隨著核電廠運轉時間的增加，以鎳基600合金材料為主的蒸汽產生器及以316不銹鋼材料為主的各種管件，應用於輕水式反應器(Light Water Reactor, LWR)時，均出現不同程度的應力腐蝕龜裂 (Stress Corrosion Cracking)。壓水式反應器一次側迴路的冷卻水，因分別添加了硼酸及氫氧化鋰控制反應度及酸鹼度，導致冷卻水有較高的導電度及氧化性，因此必須於水中添加氫氣，以提升冷卻水的還原性，並藉以抑制腐蝕。實際運轉經驗以及許多實驗結果顯示，在目前美國電力研究所（Electric Power Research Institute, EPRI）規範的溶氫濃度25-50 cc/kg H2O 與運轉溫度 320-360oC 下，鎳基合金組件材料會有應力腐蝕龜裂的現象發生，其原因為在此溶氫濃度與運轉溫度，鎳基合金會處在 Ni/NiO的相轉換交界處，使氧化層不穩定而降低保護基材的能力，從而使裂縫容易生成與成長。歐洲的學者希望將核電廠的溶氫濃度提高到75cc/kg H2O；日本的學者則是希望降低到5cc/kg H2O以節省成本。兩種方法各有優缺點，針對低溶氫的部份，硼原子被鬆散的氧化層捕捉，隨著H+濃度上升主冷卻水應力腐蝕龜裂(Primary Water Stress-Corrosion Cracking, PWSCC)發生的機率增加；高溶氫則是富鉻區不連續而導致材料的敏感性增加。另外，爐心區域因為冷卻水輻射分解效應造成氧化劑濃度較其他區域為高，可能達到10ppb以上，晶界氧化使材料敏感性增加而導致PWSCC發生。
為了瞭解 316L不銹鋼、600合金和690合金於低溶氧量、低溶氫量、高溶氫量的應力腐蝕龜裂的探討，本研究使用U 型試片，在 320oC 和 330oC 動態高溫水循環迴路中測試各個試片的應力腐蝕裂縫起始及成長。實驗結束後，透過掃描式電子顯微鏡、拉曼分析、XPS縱深分析觀察試片的表面形貌。結果顯示，三種材料在所有測試條件下均未出現明顯的應力腐蝕龜裂起始差異；因此，即使溶氫量調降至5cc/kg H2O，對於本研究所選用的三種材料仍具有防蝕保護效果。

Hydrogen addition is widely used to inhibit the generation of oxidizing radiolytic products in the coolants of light water reactors. However, hydrogen would also be one potential factor to enhance primary water stress corrosion cracking (PWSCC). Some laboratory studies showed that if 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 order to have a better understanding to the SCC in 316L SS, Alloy 600 and 690 in the coolant with a lower DH, the SCC initiation behavior was investigated via U-bend tests in 320oC and 330oC high-temperature water.
After the U-bend tests, morphologies of the surfaces of the samples were also examined by scanning electron microscopy (SEM)、XPS as well as Raman spectroscopy . The outcomes indicated that three tested materials exhibited no significant effect after 1500 hours test at a higher temperature due to a protective oxide layer formed on the surface that can inhibit crack propagation.
In other words, even though we reduced the hydrogen concentration down to 5cc/kg H2O, it remained an outstanding effect on corrosion inhibition.

[1] X. Zhong, S. C. Bali, and T. Shoji, “Effects of Dissolved Hydrogen on the Environtmentally Assisted Cracking of 316 Stainless Steel in Pwr Primary Water At 325O C,” pages 1–20, 2015.
[2] L. Marchetti, F. Martin, F. Datcharry, and J. Chêne, “Kinetics of hydrogen permeation through a Ni-base alloy membrane exposed to primary medium of pressurized water reactors,” Corrosion Sceince , November 2018, Pages 1-12.
[3] T. Kim, K. J. Choi, S. C. Yoo, and J. H. Kim, “Effects of dissolved hydrogen on the crack-initiation and oxidation behavior of nickel-based alloys in high-temperature water,” Corros. Sci., vol. 106, pp. 260–270, 2016.
[4] Takumi TERACHI , Katsuhiko FUJII and Koji ARIOKA, “Microstructural Characterization of SCC Crack Tip and Oxide Film for SUS 316 Stainless Steel in Simulated PWR Primary Water at 320oC”, Institute of Nuclear Safety System, Inc., 64 Sata, Mihama-cho, Mikata-gun, Fukui 919-1205, October 29, 2004.
[5] Junho Lee , Changheui Jang , Kyoung Soo Lee , “Evaluation of the Sensitization of 316L Stainless Steels After the Post Weld Heat Treatment”, Central Research Institute, Korea Hydro and Nuclear Power Co., Ltd., 1312 Gil, 70. Yuseongdaero, Yuseong-gu, Daejeon, 305-343, Rep. of Korea, Dept. of Nuclear and Quantum Engineering, KAIST, Daejeon, Rep. of Korea, May 29-30, 2014.
[6] G.S. Was, Corrosion 46(1990) 319-330
[7] G.S. Was, R.M Kruger, Acta Metall(1985) 841-854
[8] 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.
[9] 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
[10] 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.
[11] S. F. H. Xu, “Laboratory Investigation of PWSCC of CRDM Nozzle 3 and Its J-Groove Weld on the Davis-Besse Reactor Vessel Head,” in Proc. 12th Int. Conf. on Environmental Degradation of Materials in Nuclear Power Plant.
[12] Youwei Xu, Hongyang Jing, Y. D. Han, “Effect of δ-ferrite on stress corrosion cracking of CF8A austenitic stainless steels in a simulated pressurised water reactor environment”, Journal of Materials Research and Technology, November 2019.
[13] J. A. G. R.W. Staehle, “Quantitative Assessment of Submodes of Stress Corrosion Cracking on the Secondary Side of Steam Generator Tubing in Pressurized Water Reactors,” Corrosion, vol. 60, pp. 115–180, 2004.
[14] M. A. K. Arioka, T. Yamada, T. Miyamoto, “Intergranular Stress Corrosion Cracking Growth Behavior of Ni-Cr-Fe Alloys in Pressurized Water Reactor Primary Water,” Corrosion, vol. 70, pp. 695–707, 2014.
[15] EPRI-1014986, Pressurized Water Reactor Primary Water Chemistry Guidelines, Vol. 1, Revision 6, EPIR, Palo Alto, CA, USA, 2007
[16] 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.
[17] 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.
[18] H. W. H. Coriou, L. Grall, P. Olivier, Influence of carbon andnickel content on stress corrosion cracking of austenitic stainless alloys inpure or chlorinated water at 350◦C. Conference of fundamental aspects of stress corrosion cracking and hydrogen embrittlement of iron base alloys,page 235, Tokyo, Japan, 1969. The Iron and Steel Institute of Japan. 41, 42.
[19] P.L. Andresen, M.M. Morra,“Effects of Si on SCC of irradiated and unirradiated stainless steels and nickel alloys” , 12th International Conference on Environmental Degradation of Materials in Nuclear Power System, Salt Lake City, UT, USA (2005).
[20] G.F. Li, Y. Kaneshima, T. Shoji, “Effects of impurities on environmentally assisted crack growth of solution-annealed austenitic steels in primary water at 325°C” ,Corrosion, 56 (2000), pp. 460-469.
[21] G. Han, Z. Lu, X. Ru, J. Chen, J. Zhang, T. Shoji, “ Properties of oxide films formed on 316L SS and model alloys with modified Ni, Cr and Si contents in high temperature water” ,Corros. Sci., 106 (2016), pp. 157-171
[22] G.F. Li, Y. Kaneshima, T. Shoji, “ Effects of impurities on environmentally assisted crack growth of solution-annealed austenitic steels in primary water at 325°C” , Corrosion, 56 (2000), pp. 460-469.
[23] R.S. Dutta, R. Tewari, “Microstructural and corrosion aspects of Alloy 690”, British Corrosion, vol. 34, pp. 201–205, 1999.
[24] F.J. Meng, J.Q. Wang, E.H. Han, W. Ke, “The role of TiN inclusions in stress corrosion crack initiation for Alloy 690TT in high-temperature and high-pressure water”, Corrosion Science, vol. 52, pp. 927–932, 2010.
[25] Lee, T., Delley, B., Stampfl, C., & Soon, A, “Environment-dependent nanomorphology of TiN: the influence of surface vacancies”, Nanoscale, vol. 4, pp. 5183, 2012.
[26] 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.
[27] 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.
[28] 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.
[29] 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.
[30] 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.
[31] J.E. Maslar, W.S. Hurst, W.J. Bowers, J.H. Hendricks, M.I. Aquino, J. Electrochem. Soc. 147 (2000) 2532–2542.
[32] 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.
[33] G.G. Siu, M.J. Stokes, Phys. Rev. B 59 (1999) 3173–3179.
[34] Jongjin Kim, In situ Raman spectroscopic analysis of surface oxide films on Nibase alloy/low alloy steel dissimilar metal weld interfaces in high-temperature water, Journal of Nuclear Materials 449 (2014) 181–187.
[35] 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.
[36] J. Gui and T.M. Devine, Proc. Of the 12th Int. Corrosion Congress, NACE Internaional, Houston, 1993.
[37] J. Dunnwald and A. Otto, Corrosion Science 29, 1989.
[38] T. Ohtsuka, K. Kubo and N. Sato, Corrosion 42, 1986.
[39] T. Ohtsuka, Materials Transactions, JIM 37, 1996.
[40] N.Boucherit, P. Delicher, S. Joiret and A. Hugot-Le Goff, Materials Science Forum 51, 1989.
[41] D.Thierry, D. Persson, C. Lyegraf, D. Delichere, S.Joiret, C. Pallotta and A. Hugot Le Goff, J. Electrochem. Soc. 135, 1988.
[42] R.J. Thibeau, C.W. Rrown and R.H. Heidersbach, Applied Spectroscopy 32, 1978.
[43] Soon-Hyeok Jeon , Geun Dong Song , and Do Haeng Hur, “Effects of Hydrogen Contents on Oxidation Behavior of Alloy 690TT and Associated Boron Accumulation within Oxides in High-Temperature Water”, Volume 2018, Article ID 7845176, 12 pages.
[44] ASTM G-30, “Making and Using U-Bend Stress-Corrosion Test Specimens 1,” vol. 97, no. Reapproved 2003, pp. 1–7, 2008.
[45] H. Agrawal, P. Sharma, P. Tiwari, R. V. Taiwade, and R. Dayal, “Evaluation of Self-Healing Behaviour of AISI 304 Stainless Steel,” Trans. Indian Inst. Met., vol. 68, 2015.
[46] J.R.Davis, Corrosion of Weldments:ASM International, 2006.
[47] M. A. K. Arioka, T. Yamada, T. Miyamoto, “Intergranular Stress Corrosion Cracking Growth Behavior of Ni-Cr-Fe Alloys in Pressurized Water Reactor Primary Water,” Corrosion, vol. 70, pp. 695–707, 2014.
[48] 國立清華大學貴重儀器中心
[49] https://www.istgroup.com/tw/tech_20190221/
[50] I. K. R.W. Staehle, Anatomy of Proactivity, in: B.L. Eyre, “No Title,” in Int. Sym. on Research for Aging Management of Light Water Reactors and Its Future Trend (The 15th Anniversary of INSS), 2008, pp. 29–115.
[51] J. A. Roberts,“Structural materials in nuclear power systems,” Springer Sci. Bus. Media, 2013.
[52] A. S. M. International and A. Rights, ASM specialty handbook: nickel, cobalt, and their alloys, vol. 38, no. 11. 2013.
[53] Zhao Shen,“ An insight into PWR primary water SCC mechanisms by comparing surface and crack oxidationa,” Department of Materials, University of Oxford, Parks Road, OX1 3PH Oxford, UK, 14 December 2018.
[54] M. Michael, STAINLESS STEELSFOR DESIGN ENGINEERS.
[55] Practical guildlines for the fabrication of Austentic stainless steel
[56] Chang, Che-Jung, “Effect of Dissolved Hydrogen on the SCC initiation of Stainless steel and Ni-based Alloy in Simulated PWR primary water Environments” 2019.
[57] R. Staehle, “Definition of precursors, incubation, slow growth and propagation of SCC,” SCC Initiat. Work., vol. 2, no. 3, pp. 79–87, 2008.
[58] 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.
[59] 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 NuclearMaterials 511 (2018) 417-427.
[60] A. Turnbull. Stress corrosion cracking: mechanisms. In K. H. Jürgen Buschow, Robert W. Cahn, Merton C. Flemings, and Bernard Ilschner, editors, Encyclopedia of Materials: Science and Technology, pages 8886 – 8891. Elsevier, Oxford, 2001. 37