The reactor pressure vessel of a light water nuclear power plant is subjected to a variety of ageing processes. The prevailing phenomena are neutron irradiation embrittlement and corrosion assisted crack growth. This paper gives an overview of the underlaying mechanisms, material parameters that describe the susceptibility to those mechanisms and the resulting effects on material properties. With respect to neutron irradiation the conservatism of the presently used prediction curves could be demonstrated in the lower fluence range by investigations of materials from a decommissioned reactor pressure vessel. To maintain sufficient safety margin throughout the entire life, mitigation measures and plant experience are discussed. An important role in the safety assessment is the repeated in-service inspection and continuous monitoring of load and temperatures to ensure that the allowable limits for crack sizes are not exceeded. Test methods are presented to investigate the mechanisms of corrosion assisted crack growth and to quantify the crack growth rate and the relevant parameters.