An isothermal aging study was carried out on alloy 718 for up to 50,000 hours in a temperature range of 593 C to 704OC. Large structural transitions occurred as the material was aged for 25,000 hours at 649°C and in 5,000 hours at 704°C. As the matrix f coalesced, f, delta, and aCr precipitated and grew in the grain boundaries. The drop in yield strength with increasing time of exposure or at higher temperatures was attributed to the overaging of the y phase, while much earlier drops in Charpy impact energies was more related to the grain boundary changes.Structures found in a stress-rupture sample which had been tested at 732OC for 5,400 hours at 69 MPa and structures found in the outer rim of 28,000 hour service turbine disk showed the same overaged structures as isothermally aged alloy 718 which had 25,000 hours aging at 649OC. The stress that can be applied to alloy 7 18 depends on the degree of overaging of the 1/' phase and subsequent f, delta, and aCr formed at the grain boundaries.

Results from elevated temperature-strain controlled fatigue and constant-strain-rate tensile tests conducted on specimens of stainless steel Types 304, 304L (titanium modified), 316, as well as Incoloy 800 are reported. Specimens were irradiated to fluences of 0.4 to 5 × 1022 n/cm2, E>0.1 MeV at 700 to 750 C (1292 to 1382 F), while the postirradiation test temperature was maintained at 700 C. Reductions in tensile ductility and fatigue life occurred, with reductions in fatigue life ranging from factors of approximately 1.5 to 2.5 for the stainless steels and up to 35 for Incoloy 800 in comparison with the thermal controls. Comparisons are made between actual irradiated fatigue behavior and predictions based on several semi-empirical methods using irradiated tensile data. These methods generally provided good estimates of the irradiated fatigue behavior of these materials. Introducing tensile hold times into the fatigue cycles of irradiated and unirradiated Type 316 stainless steel resulted in substantial reductions in the fatigue life of this material. However, for tensile hold times in excess of 0.1 h a tendency towards saturation of the hold-time effect was found in both the irradiated and unirradiated material. Creep and fatigue damage for Type 316 stainless was determined and summed linearly. This total damage was found to be a function of strain range, duration of tensile hold time, and irradiation condition for Type 316 stainless steel.

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The authors would like to express their appreciation to M. D. Harper and G. A. Rigby who helped with the experimental tests. Also, they gratefully acknowledge the support provided by the U. S. Atomic Energy Commission, Fuels and Materials Branch, Division of Reactor Development and Technology.

Results are reported for strain-controlled low-cycle fatigue measurements on Vanstar-7, 8, and 9 in both the irradiated and unirradiated conditions, conducted at 400°C. Specimens were irradiated in the experimental breeder reactor II (EBR-II) to fluences of 0.18 to 6.37 × 1025 neutrons (n)/m2 (E > 0.1 MeV) at 410 to 450°C. The results indicate that neutron irradiation had little or no effect on the fatigue life of any of the alloys. Comparison with data for Types 304 and 316 austenitic stainless steels tested at 400°C shows that on the basis of strain range the irradiated Vanstar alloys are inferior or equivalent to the stainless steels below 10 000 cycles-to-failure but become superior above that point. Proportional limit and yield-strength information have been extracted from the fatigue data, where possible. These data indicate that radiation hardening occurs rapidly at low neutron fluences and that strengthening saturates or increases very slowly above a fluence of approximately 1 × 1025 n/m2.