The austenitic stainless steel DIN 1.4970 is the prime candidate fuel cladding for the future nuclear research reactor MYRRHA, which is located in Belgium. To increase the understanding of this material, the mechanical properties as well as microstructure were investigated. Tube material of 24% cold worked DIN 1.4970 and 46% cold worked DIN 1.4970 were tested in tension using two geometries: axial and ring geometry. Tensile tests were performed at a variety of temperatures and strain rates in the axial and hoop directions. An additional series of tensile tests at room temperature was performed on tube sections subjected to heat treatments with various time and temperature. A finite element analysis (FEA) was implemented to simulate gauge-less ring tensile tests that were previously found to be more reliable than gauged specimens. Difference in tensile properties between ring specimens with different cold-work levels or annealing treatments could be satisfactorily reproduced by FEA, validating of the model. The effect of sample geometry, testing temperature and ageing pre-treatments is discussed. It was found that the mechanical properties of the steel were strongly dependent on the temperature at which the material was tested, such that the samples showed lower ductility at elevated temperature, and the maximum reached load (force) was lower than at room temperature. The mechanical properties were also highly dependent on the heat treatment conditions. FE models could disclose approximate tensile properties in the hoop direction for the material. Fracture surface morphology and microstructure were investigated by scanning electron microscopy (SEM) and electron backscatter diffraction (EBSD), respectively. A ductile fracture was also observed in all samples, and the material texture could be examined.
|State||Published - 1 Jan 2018|