STP683: Response of Inconel 600 to Simulated Fusion Reactor Irradiation

    Wiffen, FW
    Oak Ridge National Laboratory, Oak Ridge, Tenn.

    Pages: 19    Published: Jan 1979


    Abstract

    Inconel 600 was irradiated in the High Flux Isotope Reactor (HFIR) to provide a partial simulation of fusion reactor service. Specimens were irradiated at 55 to 700&C to investigate swelling and postirradiation tensile properties as a function of irradiation and test temperatures under conditions of concurrent displacement damage and helium production. Helium contents from 600 to 1800 atomic parts per million (appm) and displacement levels of 4 to 9 displacements per atom (dpa) were achieved, and the results are used to estimate performance in a fusion reactor environment. The swelling was weakly dependent on temperature between 300 and 600&C, with swelling ranging from 0 to ∼ 1 percent, but increased rapidly above 600°C. The swelling values were much larger than expected from fast reactor and ion bombardment results. Cold-work was not effective in suppressing swelling of Inconel 600. Tensile property measurements and fractography on the same specimens showed strength values increased for irradiation at 55 to 400°C but decreased below unirradiated values for irradiations at 600 and 700°C. Elongation values were lowest at the temperature extremes. Total elongations below 1 percent were found only for irradiation and test temperatures of 600 and 700°C. The fractures were completely transgranular for specimens irradiated and tested at 300 and 400°C, of mixed mode but predominately intergranular at 500°C, and fully intergranular at 600 and 700°C. The results suggest that Inconel 600 does not offer any advantages over Type 316 stainless steel and does not warrant further development for fusion reactor application.

    Keywords:

    radiation, irradiation, neutron irradiation, nickel-based alloys, austenitic alloys, radiation damage, tensile properties, ductility, fractography, swelling, swelling (helium bubbles), helium, displacement damage, thermal reactors, fusion reactors


    Paper ID: STP38160S

    Committee/Subcommittee: E10.07

    DOI: 10.1520/STP38160S


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