In PWR three-dimensional transient coupling calculation software CORCA-K, the nodal Green's functionmethod and diagonal implicit Runge Kutta method are used to solve the spatiotemporal neutron dynamic diffusion equation, and the single-phase closed channel model and one-dimensional cylindricalheat conduction transient model are used to calculate the coolant temperature and fuel temperature. TheLMW, NEACRP and PWR MOX/UO2 benchmarks and FangJiaShan (FJS) nuclear power plant (NPP) transient control rod move cases are used to verify the CORCA-K. The effects of burnup, fuel effective temperature and ejection rate on the control rod ejection process of PWR are analyzed. The conclusions areas follows: (1) core relative power and fuel Doppler temperature are in good agreement with the resultsof benchmark and ADPRES, and the deviation between with the reference results is within 3.0% in LMWand NEACRP benchmarks; 2) the variation trend of FJS NPP core transient parameters is consistent withthe results of SMART and ADPRES. And the core relative power is in better agreement with the SMARTwhen weighting coefficient is 0.7. Compared with SMART, the maximum deviation is 5.08% in the rodejection condition and while 5.09% in the control rod complex movement condition