原子炉圧力容器鋼の照射ぜい化機構とモデル化

概要

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A model to describe the change in the properties of the inelastic deformation and the fracture of reactor pressure vessel steels due to neutron irradiation, i.e., irradiation embrittlement, in the ductile region is developed. First, constitutive equations for unirradiated elastic-viscoplastic-damaged materials are developed within the irreversible thermodynamics theory. To take into account the effect of hydrostatic pressure on the growth of microvoids, suitable dissipation potential is used. Then, the effect of irradiation on the material behavior is incorporated into the proposed model as function of neutron fluence Φ taking into account the interaction of irradiation induced defects and movable dislocations. Especially for the effect on damage strain threshold p_D, the mechanism of void nucleation due to stress concentration responsible for pile-up of dislocations at the inclusions in the material is proposed under unirradiated condition, then the effect of irradiation on that mechanism is considered. To demonstrate the validity of this model, it is applied to the case of uniaxial tensile loading of a low alloy steel A533B cl. 1 for the pressure vessel use of light-water reactors at 260℃. The resulting model can describe the increase of yield stress, ultimate tensile strength and the decrease of total elongation, strain hardening and strain rate dependence of yield stress due to neutron irradiation.
1998-11-15