Creep performance of fuel cladding

The understanding of the creep behavior of nuclear fuel claddings is essential to predict safely and reliably the thermal performance and mechanical integrity of fuel rods. Fuel cladding tubes experience a range of changing conditions during their reactor life, further complicating the analysis. Today's nuclear reactors widely use zirconium alloys as fuel cladding material. Zirconium alloys exhibit anisotropic creep properties and their creep behavior depends significantly on the material condition. As for envisaged Gen-IV reactors, modified austenitic stainless steels and ODS alloys are candidate materials for claddings because of higher temperatures expected in operating conditions. This paper describes the research activities related to the creep behavior of cladding materials carried out at VTT. These activities include experimental research using the newly developed Pneumatic Loading Apparatus (PLA), which is capable of testing the steady state and transient creep properties of fuel cladding specimens with internal pressure of up to 700 bar and an additional axial force of up to 4 kN in tensile or compressive direction. Furthermore, creep models for cladding materials are being developed using viscoelastic modelling approach and Logistic Creep Strain Prediction (LCSP) method.