蒸汽发生器传热管破裂事故缓解措施研究  

Study on Steam Generator Tube Rupture Mitigation

作　　者：徐苗苗 乐志东 林支康 沈永刚 Xu Miaomiao;Le Zhidong;Lin Zhikang;Shen Yonggang(China Nuclear Power Technology Research Institute Co.,Ltd,Shenzhen,518026)

机构地区：[1]中广核研究院有限公司反应堆工程设计与安全研究中心,深圳518026

出　　处：《核安全》2022年第2期62-67,共6页Nuclear Safety

摘　　要：为降低蒸汽发生器传热管破裂(SGTR)事故放射性后果,本文以华龙一号堆型为对象,分析总结了SGTR事故缓解的特点,使用成熟工程程序对SGTR事故缓解进行了优化分析。分析表明,根据放射性报警信号识别SGTR事故,并在停堆时隔离破损SG,可以显著降低破损SG的蒸汽排放量和破口流量,是降低SGTR事故放射性后果的可行方向。In order to reduce the radiological consequences induced by Steam Generator Tube Rupture(SGTR), the characteristics in SGTR mitigation is studied for Hua-long Pressurized Reactor(HPR1000) and engineering used code is used to analyze the effects of potential improvements. The analyses show that isolation of affected SG on reactor trip and high activity can largely reduce the contaminated steam release and the break flow. And it is a practicable way to optimize SGTR mitigation process.

关 键 词：SGTR 事故缓解 放射性后果 

分 类 号：TL364[核科学技术—核技术及应用]