直接注入管线失水事故非能动安全系统运行特性研究  

Experimental Study on DVI Line Break LOCA with Passive Safety System

作　　者：黄志刚[1] 张妍[1] 鲁晓东[1] 彭传新[1] 昝元锋 卓文彬[1] 闫晓[1] HUANG Zhigang;ZHANG Yan;LU Xiaodong;PENG Chuanxin;ZAN Yuanfeng;ZHUO Wenbin;YAN Xiao(CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology,Nuclear Power Institute of China,Chengdu 610213,China)

机构地区：[1]中国核动力研究设计院中核核反应堆热工水力技术重点实验室,四川成都610213

出　　处：《原子能科学技术》2021年第11期2021-2027,共7页Atomic Energy Science and Technology

摘　　要：小型模块式反应堆ACP100采用了非能动安全和模块化设计技术,可用于地区集中供暖、海水淡化和核动力商船等多个方面。其中,非能动安全设计主要包括非能动应急堆芯冷却系统、非能动余热排出系统等非能动安全系统和自动卸压等专设措施。针对ACP100非能动安全设计技术特点,在中国核动力研究设计院非能动安全系统综合性能缩比试验装置上开展了大量失水事故系统特性试验研究,根据试验数据分析,获得了非能动安全系统在直接注入管线发生破口后系统的综合响应特性,掌握了系统间的相互影响规律,并初步评估其对堆芯的冷却效果。Small modular reactor ACP100 adopts passive safety system and modular design technology and can be applied in district heating,seawater desalination and nuclear power ship,etc.The passive safety design mainly includes automatic rapid depressurization system,passive emergency core cooling system and residual heat removal system.For validation and verification of design features of the ACP100,a series of LOCA experiments were finished on a scaled integral test facility in Nuclear Power Institute of China.The system comprehensive response characteristics under DVI line break condition were obtained based on the experimental data.The influential patterns between systems were obtained and the core cooling effect was preliminarily evaluated.

关 键 词：ACP100非能动安全系统 直接注入管线 失水事故 

分 类 号：TK24[动力工程及工程热物理—动力机械及工程]