A model to describe the change in the inelastic and fracture properties of reactor pressure vessel steels due to neutron irradiation in the ductile region (i.e., irradiation embrittlement) is developed. First, constitutive equations for unirradiated elastic-viscoplastic-damaged materials are developed within the framework of the irreversible thermodynamics theory. To take into account the effect of hydrostatic pressure on the nucleation and growth of microvoids, properly defined dissipation potential is used. Then, the effect of irradiation on the material behavior is incorporated into the proposed model as a function of neutron fluence Φ by taking into account the interaction between irradiation-induced defects and movable dislocations. As regards the damage strain threshold the mechanism of void nucleation due to pile-up of dislocations at the inclusions in the material is proposed first under unirradiated-condition, and then the effect of irradiation on the mechanism is formulated. In order 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°C. The resulting model can describe the increase in yield stress and ultimate tensile strength, the decrease in total elongation and strain hardening, and the strain rate dependence of yield stress due to neutron irradiation. [S0094-4289(00)00901-4]
Modeling of Irradiation Embrittlement of Reactor Pressure Vessel Steels
Contributed by the Materials Division for publication in the JOURNAL OF ENGINEERING MATERIALS AND TECHNOLOGY. Manuscript received by the Materials Division April 13, 1998; revised manuscript received July 9, 1999. Associate Technical Editor: S. K. Datta.
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Murakami , S., Miyazaki , A., and Mizuno , M. (July 9, 1999). "Modeling of Irradiation Embrittlement of Reactor Pressure Vessel Steels ." ASME. J. Eng. Mater. Technol. January 2000; 122(1): 60–66. https://doi.org/10.1115/1.482766
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