Research on Dynamic Characteristics of PWR Nuclear Power Unit under Variable Power Loads Based on APROS
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摘要: 利用仿真软件APROS，基于模块化建模方法，搭建了包含一、二回路主要设备的百万千瓦级压水堆核电机组动态仿真模型，并进行了稳态工况与动态过程的仿真验证。在此基础上，研究了不同速率的线性降负荷与不同幅度的阶跃降负荷下核电机组主要参数的动态变化。结果表明：阶跃降负荷幅度小于等于2%满功率（FP）时，一回路平均温度波动小，不能引起控制棒的动作；当阶跃降负荷幅度大于2%FP且小于等于5%FP时，回路平均温度波动引起控制棒动作但很快回到温度死区，最终稳定的回路平均温度反而高于初始温度；负荷线性变化过程中稳压器压力波动最大可达到0.3 MPa；由于冷却剂比容与温度呈正相关，稳压器相对水位变化与回路平均温度变化趋势基本一致。本研究旨在为压水堆核电厂灵活运行提供理论参考。Abstract: A dynamic model of the million kilowatt PWR unit including the main equipment of the primary and secondary circuits was built based on the modular modeling method by using the simulation software APROS. The steady-state and dynamic processes were simulated and verified. The dynamic characteristics of the main parameters of the system under the different rates of linear load shedding and different step load shedding were studied. The results show that when the step load reduction is less than or equal to 2% full power (FP), the average temperature (Tavg) fluctuation in the primary circuit is small, which cannot cause the action of control bar; when the step load reduction is greater than 2%FP and less than or equal to 5%FP, the Tavg fluctuation causes the control bar to act but quickly returns to the temperature dead zone. Eventually, the stabilized Tavg ended up higher than the initial temperature. In the process of linear load change, the maximum pressure change of the pressurizer can reach 0.3 MPa. Since the specific volume of coolant is positively related to the temperature, the changing trend of the pressurizer water level is consistent with that of the average temperature of the loop. The purpose of this study is to provide theoretical reference for the flexible operation of PWR nuclear power plant.
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Key words:
- PWR /
- Dynamic characteristics /
- System modeling /
- Reactor power control
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图 1 核电机组系统流程简图
1—工质流动方向，表示水从汽水分离再热器出来后，进入高压加热器；1#、2#—高压加热器；3#、4#、5#、6#—低压加热器
Figure 1. Flow Chart of Nuclear Power Unit System
图 5 线性降负荷下机组主要参数变化
Figure 5. Change of Main Parameters of the Unit under Linear Load Reduction
图 6 阶跃降负荷下机组主要参数变化
Figure 6. Change of Main Parameters of the Unit under Step Load Reduction
图 7 不同线性降负荷速率下的电功率变化
Figure 7. Electric Power Change of Circuit under Different Linear Load Reduction Rates
图 8 不同线性降负荷速率下的回路平均温度变化
Figure 8. Average Temperature Change of Circuit under Different Linear Load Reduction Rates
图 9 不同线性降负荷速率下的稳压器相对水位变化
Figure 9. Pressurizer Water Level Change under Different Linear Load Reduction Rates
图 10 不同线性降负荷速率下的稳压器压力变化
Figure 10. Pressure Change of Pressurizer under Different Linear Load Reduction Rates
图 11 不同阶跃降负荷幅度下的电功率变化
Figure 11. Electrical Power Change under Different Step Load Shedding Amplitudes
图 12 不同阶跃降负荷幅度下的回路平均温度变化
Figure 12. Average Temperature Change of Circuit under Different Step Load Shedding Amplitudes
图 13 不同阶跃降负荷幅度下的稳压器相对水位变化
Figure 13. Pressurizer Water Level Change under Different Step Load Shedding Amplitudes
图 14 不同阶跃降负荷幅度下的稳压器压力变化
Figure 14. Pressurizer Pressure Change under Different Step Load Shedding Amplitudes
表 1 某核电机组设计工况的主要热力参数
Table 1. Main Thermal Parameters of Design Conditions of Nuclear Power Unit
热力参数 参数值 堆功率/MW 3400 汽轮机功率/MW 1251 主泵热功率/MW 15 回路平均温度/℃ 300.9 主蒸汽压力/MPa 5.53 额定蒸发量/(kg·s−1) 1888.6 反应堆冷却剂系统（RCS）热工最佳估算流量/(kg·s−1) 15170 
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表 2 稳态参数理论值与模拟值比较
Table 2. Comparison Between Theoretical and Simulated Values of Steady-State Parameters
参数 100%FP工况 90%FP工况 设计值 模拟值 误差/% 设计值 模拟值 误差/% 堆功率/MW 3400 3422 0.65 3113 3134.26 0.68 热管段平均温度/℃ 321.1 321.24 0.04 318.06 319.3 0.39 主蒸汽压力/MPa 5.53 5.62 1.63 5.59 5.89 5.37 主蒸汽温焓/(kJ·kg−1) 2783.7 2787.9 0.15 2783.6 2785.2 0.06 再热蒸汽压力/MPa 0.932 0.937 0.54 0.839 0.858 2.26 再热蒸汽焓/(kJ·kg−1) 2961 2964 0.10 2969.4 2971 0.05 凝汽器工作压力/kPa 4.40 4.42 0.45 4.40 4.41 0.23 主给水焓/(kJ·kg−1) 975.7 976.34 0.07 949.5 953.6 0.43
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