Assessment of creep in reactor-irradiated CuCrZr alloy intended for the ITER first wall panels

Highlights: • Assessment of in-pile creep relaxation of CuCrZr joints with CuCrZr and 316L(N)-IG (ITER Grade) stainless steel. • Irradiation in the BR2 reactor in Mol at 257 °C in water environment up to 0.7 dpa. • Thermal creep tests on the irradiated samples at 250 °C. • Reference tests after heating equivalent to the irradiation conditions. • The level of a possible creep under irradiation is below the experimental uncertainty. - Abstract: CuCrZr alloy is candidate heat sink material for the ITER blanket, first wall, and divertor. During ITER operation it will be exposed to a combination of elevated temperatures, heat flux, and intense fast neutron radiation. This environment will challenge the performance of components and joints based on CuCrZr. To address this issue, mechanical tests were performed with irradiated and reference specimens of CuCrZr and its joints with 316 L(N)-IG (ITER Grade) stainless steel made by hot isostatic pressing. The reactor exposure up to ∼0.7 dpa was performed in the BR2 reactor at SCK• CEN, in water at a temperature of 257 °C A special design was used to allow irradiation of specimens axially pre-stressed at different strain levels. The post-irradiation examination included: (i) tensile test, (ii) measurements of plastic deformation of samples axially loaded during irradiation (in situ creep test); (iii) thermal creep tests on irradiated samples. The fracture surfaces were examined in a hot cell using a Scanning Electron Microscope (SEM). The results were compared with data obtained from mechanical tests and SEM/EDX fracture surface analysis on non-irradiated reference samples. The level of a possible creep under irradiation is below the experimental uncertainty.