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    Response of Inconel 600 to Simulated Fusion Reactor Irradiation

    Published: Jan 1979

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    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.


    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

    Author Information:

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

    Committee/Subcommittee: E10.07

    DOI: 10.1520/STP38160S

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