Tensile properties of Inconel 718 after low temperature neutron irradiation T.S. Byun * , K. Farrell Metal and Ceramics Division, Oak Ridge National Laboratory, Building 5500, P.O. Box 2008, MS-6151, Oak Ridge, TN 37831, USA Abstract Tensile properties of Inconel 718 (IN718) have been investigated after neutron irradiation to 0.0006–1.2 dpa at 60– 100 °C in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The alloy was exposed in solution-annealed (SA) and precipitation-hardened (PH) conditions. Before irradiation, the yield strength of PH IN718 was about 1170 MPa, which was 3.7 times higher than that of SA IN718. In the SA condition, an almost threefold increase in yield strength was found at 1.2 dpa, but the alloy retained a positive strain-hardening capability and a uniform ductility of more than 20%. Comparisons showed that the strain-hardening behavior of the SA IN718 is similar to that of a SA 316LN austenitic stainless steel. In the PH condition, the IN718 displayed no radiation-induced hardening in yield strength and significant softening in ultimate tensile strength. The strain-hardening capability of the PH IN718 decreased with dose as the radiation-induced dissolution of precipitates occurred, which resulted in the onset of plastic instability at strains less than 1% after irradiation to 0.16 or 1.2 dpa. An analysis on plastic instability in- dicated that the loss of uniform ductility in PH IN718 was largely due to the reduction in strain-hardening rate, while in SA IN718 and SA 316LN stainless steel it resulted primarily from the increase of yield stress. Ó 2003 Elsevier Science B.V. All rights reserved. 1. Introduction Inconel 718 alloy is a precipitation hardenable, nickel-based superalloy which has decent corrosion re- sistance, high strength at ambient temperature, and ex- cellent creep and fatigue strengths at high temperature. Thus the alloy has been used for a variety of applica- tions such as gas turbines, jet engines, steam generators, and fission and fusion reactor structures [1–10]. Re- cently, the alloy has been used for the beam window and other components of the Los Alamos Neutron Science Center (LANSCE) accelerator, and was selected as a back-up material for target components of the Spalla- tion Neutron Source (SNS) under construction at Oak Ridge National Laboratory (ORNL) and other large- scale accelerators [11–16]. In those applications, the alloy has a composite structure of austenitic matrix and precipitated c 0 and c 00 phases produced by a typical heat treatment consisting of solution annealing and aging [17]. It is known that high fluence irradiation of this composite structure undergoes negligible radiation-in- duced hardening or even softening with considerable decrease in ductility [7,11–13]. Transmission electron microscopy (TEM) investigations [4,15,16] show that superlattice spots diffracted from the c 0 and c 00 precipi- tates disappeared from the diffraction patterns of the aged Inconel 718 after irradiation to a fraction of 1 dpa. The dose-dependent softening was explained by the radiation-induced dissolution of the hard precipitates [13–16]. Although the low-temperature irradiation response of solution-annealed IN718 is unknown [16], it is ex- pected to behave like other solution-annealed face-cen- tered cubic (f.c.c) materials such as 304 and 316 stainless steels [18–21], which usually demonstrate significant ra- diation-induced hardening and retain high strain-hard- ening capability after irradiation. In the IN718 alloy, if the aging process for precipitate hardening is omitted in the heat treatment [1], radiation-induced softening might * Corresponding author. Tel.: +1-865 576 7738; fax: +1-865 574 0641. E-mail address: byunts@ornl.gov (T.S. Byun). 0022-3115/03/$ - see front matter Ó 2003 Elsevier Science B.V. All rights reserved. doi:10.1016/S0022-3115(03)00006-0 Journal of Nuclear Materials 318 (2003) 292–299 www.elsevier.com/locate/jnucmat