Supercritical water-cooled reactor materials – Summary of research and open issues

• International cooperation under Generation IV led to major advances in SCWR materials and chemistry.
• The selection and qualification of a suitable fuel cladding alloy remains the biggest challenge.
• The general corrosion of iron- and nickel-based alloys in SCWR conditions is now fairly well understood.
• The SCC behavior is less clear and sometimes results of different groups are even contradictory.
• The biggest gap in knowledge remains in the behavior of the candidate alloys under irradiation.

The Supercritical Water Reactor (SCWR) is one of the six reactor concepts being investigated under the framework of the Generation IV International Forum (GIF). Research on materials and chemistry for supercritical water-cooled reactors dates back to the 1960s when a number of reactor concepts using water at supercritical temperatures but sub-critical pressures (nuclear steam) were studied. There is also significant experience available from the operation of supercritical fossil-fired power plants. In this paper, the materials requirements of the various SCWR concepts are introduced, with a focus on the European Union pressure vessel concept and the Canadian pressure tube concept. The current understanding of the key materials degradation issues is reviewed, and knowledge gaps identified.