AP1000核电厂主给水管道断裂事故瞬态特性分析
Transient Characteristics of Main Feedwater Line Rupture Accident for AP1000 Nuclear Power Plant
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摘要: AP1000是目前国际上典型的“三代”非能动核电厂，基于最佳估算程序RELAP5/MOD3.3，对AP1000核电厂系统进行了详细的建模分析，获得了主给水管道断裂事故下AP1000核电厂关键参数的瞬态特性和非能动系统响应特性。结果表明，事故过程中一、二回路的压力和温度呈现波动变化，一回路压力最大值为17.13 MPa，低于设计压力的91%，主蒸汽系统的压力也低于设计值的91%，满足验收准则的要求。
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关键词:
- RELAP5/MOD3.3程序 ,
- AP1000 ,
- 主给水管道断裂事故 ,
- 非能动核电厂
Abstract: The AP1000 is the typical “third generation” passive nuclear power plant in the world at present. The primary system of AP1000 nuclear power plant was modeled using RELAP5/MOD3.3 code, and the transient thermal-hydraulic characteristics and the response characteristics of passive system were analyzed under the accident sequence of main feedwater line rupture accident. The results show that during the accident, the primary loop pressure, secondary loop pressure and primary loop temperature are fluctuant. The RCS pressure maximum value is 17.13 MPa, less than 91% of the design pressure. The pressure of the main steam system is also less than 91% of the designed value, which satisfies acceptance criteria. -
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