研究生: |
葉育維 Yeh, Yu-Wei |
---|---|
論文名稱: |
在模擬壓水式反應器一次側水環境中溶氫對鎳基合金與沃斯田鐵系列不銹鋼之電化學行為影響分析 Effect of Dissolved Hydrogen on the Electrochemical Characteristics of Nickel-based Alloy and Austenite Stainless Steel in Simulated PWR Primary Water Environment |
指導教授: |
葉宗洸
Yeh, Tsung-Kuang 王美雅 Wang, Mei-Ya |
口試委員: |
黃俊源
Huang, Jiunn-yuan 馮克林 Fong, Clinton |
學位類別: |
碩士 Master |
系所名稱: |
原子科學院 - 工程與系統科學系 Department of Engineering and System Science |
論文出版年: | 2020 |
畢業學年度: | 108 |
語文別: | 中文 |
論文頁數: | 100 |
中文關鍵詞: | 鎳基合金 、不銹鋼 、溶氫濃度變化 、PWR一次側水迴路 、腐蝕與電化學行為 |
外文關鍵詞: | PWR primary water, Electro-chemistry |
相關次數: | 點閱:2 下載:0 |
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在壓水式反應器一次側水迴路與冷卻系統中,鎳基合金與沃斯田鐵系列不銹鋼為常見的的結構組件材料,而在一次側水迴路中的飼水端會注入氫氣,此設計的目的是用來抑制材料的均勻腐蝕、應力腐蝕龜裂現象和核反應時產生的輻射分解效應,但實際運轉經驗以及許多實驗結果顯示,在目前EPRI規範的注氫濃度25-50 cc H2/kg H2O與運轉溫度320-360 oC下,鎳基合金組件材料會有應力腐蝕龜裂的現象發生,其原因為在此注氫濃度與運轉溫度,鎳基合金會處在Ni/NiO的相轉換交界處,因此不論是鎳或是生成的氧化鎳層,都處在不穩定的狀態,而降低保護基材能力,從而使裂縫容易生成與成長。因此世界各國的核電廠決定調整注氫的濃度,歐美國家主張從原先規範的25-50 cc H2/kg H2O,提高至金屬鎳為穩定相的注氫濃度(> 75 cc H2/kg H2O),而日本的學者則認為應該降低注氫濃度使氧化鎳為穩定相(< 5 cc H2/kg H2O)。這兩種方法相對於目前的條件有一定的優缺點,而本研究將會從材料腐蝕與電化學行為的角度,來分析與探討提高與降低注氫濃度所造成的影響。
在本研究中,主要會將鎳基合金Alloy 600、Alloy 690、Alloy X750與沃斯田鐵系列不銹鋼SS316、SS316L放置於320 oC的模擬PWR一次側水環境中進行實驗,並調整實驗環境中的溶氫濃度,由低到高分別為0, 5, 30, 75 cc H2/kg H2O,透過動態電位極化掃描分析不同材料在不同環境下的腐蝕電位與腐蝕電流密度變化,並透過拉曼光譜系統與場發射掃描電子顯微鏡分析材料在不同環境下,表面氧化層結構與形貌的變化。實驗結果顯示,不論在低或高的溶氫濃度(5或75 cc H2/kg H2O),材料的腐蝕電位與腐蝕電流密度都有顯著的下降,而環境中越高的溶氫濃度會使材料有越低的腐蝕電位與腐蝕電流密度,代表氫氣對於抑制材料腐蝕行為有明顯的效果。而在表面分析的部分,SS 316與SS 316L的表面形貌與結構較為接近,Alloy 600則與Alloy X750有類似的表面形貌與結構,且在5種材料中,皆能觀察到尖晶石氧化物的比例隨著溶氫濃度的增加而增加。
Nickel-based alloys and Austenite stainless steels are major construction material used reactor coolant system of a PWR. To mitigate the problem of general corrosion and stress corrosion cracking, hydrogen is added to maintain the reducing conditions in PWR primary coolant to minimize general corrosion of material surface, and risk of stress corrosion cracking of stainless steels and nickel-based alloys. At the dissolved hydrogen concentration of 25-50 cc H2/kg H2O, the hydrogen contents may affect the nickel-based alloy surface stability due to the nickel/nickel oxide transition and lead to a higher crack growth rate. A change of hydrogen injection rate from 25-50 cc H2/kg H2O to <5 cc H2/kg H2O or to >75 cc H2/kg H2O is beneficial in avoiding hydrogen induced cracking in nickel-based alloy.
In the study, with four different dissolved hydrogen contents (0, 5, 30, 75 cc H2/kg H2O), the electrochemical behavior and the surface morphology of Alloy 600, Alloy 690, Alloy X750, SS316, and SS316L in simulated PWR primary water at 320 oC were investigated by potential dynamic test, scanning electron microscopy and Raman spectroscopy. Hydrogen injection affects the corrosion behavior significantly, the electrochemical potential become lower and the corrosion rate of alloy reduce to a smaller value. Increasing DH content results in a more negative ECP and a lower corrosion rate.
[1] Seong Sik Hwang, Review of PWSCC and mitigation management strategies of Alloy 600 materials of PWRs, Journal of Nuclear Materials 443 (2013) p.321–330, 2003.
[2] IAEA, Nuclear energy series no. NP-T-3.13, Stress corrosion cracking in light water reactors: Good practices and Lessons, IAEA, Vienna, Austria, 2011.
[3] EPRI-1007832, PWSCC of Alloy 600 Type Materials in Non-Steam Generator Tubing Applications-Survey Report through June 2002: Part 1: PWSCC in Components Other Than CRDM/CEDM Penetrations (MRP-87), EPRI, Palo Alto, CA, USA, 2003.
[4] NUREG-1823, U.S. Plant Experience With Alloy 600 Cracking and Boric Acid Corrosion of Light-Water Reactor Pressure Vessel Materials, U.S. Nuclear Regulatory Commission, Washington DC, 2005.
[5] EPRI-103696, PWSCC of Alloy 600 Materials in PWR Primary System Penetrations, EPRI, Palo Alto, CA, USA, 1994.
[6] EPRI-1014986, Pressurized Water Reactor Primary Water Chemistry Guidelines, Vol. 1, Revision 6, EPIR, Palo Alto, CA, USA, 2007.
[7] S.A. Attanasio, D.S. Morton, LM-03K049, Measurement of the Nickel/Nickel oxide transition in Ni–Cr–Fe alloys and updated data and correlations to quantify the effect of aqueous hydrogen on primary water SCC, Lockheed Martin, Schenectady, NY 12301, 2003.
[8] K. Dozaki, D. Akutagawa, N. Nagata, H. Takiguchi, K. Noring, E-journal Adv. Maint. 2 (2010) 65–76.
[9] Y.S. Lim, Hong Pyo Kim, Seong Sik Hwang, Joung Soo Kim, Dong Jin Kim, Sung Woo Kim, Man Kyo Jung, Hai Dong Cho, Whung Whoe Kim, Failure Analysis on RVH Vent Pipe in Yonggwang Unit 3, KAERI/CR-373/2010, 2010.
[10] K. Norring, B. Rosborg, J. Engstrom, J. Svenson, Influence of LiOH and H2 on primary side IGSCC of Alloy 600 steam generator tubes, in: Colloque International Fontevraud II, September 10–14, 1990, Socie´ te´ Franc, aise d’Energie Nucle´ aire: Paris, France, 1990, pp. 243–249.
[11] P.M. Scott, Prediction of Alloy 600 Component Failures in PWR Systems, Proceedings of Corrosion ‘96 Research Topical Symposia, Part 1 – Life Prediction of Structures Subject to Environmental Degradation, Denver, CO., National Association of Corrosion Engineers: Houston, TX, 1996, p. 135.
[12] T. Cassagne, B. Fleury, F. Vaillant, D.D. Bouvier, P. Combrade, An update on the influence of hydrogen on the PWSCC on Nickel base alloys in high temperature water, in: 8th International Conference on Environmental Degradation of Materials in Nuclear Power Systems, Amelia Island, Fl, USA, August 10–14, 1997.
[13] D.S. Morton, S.A. Attanasio, J.S. Fish, M.K. Schurman, Influence of dissolved hydrogen on nickel alloy SCC in high temperature water, in: Corrosion 1999 NACE Conference, Paper 447, April 1999.
[14] S. Fyfitch, Alloy 600 PWSCC : Historical Perspective, EPRI PWSCC workshop 2007.
[15] Hunt, E.S., Gross D.J., 1994. ‘PWSCC of Alloy 600 Materials in PWR Primary System Penetrations’ EPRI TR-103696.
[16] P. Scott, A review of PWSCC of Nickel-base Alloys-SCAP SCC Project Progress Meeting February 18–19, 2010 OECD Head quarter-France.
[17] Warren Bamford, Service experience with Alloy 182 Butt welds, and the weld overlay, mitigation process, 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Whistler, British Columbia August 19–23, 2007.
[18] <http://pbadupws.nrc.gov/docs/ML1126/ML11266A011.pdf.>
[19] D. D. Macdonald, Corrosion, vol. 48, p. 194, 1992.
[20] Herbert H. Uhlig & R. Winston Revie , Corrosion and Corrosion Control, Chapter 7.
[21] C. Amzallag, S. Le Hong, C. Pagès, A. Gelpi, "Stress corrosion life assessment of Alloy 600 components", Proceedings of 9th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, Newport Beach, CA, The Metallurgical Society (1999) p 243-250.
[22] EPRI-1014986, Pressurized Water Reactor Primary Water Chemistry Guidelines, Vol. 1, Revision 6, EPIR, Palo Alto, CA, USA, 2007.
[23] S.A. Attanasio, D.S. Morton, M.A. Ando, N.F. Panayotou, C.D. Thompson, Measurement of the Nickel/Nickel Oxide Phase Transition in High Temperature Hydrogenated Water Using the Contact Electric Resistance (CER) Technique, in: Proceedings of the Tenth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-water Reactors, Lake Tahoe, Nevada, USA, August 5-9, NACE, Houston, TX, 2001.
[24] T. Cassagne, F. Vaillant, P. Combrade, An Update on the Influence of Hydrogen on the PWSCC of Nickel Base Alloys in High Temperature Water, in: Proceedings of Eighth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-water Reactors, ANS, Amelia island, Florida, 1997. August 10-14.
[25] D.H. Lee, M.S. Choi, U.C. Kim, The Effect of Hydrogen on the Stress Corrosion Cracking of Alloy 600 in Simulated PWR Primary Water at 330oC, in: Proceedings of the Tenth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-water Reactors, Lake Tahoe, Nevada, USA, August 5-9, NACE, Houston, TX, 2001.
[26] D.S. Morton, S.A. Attanasio, J.S. Fish, M.K. Schurman, Influence of Dissolved Hydrogen on Nickel Alloy SCC in High-temperature Water, Paper 447, Corrosion’99, Texas, USA, April 25-30, NACE, Houston, TX, 1999.
[27] D.S. Morton, S.A. Attanasio, G.A. Young, Primary Water SCC Understanding and Characterization through Fundamental Testing in the Vicinity of the Nickel/Nickel Oxide Phase Transition, in: Proceedings of the Tenth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-water Reactors, Lake Tahoe, Nevada, USA, August 5-9, NACE, Houston, TX, 2001.
[28] Soon-Hyeok Jeon, Eun-Hee Lee, Do Haeng Hur, Effects of dissolved hydrogen on general corrosion behavior and oxide films of alloy 690TT in PWR primary water, Journal of Nuclear Materials 485 (2017) 113-121.
[29] T.M. Angeliu, G.S. Was, Grain boundary chemistry and precipitation in controlled purity Alloy 690, in: D. Cubicciotti (Ed.), Proc. 4th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, NACE, Houston, TX, 1990, pp. 5-64.
[30] G.S. Was, Corrosion 46 (1990) 319-330.
[31] J.R. Donati, M. Guttmann, Y. Rouillon, P. Saint-Paul, J.C. Van Duysen, Stress corrosion cracking behavior of nickel base alloys with 19% chromium in high temperature water, in: G.J. Theus, J.R. Week (Eds.), Proc. 3rd International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, TMS, Warrendale, PA, 1988, pp. 697-701.
[32] D. Noel, O. de Bouvier, F. Foot, T. Mignin, J.D. Mithieux, F. Vaillant, Study of the mechanisms of stress corrosion cracking of Alloys 600 and 690 in primary water reactor conditions, in: T. Magnin (Ed.), Corrosion-Deformation Interactions, CDI'96, The Institute of Materials, London, 1997, pp. 435-452.
[33] G.S. Was, R.M. Kruger, Acta Metall. 33 (1985) 841-854.
[34] K. Norring, K. Stiller, J.-O., Nilsson, Grain boundary microstructure, chemistry, and IGSCC in Alloy 600 and Alloy 690, in: Proc. 5th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, ANS, La Grange Park, IL, 1992, pp. 482-487.
[35] S.M. Bruemmer, C.H. Henager, Jr., in: Proc. 2nd International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, ANS, La Grange Park, IL, 1986, pp. 293-300.
[36] J. Hertzberg, G. Was, Metall. Mater. Trans. A, 29A (1998) 1035-1046.
[37] J. Panter, B. Viguier, J.-M. Cloué, M. Foucault, P. Combrade, E. Andrieu, Influence f oxide films on primary water stress corrosion cracking initiation of alloy 600, J. Nucl. Mater. 348 (2006) 213–221.
[38] S.Y. Persaud, A. Korinek, J. Huang, G.A. Botton, R.C. Newman, Internal oxidation of Alloy 600 exposed to hydrogenated steam and the beneficial effects of thermal treatment, Corros. Sci. 86 (2014) 108–122.
[39] G. Bertali, F. Scenini, M.G. Burke, The intergranular oxidation susceptibility of thermally-treated Alloy 600, Corros. Sci. 114 (2017) 112–122.
[40] Yun Soo Lim, Dong Jin Kim, Sung Woo Kim, Seong Sik Hwang, Hong Pyo Kim, Characterization of internal and intergranular oxidation in Alloy 690 exposed to simulated PWR primary water and its implications with regard to stress corrosion cracking, Materials Characterization 157 (2019) 109922.
[41] Y.S. Lim, S.W. Kim, S.S. Hwang, H.P. Kim, C. Jang, Intergranular oxidation of Nibased Alloy 690 in a simulated PWR primary water environment, Corros. Sci. 108 (2016) 125–133.
[42] P. Combrade, P.M. Scott, M. Foucault, E. Andrieu, P. Marcus, Oxidation of Ni base alloys in PWR water: oxide layers and associated damage to the base metal, Proc. Of the 12th Int. Conf. on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactor, 2005, pp. 883–890.
[43] P. Scott and C. Benhamou, "An overview of recent observations and interpretation of IGSCC in nickel base alloys in PWR primary water", Proceedings of 10th Int. Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Lake Tahoe, NACE International (2001).
[44] R. W. Staehle, J. A. Gorman, K. D. Stavropoulos and C. S. Welty, "Application of statistical distributions to characterizing and predicting corrosion of tubing in steam generators of Pressurized Water Reactors", Proceedings of Life Prediction of Corrodible Structures, ed. R. N. Parkins, NACE International (1994) p 1374-1439.
[45] D. Noel, O. de Bouvier, F. Foot, T. Mignin, J.D. Mithieux, F. Vaillant, Study of the mechanisms of stress corrosion cracking of Alloys 600 and 690 in primary water reactor conditions, in: T. Magnin (Ed.), Corrosion-Deformation Interactions, CDI'96, The Institute of Materials, London, 1997, pp. 435-452.
[46] R. Bandy, D. van Rooyen, Corrosion 40 (1984) 425-430.
[47] R.N. Parkins, Corrosion 46 (1990) 178-189.
[48] D. Lee, D.A. Vermilyea, Metall. Trans. 2 (1971) 2565-2571.
[49] T. M. Angeliu, D.J. Paraventi, G.S. Was, Corrosion 51 (1995) 837-848.
[50] M.M. Hall, Jr., Thermally activated dislocation creep model for primary water stress corrosion cracking of Ni-Cr-Fe Alloys, in: T. Shoji, I. Shibata (Eds.), Proc. International Symposium on Plant Aging and Life Prediction of Corrodible Structures, NACE International, Houston, TX, 1997, pp. 107-116.
[51] M.M. Hall, D.M. Symons, Hydrogen assisted creep fracture model for low potential stress corrosion cracking of Ni-Cr-Fe alloys, in: R.H. Jones (Ed.), Chemistry and Electrochemistry of Corrosion and Stress Corrosion Cracking: A Symposium Honoring the Contributions of R.W. Staehle, TMS, Warrendale, PA, 2001, pp. 447-466.
[52] J.M. Gras, Stress corrosion cracking of stream generator tubing materials, in: S.M. Bruemmer, E.I. Meletis, R.H. Jones, W.W. Gerberich, F.P. Ford, R.W. Staehle (Eds.), Parkins Symposium on Fundamental Aspects of Stress Corrosion Cracking, TMS, Warrendale, PA, 1992, pp. 411-432.
[53] J.M. Gras, Stress corrosion cracking of stream generator tubing materials, in: S.M. Bruemmer, E.I. Meletis, R.H. Jones, W.W. Gerberich, F.P. Ford, R.W. Staehle (Eds.), Parkins Symposium on Fundamental Aspects of Stress Corrosion Cracking, TMS, Warrendale, PA, 1992, pp. 411-432.
[54] Soon-Hyeok Jeon, Eun-Hee Lee, Do Haeng Hur, Effects of dissolved hydrogen on general corrosion behavior and oxide films of alloy 690TT in PWR primary water, Journal of Nuclear Materials 485 (2017) 113-121.
[55] Yun Soo Lim, Dong Jin Kim, Sung Woo Kim, Seong Sik Hwang, Hong Pyo Kim, Characterization of internal and intergranular oxidation in Alloy 690 exposed to simulated PWR primary water and its implications with regard to stress corrosion cracking, Materials Characterization 157 (2019) 109922.
[56] J. Robertson, Corros. Sci. 32 (1991) 443-465.
[57] Takumi Terachi , Takuyo Yamada , Tomoki Miyamoto , Koji Arioka & Koji Fukuya, Corrosion Behavior of Stainless Steels in Simulated PWR Primary Water—Effect of Chromium Content in Alloys and Dissolved Hydrogen, Journal of Nuclear Science and Technology (2008), 45:10, 975-984.
[58] Xiangyu Zhong, Shuang Xia b, Jian Xu, Tetsuo Shoji, The oxidation behavior of 316L in simulated pressurized water reactor environments with cyclically changing concentrations of dissolved oxygen and hydrogen, Journal of Nuclear Materials 511 (2018) 417-427.
[59] Jongjin Kim, In situ Raman spectroscopic analysis of surface oxide films on Ni-base alloy/low alloy steel dissimilar metal weld interfaces in high-temperature water, Journal of Nuclear Materials 449 (2014) 181–187.
[60] Ji Hyun Kim, Il Soon Hwang, Development of an in situ Raman spectroscopic system for surface oxide films on metals and alloys in high temperature water, Nuclear Engineering and Design 235 (2005) 1029–1040.
[61] Maslar, J.E., Hurst, W.S., Bowers W.J.Jr., Hendricks, J.H., Aquino, M.I., 2002a, In situ Raman spectroscopic investigation of nickel hydrothermal corrosion, Corrosion 58, 225.
[62] Maslar, J.E., Hurst, W.S., Bowers Jr., W.J., Hendricks, J.H., 2002b, In situ Raman spectroscopic investigation of stainless steel hydrothermal corrosion, Corrosion 58, 739, 2002.
[63] J.E. Maslar, W.S. Hurst, W.J. Bowers, J.H. Hendricks, M.I. Aquino, J. Electrochem. Soc. 147 (2000) 2532–2542.
[64] J.E. Maslar, W.S. Hurst, W.J. Bowers Jr., J.H. Hendricks, M.I. Aquino, I. Levin, Appl. Surf. Sci. 180 (2001) 102–118.
[65] G.G. Siu, M.J. Stokes, Phys. Rev. B 59 (1999) 3173–3179.
[66] 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.
[67] Oh, S.J., D. Cook, and H. Townsend, Characterization of iron oxides commonly formed as corrosion products on steel. Hyperfine interactions, 1998. 112 p. 59-66.
[68] J. Gui and T.M. Devine, Proc. Of the 12th Int. Corrosion Congress, NACE Internaional, Houston, 1993.
[69] J. Dunnwald and A. Otto, Corrosion Science 29, 1989.
[70] T. Ohtsuka, K. Kubo and N. Sato, Corrosion 42, 1986.
[71] T. Ohtsuka, Materials Transactions, JIM 37, 1996.
[72] N.Boucherit, P. Delicher, S. Joiret and A. Hugot-Le Goff, Materials Science Forum 51, 1989.
[73] D.Thierry, D. Persson, C. Lyegraf, D. Delichere, S.Joiret, C. Pallotta and A. Hugot-Le Goff, J. Electrochem. Soc. 135, 1988.
[74] R.J. Thibeau, C.W. Rrown and R.H. Heidersbach, Applied Spectroscopy 32, 1978
[75] M.E. Indig, C. Groot, Corrosion 26 (1970) 171-176.
[76] W.K. Lai, Z. Szklarska-Smialowska, Corrosion 47 (1991) 40-47.
[77] S.A. Attanasio, D.S. Morton, M.A. Ando, CORROSION 2002, Paper 02517, NACE, 2002.
[78] Yubing Qiu, Tetsuo Shoji, Zhanpeng Lu, Effect of dissolved hydrogen on the electrochemical behaviour of Alloy 600 in simulated PWR primary water at 290 oC, Corrosion Science 53 (2011) 1983–1989