Early corrosion behaviour of irradiated FeCrAl alloy in a simulated pressurized water reactor environment

Hydrothermal Corrosion of Latest Generation of FeCrAl Alloys for Nuclear Fuel Cladding

“…In the case of the model alloy Fe17Cr5.5Al flat coupon specimens, a surface Fe-Cr oxide spinel was observed, followed by enriched Cr and Al oxides (Figure 10). It is believed that the continuous Cr oxide layer underneath the spinel provides a low partial pressure of oxygen activity [26] that is needed to stabilize Al oxide [27]. While the FA-SMT and PM-C26M tube specimens are fabricated by powder metallurgy, the model alloy flat coupons are manufactured by VIM.…”

Review on synergistic damage effect of irradiation and corrosion on reactor structural alloys

“…However, based on the available study data [19-21, 74, 76, 82, 85-89], it has been observed that irradiation affects the quantitative chemistry, morphology, and thickness of the oxide layers formed on the alloys. In irradiated materials, the Cr content in the inner layer has been observed to be high [19,[76][77][78]85]. In contrast, a lower Cr content has been observed in the inner oxides of 316 L stainless steel irradiated with protons exposed to high-temperature water, due to the dissolution of Cr-rich spinel oxides in irradiated water with added hydrogen [90].…”

“…For a comprehensive understanding of the composition and structure of the oxide formed on the metal surface, as well as the depth of the inner oxide layer following irradiation, extensive studies have been conducted, which have been summarized in Table 6 [19-21, 76, 82, 83]. Although Deng et al [20] have already reported that a finer oxide grain in the inner oxide film was formed on a 3-dpa proton-irradiated 316 L austenitic stainless steel after exposure to the primary water of the simulated PWR for 500 h, it is generally accepted that irradiation does not alter the crystal structure and the qualitative chemistry of the inner and outer oxides formed on materials, which can be characterized using high-resolution TEM (HRTEM) [19,21,74,77,78,82,84,85]. Similar to the unirradiated sample, the oxide layer maintains its characteristic double-layer structure when stainless steel is irradiated.…”

“…Generally, these irradiation-induced defects accelerate corrosion primarily in two ways [19,20,74,83,[85][86][87][88][89]:…”

Friction and Wear Properties of Cr-Nx Coatings for Nuclear Fuel Cladding

“…In addition to replacing UO 2 fuel by using materials such as U 3 Si 2 [7], UN [8], etc., with higher thermal conductivity, there are two main ways to improve the performance of fuel cladding in ATF systems. The first method is to replace the Zr alloy with untraditional fuel-cladding materials, such as SiC [9][10][11] and FeCrAl [4,[12][13][14]). The second one is to deposit a layer of coating on the surface of the Zr alloy to improve its oxidation resistance and mechanical performance under accident conditions [15].…”