The results presented below are related to potential cladding materials for fast reactor fuel elements, that is, austenitic stainless steels and precipitation-hardened nickel alloys.
These materials were irradiated to a dose of 3 × 1021 nvt (E > 1 MeV) in different thermal reactors and to a dose of 1.6 × 1022 nvt (total dose) in a fast reactor. The irradiation temperature was about 700 C in thermal reactors and about 500 to 600 C in a fast reactor. After irradiation, tension tests and uniaxial stress rupture tests were performed at temperatures in the 600 to 800 C range.
For all alloys, high-temperature tests after irradiation have shown a more or less pronounced embrittlement, not recoverable by any post-irradiation annealing. This embrittlement is associated with intergranular wedge-type cracks initiating at triple points. For a given alloy, at a given temperature, the lower the rate of deformation, the more pronounced the embrittlement.
For Type 316 austenitic steels, we observed that the radiation induced hardening was low and disappeared at a maximum temperature of 800 C. The embrittlement was associated with a decrease in the nonuniform part of the elongation in tension tests and with a decrease in the elongation and duration in the tertiary stage of stress-rupture tests. Two types of factors were studied: 1. Those concerning the structural parameters: grain size, precipitation, prestraining, and chemical composition. 2. These concerning the actual conditions of irradiation.