The fatigue assessment of safety relevant components is of importance for ageing management with regard to safety and reliability of nuclear power plants. Austenitic stainless steels are often used for reactor internals due to their excellent mechanical and technological properties as well as their corrosion resistance. During operation reactor internals are subject to mechanical and thermo-mechanical loading which induce low cycle (LCF), high cycle (HCF) and even very high cycle (VHCF) fatigue. While the LCF behavior of austenitic steels is already well investigated the fatigue behavior in the VHCF regime has not been characterized in detail so far. Accordingly, the fatigue curves in the applicable international design codes have been extended from originally 106 to the range of highest load cycles up to 1011 load cycles by extrapolation. Nevertheless, the existing data base for load cycles above 107 is still highly insufficient. The aim of the cooperative project of the Institute of Materials Science and Engineering (WKK) at University of Kaiserslautern, Materials Testing Institute (MPA) Stuttgart and Framatome GmbH, Germany is to create a comprehensive database up to the highest load cycles N = 2·109 for austenitic stainless steels at ambient and elevated temperature. Hence, an elastic-plastic material model was developed to convert the displacement-controlled fatigue tests performed on an ultrasonic fatigue testing system to a total strain amplitude, which can be integrated into existing design codes and used to assess critical components.