An overall evaluation of several in-reactor creep and creep-rupture experiments in BR2 and FFTF on pressurized tubes of the stabilized austenitic stainless steels 1.4970, 1.4981, 1.4988 and the Ni-base alloy Hastelloy-X is given in this report. Even at temperatures around and above 0.5 T_m the analysis of the data implies that for low stress levels the stress-induced preferred absorption (SIPA) creep process is the predominant deformation mechanism. The results show that the in-reactor rupture lives of the austenitic steels as well as the Ni-base alloy Hastelloy-X are reduced compared with the rupture lives of the corresponding material in the unirradiated and post-irradiation tested conditions, so that an additional damage mechanism via helium has to be assumed to account for the reduction in rupture life.

Analytical models have been developed, which accounts for the critical variables stress, temperature, and dpa rate and which describes in a more general way the in-reactor creep and creep-rupture life behavior of the materials mentioned above. The analysis of the results indicates clearly that the use of the unirradiated and post-irradiation data for predicting in-reactor creep and failures is invalid for these materials.