摘要
AP/CAP系列核电厂设计了安全壳非能动冷却系统(PCCS),可以实现事故后72 h内对安全壳非能动冷却。但是,72 h后如顶部水箱不能及时补水,仅靠安全壳自身的散热能力很难将全部的余热带走,安全壳仍有超压风险。针对目前核电厂安全壳余热导出能力有限时长的短板,对一套创新的安全壳内热量非能动导出系统搭建试验台架以验证其载热性能。在设计基准事故(DBA)条件下,开展安全壳内不同压力、温度和气体组分条件下系统载热性能的试验研究。结果表明,DBA条件下该套系统的载热能力完全满足设计要求。本文进一步给出了适用于低过冷度条件的含不凝性气体管外冷凝换热系数关联式。
There is a passive containment cooling system (PCCS) in the AP / CAP series nuclear power plants, to cool the containment passively within 72 hours after accidents. However, after 72 hours, if the top tank is not able to replenish water in time, it is difficult to take all the residual heat away by the containment itself, and there exists potential overpressure for the containment. To solve the problem of the time limit in the residual heat removal of the containment in nuclear power plants, an experiment system for an innovative passive residual heat removal system for the containment is built, to study the heat transfer performance under different pressure, temperature and gas composition under design basis accident (DBA) conditions. The results show that the heat capacity of the system completely meets the design. Further, the correlation formula of the external tube condensation heat transfer coefficient with non-condensable gas for low subcooling condition is given.
作者
孟现珂
费立凯
高彬
张圣君
何丹丹
Meng Xianke;Fei Likai;Gao Bin;Zhang Shengjun;He Dandan(State Nuclear Power Technology R&D Center,Future S&T City,Beijing,102209,China)
出处
《核动力工程》
EI
CAS
CSCD
北大核心
2019年第4期39-43,共5页
Nuclear Power Engineering
基金
国家重大科技专项课题(2017ZX06004002-007-002)
