环路流量偏差对核电1000MW机组压水反应堆焓增特性影响的模拟研究

Simulation Study of Enthalpy Gain Characteristics on Nuclear Power 1000 MW Unit Pressurized Water Reactor Loop Flow Deviation

使用标准的k-?双方程、有限体积离散法,对一台1 000 MW核电机组反应堆流场、温度场及堆芯焓增特性进行了建模研究;结合堆芯测量仪器及CLP和IN-CORE软件进行了反应堆实际物理试验;将模拟试验和实际物理试验进行了耦合研究。研究了反应堆正常运行及失流等工况下流量偏差对反应堆堆芯焓增特性的影响,并对实际运行流量偏差工况进行了分析。模拟和物理试验结果表明:堆芯流量分配板和下栅格板能够很好地调整进入堆芯的冷却剂流量偏差,堆芯内部温度场具有较强的流量偏差自调节功能。环路流量减少时,反应堆流场、温度场及堆芯焓增特性变化比较明显;环路流量稍有增加时变化不明显,环路流量增加值不大于5%时,反应堆内部流场及温度场偏差无明显变化。试验数据为反应堆的安全运行提供了理论数据。

The flow field, temperature field and enthalpy gain of a 1 000 MW unit nuclear reactor were numerically studied with standard k-？ model and finite volume discretization method. The operation physical tests were done combining the core measuring instrument with CLP and IN-CORE software, then the simulation and physical tests were coupling studied. The flow deviation influence of normal and loss of flow conditions on the core enthalpy gain characteristic has been done, and the flow deviation under operating condition was analyzed. Simulation and physical test results show that, the core flow distribution plate and the under grid plate can adjust the coolant flow deviation into the reactor core effectively. The internal core temperature field can adjust the flow deviation single. The core flow field, temperature field and core enthalpy gain characteristic change obviously when the loop flow reduces, which change not obvious when the loop flow increases slightly. The core flow field and temperature field almost do not change when the loop increase flow is less than 5 percent. The experimental data provides theoretical data for the safe operation of the reactor.