Zirconium and zirconium based alloys have been used for many decades in the thermal reactor industry. Their main advantages over several other common engineering materials include: excellent corrosion resistance in aqueous environments; acceptable high-temperature mechanical properties; and low thermal neutron absorption cross section. However, it would appear that the fusion community has shown little interest in this very well characterised and useful material. Whatever the reasons for its neglect, this paper is intended as a reassessment of zirconium-alloys and of their potential application in fusion power plants. It is hoped that it will promote further discussion and, perhaps, reopen the case for zirconium.

The metallurgical and mechanical properties of zirconium and zirconium-alloys have been reviewed in the light of forty years of accumulated fission reactor experience. Zirconium-based materials appear to have a good combination of thermo-physical-mechanical properties that make them potentially attractive for some fusion uses. Using simple neutronics models, the FISPACT inventory code and the anticipated first wall conditions in two different tokamak concepts, calculations have been performed from which it would appear that corrosion, hydride formation and activation characteristics do not bar zirconium/zirconium-alloys from use in near-term fusion power plants (including possibly ITER). However, there are still doubts over the lifetime performance of zirconium-alloy under combined creep, fatigue and irradiation conditions, typical of the first wall of a fusion power plant.

* This work was funded jointly by the UK Department of Trade and Industry and Euratom.