Interaction of neutron and thermal environmental factors in the embrittlement of selected structural alloys for advanced reactor applications

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An experimental irradiation was conducted in the Big Rock Point Reactor (BRPR) to obtain comparisons of the individual irradiation response of five structural steels of current and potential application for nuclear reactor pressure vessels. The steels included A302-B, A543, A517-E, 7−12%Ni-Cr-Mo and 12Ni-5Cr-3Mo (2 heats) in plate form. The temperature and duration of the irradiation exposure were 585°F (307°C) and 9,726 hr respectively. The exposure was conducted under typical boiling water reactor conditions.Charpy-V (CV) specimens of the five steels were evaluated in the as-irradiated condition and in the postirradiation annealed condition. Non-irradiated CV specimens aged at 575°F (302°C) for 9,726 hr were also evaluated. Additional non-irradiated CV specimens were aged at 700 and 800°F (371 and 427°C) for 168 hr as a reference for evaluating the postirradiation annealed specimens.Postirradiation CV 30 ft-lb transition temperatures of all materials were at or below boiling water temperatures; individual transition temperature increases ranged from 95 to 205°F (53 to 114°C) for fluences of 4.7 to 7.4 × 1019 n/cm2 > 1 MeV. Postirradiation annealing was found to be generally less effective with higher initial yield strength steels. The effect of long-term 575°F (302°C) aging on notch ductility behavior was negligible for the lower yield strength steels but produced transition temperature increases of 55 to 150°F (31 to 80°C) with the higher yield strength steels. Aging of unirradiated CV specimens for 168 hr at 700 and 800°F (371 and 427°C) resulted in transition temperature increases for most steels that were very close to the transition temperatures after postirradiation annealing.Tensile properties were found altered only slightly by the irradiation exposure. Small increases in yield and tensile strength were accompanied by small decreases in tensile ductility.In terms of postirradiation strength and ductility, the three lower yield strength steels appear to be more acceptable for application in reactor pressure vessels than the two higher strength steels. The data suggest that the 12-5-3 maraging steel is not suited for long-term elevated temperature reactor service due to thermal instability.