Numerical Simulation Research on Natural Circulation Flow of the Reactor Coolant System
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摘要: 采用计算流体动力学（CFD）程序建立了包含反应堆、蒸汽发生器、主泵和主管道在内的三环路反应堆冷却剂系统的高保真三维数值模型，开展了低功率运行工况下系统级热工水力现象的三维数值分析，获得了不同区域的冷却剂温度，并与核电厂实测数据对比，验证了数值模型的合理性。分析结果表明：该功率水平下的自然循环流量为满功率运行流量的4.5%，堆芯出口温度稳定，可以有效导出堆芯热量；局部热对流现象使不同环路的冷却剂产生更充分搅混；顶盖腔室内存在热分层现象，现有的顶盖温度测点读数不是该区域内的最高温度；主泵出口产生旋转流，并且靠近主管道管壁区域切向速度较大，中心区域形成局部对流。该研究工作可以进一步提升设计者对核电厂复杂系统级三维热工水力现象的认识。
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关键词:
- 计算流体动力学（CFD） /
- 反应堆冷却剂系统 /
- 自然循环
Abstract: Computational Fluid Dynamics(CFD) program is employed to enable the high-fidelity modeling of the reactor coolant system (RCS) for a typical three-loop pressurized water reactor, and the completed model of the RCS is build including reactor vessel and internals, core, steam generator, primary pump and linking pipe. The three-dimensional, global and localized flow features have been investigated under natural circulation flow condition with lower core thermal power, and the temperature at different locations are compared with the measured values from the operating nuclear power plant in order to verify the accurate description of the developed CFD model. The results show that the natural circulation flow rate is about 4.5% of the full power flow rate while the temperature of the core outlet is stable, and the residual core heat could be effectively removed. The phenomenon of the thermal stratification in the reactor pressure vessel head dome shows that the measured temperature value of the detector position in nuclear power plant could not provide the highest value. The coolant from different loops could be more fully mixed due to the local convection flow. There is a swirling flow at the outlet of primary pump, and the tangential velocity near the pipe wall is large while the local convection occurs at the central area. This analysis practice provides an effective evaluation for the system-level three-dimensional thermal hydraulics phenomena of the reactor coolant system. -
图 3 不同网格数量下的回路流量和顶盖旁流份额变化
Figure 3. Comparison of Loop Flow and Head bypass Percentage for Different Mesh
表 1 不同区域温度的计算值和实测值对比
Table 1. Temperature Comparison between CFD Values and Measured Values
参数名 冷管段
温度/℃热管段
温度/℃反应堆压力容器顶盖腔室
温度/℃堆芯出口
温度/℃实测值 294.8 295.6 294 296 计算值 294.7 296 294.8 296.1 相对偏差/% −0.03 0.14 0.27 0.03 
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表 2 不同区域温度的计算值和实测值对比
Table 2. Temperature Comparison between CFD Values and Measured Values
参数名 冷管段
温度/℃热管段
温度/℃反应堆压力容器顶盖腔室
温度/℃堆芯出口
温度/℃实测值 293 317.5 312 317.5 计算值 296.2 316.7 305.5 317.4 相对偏差/% 1.1 −0.25 −2.08 −0.03
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