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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。
Using the standard k-? Two-equation and finite-volume method, the flow field, temperature field and core enthalpy-increasing characteristics of a 1 000 MW nuclear reactor are modeled. Combining the core measuring instruments with CLP and IN- CORE software carried out the actual reactor physical experiment; the simulation test and the actual physical test conducted a coupling study. The effect of flow rate deviation on the enthalpy-increasing characteristics of reactor core under normal operation and de-flow conditions was studied. The deviation of actual operation flow was analyzed. Simulation and physical test results show that the core flow distribution plate and the lower grid plate can well adjust the coolant flow deviation into the core, and the inner temperature field of the core has strong self-adjusting function of flow deviation. When the loop flow decreases, the changes of reactor flow field, temperature field and core enthalpy change obviously; the loop flow does not change obviously when the flow increases slightly; when the loop flow increase is less than 5%, the internal flow field and temperature No significant changes in field deviations. The experimental data provided theoretical data for the safe operation of the reactor.