The stabilized austenitic steel DIN 1.4970 is the candidate cladding material for the fast flux materials testing reactor MYRRHA in development at the Belgian Nuclear Research center (SCK-CEN). As the first containment barrier of the nuclear fuel and fission products, geometrical stability and maintenance of good mechanical properties of the cladding under high neutron flux and elevated temperatures is paramount. Between the 1970's and the 1990's, the composition of DIN 1.4970 was tailored for optimal irradiation swelling resistance. The main microstructural features found to be responsible were coherent TiC nanoprecipitates, <10 nm in size, present on dislocations, which act as recombination centers of vacancies and interstitials which are damage of the steel crystal lattice caused by neutron bombardement and the cause of swelling and embrittlement. These beneficial TiC nanoprecipitates were found to form in highly cold worked material during reactor operation or after short ageing heat treatments between 600 and 950 degrees, and were shown to be highly resistant to coarsening. For optimal swelling resistance, a high number density of small precipitates - and hence a high number of dislocations acting as nucleation sites - is desired.
Under MYRRHA conditions, TiC nanoprecipitates are not expected to precipitate in pile. For this reason, a preliminary ageing heat treatment may be envisioned. Ageing at elevated temperature may induce unwanted microstructural changes such as recovery and recrystallization, while ageing at low temperature may result in the precipitation of other phases besides the nanoprecipitates. All these may influence the mechanical properties, which is why it is important to map out how the microstructure changes with heat treatment.