**1. Introduction**

Ferritic/martensitic steels with a chromium content of 9~12 wt.% have been considered as candidate structural materials in future advanced nuclear reactors due to their higher resistance to irradiation swelling, lower thermal expansion coefficients, and higher thermal conductivity [1–4]. Ferritic/martensitic steels are usually characterized by the tempered martensitic structure, consisting of a high density of tangled dislocations within laths and dispersion of carbides along their boundaries and within their matrix. During the long-term exposure at temperature above 823 K, reduction in the dislocation density and coarsening of carbide would lead to the recovery of martensitic structure and a significant reduction in creep strength [5–7]. Afterwards, dispersed oxide nano-particles with a high number density are introduced into the ferrite matrix to develop the oxide dispersion strengthened (ODS) steels [8–10]. These highly stabilized oxide nano-particles are responsible for the excellent tensile strength and creep properties at 923~1173 K. The ODS steels are produced by much more complicated and expensive powder-metallurgy techniques.

Moreover, intermetallic precipitates are potentially considered as the strengthening phase during the design of high-temperature ferritic alloys. K. Yamamoto et al. reported the Fe-Cr-Nb ferritic heat-resistant alloy strengthened by the Fe2Nb Laves phase, and found that the presence of Fe2Nb phase could significantly improve the high-temperature strength [11]. D.G. Morris et al. reported a Fe-Al-Zr ferritic alloy with coherent Fe3Zr phase, and Fe3Zr phase with excellent stability contributed to the grea<sup>t</sup> improvement in creep strength at 973 K [12,13]. Thermodynamic modeling of Fe-Cr-Zr system through the Calphad approach found that Fe-Zr Laves phase could form in the ferritic Fe-Cr-Zr system [14]. Our recent investigation found that dispersed Fe2Zr phase were introduced

**Citation:** Chen, S.; Rong, L. Oxidation Behavior of Intermetallic Phase and Its Contribution to the Oxidation Resistance in Fe-Cr-Zr Ferritic Alloy. *Metals* **2022**, *12*, 827. https://doi.org/10.3390/ met12050827

 Academic Editor: Renato Altobelli Antunes

Received: 22 April 2022 Accepted: 9 May 2022 Published: 11 May 2022

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into equiaxed α-Fe matrix in the Fe-9Cr-2W alloy with the Zr content of 7~10 wt.%, and enhanced creep-rupture properties was achieved up to 973 K in comparison with the typical 9Cr2WVTa ferritic/martensitic steel [15]. L. Tan et al. found that only a small amount of radiation-induced precipitates was observed in Fe-9Cr-1W-11Zr alloy after Fe ion irradiation to ~50 dpa at 673 K, demonstrating promising radiation resistance [16,17]. It was found that the Fe-Cr-Zr alloy presented better high-temperature creep properties and superior radiation resistance in comparison with the ferritic/martensitic steel. In addition, corrosion is a life-limiting property when the alloy is exposed to the service environment. Liquid metal are the primary coolants for the advanced generation IV nuclear reactors. For instance, liquid Pb-Bi eutectic are the coolant for the advanced lead fast reactors. The liquid Pb-Bi eutectic is very corrosive towards the structural material [18–24]. Therefore, the corrosion properties in contact with liquid Pb-Bi eutectic should be considered. Until now, the corrosion properties of ferritic alloy with the dispersed intermetallic phase have not been clearly understood. In this study, the high-temperature oxidation resistance of Fe-Cr-Zr alloy in air and in static liquid Pb-Bi eutectic were investigated in view of the service environment in the advanced generation IV nuclear reactors.

#### **2. Materials and Methods**
