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    Phosphorus Segregation and Intergranular Embrittlement in Thermally Aged and Neutron Irradiated Reactor Pressure Vessel Steels

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    A study on phosphorus (P) segregation at grain boundaries and intergranular embrittlement of reactor pressure vessel steels due to thermal aging and neutron irradiation has been performed for A533B steel plates and weld metals doped with different bulk levels of P. The initial P concentration at grain boundaries of these materials determined by scanning Auger microscopy ranged from 9 to 33 %. The materials thermally aged at temperatures ranging from 450 to 550°C for up to 10000 h exhibited increases in P concentration described by the McLean's equation. A linear correlation was obtained between P concentration and Charpy 41 J transition temperature (T41J). Fracture toughness tests in the ductile-tobrittle transition temperature region have been conducted by the Master Curve method using 1T-CT specimens. The material subject to intergranular embrittlement exhibited the dual fracture toughness distributions typically observed in inhomogeneous materials. This was probably due to two types of initiation of cleavage and intergranular fracture. The lower tail analysis based on the Master Curve method enabled conservative fracture toughness estimates by fitting the lower fracture toughness arising from intergranular fracture. The relation between median KJc values estimated from the lower tail analysis and temperature appeared to follow the Master Curve. For neutron irradiated materials up to a fluence of 6.9×1019 n/cm2 (E>1 MeV) at 290°C by the Japan Materials Testing Reactor, an increase in grain boundary P segregation associated with decrease in carbon segregation was observed with increasing fluence. Relative significant hardening was recognized for the material with high bulk P content. Shifts in the reference temperature T0 and T41J were comparable for the thermally aged and irradiated materials.


    phosphorus segregation, grain boundary, thermal aging, neutron irradiation, embrittlement, fracture toughness, Master Curve, reactor pressure vessel

    Author Information:

    Nishiyama, Yutaka
    Japan Atomic Energy Agency, Ibaraki-Ken,

    Onizawa, Kunio
    Japan Atomic Energy Agency, Ibaraki-Ken,

    Suzuki, Masahide
    Japan Atomic Energy Agency, Ibaraki-Ken,

    Committee/Subcommittee: E10.07

    DOI: 10.1520/STP46566S