Corrosion Behavior and Mechanism of Irradiated 304 Nuclear Grade Stainless Steel in High-Temperature Water

doi:10.1007/s40195-020-01123-y

Abstract

Abstract:

Corrosion behavior and mechanism of irradiated 304 nuclear grade stainless steel were studied in simulated pressurized water reactor primary water. The microstructure of the oxide formed on the steel irradiated to different doses over an exposure period range of 25-1500 h was analyzed and compared. It was found that the general and intergranular corrosion rates of the steel were increased with irradiation dose, in correspondence with an evolution of the general oxide and the oxide formed at the grain boundary. Correlation of the oxide evolution with the corrosion kinetics and mechanism has been discussed in detail.

Key words: Stainless steel, Irradiation, High-temperature corrosion, Intergranular corrosion

Ping Deng, En-Hou Han, Qunjia Peng, Chen Sun. Corrosion Behavior and Mechanism of Irradiated 304 Nuclear Grade Stainless Steel in High-Temperature Water[J]. Acta Metallurgica Sinica (English Letters), 2021, 34(2): 174-186.

Figures/Tables 12

Table 1 Chemical composition (wt%) of 304NG stainless steel

C	Mn	Si	S	P	Ni	Cr	Co	Fe
0.04	1.73	0.27	0.002	0.021	8.87	19.51	0.04	Bal

Fig. 1 Surface morphologies of the oxide scales formed on solution-annealed a1, b1, c1, d1, 0.5 dpa a2, b2, c2, d2, 1.5 dpa a3, b3, c3, d3 and 3 dpa b4, c4, d4 irradiated specimens following the exposure in simulated primary PWR water at 320 °C for the period of 25 h a1, a2, a3, 500 h b1, b2, b3, b4, 1000 h c1, c2, c3, c4 and 1500 h d1, d2, d3, d4

Fig. 2 XPS composition profiles of the oxide scales formed on 0.5 and 1.5 dpa irradiated specimens following the exposure for 500 h a, b, 1000 h c, d and 1500 h e, f. The black vertical lines located at half of the highest O concentration display the thickness of the oxide film

Fig. 3 Thickness of the oxide scales on 304NG SS as a function of irradiation dose and exposure time

Fig. 4 TEM observation and analysis of the cross section of the oxide scales formed on 1.5 dpa irradiated specimens of 304NG SS following the 100- and 1000-h exposure in simulated primary PWR water. a and b TEM observations of the cross section of the oxide scales. c and d high-resolution observations and analysis of the inner oxide. e and f corresponding point scans collected along the red lines shown in a and b, respectively

Fig. 5 Time-dependent inner oxide thickness obtained on 0.5 and 1.5 dpa irradiated specimens following the exposure to primary PWR water at 320 °C. The experimental data were adjusted with a power fit, and the error bars mean the maximum and minimum values

Fig. 6 a, b TEM observation of the cross section of a grain boundary on a 0.5- and b 1.5 dpa irradiated specimens following the 25-h exposure in primary PWR water at 320 °C. c, d The corresponding EDX mappings for O, Cr, Fe and Ni, respectively

Fig. 7 a TEM observation of the cross section of a grain boundary on 0.5 dpa irradiated specimen following the 100-h exposure in PWR primary water at 320 °C. b The corresponding EDX mappings for O, Cr, Fe and Ni, respectively. c and d EDX point scans across the grain boundary collected along the red lines shown in a

Fig. 8 a TEM observation of the cross section of a grain boundary on 1.5 dpa irradiated specimen following the 100-h exposure in PWR primary water at 320 °C. b The corresponding EDX mappings for O, Cr, Fe and Ni, respectively. c and d EDX point scans across the grain boundary collected along the red lines shown in a

Fig. 9 a TEM observation of the cross section of a grain boundary on 0.5 dpa irradiated specimen following the 1500-h exposure in PWR primary water at 320 °C. b The corresponding EDX mappings for O, Cr, Fe and Ni, respectively. c and d EDX point scans across the grain boundary collected along the red lines shown in a

Fig. 10 a TEM observation of the cross section of a grain boundary on 1.5 dpa irradiated specimen following the 1500-h exposure in PWR primary water at 320 °C. b The corresponding EDX mappings for O, Cr, Fe and Ni, respectively. c and d EDX point scans across the grain boundary collected along the red lines shown in a

Fig. 11 Schematics showing the evolution of the oxide scale formed on a1-a4 solution-annealed and b1, b2 irradiated specimens of 304NG SS following the exposure to PWR primary water at 320 °C

References 43

[1]	K.J. Stephenson, G.S. Was, J. Nucl. Mater. 444, 331(2014)
[2]	H. Nishioka, K. Fukuya, K. Fujii, T. Torimaru, J. Nucl. Sci. Technol. 45, 1072(2008)
[3]	Y. Chen, B. Alexandreanu, W.K. Soppet, W.J. Shack, K. Natesan, A.S. Rao, Slow strain rate tensile tests of irradiated austenitic stainless steels in simulated PWR environment. Paper presented at the 15th international conference on environmental degradation of materials in nuclear power systems-water reactors, Colorado, 1277(2011)
[4]	Z. Jiao, G.S. Was, J.T. Busby, The role of localized deformation in iascc of proton-irradiated austenitic stainless steels. Paper presented at the 13th international conference on environmental degradation of materials in nuclear power systems, Whistler, British Columbia, 19 (2007)
[5]	Z. Jiao, G.S. Was, T. Miura, K. Fukuya, J. Nucl. Mater. 452, 328(2014)
[6]	Z. Jiao, G.S. Was, J. Nucl. Mater. 408, 246(2011)
[7]	Z. Jiao, G.S. Was, J. Nucl. Mater. 407, 34(2010)
[8]	Z. Jiao, G.S. Was, J. Nucl. Mater. 382, 203(2008)
[9]	K. Fukuya, H. Nishioka, K. Fujii, T. Miura, T. Torimaru, J. Nucl. Mater. 417, 958(2011)
[10]	T. Miura, K. Fujii, K. Fukuya, Y. Ito, J. Nucl. Mater. 386-388, 210(2009)
[11]	K. Fukuya, H. Nishioka, K. Fujii, T. Miura, Y. Kitsunai, J. Nucl. Mater. 432, 67(2013)
[12]	M.D. McMurtrey, G.S. Was, Role of slip behavior in the irradiation assisted stress corrosion cracking in austenitic steels. Paper presented at the 15th international symposium on environmental degradation of materials in nuclear power systems-water reactors, Colorado, 1383(2011)
[13]	I. Ioka, Y. Ishijima, K. Usami, N. Sakuraba, Y. Kato, K. Kiuchi, J. Nucl. Mater. 417, 887(2011)
[14]	Z.C. Zheng, Y.X. Yu, W.P. Zhang, Z.Y. Shen, F.F. Luo, L.P. Guo, Y.Y. Ren, R. Tang, Acta Metall. Sin.-Engl. Lett. 30, 89(2017)
[15]	A.B. Du, W. Feng, H.L Ma, T. Liang, D.Q. Yuan, P. Fan, Q.L. Zhang, C. Huang, Acta Metall. Sin.-Engl. Lett. 30, 1049(2017)
[16]	P. Deng, C. Sun, Q.J. Peng, E.H. Han, W. Ke, Z. Jiao, Acta Metall. Sin. 55, 349(2019)
[17]	A. Turnbull, K. Mingard, J.D. Lord, B. Roebuck, D.R. Tice, K.J. Mottershead, N.D. Fairweather, A.K. Bradbury, Corros. Sci. 53, 3398(2011)
[18]	G.S. Was, Recent Developments in understanding irradiation assisted stress corrosion cracking. Paper presented at the 11th international symposium on environmental degradation of materials in nuclear power systems-water reactors, Washington, 965(2003)
[19]	P. Deng, Q.J. Peng, E.H. Han, W. Ke, C. Sun, Z. Jiao, Corros. Sci. 127, 91(2017)
[20]	S. Perrin, L. Marchetti, C. Duhamel, M. Sennour, F. Jomard, Oxid. Met. 80, 623(2013)
[21]	M. Dumerval, S. Perrin, L. Marchetti, M. Sennour, F. Jomard, S. Vaubaillon, Y. Wouters, Corros. Sci. 107, 1(2016)
[22]	C.S. Liu, J.Q. Wang, Z.M. Zhang, E.H. Han, W. Liu, D. Liang, Z.T. Yang, Sin Acta Metall. Sin.-Engl. Lett. 32, 506(2019)
[23]	P. Deng, Q.J. Peng, E.H. Han, W. Ke, C. Sun, H.H. Xia, Z. Jiao, Acta Metall. Sin. 53, 1588(2017)
[24]	Z. Jiao, G.S. Was, Acta Mater. 59, 1220(2011)
[25]	P. Ampornrat, Y.B. Chen, L.M. Wang, G.S. Was, Microstructure and oxidation mechanisms of ferritic-martensitic alloy HCM12A in supercritical water. Paper presented at the 14th international symposium on environmental degradation of materials in nuclear power systems-water reactors, Virginia Beach 1751 (2009)
[26]	Y.L. Han, E.H. Han, Q.J. Peng, W. Ke, Corros. Sci. 121, 1(2017)
[27]	X. Gao, X. Wu, Z.E. Zhang, H. Guan, E.H. Han, J. Supercrit. Fluids 42, 157 (2007)
[28]	F. Huang, J.Q. Wang, E.H. Han, W. Ke, Corros. Sci. 76, 52(2013)
[29]	W.J. Kuang, X. Wu, E.H. Han, Corros. Sci. 52, 4081(2010) DOI URL
[30]	W.J. Kuang, E.H. Han, X. Wu, J. Rao, Corros. Sci. 52, 3654(2010)
[31]	Y.J. Kim, Corrosion 51, 849 (1995)
[32]	J. Robertson, Corros. Sci. 29, 1275(1989)
[33]	J. Robertson, Corros. Sci. 32, 443(1991)
[34]	R.J. Lemire, G.A. McRae, J. Nucl. Mater. 294, 141(2001)
[35]	P. Deng, Q.J. Peng, E.H. Han, W. Ke, Corrosion 73, 1237 (2017)
[36]	S. Lozano-Perez, D.W. Saxey, T. Yamada, T. Terachi, Scr. Mater. 62, 855(2010)
[37]	K. Kruska, S. Lozano-Perez, D.W. Saxey, T. Terachi, T. Yamada, T. Smith, D.W. George, Corros. Sci. 63, 225(2012)
[38]	W.J. Kuang, G.S. Was, Acta Mater. 182, 120(2020)
[39]	C. Ma, E.H. Han, Q.J. Peng, W. Ke, Appl. Surf. Sci. 442, 423(2018)
[40]	Y.L. Han, J.N. Mei, Q.J. Peng, E.H. Han, W. Ke, Corros. Sci. 98, 72(2015)
[41]	S. Ghosh, M.K. Kumar, V. Kain, Appl. Surf. Sci. 264, 312(2013)
[42]	D.D. Macdonald, M.U. Macdonald, J. Electrochem. Soc. 137, 2395(1990)
[43]	L. Marchetti, S. Perrin, F. Jambon, M. Pijolat, Corros. Sci. 102, 24(2016)