Improved irradiation resistance of accident-tolerant high-strength FeCrAl alloys with heterogeneous structures

Abstract

Post–neutron irradiation examination is performed on advanced accident-tolerant fuel (ATF) cladding iron-chromium-aluminum (FeCrAl) alloys with ∼10–13at. % Cr, ∼10–12 at. % Al, ∼1 at. % Mo, and minor alloying elements including Y irradiated to a damage level of 7 displacements per atom (dpa) at irradiation temperatures of 267–282 °C. A compositional dependency of the Cr and Al content is observed on the ratio of sessile and glissile dislocation loops, where the density of a⟨100⟩ type loops is somewhat higher than the a/2⟨111⟩ type loops. The α′ precipitate number density is inversely correlated to the starting Cr concentration of the alloys of interest. The irradiation to a higher dose of 7 dpa results in a higher density of dislocation loops and α′ precipitates for the same alloys at a lower irradiation dose, such as 1.8 dpa. In this work, the effect of α′ precipitates on the dislocation loop density is discussed, and the presence of α′ appears to inhibit the nucleation of loops. Compared with first-generation FeCrAl alloys, these advanced alloys with heterogeneous structure exhibit a lower Cr concentration in α′ precipitation at the same dose level; they act as weaker obstacles deviating from the primary hardening contribution from the mature α′. Hence, the overall irradiation-induced hardening decreases; our alloys show improved radiation resistance because of their stronger sink strengths. The results presented in this paper could provide insights for the design and optimization of ATF cladding materials for future fission and space applications.

Introduction

Over the past decade, iron-chromium-aluminum (FeCrAl) alloys have been proposed as one of the accident-tolerant fuel (ATF) cladding materials for nuclear reactors. Following the Fukushima Daiichi nuclear accident, ATF concepts have focused on the use and application of advanced materials and alloys for nuclear applications. These alloys have superior properties over the conventionally used Zr alloys in the high-temperature environments of the current light-water reactor (LWR) fleet, especially in station blackout scenarios [1], [2], [3], [4], [5]. These ATF cladding materials also show negligible interaction rates with high-temperature water and thus generate little hydrogen, which is a major cause of the hydrogen/steam explosion during a meltdown accident [6], [7], [8]. They are also designed to withstand the extreme conditions in nuclear reactors such as intense neutron irradiation, high temperature, oxidation, and corrosion while showing good high-temperature thermal and mechanical properties [9,10]. FeCrAl alloys have displayed all of the aforementioned properties while maintaining tolerable radiation resistance [11]. In the initial development stage of the FeCrAl alloys, their excellent corrosion resistance is one of the key performance factors that makes them suitable for ATF cladding: because of the formation of passivating chromia and alumina layers formed by the high-temperature steam; further reaction of the fuel cladding and coolant is effectively prevented [12], [13], [14], [15]. The addition of minor elements can improve properties of the ternary FeCrAl system; for example, Y was added to slow the oxidation rate [16], [17], [18], [19], [20], whereas Si, Nb, and Mo were added to enhance the solid solution strengthening and grain refinement for better high-temperature strength [12,[20], [21], [22]].

Despite the advantages of FeCrAl alloys, the existence of Fe-Cr phase separation into α and αʹ phases at modest to low temperatures is a challenging issue. It requires significant effort to balance the alloy composition and reduce the formation of radiation-enhanced Cr-rich clusters: αʹ precipitates have been demonstrated to be a major cause of the irradiation-induced hardening and embrittlement in the FeCr/FeCrAl alloy systems [23]. Several studies performed via pure thermal aging illustrated that the addition of Al and reduction of the Cr content could decrease the propensity for αʹ precipitation and improve oxidation resistance. This trend has also been observed in first-generation (Gen I) FeCrAl alloys under neutron irradiation up to 7 displacements per atom (dpa) and second-generation (Gen II) FeCrAl alloys at 1.8 dpa under neutron irradiation through atom probe tomography (APT) [16,17]. Complementary studies via small-angle neutron scattering (SANS) have shown similar trends. The density of αʹ precipitation was shown to decrease as the Cr content decreases and that of Al increases (including the two commercial FeCrAl alloys Alkrothal 720 and Kanthal APMT) at 0.3–7 dpa [25], [26], [27]. Moreover, a series of ion irradiation experiments demonstrated the dislocation loop evolution up to 16 dpa [28], [29], [30]. Since these studies focused on a higher dose rate and less irradiation time compared to those of the neutron irradiation studies, αʹ precipitates were not observed, and the authors aimed to separate the effect of the irradiation dose and temperature from the presence of the αʹ precipitates. Furthermore, the neutron irradiation data only investigated the dislocation structure up to 1.8 dpa, and the loop formation and structure are still unknown at higher doses.

This work aims to enhance the current knowledge in the microstructural evolution trends in neutron-irradiated FeCrAl alloys and demonstrate the irradiation-induced transition patterns with respect to chemistry, defect sinks, and irradiation conditions. (It is noted that there are also significant ion-beam-irradiation studies available in the literature, but here we focus on neutron-irradiation microstructural evolution [28], [29], [30], [31], [32], [33], [34].) Here, we report the first data of post irradiated microstructures of the second generation (Gen II) wrought FeCrAl alloys is provided via advanced electron microscopy to address the lack of loop evolution data at higher neutron irradiation dose levels. We compare our results on irradiation-induced microstructure in FeCrAl alloys with previous studies. Our work emphasizes the impact of Cr-rich αʹ precipitation on the trend of dislocation evolution under the irradiation conditions examined here. The radiation-enhanced diffusion (RED) perspective is then provided as an alternative way to elucidate the potential mechanisms driving defect nucleation and growth. The irradiation-hardening mechanism is discussed to reflect the barrier strength of each defect, allowing for optimization of such ATF cladding for extreme conditions.