Analysis of Impact of Medium Pressure Safety Injection on Severe Accident Process under RRA Connected Mode
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摘要: 针对中国改进型百万千瓦级压水堆（CPR1000）核电机组在中间停堆反应堆余热排出系统（RRA）连接模式下失去高低压安注和喷淋的冷却剂丧失事故（LOCA），采用MAAP5程序对参考机组的反应堆堆芯、反应堆冷却剂系统以及安全壳系统进行模拟计算，同时结合计算结果分析中压安注系统对该严重事故序列进程的影响，并研究其对事故的缓解作用。分析结果表明，在RRA连接模式下出现LOCA导致的堆芯裸露和升温过程中，中压安注的及时注入能有效地限制堆芯的升温行为，并可对严重事故进程起到重要的缓解作用，甚至为事故工况下失去高低压安注和喷淋时避免堆芯完整性遭到破坏提供可能。最后，根据分析结果针对现行核电机组的运行规程提出改进建议：对于中压安注箱的行政隔离行为，只对其电气开关做相应的隔离操作，而对安全壳厂房内的阀门就地部分做挂牌警示，不做现场挂锁的操作，这样不仅可避免在正常运行工况下中压安注箱误注入行为的发生，同时能够在RRA连接模式下发生LOCA时有效地保障堆芯的完整性，在保证电厂正常安全运行的同时，提高了机组在该模式下发生严重事故的缓解能力。
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关键词:
- 严重事故 /
- 中压安注 /
- 反应堆余热排出系统（RRA）连接模式 /
- MAAP5 /
- 冷却剂丧失事故(LOCA)
Abstract: Aiming at the loss of coolant accident (LOCA) of China's improved million kilowatt PWR (CPR1000) Nuclear Power Unit losing high and low pressure safety injection and spray in reactor residual heat removal system (RRA) connected mode of intermediate shutdown, the reactor core, reactor coolant system and containment system of the reference unit are simulated and calculated by using MAAP5 program, at the same time the impact of medium pressure safety injection on the severe accident process is analyzed by combining with the calculation results, and the mitigation effect on the accident is studied. The results show that the timely injection of medium pressure safety injection can effectively limit the temperature rising behavior of the core, and play an important role in mitigating the process of severe accidents during the process of core exposure and temperature rise caused by LOCA under RRA connected mode, and even it is possible to avoid core integrity damage when high and low pressure safety injection and spray is lost under accident conditions. Finally, according to the analysis results, the improvement suggestions are put forward for the current operation regulations of nuclear power units. For the administrative isolation behavior of the medium pressure safety injection tank, only the corresponding isolation operation is performed for its electrical switch, while the local part of the valves in the containment building is provided with a tag warning, and the on-site padlock operation is not performed, which can not only avoid the misinjection of the medium pressure safety injection tank under normal operation conditions, the integrity of the reactor core can be also effectively guaranteed after the occurrence of LOCA under RRA connected mode. The normal and safe operation of the power plant can be ensured, and the mitigation ability of severe accidents of the unit under this mode can be improved. -
图 1 堆芯燃料（7环）模拟俯视图
a—径向第1环节点；b—径向第2环节点；c—径向第3环节点；d—径向第4环节点；e—径向第5环节点；f—径向第6环节点；g—径向第7环节点，下同
Figure 1. Top View of Core Fuel (7 Ring) Simulation
图 3 反应堆冷却剂系统控制体模拟图
Figure 3. Control Volume Simulation Diagram of Reactor Coolant System
图 5 安全壳控制体模拟图
1—反应堆堆坑；2—安全壳下部空间；3—安全壳中部空间；4—安全壳上部空间；5—稳压器卸压隔间；6—安全壳环廊；7—稳压器隔间；8—环路1蒸汽发生器隔间；9—辅助厂房隔间；10—环路1主泵隔间；11—环路2蒸汽发生器隔间；12—控制室隔间；13—环路2主泵隔间；14—环路3蒸汽发生器隔间；15—换料/乏燃料水池隔间；16—环路3主泵隔间
Figure 5. Simulation Diagram of Containment Control Volume
图 6 堆芯水位变化（无中压安注）
Figure 6. Changes in Core Water Level without Medium Pressure Safety Injection
图 7 燃料最高温度（无中压安注）
Figure 7. Maximum Temperature of Fuel without Medium Pressure Safety Injection
图 8 包壳最高温度（无中压安注）
Figure 8. Maximum Temperature of Cladding without Medium Pessure Safety Injection
图 9 堆芯水位变化（有中压安注）
Figure 9. Changes in Core Water Level with Medium Pressure Safety Injection
图 10 燃料最高温度（有中压安注）
Figure 10. Maximum Temperature of Fuel with Medium Pressure Safety Injection
图 11 包壳最高温度（有中压安注）
Figure 11. Maximum Temperature of Cladding with Medium Pressure Safety Injection
表 1 事故主要参数及假设
Table 1. Main Parameters and Assumptions of Accident
参数 假设条件 初始功率 冷停堆，衰变功率36 MW 一回路压力 RRA连接模式运行压力，2.6 MPa.a（绝对压力） 一回路平均温度 RRA连接模式运行温度，约170℃ 破口位置 一环路热段破口 破口大小 100%（大破口，当量直径0.64 m） 堆芯冷却措施 无（高、低压安注不可用） 专设安全设施 中压安注箱行政隔离 
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表 2 安注箱相关行政隔离设备清单
Table 2. List of Administrative Isolation Equipment Related to Safety Injection Tank
标志 设备名称 状态 位置 RIS001VP RIS001BA隔离阀 关闭上锁（LC） R111 RIS002VP RIS002BA隔离阀 LC R121 RIS003VP RIS003BA隔离阀 LC R131 LLE 308 RIS001VP电源开关 断开上锁（LD） L406 LLE 306 RIS002VP电源开关 LD L406 LLD 501 RIS003VP电源开关 LD L408
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[1] 国家核安全局. 核动力厂设计安全规定: HAF 102—2004[S]. 北京: 国家核安全局, 2004. [2] LUTZ ROBERT J. Westinghouse owners group core damage assessment guidance: CAP-14696-A[R]. Pittsburgh: Westinghouse Electric Corporation, 1999. [3] IAEA. Generic assessment procedures for determining protective actions during a reactor accident: IAEA-TECDOC-955[R]. Vienna: IAEA, 1997. [4] 苏光辉. 轻水堆核电厂严重事故现象学[M]. 北京: 国防工业出版社, 2016. [5] 佟立丽. 核反应堆严重事故机理研究[M]. 上海: 上海交通大学出版社, 2016. -
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