研究生: |
陳柏臻 Chen, Po-Chen |
---|---|
論文名稱: |
氧化鋯與氧化鈦被覆對304與304L不鏽鋼之防蝕效益及耐久度研究 Influence and Durability of ZrO2 and TiO2 Coating on Type 304 and 304L Stainless Steels |
指導教授: |
葉宗洸
Yeh, Tsung-Kuang |
口試委員: |
歐陽汎怡
黃俊源 馮克林 |
學位類別: |
碩士 Master |
系所名稱: |
原子科學院 - 核子工程與科學研究所 Nuclear Engineering and Science |
論文出版年: | 2013 |
畢業學年度: | 101 |
語文別: | 中文 |
論文頁數: | 184 |
中文關鍵詞: | 熱水沉積被覆 、極化掃描 、耐久度 、不同流速 、光激發效應 |
相關次數: | 點閱:4 下載:0 |
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BWR許多材料組件均為304和304L不鏽鋼,長期運轉下容易發生沿晶應力腐蝕龜裂(Intergranular Stress Corrosion Cracking, IGSCC)的問題。目前世界均採用加氫水化學(Hydrogen Water Chemistry, HWC)和貴重金屬添加(Noble Metal Chemical Addition, NMCA)技術,來防制IGSCC的發生。但這些方法有副作用,因此近年抑制性被覆(Inhibitive Protective Coatings, IPC)逐漸發展。
根據文獻,利用熱水沉積法被覆氧化鋯和氧化鈦在304不鏽鋼的防蝕效果已被證實。本實驗的主要目的,就是要測試氧化鋯和氧化鈦被覆層在304和304L不鏽鋼上的耐久度。測試的方法,是將試片放置在除氧、濃度為2 mM K3Fe(CN)6 與K4Fe(CN)6混合溶液中,進行高流速沖刷4周,以電化學方法測試被覆層是否會剝落。結果顯示,被覆過後試片的腐蝕速度較未被覆慢,證明被覆層抑制金屬腐蝕效益。且經過長期高流速耐久度的沖刷測試,被覆試片的電化學性質均沒有明顯改變,由此可見抑制性被覆層在304與304L不鏽鋼上,有四周的耐久度。
1. Barberis, P., T. Merle-Mejean, and P. Quintard, On Raman Spectroscopy of Zirconium Oxide Films. Journal of Nuclear Materials, 1997. 246: p. 232-243.
2. Jayaweera, P., S. Hettiarachchi, and H. Ocken, Determination of High Temperature Zeta Potential and pH of Zero Charge of Some Transition Metal Oxides. Colloids and Surface A: Physicochemical and Engineering Aspects, 1994. 85: p. 19-27.
3. Keeling, D.P.T.a.D.R., www.esrl.noaa.gov/gmd/ccgg/trends/.
4. IEA, Energy Balances of OECD Countries, 2012 ; Energy Balances of NON-OECD Countries, 2012, Global Energy Network Institute.
5. Commission, U.S.N.R., Cracking of Vertical Welds in the Core Shroud and Degraded Repair. NRC Information Notice 97-17, , April 4, 1997.
6. Andresen, P.L., Factors Governing the Prediction of LWR Component SCC Behavior From Laboratory Data, in National Association of Corrosion Engineers1999, NACE Interantional: San Antonio, Tx.
7. 楊文斗, 反應堆材料學. 原子能出版社, 2000: p. 125-130.
8. 行政院原子能委員會, 核二廠一號機反應爐支撐裙板錨定螺栓斷裂事件暨修復後運轉安全評估報告, 2012.
9. Andresen, P., Emerging Issues and Fundamental Processes in Environmental Cracking in Hot Water, in National Association of Corrosion Engineers2007, NACE International: Nashville, Tennessee.
10. Jr., W.D.C., Materials Science and Engineering - An Introduction, 5th Edition. 1999.
11. Chung, P.C.-K., Quantitative Study of the Degree of Sensitization of Austenitic Stainless Steel by Electrochemical Measurements. 1979: Ohio State University.
12. Busby, J.T., G.S. Was, and E.A. Kenik, Isolating the effect of radiation-induced segregation in irradiation-assisted stress corrosion cracking of austenitic stainless steels. Journal of Nuclear Materials, 2002. 302(1): p. 20-40.
13. Wilde, B.E., Stress Corrosion Cracking. Vol. Failure Analysis and Prevention. 1986: American Society for Metals. 203.
14. Ford, P., Development and Use of a Predictive Model of Crack Propagation in 304/316L, A533B/A508 and Inconel 600/182 Alloys in 288oC water, in Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors1988.
15. Ford, P., Stress Corrosion Cracking of Low Alloy Steels Under BWR Conditions; Assessment of CHR Algorithms, in International Symposium on Environment Degradation of Materials in Nuclear Power Systems-Water Reactors1999.
16. Angeliu, T.M., P.L. Andresen, and F.P. Ford, Apolying Slip-Oxidation to the SCC of Austenitic Materials in BWR/PWR Environments in National Association of Corrosion Engineers1998, NACE International: San Diego Ca.
17. 鍾自強, 焊接奧斯田鐵不锈鋼的問題及解決方法. 機械月刊, 中華民國七十年10月號. 第七卷第十期.
18. Ford, F. and P. Andresen, Corrosion in Nuclear Systems: Environment Assisted Cracking in Light Water Reactors, 1994: New York. p. 501-546.
19. Kazushige, I., W. Yoichi, and T. Masahiko, Hydrazine and Hydrogen Co-injection to Mitigate Stress Corrosion Cracking of Structural Materials in Boiling Water Reactors, (II) Reactivity of Hydrazine with Oxidant in High Temperature Water under Gamma-irradiation. Journal of Nucler Science and Technology, 2006.
20. Kaesche, H., Die Korrosion der Metalle physikalisch-chemische Prinzipien und aktuelle Probleme. 1990: Springer.
21. Winkler, R., H. Lehmann, and F. Michel, VGB Kraftwerkstechnik, 1989. 69: p. 527-531.
22. Asakura, Y., et al., Relationships Between Corrosion Behavior of AISI 304 Stainless Steel in High-Temperature Water and Its Oxide Film Structure. Corrosion, 1989. 45: p. 119-124.
23. Winkler, R., F. Huttner, and F. Michel, Senkung der Korrosionsrate im Primarkreislauf von Druckwasserreaktoren zur Begrenzung radioaktiver Ablagerungen. VGB Kraftwerkstechnik, 1989. 69: p. 524-531.
24. Kim, Y.J., Effect of Water Chemistry on Corrosion Behavior of 304 SS in 288℃ Water, in International Water Chemistry Conference2004: San Francisco.
25. Kim, Y.J., Analysis of Oxide Film Formed on Type 304 Stainless Steel in 288oC Water Containing Oxygen, Hydrogen, and Hydrogen Peroxide. Corrosion, 1999. 55.
26. Miyazawa, T., S. Uchida, and T. Satoh, Effects of Hydrogen Peroxide on Corrosion of Stainless Steel, (IV) Determination of Oxide Film Properties with Multilateral Surface Analyses. Journal of Nuclear Science and Technology, 2005. 42(2): p. 233-241.
27. Murayama, Y., T. Satoh, and S. Uchida, Effects of Hydrogen Peroxide on Intergranular Stress Corrosion Cracking of Stainless Steel in High Temperature Water, (V). Journal of Nuclear Science and Technology, 2002. 39: p. 1199-1206.
28. Miyazawa, T., T. Terachi, and S. Uchida, Effects of Hydrogen Peroxide on Corrosion of Stainless Steel, (V) Characterization of Oxide Film with Multilateral Surface Analyses. Journal of Nuclear Science and Technology, 2006. 43(8): p. 884-895.
29. Wada, Y., A. Watanabe, and M. Tachibana, Effects of Hydrogen Peroxide on Intergranular Stress Corrosion Cracking of Stainless Steel in High Tmeperature Water, (IV) Effects of Oxide Film on Electrochemical Corrosion Potential. Jounal of Nuclear Science and Technology, 2001. 38: p. 183-192.
30. Macdonald, D., Theoretical Estimation of Crack Growth Rates in Type 304 Stainless Steel in Boiling-Water Reactor Coolant Environment. Corrosion, 1996. 52.
31. Uhlig, H.H. and R.W. Revie, Corrosion and Corrosion Control. p. Chapter 7.
32. Kim, Y.J., In-Situ Electrochemical Impedance Measurement of Oxide Film Formed on Type 304 Stainless Steel in High-Temperature Water. Corrosion, 2000. 56: p. 389-394.
33. Stellwag, B. and R. Kilian, Investigations into Chemistry-Related Alternatives to Hydrogen Water Chemistry in BWR Plants, in International Workshop on LWR Cooloant Water Radiolysis & Electrochemistry2000.
34. Cowan, R., The mitigation of IGSCC of BWR internals with hydrogen water chemistry. Nuclear energy, 1997. 36(4): p. 257-264.
35. Lin, C.C., F.R. Smith, and R.L. Cowan, Effects of Hydrogen Water Chemistry on Radiation Field Buildup in BWRs. Nuclear Engineering and Design, 1996. 166: p. 31-36.
36. Hunter, R.J., Zeta Potential in Colloid Science : Principles and Applications. 1988: Academic Press.
37. Zhou, Z.F., E. Chalkova, and S.N. Lvov, Development of a hydrothermal deposition process for applying zirconia coatings on BWR materials for IGSCC mitigation. Corrosion Science, 2007. 49: p. 830-843.
38. Atik, M. and M.A. Aegerter, Corrosion resistant sol-gel ZrO2 coatings on Stainless Steel. Journal of Non-Crystalline Solids, 1992. 147&148: p. 813-819.
39. Kim, Y.J. and P.L. Andresen, Application of Insulated Protective Coatings for Reduction of Corrosion Potential of Type 304 Stainless Steel in High-Temperature Water. Corrosion, 1998. 54: p. 1012-1017.
40. Yeh, T.K., M.Y. Lee, and C.H. Tsai, Intergranular Stress Corrosion Cracking of Type 304 Stainless Steels Treated with Inhibitive Chemicals in Simulated Boiling Water Reactor Environments. Journal of Nuclear Science and Technology, 2002. 39: p. 531-539.
41. Yeh, T.K., C.T. Liu, and C.H. Tsai, The Influence of ZrO2 Treatment on the Electrochemical Behavior of Oxygen and Hydrogen on Type 304 Stainless Steels in High Temperature Water. Journal of Nuclear Science and Technology, 2005. 42: p. 809-815.
42. Yeh, T.K., C.H. Tsai, and Y.H. Cheng, The Influence of Dissolved Hydrogen on the Corrosion of Type 304 Stainless Steels Treated with Inhibitive Chemicals in High Temperature Pure Water. Journal of Nuclear Science and Technology, 2005. 42: p. 462-469.
43. Yeh, T.K., Y.C. Chien, and B.Y. Wang, Electrochemical Characteristics of Zirconium Oxide Treated Type 304 Stainless Steels of Different Surface Oxide Structures in High Temperature Water. Corrosion Science, 2008. 50: p. 2327-2337.
44. Zhou, Z.F., Optimization of Zirconium Oxide Coating Technology to Mitigate IGSCC in BWRs, in BWRVIP Mitigation Committee Meeting2002.
45. Yeh, T.-K. and P.-I. Wu, Corrosion of ZrO2 treated type 304 stainless steels in high temperature pure water
with various amounts of hydrogen peroxide. Progress in Nuclear Energy, 2012: p. 62-70.
46. Ono, S., M. Miyano, and M. Hishida, ECP decrease of ceramic coated stainless steel by Radiation Induced Surface Activation, in International Symposium on Mechanism and Application of Radiation Induced Surface Activation2005: Tokyo, Japan.
47. G. X. Shen, Y. C. Chen, C. J. Lin, Thin Solid Films, 489 (2005) 130-136
48. Y. Ohko, S. Saitoh, T. Tatsuma, and Akira Fujishima, Journal of The Electrochemical Society, 148 (1) B24-B28 (2001).
49. Masato Okamura et al. Corrosion Mitigation of BWR Structural Materials by the Photoelectric Methods with TiO2 -Laboratory Experiments of TiO2 Effect on ECP Behavior and Materials Integrity, ENVIRONMENTAL DEGRADATION OF MATERIALS IN NUCLEAR SYSTEMS-WATER REACTORS (2005).
50. M. Okamura et al., SCC Mitigation Method of BWR Structural materials by TiO2 technique, Proc. Symposium on Water Chemistry and Corrosion of Nuclear Power Plants in Asia, Taipei, Taiwan, September 26-28 p.117 (2007).
51. K.Takamori et. al, Corrosion Mitigation of BWR Structural Materials by the Photoelectric Method with TiO2 - A SCC Mitigation Technique and its Feasibility Evaluation , 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 2005.
52. K.Takamori et. al, DEVELOPMENT OF BWR COMPONENTS SCC MITIGATION METHOD BY THE TiO2 TREATING TECHNIQUE, 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, British Columbia, (2007).
53. C. X. Shan, X. Hou, K.-L. Choy, Surface & Coatings Technology, 202 (2008) 2399–2402.
54. Yeh, T.-K., et al., Hydrothermal treatments of TiO< sub> 2</sub> on Type 304 stainless steels for corrosion mitigation in high temperature pure water. Nuclear Engineering and Design, 2013. 254: p. 228-236.
55. C. C. Lin, F. R. Smith, Decomposition of Hydrogen Peroxide at Elevated Temperature, EPRI Report NP-6733, 1990.
56. D. D. Macdonald et al., Corrosion Potential Measurements on Type 304 SS and Alloy 182 in Simulated BWR Environments,Corrosion-January , 8-16, 1993.
57. Anders Molander and Mats Ullberg, "The Corrosion Potential of Stainless Steel in BWR Environment Comparison of Data and Modeling Results, Symposium on Water Chemistry and Corrosion in Nuclear Power Plants in Asia,2003.
58. Fong, C., FLOW EFFECTS ON THE STRESS CORROSION CRACKING OF SENSITIZED STAINLESS STEELS UNDER BWR ENVIRONMENTS. 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems, 2007.
59. N. Hiraide, K. Sugimoto, Degradation of nitride, carbide and oxide ceramics materials in high temperature aqueous solutions, Boshoku Gijyutsu, 37, 415, 1988.
60. T. Kawakubo, H. Hirayama, A. Goto et al., Corrosion Behavior of Ceramics in High Purity Water at 290oC, Zairyo, 1989.
61. Young-Jin Kim and Peter L. Andresen, Application of Insulated Protective Coating for 304 SS SCC Mitigation in 288oC Water, 11th Int. Conf. Environmental Degradation of Materials in Nuclear Systems , 2003.
62. Kazushige ISHIDA et al., Effects of Flow Rate on Dissolution of Monocrystal Aluminaand Monocrystal Yttria-Stabilized Zirconia in High-Temperature Pure Water, Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 46, No. 12, p. 1120–1128, 2009.
63. R. Katsura et al., DL-EPR Study of Neutron Irradiation in Type 304 Stainless Steel, Corrosion 48,1992.
64. Ohsaka, T., F. Izumi, and Y. Fujiki, Raman spectrum of anatase, TiO2. Journal of Raman Spectroscopy, 1978. 7(6): p. 321-324.