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    Fracture Toughness of the Ni-Modified A302-B Plate of the BR3 Reactor Pressure Vessel


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    The present paper aims to compare regulatory predictions of fracture toughness to direct measurements using precracked Charpy specimens tested in three-point bending. Trepans were extracted from the vessel of the BR3 reactor in order to perform fracture toughness tests in addition to Charpy impact and tensile testing. This vessel was successfully wet-annealed at 343°C during 1 week after a maximum neutron fluence of 3.3 1019 n/cm2. The reactor is presently under decommissioning after a maximum neutron fluence of 4.0 1019 n/cm2. The trepans (base metal and weld) were taken from two regions: vessel mid-plane and nozzle elevation (no neutrons, negligible thermal aging). Here, the base metal (Ni-modified A-302 grade B) is investigated: Tensile, Charpy impact and fracture toughness tests were performed at thickness ranging from 1/4T to 1/2T.

    The determination of the reference temperature, T0, is performed according to the ASTM E1921 standard.

    It is found that the 41J-shift of the Charpy impact transition curve is similar to the T0-shift of the master curve: ΔT41J= 96°C and ΔT0= 115°C. This is an important result as current regulation is based on the 41J-shift. This regulation indexes the ASME fracture toughness curve to the initial RTNDT for the baseline condition and is increased by the 41J-energy shift for the irradiated condition. This results in an over-conservatism of about 40°C. It is found also that the T41J and T0 shifts fall within the regulatory predictive bounds based on the chemistry factor (Cu and Ni-content) and neutron fluence.


    fracture toughness, Charpy impact, tensile, transition, irradiation, Ni-modified A-302 grade B reactor pressure vessel steel, reference temperature

    Author Information:

    Chaouadi, R
    SCK•CEN, Reactor Materials Research, Mol,

    Fabry, A
    SCK•CEN, Reactor Materials Research, Mol,

    van Walle, E
    SCK•CEN, Reactor Materials Research, Mol,

    Van de Velde, J
    SCK•CEN, Reactor Materials Research, Mol,

    Committee/Subcommittee: E10.08

    DOI: 10.1520/STP12389S