Supercritical Water-cooled Reactors (SCWR) is one of the Gen IV reactor systems being developed through an international treaty-level collaboration led by Generation IV International Forum (GIF). Following the release of the core concepts from the Japanese and the EU members, Canada published its own pressure-tube based conceptual design in 2015. With an outlet temperature of 625 C and a core pressure of 25 MPa, the Canadian SCWR concept requires cladding materials that can sustain extremely harsh in-core physical and chemical conditions. Based on initial calculations using stainless steels and nickel alloys, the maximum cladding surface temperature is predicted to be as high as 825 C; the irradiation dose can reach as much as 10 dpa and the supercritical coolant can be very oxidizing due to the production of oxygen and hydrogen peroxide by radiolysis of light water. Under these conditions, the cladding can be readily degraded by corrosion, stress-corrosion, creep, or any of the radiation-related processes such as void-swelling and embrittlement. In the course of the Canadian program (2007-2015) on R&D of in-core materials, unique experimental and computational facilities were set up specifically to probe into the behaviours of candidate alloys under these extreme conditions. Novel, and frequently surprising, results have been achieved. Key highlights of this collaborative R&D effort are presented in this paper and the challenges for the future are also discussed.