The mechanical properties of 316L/304L stainless steels, alloy 718 and MOD 9CR-1MO after irradiation in a spallation environment

The accelerator production of tritium (APT) project proposes to use a 1.0 GeV, 100 mA proton beam to produce neutrons via spallation reactions in a tungsten target. The neutrons are multiplied and moderated in a lead/aluminium/water blanket and then captured in 3He to form tritium. The materials in the target and blanket region are exposed to protons and neutrons with energies into the GeV range. The effect of irradiation on the tensile and fracture toughness properties of candidate APT materials, 316L and 304L stainless steel (annealed), modified (Mod) 9Cr-1 Mo steel, and Alloy 718 (precipitation hardened), was measured on tensile and fracture toughness specimens irradiated at the Los Alamos Neutron Science Center accelerator, which operates at an energy of 800 MeV and a current of 1 mA. The irradiation temperatures ranged from 50-164 deg C, prototypic of those expected in the APT target/blanket. The maximum achieved proton fluence was 4.5 x 1021 p/em2 for the materials in the centre of the beam. This maximum exposure translates to a dpa of 12 and the generation of 10 000 appm H and 1 000 appm He for the type 304L stainless steel tensile specimens. Specimens were tested at the irradiation temperature of 50-164 deg C. Less than 1 dpa of exposure reduced the uniform elongation of the Alloy 718 (precipitation hardened) and mod 9Cr-1Mo to less than 2%. This same dose reduced the fracture toughness by 50%. Approximately 4 dpa of exposure was required to reduce the uniform elongation of the austenitic stainless steels (304L and 316L) to less than 2%. The yield stress of the austenitic steels increased to more than twice its non-irradiated value after less than 1 dpa. The fracture toughness reduced significantly by 4 dpa to ∼ 100 MPa.m1/2. These results are discussed and compared with results of similar materials irradiated in fission reactor environments. (authors)