Review of the Effect of Irradiation-Assisted Stress Corrosion Cracking on Stainless Steel in Light Water Reactor Environments
XU Chaoliang1,2,4, QUAN Qiwei2,4, WU Huanchun2,4, LI Yuanfei2,4, JIA Wenqing2,4, YIN Jian2,4, LI Shilei3, SONG Miao1, ZHANG Lefu1, LIU Xiangbing2,4,*, GUO Xianglong1,*
1 School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240, China 2 Material Engineering Center, Suzhou Nuclear Power Research Institute Co., Ltd., Suzhou 215004, Jiangsu, China 3 State Key Laboratory for Advanced Metals and Materials, University of Science and Technology Beijing, Beijing 100083, China 4 National Engineering Research Center for Nuclear Power Plant Safety & Reliability, Suzhou 215004, Jiangsu, China
Abstract: Stainless steel materials used in the internal components of light water reactors (LWRs) operate under conditions of high temperature, high irradiation,and reactor primary water environment. Irradiation-assisted stress corrosion cracking (IASCC) is the primary failure mechanism, leading to multiple fracture incidents in reactor internals. Through investigation and analysis, this summary introduces the irradiation environment in LWRs. Then an analysis is given about the impact of irradiation on the microstructure and hardness, micro-regional chemical elements, electrochemical corrosion potential (ECP) of the primary water, temperature, and stress in stainless steel. Furthermore, the influence patterns of these factors on IASCC is summarized.
1 Was G S, Ashida Y, Andresen P L. Corrosion Reviews, 2011, 29, 7. 2 International Atomic Energy Agency. Assessment and management of ageing of major nuclear power plant components important to safety: PWR vessel internals. IAEA, 2007. 3 Somville F, Gerard R, Bosch R W, et al. In: 2016 International LWR Material Reliability Conference and Exhibition. Chicago, 2016, pp. 1. 4 Qin Wei, Xu Chaoliang, Liu Xiangbing, et al. Journal of Physics: Conference Series, 2025, 2941, 012006. 5 Robert Gérard, Frédéric Somville. In: Proceedings of the 17th International Conference on Nuclear Engineering. Brussels, 2009, pp. 521. 6 Kameyama Masashi, Chigusa Naoki, Kubo Noboru, et al. Engineering, Materials Science Quarterly Journal of the Japan Welding Society, 2005, 23, 77. 7 Was G S. Fundamentals of radiation materials science-metals and alloys. Springer, USA, 2017, pp. 951. 8 Kenik E A, Busby J T. Materials Science and Engineering R, 2012, 73, 67. 9 Garner F A, Greenwood L R. In:Proceedings of the 11th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. 2003, pp.887. 10 Garner F A. Radiation-induced damage in austenitic structural steels used in nuclear reactors, Elsevier, USA, 2020, pp. 63. 11 Bruemmer S M, Simonen E, Scott P M, et al. Journal of Nuclear Materials, 1999, 274, 299. 12 Deng Ping, Sun Chen, Peng Qunjia, et al. Journal of Chinese Society for Corrosion and Protection, 2015, 35(6), 479 (in Chinese). 邓平, 孙晨, 彭群家, 等. 中国腐蚀与防护学报, 2015, 35(6), 479. 13 Zinkle S J, Maziasz P J, Stoller R E. Journal of Nuclear Materials, 1993, 206, 266. 14 Yu Jinnan. Radiation effects of materials, Chemical Industry Press, China, 2007, pp. 416 (in Chinese). 郁金南. 材料辐照效应, 化学工业出版社, 2007, pp. 416. 15 Chopra O K, Rao A S. Journal of Nuclear Materials, 2011, 409, 235. 16 Robertson J P, Ioka I, Rowcliffe A F, et al. In: Effects of Radiation on Materials: 18th International Symposium. Hyannis, 1999, pp. 671. 17 Seeger A K. On the theory of radiation damage and radiation hardening, Max-Planck-Institure, 1959. 18 Pokor C, Bréchet Y, Dubuisson P, et al. Journal of Nuclear Materials, 2004, 326, 30. 19 Mme Jyoti Gupta. Intergranular stress corrosion cracking of ion irradiated 304l stainless steel in PWR environment. Ph. D. Thesis, Université De Toulouse, France, 2016. 20 Lucas G E. Journal of Nuclear Materials, 1993, 206, 287. 21 Tsay K V, Maksimkin O P, Turubarova L G, et al. Journal of Nuclear Materials, 2013, 439, 148. 22 Garner F A. Radiation damage in austenitic steels, Elsevier, USA, 2012, pp. 34. 23 Konings R J M. Comprehensive nuclear materials, Elsevier, USA, 2012, pp. 1. 24 Was G, Andresen P. Corrosion, 2007, 63, 19. 25 Terachi T, Yamada T, Miyamoto T, et al. Journal of Nuclear Materials, 2012, 426(1-3), 59. 26 Zinkle S J, Was G S. Acta Materialia, 2013, 61(3), 735. 27 Stephenson K J, Was G S. Journal of Nuclear Materials, 2015, 456, 85. 28 Zhou R, West E A, Jiao Z J, et al. Journal of Nuclear Materials, 2009, 395, 11. 29 Jiao Z, Was G S. Journal of Nuclear Materials, 2010, 407, 34. 30 Jiao Z, Was G S. Journal of Nuclear Materials, 2011, 408, 246. 31 Lam N Q, Kumar A, Wiedersich H. Proceedings of the eleventh international symposium, Scottsdale, 1982, pp. 985. 32 Toyama T, Nozawa Y, Van Renterghem W, et al. Journal of Nuclear Materials, 2011, 418, 62. 33 Anthony T R. Acta Metallurgica, 1969, 17(5), 603. 34 Simonen E P, Bruemmer S M. Journal of Nuclear Materials, 1996, 239, 185. 35 Allen T R, Was G S, Kenik E A. Journal of Nuclear Materials, 1997, 244, 278. 36 Johnson R A, Lam N Q. Physical Review B, 1977, 15, 1794. 37 Konings R J M, Stoller R E. Comprehensive nuclear materials, Elsevier, USA, 2020, pp. 1. 38 Pathania. Analytical transmission electron microscopy (ATEM) characterization of stress-corrosion cracks in LWR-irradiated austenitic stainless steel core components. EPRI, 2003. 39 Pathania R S, Nelson J L. The use of proton irradiation to understand IASCC in LWR cores. EPRI, 2001. 40 Was G S, Busby J T. Use of proton irradiation to determine IASCC mechanisms in light water reactors: solute addition alloys. Electric Power Research Institute, 2003. 41 Jenssen A, Ljungberg L G, Walmsley J, et al. Corrosion, 1998, 54, 48. 42 Jiao Z, Was G S. Acta Materialia, 2011, 59, 1220. 43 Gupta J, Hure J, Tanguy B, et al. Journal of Nuclear Materials, 2018, 501, 45. 44 Kuang W J, Feng X Y, Du D H, et al. Corrosion Science, 2022, 199, 110187. 45 Swaminathan S, Sun K, Was G S. Journal of Nuclear Materials, 2023, 585, 154604. 46 Du D H, Sun K, Was G S. Corrosion Science, 2021, 193, 109902. 47 Deng P, Peng Q J, Han E H. Scientific Report, 2021, 11, 1371. 48 Yonezawa T, Fujimoto K, Nakada S, et al. In: Proc. 8th International Symposium Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. La Grange Park, USA, 1997, pp. 823. 49 Li G F, Kaneshima Y, Shoji T. Corrosion, 2000, 56(5), 460. 50 Wang Shenkai, Feng Xinyu, Zhang Shihao, et al. In: Abstracts Collection of the 11th National Conference on Corrosion and Protection. Liaoning, 2021, pp. 1 (in Chinese). 王生凯, 冯兴宇, 张世豪, 等. 第十一届全国腐蚀与防护大会论文摘要集, 2021, pp. 1. 51 Wang Shengkai, Zhang Shihao, Xie Jiayu, et al. Acta Materialia, 2022, 241, 118408. 52 Gao Junxuan, Cao Han 1, Kuang Wenjun, et al. Journal of Chinese Society for Corrosion and Protection, 2024, 44(4), 835 (in Chinese). 高俊宣, 曹晗, 匡文军, 等. 中国腐蚀与防护学报, 2024, 44(4), 835. 53 Andresen P L, Emigh P W, Morra M M. In: NACE Corrosion. Houston, 2005, pp. NACE-05592. 54 Hashimoto N, S. Zinkle S, A. Rowcliffe A, et a. Journal of Nuclear Materials, 2000, 283-287, 528. 55 Stephenson K. The role of dislocation channeling in IASCC initiation of neutron irradiated austenitic stainless steel. Ph. D. Thesis, The University of Michigan, USA, 2016. 56 Drew Coulson Johnson. Origin and characterization of the local stress state at dislocation channel-grain boundary interaction sites and its role in irradiation assisted stress corrosion cracking. Ph. D. Thesis, University of Michigan, USA, 2020. 57 Scott P. Journal of Nuclear Materials, 1994, 211, 101. 58 Kim Y. Corrosion, 2002, 58, 208. 59 Ishigure K, Nukii T, Ono S. Journal of Nuclear Materials, 2006, 350(1), 56. 60 Justin Ryan Hesterberg. Processes responsible for the mitigation of iascc susceptibility following the post-irradiation annealing of austenitic stainless steels. Ph. D. Thesis, University of Michigan, USA, 2019. 61 Jacobs A J, Hale D A, Siegler M. Unpublished data on SCC of irradiated SS in 288 ℃ water and inert gas. GE Nuclear Energy, 1986. 62 Ljungberg L G, Atom A B B, Sweden V. Personal communication, March 1991. 63 Ashida Y, Flick A, Andresen P L, et al. In: 15th International Confe-rence on Environmental Degradation of Materials in Nuclear Power Systems-water Reactors. Colorado, 2011, pp. 1241. 64 Onchi T, Dohi K, Soneda N, et al. Journal of Nuclear Materials, 2005, 340, 219. 65 Miwa Y, Tsukada T, Tsuji H, et al. Journal of Nuclear Materials, 2002, 307-311, 347. 66 Tsukada T, Jitsukawa S, Shiba K, et al. Journal of Nuclear Materials, 1993, 207, 159. 67 Takashi Tsukada, Yukio Miwa, Shiro Jitsukawa, et al. Journal of Nuclear Materials, 2004, 329-333, 657. 68 Andresen P L. Corrosion, 1993, 49 (9), 714. 69 Zhu Ruolin, Zhang Litao, Wang Jianqiu, et al. Journal of Chinese Society for Corrosion and Protection, 2018, 38(1), 54 (in Chinese). 朱若林, 张利涛, 王俭秋, 等. 中国腐蚀与防护学报, 2018, 38(1), 54. 70 McMurtrey M D, Cui B, Robertson I, et al. Current Opinion in Solid State and Materials Science, 2015, 19, 305. 71 Mo-Rigen He, Drew C. Johnson, Gary S. Was, et al. Acta Materialia, 2017, 138, 61. 72 Evrard P, Sauzay M. Journal of Nuclear Materials, 2010, 405(2), 83. 73 Deng Ping. Irradiation assisted corrosion and stress corrosion of nuclear-grade 304 stainless steel in high temperature and high pressure water. Ph. D. Thesis, University of Science and Technology of China, China, 2018 (in Chinese). 邓平. 核级304不锈钢辐照促进高温高压水环境腐蚀与应力腐蚀研究. 博士学位论文, 中国科学技术大学. 2018. 74 Garner F, Simonen E, Oliver B, et al. Journal of Nuclear Materials, 2006, 356, 122. 75 Allen T, Cole J, Trybus C, et al. Journal of Nuclear Materials, 2006, 348, 148. 76 Garner F, Toloczko M, Sencer B. Journal of Nuclear Materials, 2000, 276, 123. 77 Neustroev V, Garner F. Journal of Nuclear Materials, 2009, 386-388, 157. 78 Chung H. Assessment of void swelling in austenitic stainless steel core internals, Argonne National Laboratory, 2006. 79 Heald P T, Speight M V. Philosophical Magazine, 1974, 29(5), 1075. 80 Brailsford A D, Bullough R. Philosophical Magazine, 1973, 27(1), 49. 81 Malcolm Griffiths. Materials, 2021, 14, 2622. 82 Was G S, Ashida Y, Andresen P L. Corrosion Reviews, 2011, 29, 7. 83 Warren Bilanin. BWRVIP-45: BWR vessel and internals project weldability of irradiated LWR structural components. EPRI, 1997. 84 Brager H R, Straalsund J L. Journal of Nuclear Materials, 1973, 46(2), 134. 85 Ullmaier H. Radiation Effects, 1983, 78(1-4), 1. 86 By S, Li M L, Grossbeck Z, et al. Welding Journal, 2011, 90(1), 19. 87 West E A, Was G S. Journal of Nuclear Materials, 2011, 408, 142. 88 Johnson D C, Kuhr B, Farkas D, et al. Acta Materialia, 2019, 170, 166. 89 Tang H T. Materials reliability program: PWR internals material aging degradation mechanism screening, and threshold values (MRP-175), EPRI, 2005. 90 Faïza Sefta. Proposed revision of the IASCC sensitivity criterion(H-B60-2015-00404-EN). Materials aging institute. 2015. 91 Was G S, Gussev M N, Rosseel T M. Developing a mechanistic understanding of irradiation-assisted stress corrosion crack initiation. lwrs news letter. 2020. 92 中国核能行业协会. 离子辐照模拟中子辐照损伤实验研究的技术指南(T/CNEA 180-2024), 2024. 93 ASTM. Standard practice for investigating the effects of neutron radiation damage using charged-particle irradiation(ASTM E521-16), 2016.
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