| 小型铅-铋冷却快堆提棒事故核热耦合研究 |
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| 引用本文： | 杨冬梅,刘晓晶,张滕飞,程旭.小型铅-铋冷却快堆提棒事故核热耦合研究[J].核动力工程,2019,40(2):184-188. |
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| 作者姓名： | 杨冬梅 刘晓晶 张滕飞 程旭 |
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| 作者单位： | 上海交通大学,上海,200240;上海交通大学,上海,200240;上海交通大学,上海,200240;上海交通大学,上海,200240 |
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| 摘 要： | 基于热工程序COBRA-YT和物理程序SKRTCH-N，利用并行虚拟机（PVM）平台开发了核热耦合工具:COBRA-YT将冷却剂密度和燃料温度等热工参数传递给物理程序，用以更新截面；SKETCH-N执行物理计算，并将功率分布反馈给热工程序；最后，应用该耦合程序分析铅-铋冷却快堆的提棒事故。计算结果显示控制棒提起后，功率迅速升高，在1.42?s后达到最大值；5?s后包壳温度达到峰值1264℃，超出了设计限值。结果表明:在提棒事故后，均一化布置堆芯的安全会在极短时间内受到严重威胁，故该堆芯应采用分区布置。
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| 关 键 词： | 铅-铋冷却快堆 热工程序开发 耦合程序开发 提棒事故 |
Coupled Neutronics and Thermal-Hydraulics Simulation of RIA for Small LBE-Cooled Fast Reactor |
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| Abstract: | The coupled tool based on neutronics code SKETCH-N and thermal-hydraulics code COBRA-YT has been developed via Parallel Virtual Machine?(PVM) software platform. COBRA-YT code performs the thermal-hydraulics calculation and transfers its results such as coolant density and fuel temperature to the neutronics code SKETCH-N to update the cross-section; then SKETCH-N carries out the neutron-physical simulation of the reactor and provides the power density to the thermal-hydraulics code COBRA-YT as boundary conditions. Finally, this coupled code platform is used in?the lead-bismuth fast reactor design to simulate some transient and control rod withdrawal accidents. The reactor power increases rapidly and reaches the peak at 1.42s after the control rod withdrawal. Meanwhile, the cladding temperature reaches the maximum 1264℃, exceeding its design limit. The results achieved so far indicates that?the?control rod withdrawal accident?poses a threat to the core with the same enrichment,?and the optimization work on the core zoning scheme should be done. |
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