論文名稱 核電廠元件疲勞壽命模擬分析
(Simulation of Fatigue Life For Nuclear Power Plant Components)
相關論文
★ 晶圓針測參數實驗與模擬分析★ 車銑複合加工機床面結構最佳化設計
★ 精密空調冷凝器軸流風扇葉片結構分析★ 第四代雙倍資料率同步動態隨機存取記憶體連接器應力與最佳化分析
★ PCB電性測試針盤最佳鑽孔加工條件分析★ 鋰-鋁基及鋰-氮基複合儲氫材料之製程開發及研究
★ 合金元素(錳與鋁)與球磨處理對Mg2Ni型儲氫合金放電容量與循環壽命之影響★ 鍶改良劑、旋壓成型及熱處理對A356鋁合金磨耗腐蝕性質之影響
★ 可撓式OLED封裝薄膜和ITO薄膜彎曲行為分析★ MOCVD玻璃承載盤溫度場分析
★ 不同環境下之沃斯回火球墨鑄鐵疲勞裂縫成長行為★ 不同環境下之Custom 450不銹鋼腐蝕疲勞性質研究
★ AISI 347不銹鋼腐蝕疲勞行為★ 環境因素對沃斯回火球墨鑄鐵高週疲勞之影響
★ AISI 347不銹鋼在不同應力比及頻率下之腐蝕疲勞行為★ 電子構裝用無鉛銲錫之低週疲勞行為研究
檔案 [Endnote RIS 格式]    [Bibtex 格式]    [相關文章]   [文章引用]   [完整記錄]   [館藏目錄]   至系統瀏覽論文 ( 永不開放)
摘要(中) 由於爐心噴灑管嘴對熱疲勞非常的敏感,所以在核反應爐中視為關鍵的組件,本研究主旨在計算核反應爐內爐心噴灑管嘴在不同狀態下之熱應力分佈及疲勞壽命。疲勞壽命計算方法之一是使用多軸向應力疲勞分析,其疲勞壽命的計算方式係考慮六個應力分量的變化,與ASME NB-3200規範所述一致,而另一疲勞壽命計算方式則採用多軸疲勞之臨界平面法,並藉由商用軟體FE-SAFE加以分析,進一步比較兩種不同方法的分析結果,並選擇合適的疲勞壽命評估模式。
分析結果顯示,使用ASME規範計算,發現有六個分析位置點顯示有限的壽命,而使用商用軟體FE-SAFE時,則有四個分析位置點顯示有限的壽命,進而得知ASME規範的多軸向應力疲勞分析較商用軟體FE-SAFE的臨界平面法保守。因此對核反應爐中的爐心噴灑管嘴而言, ASME規範的疲勞壽命計算方式是較為保守的評估模式。
摘要(英) One of the critical components or locations in a nuclear reactor is the core spray nozzle because it is very sensitive to thermal fatigue. The fatigue life of a core spray nozzle is calculated using a new multiaxial stress-based fatigue (SBF) analysis module during transient events. Fatigue calculation considers all six components of stress in accordance with ASME Subarticle 3200 involving modified stress cycle pairing method. The fatigue life is also calculated using a critical plane approach by means of a multiaxial fatigue software, FE-SAFE. The comparison between these two approaches provides some insight into selection of a suitable fatigue life assessment model for the core spray nozzle in a nuclear power plant.
The investigated points in the core spray nozzle using ASME code shows finite life at six points, while the FE-SAFE results show four points of finite life. The ASME code gives more conservative results than does the critical plane approach in the FE-SAFE. ASME code is suitable for fatigue life assessment of the core spray nozzle in a nuclear power plant for a more conservative consideration.
關鍵字(中) 關鍵字(英) ★ Fatigue Life
★ Core Spray Nozzle
論文目次 TABLE OF CONTENTS
Page
LIST OF TABLES V
LIST OF FIGURES VI
NOMENCLATURE IX
1. INTRODUCTION 1
1.1 Core Spray Nozzle 1
1.2 Thermal Stress Analysis 2
1.3 Fatigue Analysis 5
1.3.1 ASME code 5
1.3.2.1 Rubberband peak and valley detection (PV detection) 9
1.3.2.2 Rainflow 3D 10
1.3.2 Multiaxial fatigue analysis: critical plane approach 11
1.4 Purposes and Scope 13
2. MODELING 15
2.1 Finite Element Model 15
2.2 Material Properties 16
2.3 Boundary Conditions 16
2.4 Investigated Cases 17
2.5 Fatigue Life Analysis 18
2.5.1 ASME code 18
2.5.2 Multiaxial fatigue analysis: crtitical plane approach 18
3. RESULTS AND DISCUSSION 20
3.1 Thermal Stress Analysis 20
3.1.1 Temperature distribution 20
3.1.2 Thermal stress distribution 21
3.2 Fatigue Analysis 24
3.2.1 ASME code 24
3.2.2 Critical plane approach 24
4. CONCLUSIONS 26
REFERENCES 27
TABLES 30
FIGURES 34
APPENDIX: COMPUTER CODES FOR RUBBERBAND PV DETECTION AND
RAINFLOW 3D 83
參考文獻 REFERENCES
1. J. K. Shulfis and R. E. Faw, Fundamental of Nuclear Science and Engineering, 2nd Ed., CRC Press, Taylor & Francis Group, Boca Raton, FL, USA, 2010.
2. G. E. Paredes, A. N. Carrera, A. V. Rodriguez, E. G. Espinosa-Martinez, “Modeling of the High Pressure Core Spray Systems with Fuzzy Cognitive Maps for Operational Transient Analysis in Nuclear Power Reactors,” Progress in Nuclear Energy, Vol. 51, pp. 434-442, 2009.
3. “BWR/4 Technology Manual (R-104B),” United States Nuclear Regulatory Commission Technical Training Center, Chattanooga, TN, USA, 2010.
4. “Internal Core Spray Piping and Sparger Replacement Design Criteria (BWRVIP-16NP),” BWR Vessel and Internals Project, Boiling Water Reactor Vessel & Internals Project Repair Committee, Electric Power Research Institute, Palo Alto, California, USA, 2000.
5. I. S. Meza, L. H. Hernández-Gómez, G. U. Sosa, J. A. Beltrán-Fernández, E. A. Merchán-Cruz, J. M. Sandoval-Pineda, and A. T. Velásquez-Sánchez, “Thermal Fatigue Analysis of an Emergency Core Cooling System Nozzle of a BWR Reactor, by Finite Element Method,” Cientifica, Vol. 11, pp. 113-119, 2007.
6. “Sizing Calculation for Nozzle Core Spray (N5 Nozzle),” Chinsan Nuclear Power Plant Project Sheet, Muroran Plant, The Japan Steel Works, Ltd, Atomic Energy Department, Muroran, Hokkaido, Japan, 1972.
7. R. Ciceroa, S. Cicerob, I. Gorrochateguic, and R. Lacalle, “Considerations on Fatigue Stress Range Calculations in Nuclear Power Plants Using On-line Monitoring Systems and the ASME Code,” Nuclear Engineering and Design, Vol. 240, pp. 47–56, 2010.
8. R. E. Hoffman and T. Ariman, “Thermal and Mechanical Stresses in Nuclear Reactor Vessels,” Nuclear Engineering and Design, Vol. 20, pp. 31-55, 1972.
9. “Rules for Construction of Nuclear Power Plant Components,” ASME Boiler and Pressure Vessel Code, Division-1 Subsection NB, Class 1 Components, American Society of Mechanical Engineers, New York, USA, 1998.
10. D. R. Moss, Pressure Vessel Design Manual: Illustrated Procedures for Solving Major Pressure Vessel Design Problems, 3rd. Ed. Gulf Professionals Publishing, Burlington, MA, USA, 2004.
11. “Stress-Based Fatigue Monitoring, Methodology for Fatigue Monitoring of Class 1 Nuclear Components in a Reactor Water Environment,” EPRI 2011 Technical Report, Electric Power Research Institute, Palo Alto, CA, USA, 2011.
12. S. R. Hill, Supporting New Build and Nuclear Manufacturing in South Africa: Section III – Component Design and Construction, Westinghouse Electric Company, Sandton, South Africa, 2008.
13. T. Gilman, Stress-Based Fatigue Monitoring: Methodology for Fatigue Monitoring of Class 1 Nuclear Components in a Reactor Water Environment (EPRI Technical Report 1022876)-NRC Fatigue Meeting Presentation, Electric Power Research Institute, Palo Alto, CA, USA, 2011.
14. ABAQUS User’s Manual V6.11, Dassault Systèmes Simulia Corp., Providence, RI, USA, 2011.
15. “Material, Part D–Properties,” 1998 ASME Boiler and Pressure Vessel Code, an International Code, American Society of Mechanical Engineers, New York, USA, 1998.
16. D. F. Socie and G. B. Marquis, Multiaxial Fatigue, Society of Automotive Engineers, Inc., Warrendale, Pa, USA, 2000.
17. “Fatigue Theory Reference Manual,” FE-SAFE 6.2 User Manual, Safe Technology US Limited, Livonia, MI, USA, 2011.
18. J. Draper, Modern Metal Fatigue Analysis, EMAS Publishing, Warrington, WA3 6FW, UK, 2011.
19. J. L. T. Santos, M. D. Freitas, B. Li, and T. P. Trigo, “Fatigue Assessment of Mechanical Components under Complex Multiaxial Loading,” pp. 463–482 in Biaxial/Multiaxial Fatigue and Fracture, ESIS Publication 31, edited by A. Carpinteri, M. D. Freitas, A. Spagnoli, Elsevier, Oxford, UK, 2003.
20. J. Liu and H. Zenner, “Fatigue Limit of Ductile Metal under Multiaxial Loading,” pp. 147–164 in Biaxial/Multiaxial Fatigue and Fracture, ESIS Publication 31, edited by A. Carpinteri, M. D. Freitas, A. Spagnoli, Elsevier, Oxford, UK, 2003.
21. A. R. Kallmeyer, A. Krgo, and P. Kurath, “Evaluation of HCF Multiaxial Fatigue Life Prediction Methodologies for Ti-6Al-4V,” ASME Journal of Engineering Materials and Technology, Vol. 124, No. 2, 2002, pp. 229-237.
22. “Tutorials,” FE-SAFE 6.2 User Manual, Safe Technology US Limited, Livonia, MI, USA, 2011.
23. “Rules for Construction of Pressure Vessel,” ASME Boiler and Pressure Vessel Code, Division-2, American Society of Mechanical Engineers, New York, USA, 1998.
24. Y.-T. Chiu, “Fortran Program for Rubberband PV Detection and Rainflow 3D,” National Central University, 2012.
25. “Boiling Water Reactor (BWR) System,” United States Nuclear Regulatory Commission Technical Training Center, Chattanooga, TN, USA, 2010.
26. W. C. Patterson, “Nuclear Power,” 2nd Edition, Penguin Books, Ltd., Middlesex, UK, 1983.
27. “Reevaluation of Station Blackout Risk at Nuclear Power Plants,” Analysis of Loss of Offsite Power Events: 1986-2004, Idaho National Laboratory, U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research, Washington, DC, USA, 2005.
28. O. Coata and L. Cizelj, "Thermal Fatigue Assessment: A Two-Dimensional Approach,” Proceedings of the International Conference of Nuclear Energy for New Europe, Bovec, Slovenia, 2011.
29. Y. A. Cengel and M. A. Boles, “Thermodynamics: An Engineering Approach,” 5th Edition, McGraw-Hill Inc., Boston, MA, USA, 2006.
30. A. M. Clayton, “Thermal Shock in Nuclear Reactor” Progress in Nuclear Energy, Vol. 12, pp. 57-83, 1983.
31. N. M. Skliarov, “The Great Soviet Encyclopedia,” 3rd Edition, The Gale Group, Inc., Michigan, USA, 2010.
指導教授 林志光
(Chih-Kuang Lin、Anindito Purnowidodo)
審核日期 2013-1-30
推文 facebook   plurk   twitter   funp   google   live   udn   HD   myshare   reddit   netvibes   friend   youpush   delicious   baidu   
網路書籤 Google bookmarks   del.icio.us   hemidemi   myshare   

若有論文相關問題,請聯絡國立中央大學圖書館推廣服務組 TEL:(03)422-7151轉57407,或E-mail聯絡  - 隱私權政策聲明