Research Progress of the Niobium-Stabilized Austenitic Stainless Steels for Generation-IV Nuclear Reactors

doi:10.11900/0412.1961.2025.00203

Acta Metall Sin

2026, Vol. 62

Issue (7): 1175-1188 DOI: 10.11900/0412.1961.2025.00203

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Research Progress of the Niobium-Stabilized Austenitic Stainless Steels for Generation-IV Nuclear Reactors

CHEN Shenghu¹^,²(

), XIE Ang¹, YANG Binbin², LIU Yongtao², CHEN Sihan¹, ZHAO Mingjiu¹^,², JIANG Haichang¹^,², RONG Lijian¹^,²(

)

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

Cite this article:

CHEN Shenghu, XIE Ang, YANG Binbin, LIU Yongtao, CHEN Sihan, ZHAO Mingjiu, JIANG Haichang, RONG Lijian. Research Progress of the Niobium-Stabilized Austenitic Stainless Steels for Generation-IV Nuclear Reactors. Acta Metall Sin, 2026, 62(7): 1175-1188.

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Abstract

Austenitic stainless steels are utilized in the in-core and out-of-core structural components of various Generation-IV nuclear reactors owing to their excellent comprehensive properties, mature manufacturing processes, and decades of service experience in pressurized water reactors. However, after prolonged exposure to high temperature and high-intensity irradiation, these steels gradually exhibit some performance limitations, including lower creep strength, inadequate microstructural stabilities, and higher radiation embrittlement sensitivities. Recently, alloying with Nb has been proven to effectively regulate the precipitation behavior of secondary phases and the formation and evolution of irradiation defects in austenitic stainless steels, promising simultaneous improvements in creep strength and irradiation tolerance. Therefore, Nb alloying has emerged as a key pathway for the development of next-generation austenitic stainless steels with enhanced high-temperature performance, improved irradiation resistance, and extended service life. This paper systematically reviews the research progress of Nb-stabilized austenitic stainless steels for Generation-IV nuclear power applications, including the compositional design and optimization to improve the creep properties and irradiation resistance, the effects of Nb addition on austenite stability, the formation mechanisms of primary NbC, and the precipitation behavior of Nb-bearing secondary phases. Additionally, the mechanisms by which Nb alloying influences the strength-ductility balance, creep strength, high-temperature microstructural stability, and irradiation-induced defect evolution are discussed. Finally, some future research prospects of Nb-stabilized austenitic stainless steels for Generation-IV nuclear reactors are discussed.

Key words: niobium-stabilized austenitic stainless steel strength-ductility balance creep strength high-temperature microstructural stability irradiation-induced defect Nb-bearing secondary phase

Received: 15 July 2025

ZTFLH:

TG142.25

Fund: Strategic Priority Research Program of Chinese Academy of Sciences(XDA0410000);LingChuang Research Project of China National Nuclear Corporation

Corresponding Authors: CHEN Shenghu, professor, Tel: (024)23971981, E-mail: chensh@imr.ac.cn；

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	Articles by authors
	CHEN Shenghu
	XIE Ang
	YANG Binbin
	LIU Yongtao
	CHEN Sihan
	ZHAO Mingjiu
	JIANG Haichang
	RONG Lijian

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https://www.ams.org.cn/EN/10.11900/0412.1961.2025.00203 OR https://www.ams.org.cn/EN/Y2026/V62/I7/1175

Table 1 Summary of the reactor core environments of the six Generation-IV reactor systems^[6-10]

Table 2 Chemical compositions of niobium-stabilized austenitic stainless steels for advanced nuclear reactors^[44-46]

Fig.1 SEM images of as-cast Fe-15Cr-10Ni austenitic stainless steels with the Nb contents (mass fraction) of 0.01% (a), 0.3% (b), and 0.9% (c)

Fig.2 Thermo-Calc calculation of NbC formation in Fe-15Cr-10Ni ausenitic stainless steels with different Nb contents (a) and OM image of as-cast samples with 0.9%Nb (b) (Inset in Fig.2b is the high magnification image of the primary NbC)

Fig.3 Nanoindentation load-depth curves of primary NbC and austenite (a) and TEM images near primary NbC after deformation to a true strain of 10% (b) in as-cast sample (SF—stacking fault. Inset in Fig.3b refers to the corresponding SAED pattern of NbC)^[62]

Fig.4 Three-dimensional (3D) view of primary NbC in the as-cast Fe-15Cr-10Ni-0.9Nb ausenitic stainless steel (a) and SEM image of the fracture surface morphology (b) and corresponding EDS mapping of Nb (c)

Fig.5 Creep-rupture stresses of several austenitic stainless steels (LMP—Larson-Miller parameter; T—temperature, K; t_r—creep rupture time, h; C—material constant, C = 20)

Fig.6 TEM images of secondary phase in Alloy 709 after aging at 650 ^oC for 500 h (Inset is the high magnification image of the dislocation (arrowed in Fig.6a)) (a)^[76] and HT-UPS steel after creep-rupture test at 700 ^oC, 170 MPa for 18000 h (Inset is the analytical electron microscopy of MC carbides (arrowed in Fig.6b)) (b)^[77]

Fig.7 Temperature-time-precipitation phase diagrams of 316L steel (a)^[80] and Fe-20Cr-25Ni-Nb stabilized austenitic stainless steel (b)^[81]

Fig.8 Schematics of formation mechanisms of G-phase and σ-phase in Nb-stabilized austenitic stainless steel (a, b) transformation mechanism of Nb(C, N) to G-phase (c, d) formation mechanism of σ-phase adjacent to Nb(C, N)

Fig.9 Variation in void number density (a) and mean void diameter (b) in Fe-16Ni-15Cr and Fe-16Ni-15Cr-0.6Nb alloys after irradiation to 50 dpa^[92]

Fig.10 Schematic illustration of the 3D structure after irradiation (a) and evolution of irradiation-induced defects along the NbC/matrix interface (b)^[96] (SIA—self-interstitial atom)

Fig.11 Property ranking chart of Nb stabilized austenitic stainless steels^[104]

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