In PWR three-dimensional transient coupling calculation software CORCA-K, the nodal Green's function method 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 cylindrical heat conduction transient model are used to calculate the coolant temperature and fuel temperature. The LMW, 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 are as follows: (1) core relative power and fuel Doppler temperature are in good agreement with the results of benchmark and ADPRES, and the deviation between with the reference results is within 3.0% in LMW and NEACRP benchmarks; 2) the variation trend of FJS NPP core transient parameters is consistent with the results of SMART and ADPRES. And the core relative power is in better agreement with the SMART when weighting coefficient is 0.7. Compared with SMART, the maximum deviation is −5.08% in the rod ejection condition and while −5.09% in the control rod complex movement condition.

在压水堆三维瞬态耦合计算软件CORCA-K中，采用节点格林函数法和对角隐式龙格库塔法求解时空中子动力扩散方程，单相封闭通道模型和一维圆柱热传导瞬态模型用于计算冷却液温度和燃油温度。LMW、NEACRP 和 PWR MOX/UO ₂基准和方家山（FJS）核电站（NPP）瞬态控制棒移动案例被用来验证CORCA-K。分析了燃耗、燃料有效温度和弹射率对压水堆控制棒弹射过程的影响。结论如下： (1)堆芯相对功率和燃料多普勒温度与基准和ADPRES的结果吻合较好，LMW和NEACRP基准与参考结果的偏差在3.0%以内；2) FJS核电厂堆芯暂态参数变化趋势与SMART和ADPRES结果一致。当加权系数为0.7时，核心相对功率与SMART的一致性较好。与SMART相比，在棒弹射条件下最大偏差为-5.08%，而在控制棒复杂运动条件下为-5.09%。