The precipitation strengthened alloys have the potential for use in fuel cladding and duct applications for liquid metal reactors due to their high strength and low swelling rate. Unfortunately, these high strength alloys tend to exhibit poor fracture toughness, and the effects of neutron irradiation on the fracture properties of the material are of concern. Compact tension specimens of alloy D21 were irradiated in the Experimental Breeder Reactor II to a fluence of 2.7 x 1022 neutrons (n)/cm2 (E > 0.1 MeV) at 425, 500, 550, and 600°C. Fracture toughness tests on these specimens were performed using electric potential techniques at temperatures ranging from 205 to 425°C. The material exhibited low post-irradiation fracture toughness which increased with either increasing test or irradiation temperature. The tearing modulus, however, increased with increasing irradiation temperature but decreased with increasing test temperature. Results were analyzed using the J-integral approach. The fracture toughness of irradiated D21 was evaluated essentially following the procedure recommended in ASTM Test Method E813. It was found that the data elimination limits illustrated in E813 were too large for the specimens tested, although the thickness criterion was satisfied. The precautions needed to determine JIc based on a reduced data qualification range were discussed.

The alloy V-15Cr-5Ti was cyclotron-implanted with 80 appm He and subsequently irradiated in the Experimental Breeder Reactor (EBR-II) to 30 displacements per atom (dpa). The same alloy was also irradiated in the 10, 20, and 30% cold-worked conditions. Irradiation temperatures ranged from 400 to 700°C. No significant effects of helium on mechanical properties were found in this temperature range although the neutron irradiation shifted the temperature of transition from cleavage to ductile fracture to about 625°C. Ten percent cold work was found to have a beneficial effect in reducing the tendency for cleavage fracture following irradiation, but high levels (20%) were observed to reduce ductility. Still higher levels (30%) improved ductility by inducing recovery during the elevated-temperature irradiation. Swelling was found to be negligible, but precipitates--titanium oxides or carbonitrides--contained substantial cavities.