Research Progress on Irradiation Assisted Stress Corrosion Cracking Behavior and Mechanism of Austenitic Steel

doi:10.11902/1005.4537.2023.287

Journal of Chinese Society for Corrosion and protection

2024, Vol. 44

Issue (4): 835-846 DOI: 10.11902/1005.4537.2023.287

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Research Progress on Irradiation Assisted Stress Corrosion Cracking Behavior and Mechanism of Austenitic Steel

GAO Junxuan¹, CAO Han¹(

), KUANG Wenjun²(

), ZHENG Quan¹, ZHANG Peng¹, ZHONG Weihua¹

1. China Institute of Atomic Energy, Beijing 102413, China
2. School of Materials Science and Engineering, Xi'an Jiaotong University, Xi'an 710049, China

Cite this article:

GAO Junxuan, CAO Han, KUANG Wenjun, ZHENG Quan, ZHANG Peng, ZHONG Weihua. Research Progress on Irradiation Assisted Stress Corrosion Cracking Behavior and Mechanism of Austenitic Steel. Journal of Chinese Society for Corrosion and protection, 2024, 44(4): 835-846.

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Abstract

Austenitic stainless steel is the main material for reactor core components. In the service process, under the joint action of neutron irradiation, tensile stress and high-temperature and high-pressure water medium in the primary coolant circuit, the low stress brittle cracking phenomenon, namely irradiation assisted stress corrosion cracking (IASCC) can occur, which will lead to component fracture and failure, thus affecting the structural integrity of reactor core components. This has become one of the key issues affecting the safety and economic operation of nuclear power plants. In this paper, the crack-initiation, crack-growth and cracking-sensitivity of IASCC are summarized. Then the main influence mechanisms of localized deformation, grain boundary oxidation, irradiation hardening, irradiation induced segregation (RIS), irradiation swelling, irradiation creep and irradiation induced phase transformation on IASCC are analyzed. Meanwhile, the application of post-irradiation annealing (PIA) for resolving IASCC is summarized. Finally, the prospect of research on IASCC mechanism is discussed. This review is aimed to provide a reference for understanding the issue of neutron irradiation assisted stress corrosion cracking of austenitic stainless steels.

Key words: crack initiation and growth IASCC sensitivity localized deformation GB oxidation irradiation hardening irradiation induced segregation irradiation swelling and irradiation creep irradiation induced phase transformation post-irradiation annealing

Received: 10 September 2023 32134.14.1005.4537.2023.287

ZTFLH:

TG172

Fund: National Key Research and Development Program of China(2019YFB1900902)

Corresponding Authors: CAO Han, E-mail: caohan@ciae.ac.cn;

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	GAO Junxuan
	CAO Han
	KUANG Wenjun
	ZHENG Quan
	ZHANG Peng
	ZHONG Weihua

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https://www.jcscp.org/EN/10.11902/1005.4537.2023.287 OR https://www.jcscp.org/EN/Y2024/V44/I4/835

Fig.1 Dose dependence of CGR and yield stress in IASCC datas^[15]: (a) CGR versus dose plateau behavior of JCB and earlier IASCC data, (b) Irradiated yield stress model and data at 288 ℃ for cold-worked types 316 and 316Ti and types 304 and 304 L stainless steel

Fig.2 Variation of CGR with stress intensity factor (K) (a) and irradiation temperature (b) in IASCC datas^[16](The matrix of samples with different numbers in the figure is high purity 316L stainless steel, and the contents in brackets represent the alloying elements adulterated or removed), The K dependence of the Hf addition alloy samples PS01 (round symbols) and PS02 (square symbols) irradiated to 9.6 dpa and tested in NWC, HWC, and PWR water CGR as a function of temperature for solute addition alloys tested at NRI, SCK-CEN, and U-M in PWR at 288^oC and 320^oC

Fig.3 Variation of IASCC sensitivity parameters with dose of neutron irradiation: (a) Yield strength and total elongation as a function of neutron dose for solution-annealed Type 304, 304L, 316L, and 347 stainless steels at 270-380^oC^[10], (b) Dose dependency of IASCC susceptibility of irradiated 316 stainless steel in a PWR environment with neutron exposure^[19]

Fig.4 Statistics of normal stress and cracking at DC-GB position : (a) Total calculated stress acting normal to the grain boundary as a function of the crystallographic orientation of the deforming grain and the geometric orientation of the grain boundary normal vector^[24], (b) Cracking statistics at different DC-GB locations^[23]

Fig.5 Mechanism diagram of effect of δ phase on crack growth (a) Diagram of crack growth with network ferrite^{(b) Diagram of crack growth along the γ/δ phase boundary in the presence of Ti(CN) particles^[74]}

Fig.6 Plots of IASCC sensitivity and radiation-induced characteristics of the material as a function of annealing conditions (The five material states from left to right are: 5.9 dpa, 5.9 dpa + 500℃ annealing: 1 h and 5.9 dpa + 550℃ annealing: 1, 5, 20 h)^[86]

Fig.7 Variation of microscopic defects and mechanical properties with annealing conditions: (a) size and number density of dislocation loops in 304L stainless steel irradiated to 5.9 dpa under different annealing conditions^[85], (b) recovery of yield strength ( $σ 0.2$ ) and ductility as a function of annealing temperature^[79,85]

Fig.8 CGR of 304L stainless steel irradiated to 5.9 dpa under different annealing conditions and different water environments^[54]

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