SCWR-FQT回路的冷却剂流量丧失事故研究  

Loss of Coolant Flow Accident Analysis for SCWR-FQT Loop

作　　者：周翀[1] 杨燕华[1] 

机构地区：[1]上海交通大学核科学与工程学院,上海200240

出　　处：《原子能科学技术》2013年第12期2238-2243,共6页Atomic Energy Science and Technology

摘　　要：超临界水冷堆燃料验证实验(SCWR-FQT)将对1个小型燃料组件在超临界水环境下进行堆内性能测试。为了对该实验回路进行系统设计和安全分析,应用修改过的ATHLET程序建立实验回路计算模型,对两种造成燃料组件实验段冷却剂流量部分或全部丧失的设计基准事故进行模拟分析,即由于装载实验段的压力管内部的导向管破裂导致流经实验段的冷却剂旁通和主冷却剂泵卡轴事故。计算结果显示:实验段冷却剂旁通事故中,燃料包壳温度在事故初期出现约920℃的峰值;而主泵卡轴事故中,燃料包壳温度未明显升高。计算结果表明,现有的安全系统设计能保证在事故情况下维持燃料组件实验段的有效冷却。The supercritical water-cooled reactor fuel qualification test （SCWR-FQT） intends to test a small scale fuel assembly under supercritical water environment in a research reactor. For the safety system design and safety analysis of the test loop, the modified ATHLET code was applied to model the loop, and simulate two design basis accidents, which caused partial or total loss of coolant flow in the fuel assembly test section. The calculated accidents were coolant bypassing the test section due to cracks in the internal structures of the pressure tube, and primary pump seizure. The calculation results show that, in the first accident the fuel cladding temperature shows a peak value of about 920 ℃ in the early stage of the transient, while in the second accident the fuel cladding temperature suffers no obvious increase. It is suggested by the results that the current safety system design is able to provide effective cooling to the test section in the two simulated accidents.

关 键 词：超临界水冷堆 ATHLET 冷却剂流量丧失事故 

分 类 号：TL333[核科学技术—核技术及应用]