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Acta Metall Sin  2019, Vol. 55 Issue (3): 349-361    DOI: 10.11900/0412.1961.2018.00359
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Study on Irradiation Assisted Stress Corrosion Cracking of Nuclear Grade 304 Stainless Steel
Ping DENG1,2,Chen SUN3,Qunjia PENG1,4(),En-Hou HAN1,Wei KE1,Zhijie JIAO5
1. CAS Key Laboratory of Nuclear Materials and Safety Assessment, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016, China
2. School of Materials Science and Engineering, University of Science and Technology of China, Shenyang 110016, China
3. State Power Investment Corporation Research Institute, Beijing 102209, China
4. Suzhou Nuclear Power Research Institute, Suzhou 215004, China
5. Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109, USA
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Abstract  

Irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steel core components is one major concern for maintenance of nuclear power plants. Previous studies on the IASCC had mainly focused on the effect of irradiation on changes in deformation modes and interaction of dislocation channels with grain boundary. The role of corrosion in IASCC, however, has not received sufficient attentions. In the process of stress corrosion cracking (SCC), corrosion occurs simultaneously with localized deformation in the vicinity of the crack tip. This indicates that corrosion is one of the potential contributors to IASCC. In this work, IASCC of proton-irradiated nuclear grade 304 stainless steel (304SS) was investigated. The IASCC tests were conducted by interrupted slow strain rate tensile (SSRT) tests at 320 ℃ in simulated primary water of pressurized water reactor containing 1200 mg/L B as H3BO3 and 2.3 mg/L Li as LiOH·H2O, with a dissolved hydrogen concentration of 2.6 mg/L. Following the SSRT tests, the localized deformation, corrosion and IASCC of the specimens were characterized. The results revealed that increasing the irradiation dose promoted residual strain accumulation at slip steps and grain boundaries of nuclear grade 304SS. Since the slip step usually transmitted or terminated at the grain boundary, it eventually promoted localized deformation at the grain boundary. Specially, the slip step transmitted at grain boundary led to slip continuity at the grain boundary. In contrast, a slip discontinuity was observed at the grain boundary where the slip step terminated, which caused a much higher strain accumulation by feeding dislocations to the grain boundary region. Further, formation of the slip discontinuity was related to the Schmidt factor pair type of the adjacent grains. The irradiation resulted in a depletion of Cr and an enrichment of Ni at grain boundary, while the magnitude of Cr depletion and Ni enrichment increased with increasing the irradiation dose. Following the SSRT tests, intergranular cracking was observed on surfaces of the irradiated specimens, while the number of the cracks was increased by a higher irradiation dose and applied strain. This suggested a higher IASCC susceptibility of nuclear grade 304SS in the primary water. Meanwhile, significant intergranular oxidation ahead of the crack tip was observed, while both the width and length of the oxide were larger at a higher irradiation dose. The synergic effect of irradiation-promoted deformation and intergranular corrosion was the primary cause for the IASCC of the irradiated steel.

Key words:  nuclear grade stainless steel      proton irradiation      localized deformation      corrosion      irradiation assisted stress corrosion cracking     
Received:  31 July 2018     
ZTFLH:  TG139.4  
Fund: International Science & Technology Cooperation Program of China(2014DFA50800);National Natural Science Foundation of China(51571204);Essential Research Fund by State Nuclear Power Technology Corporation(2015SN010-007)
Corresponding Authors:  Qunjia PENG     E-mail:  pengqunjia@cgnpc.com.cn

Cite this article: 

Ping DENG,Chen SUN,Qunjia PENG,En-Hou HAN,Wei KE,Zhijie JIAO. Study on Irradiation Assisted Stress Corrosion Cracking of Nuclear Grade 304 Stainless Steel. Acta Metall Sin, 2019, 55(3): 349-361.

URL: 

https://www.ams.org.cn/EN/10.11900/0412.1961.2018.00359     OR     https://www.ams.org.cn/EN/Y2019/V55/I3/349

Fig.1  Schematic of geometry and dimension of the slow strain rate tensile (SSRT) specimen (unit: mm)
Fig.2  EBSD analyses of stress corrosion cracking (SCC) area for the 0.5 dpa irradiated specimens following a strain of 3% by slow strain rate tensile (SSRT) test in the primary water (The cracked grain boundaries are highlighted by black arrows, LAB—low-angle grain boundary, RGB—random grain boundary, CSL—coincidence site lattice special boundary)(a) SEM image of the analyzed area (b) color coded local misorientation (ML) map showing residual strain (c) distribution of local misorientations (d) crystal orientation (e) grain boundary character (f) Schmidt factor
Fig.3  EBSD analyses of SCC area for the 5.0 dpa irradiated specimens following a strain of 3% by SSRT test in the primary water (The cracked grain boundaries are highlighted by black arrows)(a) SEM image of the analyzed area (b) color coded ML map showing residual strain (c) distribution of local misorientations (d) crystal orientation (e) grain boundary character (f) Schmidt factor
Fig.4  EBSD analyses of the surface steps in grain boundary area for the 5.0 dpa irradiated specimens following a strain of 3% by SSRT test in the primary water(a) SEM image of the analyzed area (b) color coded ML map showing residual strain
Fig.5  Schmidt factor distributions of grains in the nuclear grade 304 stainless steel (The low and high categories each contain 50% of the total grain population) (a), and plots of the co-dependence of irradiation assisted SCC (IASCC) cracking propensity on Schmidt factor pair type (LL—a grain boundary adjacent to two grains with low Schmid factors, LH—a grain boundary adjacent to a low and high Schmid factor, HH—a grain boundary adjacent to two grains with high Schmid factors, GBs—grain boundaries) (b)
Fig.6  TEM images and analyses of deformation structures for the 0.5 dpa irradiated specimens following a strain of 3% by SSRT test in the primary water(a) image of the materials confirming the presence of cleared channels (b, c) accumulation of several stacking faults and dislocations in a channel (d) a channel demonstrated several stages of the evolution process, with the head of the channel marked by several dislocations, followed by planar stacking faults, and finally left behind a channel with reduced defect density (e) grain boundary regions experiencing more signification straining by the interaction of dislocation channels (f) high magnification view of a channel with feeding dislocations to the grain boundary region
Fig.7  SEM images of stress corrosion cracking at surfaces of nuclear grade 304 stainless steel following a strain of 3% by the SSRT tests (a~d) and corresponding schematics showing the interaction between slip step and grain boundary crack (e~h)(a, e) solution annealed (b, f) 0.5 dpa (c, g) 1.5 dpa (d, h) 5.0 dpa
Fig.8  Summary of SCC cracks on surfaces of nuclear grade 304 stainless steel following interrupted SSRT test in the primary water
Fig.9  SEM images of secondary cracking on the surface of nuclear grade 304 stainless steel tested to fracture by SSRT in the primary water (a~c) and corresponding high magnification SEM images of the selected region (d~f) (TGSCC—transgranular SCC, IGSCC—intergranular SCC)(a, d) solution annealed (b, e) 0.5 dpa (c, f) 5.0 dpa
Fig.10  TEM image and analyses of crack tip for the 0.5 dpa irradiated specimens following a strain of 3% by SSRT test in the primary water(a) picture showing the area analyzed (The inset shows the SAED pattern collected from the crack tip region highlighted by black circle)(b) EDS results collected along line A across the oxide shown in Fig.10a (c) EDS results collected along line B ahead the oxide tip with a distance of about 200 nm shown in Fig.10a
Fig.11  TEM image and analyses of crack tip for the 5.0 dpa irradiated specimens following a strain of 3% by SSRT test in the primary water(a) picture showing the area analyzed (The inset shows the SAED pattern collected from the crack tip region highlighted by black circle)(b) EDS results collected along line A across the oxide shown in Fig.11a(c) EDS results collected along line B ahead the oxide tip with a distance of about 200 nm shown in Fig.11a
Fig.12  SEM images of the fracture surface of the 0.5 (a, c, e) and 5.0 dpa (b, d, f) irradiated specimen tested by SSRT in the primary water (a, b) and corresponding high magnification SEM images of the selected region (c~f)
Fig.13  Schematics showing the strain concentration at grain boundary promoted by Schmidt factor and slip step(a) nucleation of dislocations by irradiation (b) formation of slip channel under stress(c) occurrence of slip continuity at a HH Schmidt factor type grain boundary, leading to a low strain concentration(d) occurrence of slip discontinuity at a LL Schmidt factor type grain boundary, leading to a high strain concentration
Fig.14  Schematics showing the IASCC mechanism promoted by localized deformation and localized corrosion at grain boundary (RIS—radiation induced segregation)(a) oxidation occurrence at grain boundary simultaneously with localized deformation(b) cracking initiation within the oxide at the grain boundary(c) crack growth promoted by further deformation and corrosion at the grain boundary
[1] Ge?rard R, Somville F. Situation of the baffle-former bolts in Belgian units [A]. Proceeding of the 17th International Conference on Nuclear Engineering [C]. Brussels, Belgium: American Society of Mechanical Engineers, 2009: 521
[2] Andresen P L, Ford F P, Murphy S M, et al. State of knowledge of radiation effects on environmental cracking in light water reactor core materials [A]. Proceeding of the 4th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors [C]. Houston: National Association of Corrosion Engineers, 1989: 83
[3] Nishioka H, Fukuya K, Fujii K, et al. IASCC initiation in highly irradiated stainless steels under uniaxial constant load conditions [J]. J. Nucl. Sci. Technol., 2008, 45: 1072
[4] Zhou R S, West E A, Jiao Z J, et al. Irradiation-assisted stress corrosion cracking of austenitic alloys in supercritical water [J]. J. Nucl. Mater., 2009, 395: 11
[5] Was G S, Bruemmer S M. Effects of irradiation on intergranular stress corrosion cracking [J]. J. Nucl. Mater., 1994, 216: 326
[6] Stephenson K J, Was G S. Comparison of the microstructure, deformation and crack initiation behavior of austenitic stainless steel irradiated in-reactor or with protons [J]. J. Nucl. Mater., 2015, 456: 85
[7] Jiao Z, Was G S. Impact of localized deformation on IASCC in austenitic stainless steels [J]. J. Nucl. Mater., 2011, 408: 246
[8] Jiao Z, Was G S. Localized deformation and IASCC initiation in austenitic stainless steels [J]. J. Nucl. Mater., 2008, 382: 203
[9] Jiao Z, Was G S, Busby J T. The role of localized deformation in IASCC of proton-irradiated austenitic stainless steels [A]. Proceeding of the 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors [C]. Whister, British Columia, 2007 (CD-ROM)
[10] Jiao Z, Was G S. The role of irradiated microstructure in the localized deformation of austenitic stainless steels [J]. J. Nucl. Mater., 2010, 407: 34
[11] Was G S, Busby J T. Role of irradiated microstructure and microchemistry in irradiation-assisted stress corrosion cracking [J]. Philos. Mag., 2005, 85: 443
[12] West E A, Was G S. Strain incompatibilities and their role in intergranular cracking of irradiated 316L stainless steel [J]. J. Nucl. Mater., 2013, 441: 623
[13] McMurtrey M D, Was G S, Cui B, et al. Strain localization at dislocation channel-grain boundary intersections in irradiated stainless steel [J]. Int. J. Plast., 2014, 56: 219
[14] Was G S, Farkas D, Robertson I M. Micromechanics of dislocation channeling in intergranular stress corrosion crack nucleation [J]. Curr. Opin. Solid State Mater. Sci., 2012, 16: 134
[15] Fukuya K, Nishioka H, Fujii K, et al. Charateriation of IASCC crack tip in highly irradiated stainless steels [A]. Proceeding of the 14th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors [C]. Virginia Beach, Virginia, 2009: 1248 (CD-ROM)
[16] Thomas L E, Bruemmer S M. Microstructural and microchemical characterization of intergranular stress corrosion cracks in irradiated type 304SS removed from a BWR top guide [A]. Proceeding of the 11th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors [C]. Stevenson, WA, 2003: 1049 (CD-ROM)
[17] Edwards D J, Thomas L E, Asano K, et al. Microstructure, microchemistry and stress corrosion crack characteristics in a BWR 316L SS core shroud weld [A]. Proceeding of the 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors [C]. Whister, British Columia, 2007 (CD-ROM)
[18] Thomas L E, Edwards D J, Asano K, et al. Crack-tip characteristics in BWR service components [A]. Proceeding of the 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors [C]. Whister, British Columia, 2007 (CD-ROM)
[19] Deng P, Peng Q J, Han E H, et al. Effect of irradiation on corrosion of 304 nuclear grade stainless steel in simulated PWR primary water [J]. Corros. Sci., 2017, 127: 91
[20] Kamaya M. Characterization of microstructural damage due to low-cycle fatigue by EBSD observation [J]. Mater. Charact., 2009, 60: 1454
[21] Deng P, Peng Q J, Han E H, et al. Study of irradiation damage in domestically fabricated nuclear grade stainless steel [J]. Acta Metall. Sin., 2017, 53: 1588
[21] 邓 平, 彭群家, 韩恩厚等. 国产核用不锈钢辐照损伤研究 [J]. 金属学报, 2017, 53: 1588
[22] Shen Z, Arioka K, Lozano-Perez S. A mechanistic study of SCC in Alloy 600 through high-resolution characterization [J]. Corros. Sci., 2018, 132: 244
[23] Lu Y H, Peng Q J, Sato T, et al. An ATEM study of oxidation behavior of SCC crack tips in 304L stainless steel in high temperature oxygenated water [J]. J. Nucl. Mater., 2005, 3477: 52
[24] Lozano-Perez S, Kruska K, Iyengar I, et al. The role of cold work and applied stress on surface oxidation of 304 stainless steel [J]. Corros. Sci., 2012, 56: 78
[25] Gertsman V Y, Bruemmer S M. Study of grain boundary character along intergranular stress corrosion crack paths in austenitic alloys [J]. Acta Mater., 2001, 49: 1589
[26] Hu C L, Xia S, Li H, et al. Effect of grain boundary network on the intergranular stress corrosion cracking of 304 stainless steel [J]. Acta Metall. Sin., 2011, 47: 939
[26] 胡长亮, 夏 爽, 李 慧等. 晶界网络特征对304不锈钢晶间应力腐蚀开裂的影响 [J]. 金属学报, 2011, 47: 939
[27] Ming H L, Zhang Z M, Wang J Q, et al. Microstructure and local properties of a domestic safe-end dissimilar metal weld joint by using hot-wire GTAW [J]. Acta Metall. Sin., 2017, 53: 57
[27] 明洪亮, 张志明, 王俭秋等. 国产核电安全端异种金属焊接件的微观结构及局部性能研究 [J]. 金属学报, 2017, 53: 57
[28] Hou J, Peng Q J, Shoji T, et al. Effects of cold working path on strain concentration, grain boundary microstructure and stress corrosion cracking in Alloy 600 [J]. Corros. Sci., 2011, 53: 2956
[29] Hou J, Peng Q J, Lu Z P, et al. Effects of cold working degrees on grain boundary characters and strain concentration at grain boundaries in Alloy 600 [J]. Corros. Sci., 2011, 53: 1137
[30] Alexandreanu B, Was G S. Grain boundary deformation-induced intergranular stress corrosion cracking of Ni-16Cr-9Fe in 360 ℃ Water [J]. Corrosion, 2003, 59: 705
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