小型铅-铋冷却快堆提棒事故核热耦合研究

Coupled Neutronics and Thermal-Hydraulics Simulation of RIA for Small LBE-Cooled Fast Reactor

基于热工程序COBRA-YT和物理程序SKRTCH-N,利用幵行虚拟机(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,幵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1.42s后达到最大值;5s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。

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 1.42 s 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.