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典型严重事故非能动安全壳冷却系统效果分析

Assessment of Passive Containment Cooling System Performance During Severe Accident

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【作者】 邹杰佟立丽曹学武

【Author】 ZOU Jie;TONG Li-li;CAO Xue-wu;School of Mechanical Engineering,Shanghai Jiao Tong University;

【机构】 上海交通大学机械与动力工程学院

【摘要】 先进压水堆采用非能动安全壳冷却系统(PCCS)在事故下维持安全壳完整性,包括重力喷洒形成安全壳外部水膜冷却和空气冷却流道中空气对流传热。针对严重事故下PCCS效果研究,建立了非能动压水堆安全壳及非能动安全壳冷却系统的传热分析模型(包括对流传热及蒸发/冷凝传热),并耦合反应堆主系统模型及专设安全设施模型。通过与西屋公司PCCS大尺度试验结果的比对验证了模型的可用性,进而针对非能动先进压水堆选取全厂断电、热段小破口失水始发事故作为典型严重事故序列,模拟了事故进程、主系统响应及安全壳的响应,分析了PCCS对安全壳的降温、降压作用。结果表明,安全壳压力72h内未超过安全限值,保持安全壳完整性。

【Abstract】 Advanced passive PWR adopts passive containment cooling system(PCCS)to maintain containment integrity during postulated accident,including water film cooling induced by gravity water spray and air conviction heat transfer in air cooling channels.The performance of PCCS during severe accident for advanced passive PWR was assessed using integrated safety analysis code.The analysis model including heat transfer models(conviction heat transfer and condensation/evaporation heat transfer)of containment and PCCS,coupled with primary system model and engineered safety features model was built.These models were validated against the Westinghouse PCCS largescale test results.The station blackout accident and hot leg small LOCA were selected.The accident process,primary system response,and containment response were calculated.PCCS performance was evaluated by containment pressure and temperaturechange under PCCS cooling.The result shows that containment pressure is controlled under design limit.Containment integrity is maintained in 72 h with PCCS water cooling.

【基金】 国家自然科学基金资助项目(11205099)
  • 【会议录名称】 北京核学会第十届(2014年)核应用技术学术交流会论文集
  • 【会议名称】北京核学会第十届(2014年)核应用技术学术交流会
  • 【会议时间】2014-11-27
  • 【会议地点】中国湖北武汉
  • 【分类号】TL364.3
  • 【主办单位】北京核学会、《原子能科学技术》编辑部
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