The　safety　of　nuclear　power　plants　has　put　forward　new　requirements　for　the　corrosion　resistance　of　fuel　cladding　with　the　advancement　of　industrialization.　FeCrAl　alloys　are　expected　to　be　a　substitute　for　Zircaloy.　This　work　summarizes　the　corrosion　of　Zircaloy　and　presents　a　comprehensive　review　on　the　corrosion　behavior　of　FeCrAl　alloys　including　oxidation　kinetics,　corrosion　processes　and　influencing　factors　in　nuclear　reactors.　Corrosion　performance　of　Zircaloy　in　nuclear　water　chemistry　is　definite　while　influencing　mechanisms　of　relevant　parameters　(e.g.　coolant　pH)　are　still　unclear.　A　small　change　of　LiOH　concentration　may　cause　contrary　effect　on　corrosion　behavior.　Different　from　Zircaloy　cladding,　FeCrAl　cladding　shows　various　kinetics　in　corrosive　environments.　A　single,　duplex　or　triple　oxide　layer　is　all　possibly　observed.　Specific　corrosion　behavior　depends　on　the　properties　of　water　chemistry　and　chemical　reactions　of　main　alloying　elements.　Developing　challenges　and　prospects　of　FeCrAl　alloys　are　concluded　to　expand　its　commercial　scale　in　nuclear　fuel　cladding.　For　successful　commercial　application　of　FeCrAl　cladding,　a　systematic　evaluation　considering　corrosion　resistance　and　other　properties　(e.g.,　mechanical　strength,　thermal　hydraulic　characteristics　and　neutron　economy)　is　required.　©　2021　Elsevier　B.V.