[en] A reactor pressure vessel may be exposed to the most severe loading during its operational life, when in emergency cooling cold water is injected into it. The very high thermal stresses combined with the stresses due to internal pressure may cause initiation of an existing crack and its propagation into the pressure vessel wall. A joint pressure vessel integrity research program between three partners has been going on since 1990. The partners are the Prometey Institute from Russia, the Imatran Voima Oy (IVO) from Finland and the Technical Research Centre of Finland (VTT). The main objective of the research program is to increase the reliability of the VVER-440 reactor pressure vessel safety analysis. This is achieved by providing the material property data for the VVER-440 pressure vessel steel and by producing experimental knowledge of the crack behaviour in pressurized thermal shock loading for the validation of different fracture assessment methods. The program is divided into four parts: pressure vessel test, material characterization, computational fracture analyses and evaluation of the analysis methods. The testing program comprises tests on two model pressure vessels with axial surface flaws. The second model vessel has an austenitic steel cladding. A special heat treatment is applied to the vessels prior to the tests in order to simulate the end of life toughness state of a real reactor pressure vessel. In this paper, the results of the test and the computational analyses considering the first (uncladded) model vessel are discussed. The fracture behaviour of the model vessel based on fractographic examinations and test measurements is described. Both the results of pre-test analyses using initial material properties and post-test analyses using actual material properties are compared with the experimental observations. Finally, evaluation of the fracture assessment methods is performed. 12 refs., 12 figs., 4 tabs