Study on Irradiation Assisted Stress Corrosion Cracking of Nuclear Grade 304 Stainless Steel

doi:10.11900/0412.1961.2018.00359

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 DENG^1,²,Chen SUN³,Qunjia PENG^1,⁴(

),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

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.

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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)

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	Ping DENG
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	Wei KE
	Zhijie JIAO

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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 (M_L) 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 M_L 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 M_L 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

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