原子能科学技术2024，Vol.58Issue(z2)：16-32,17.DOI:10.7538/yzk.2024.youxian.0127
池式铅冷快堆SGTR事故多组分多相流动过程数值模拟研究
Numerical Investigation on Multi-phase Flow of Pool-type Lead-cooled Fast Reactor under SGTR Accident
摘要
Abstract
Steam generator tube rupture(SGTR)accident is one of the accident scenarios that must be considered during the design and safety analysis process of lead-cooled fast reactors(LFR).Focused on this topic,numerous experiments and theoretical research have been conducted in recent decades,which could be generally categorized into two aspects:on the one hand,separation effect experiments and theoretical analysis focusing on the specific phenomena of lead-water interaction process;on the other hand,engineering test facilities and numerical simulations focusing on the overall process of LFR SGTR accident,mainly concentrated on the evolution of macro parameters such as system pressure and lead pool temperature during SGTR accident.In this paper,firstly,the basic mathematical and physical models of ACENA code was introduced;then the code was validated against the LBE-N2 two-phase flow experiment HESTIA-2,the KYILN-Ⅱ-S LBE-H2O interaction experiment and the analytical solution of the point-kinetic neutronic equations;after that the Eulerian multi-phase analysis code ACENA was applied to simulate the postulated unprotected SGTR process of European pool-type lead-cooled fast reactor ALFRED.Attention was paid to the simulated migration process of steam bubbles in the lead pool as well as the evolution of safety parameters such as maximum cladding temperature,fuel centerline temperature,relative nuclear power and reactor vessel pressure.Furthermore,the sensitivity of ruptured tube quantities,lead coolant circulation paths and neutron dynamics parameters on calculation results were analyzed.This paper's study demonstrates that the Ishii-Chawla-Suzuki interfacial drag coefficient model could accurately predict the diffusion and migration characteristics of rising bubbles in circular/annular LBE flow channels when combined with the interfacial area concentration transport model proposed by Ishii et al;The simulation of KYLIN-Ⅱ-S experiment shows that ACENA code could reasonably reproduce phenomena such as system pressure fluctuations and temperature transients of liquid lead during the interaction process between LBE and water;the calculated results of ALFRED reactor's key thermal-hydraulic parameters under hot full power condition by ACENA code are generally consistent with results obtained by internationally recognized one-dimensional system code TRACE/FRED,proving the reliability of ACENA code when applied to full reactor system level;The calculation of the hypothetical SGTR accident of the ALFRED reactor verifies the ACENA code's capability of simulating the complex multi-component multi-phase flow,and the calculation results reveal that designing the primary coolant flow path rationally and reducing the number of ruptured pipes are of great significance to mitigating the potential consequences of LFR SGTR accidents.This work can provide reference for the safety analysis work of China's pool-type lead-cooled fast reactors.关键词
铅冷快堆/ALFRED/蒸汽发生器传热管破裂/ACENA程序/多相流动/机理模型/安全分析Key words
lead-cooled fast reactor/ALFRED/steam generator tube rupture/ACENA code/multi-phase flow/mechanism model/safety analysis分类
能源科技引用本文复制引用
陈宇彤,张大林,林悦,张熙司,田文喜,秋穗正,苏光辉..池式铅冷快堆SGTR事故多组分多相流动过程数值模拟研究[J].原子能科学技术,2024,58(z2):16-32,17.基金项目
国家自然科学基金(12075184) （12075184）
中核集团"青年英才"项目 （）
