Advanced materials are needed to withstand higher operating temperatures for longer service life providing more efficient, safe, and reliable structures for next generation nuclear reactors. Alloy 709 is an advanced austenitic stainless steel with excellent mechanical properties that make it a preferred candidate structural material for Sodium Fast Reactor (SFR) which has higher technology readiness level for deployment. Tensile tests of Alloy 709 were carried out in a wide range of temperatures 25–800 °C and nominal strain rates of 10-5–10-3 s-1 to investigate the serrated flow behavior which is a manifestation of dynamic strain ageing (DSA). Serrations in the Alloy 709 were found to occur at two temperature regimes identified as low temperature serrated flow (LT-SF) at 200–350 °C and high temperature serrated flow (HT-SF) at 350–700 °C separated by mild serrations or smooth flow depending on the strain rate. Different methods for the determination of activation energy for the serrated flow were employed that yielded values of 106 kJ/mole and 194 kJ/mole for the LT-SF and HT-SF regimes respectively. Based on the activation energy values and the dependence of the critical strain on the temperature and strain rate, diffusion of interstitial atoms such as nitrogen and/or carbon have been suggested to be responsible for serrated flow in the LT-SF regime while the diffusion of substitutional atoms such as chromium is responsible for DSA in the HT-SF regime. Other manifestations of DSA in the Alloy 709 are observed including peaks and/or plateaus in flow stresses, negative strain rate sensitivity and planar substructure in deformed samples at intermediate temperatures. However, no loss in ductility is found to occur in the DSA regime in the Alloy 709. The results are discussed and compared with relevant austenitic stainless steels used for SFR applications.