A series of in-pile tests simulating a Loss Of Coolant Accident (LOCA) in a Pressurised Water Reactor (PWR) were performed by the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) from 1981 to 1984 in the PHEBUS reactor, at Cadarache Nuclear Centre (France). The objectives of the two PHEBUS-LOCA tests 215-R and 218 were to investigate phenomena occurring inside a bundle of fresh fuel rods under various representative LOCA conditions. The FRETA-F code was initially developed by the Japan Atomic Energy Research Institute (JAERI) to analyse the thermal and mechanical behaviour of fuel rods in a PWR during a LOCA transient. Recently, the code was updated and improved by IRSN in order to extend its capability (V1.1 version). This paper presents the assessment of this new version against the two PHEBUS-LOCA tests. The analyses mainly focused on the physical behaviour of the cladding during the transients : temperature histories, clad deformation and burst, flow blockage of the fluid section and zircaloy cladding oxidation. The calculations indicated an overprediction of the cladding hoop strain at burst. This analysis highlights the need for modelling the thermal and mechanical interactions between the rods and therefore for developing an advanced 3D multi-rod code associated with experimental activities for the assessment of the new models.