In nuclear reactors, stress corrosion cracking (SCC) behavior in primary water directly restricts the service life and structural reliability of the equipment, therefore, which has attracted much attention of many scholars and engineers. In order to study the rate of the stress corrosion crack propagation of alloy 690 weld joints, the experimental and numerical methods were presented in this paper. The results of this study have a significance to improve the reliability of nuclear reactor. The crack growth rate tests using half compact-tension specimens, which was applied to obtain the crack growth rate of alloy 690 in a dissimilar welding (SA508/Alloy690/316L), were conducted in 325℃ water, with different chloride concentrations, oxygen concentrations, cold deformations and stress intensity factors. The tests were composed of three stages, i.e., pre-existing crack in air, fatigue crack and environment crack in water. The experimental results showed that the cold working induced a comparatively high crack growth rate, even up to 10-8mm/s. In addition, the combined effect of dissolved chloride and oxygen in primary water, the cold deformation and the stress intensity factor would further accelerate the stress corrosion cracking growth rates. Besides, a semi-phenomenological numerical method for predicting the stress corrosion crack growth rates was generated to fill the gap between engineering demands and what current experimentation can reach out to. The key idea here is to calibrate the parameters of a partial differential equation model presumably to work during the whole stress corrosion cracking life span against experimentally obtainable information, and then to make predictions over stress corrosion cracking indices that are not experimentally reachable. Additionally, several representative examples were provided to verify the accuracy and validity of this method.