HIGH TEMPERATURE CYCLIC DEFORMATION BEHAVIOR OF AN ADVANCED AUSTENITIC STAINLESS STEEL (ALLOY 709) Zeinab Y. Alsmadi 1* , Abdullah S. Alomari 1 , N. Kumar 2 and K.L. Murty 1 1 Department of Nuclear Engineering, North Carolina State University, Raleigh, NC, USA 2 Metallurgical and Materials Engineering, The University of Alabama, Tuscaloosa, AL *Corresponding author: zalsmad@ncsu.edu Advanced structural materials are needed to improve the efficiency, safety and reliability of next-generation nuclear reactors. Preliminary investigation on mechanical properties of a Fe-25wt.%Ni-20Cr austenitic stainless steel (Alloy 709) suggests it might be a potential candidate for Sodium-Cooled Fast Reactor. High temperature cyclic deformation behavior including fatigue and creep-fatigue for the Alloy 709 is considered for design and safety purposes. Strain-controlled low- cycle fatigue tests were performed at strain amplitudes ranging from 0.15% to 0.6% at 750 °C in air following ASTM standard E2714–13, with a constant strain rate of 2 × 10 -3 s -1 . In addition, different hold times of 1, 10, 30 and 60 minutes were introduced at the maximum tensile strain to investigate the effect of the creep damage on the fatigue-life at strain amplitude of 0.5% at 750 °C, while 30 and 60 minutes were introduced at the maximum tensile strain at strain amplitude of 0.3% at 750 °C. During continuous cyclic loading, fatigue life is found to decrease with increase in strain amplitude, while the creep-fatigue life is found to decrease with increasing hold time indicating rapid initiation and propagation of cracks. The fractographs of the deformed samples are presented and discussed. I. INTRODUCTION Generation IV nuclear reactors are designed to be safer, more efficient and reliable, and have a longer design lifetime than current nuclear reactors. While current reactors have an average operational life of 30-40 years, however, Generation IV nuclear reactors can have a longer lifetime of 60-80 years and possibly more. One of the Generation IV nuclear reactors is Sodium-Cooled Fast Reactor (SFR) that operates at temperatures above 550 o C and uses sodium as coolant and moderator. The structural materials of SFR need to have mechanical properties that enable it to exhibit high tolerance at high temperatures, corrosive environments and relatively higher radiation doses (typically 80-200 dpa). One of the candidate structural materials for SFR is Fe-25Ni-20Cr (wt.%) advanced austenitic stainless steel (known as Alloy 709) stabilized by nitrogen and strengthened by niobium. Alloy 709 is an excellent candidate as structural material for SFR because of its high corrosion-resistance, sodium compatibility, thermal stability, high temperature strength and good creep properties. In order to understand the plastic deformation of this alloy and predict its response under typical SFR operating conditions, different mechanical tests should be performed and thus explain the damage mechanisms at similar operating conditions expected in SFR. Creep- fatigue damage is expected to occur in many components such as reactor cladding, pressure vessels and gas turbines operating at high temperatures in next-generation nuclear reactors [1-5]. In this paper, the creep-fatigue interaction of the Alloy 709 is investigated by conducting strain-controlled low-cycle fatigue (LCF) tests at strain amplitudes ranging from 0.15% to 0.6% at 750 °C and creep-fatigue tests with hold times of 60, 600, 1,800 and 3,600 seconds at strain amplitude of 0.5% at 750 °C, while 1,800 and 3,600 seconds at strain amplitude of 0.3% at 750 °C. As shown in Fig. 1a, when no hold time is introduced, a triangle waveform is produced while a trapezoidal waveform is produced with imposed hold times (Fig. 1b). Also shown in Fig. 1 is the corresponding variation of the stress and hysteresis loops during loading cycles. II. EXPERIMENTAL SETUP II.A. Material The investigated material is a Fe-25Ni-20Cr advanced austenitic stainless steel (Alloy 709) received in the form