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

被引：0
|
作者：
Yang D. [1 ]
Liu X. [1 ]
Zhang T. [1 ]
Cheng X. [1 ]
机构：
[1] Shanghai Jiaotong University, Shanghai
来源：
Hedongli Gongcheng/Nuclear Power Engineering | 2019年 / 40卷 / 02期
关键词：
Control rod withdrawal accident; Coupled code development; Lead-bismuth eutectic cooled fast reactor; Thermal-hydraulics code development;
D O I：
10.13832/j.jnpe.2019.02.0184
中图分类号：
学科分类号：
摘要：
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 © 2019, Editorial Board of Journal of Nuclear Power Engineering. All right reserved.
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页码：184 / 188
页数：4
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