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    Fracture Toughness and Tensile Properties of Irradiated Reactor Pressure Vessel Cladding Material

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    A comprehensive testing programme was undertaken to evaluate the effects of irradiation, thermal ageing, specimen orientation, and test temperature on the fracture oughness and tensile behaviour of a Type 309L/308L stainless steel strip clad deposit. Specimens were irradiated in the high flux reactor (HFR) Petten to nominal neutron doses of 0.05 dpa and 0.1 dpa at 295°C. Testing was performed at room temperature, 100, 200 and 295°C. The fracture resistance properties of the clad material were unaffected by irradiation to 0.1 dpa, which is approximately twice the predicted end-of-life dose. The same neutron dose resulted in an increase in 0.2% yield stress (15-20 MPa) and a small loss of ductility (3%). Thermal ageing to the simulated pressurised water reactor (PWR) end-of-life thermal condition (1000 hrs at 400°C) had no significant effect on the fracture resistance behaviour or tensile properties of clad material in either irradiated or unirradiated condition. Clad fracture resistance properties were unaffected by orientation in the plane of the clad layer and were only dependent on test temperature.


    pressure vessel cladding, fracture resistance, neutron irradiation, tensile, thermal ageing, orientation, temperature

    Author Information:

    Horsten, MG
    Materials scientist, NRG, ZG Petten,

    Belcher, WPA
    Metallurgist, British Energy Generation Ltd., Barnwood, Gloucester,

    Committee/Subcommittee: E10.07

    DOI: 10.1520/STP10542S