Study on Characteristics of Xi’an Pulsed Reactor in Reflood Stage
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摘要: 再淹没过程是失水事故重要环节，明确再淹没过程中堆芯瞬态热工特性对破口失水事故安全分析极为关键。本研究基于子通道分析程序CTF，针对西安脉冲反应堆典型大、小破口失水事故中再淹没过程开展了瞬态分析计算。分析中将堆芯依据功率划分为8个通道，获得了骤冷前沿和堆芯温度的时空分布特性，分析了不同位置燃料棒在各高度的包壳与燃料棒中心温度变化规律。研究结果表明，外围燃料棒相较于中心最热棒温度更低且更早被完全冷却；大、小破口失水事故工况下冷却剂均能完全淹没堆芯，使堆芯完全冷却，大破口失水事故工况下因衰变功率更大，堆芯被完全冷却所需时间更长；大破口失水事故工况下包壳失效风险较小，小破口失水事故包壳失效风险相较于大破口失水事故工况更高，大、小破口失水事故工况下均不会发生燃料芯块熔毁。Abstract: Reflood process is an important part of LOCA, and it is crucial to identify the transient thermal characteristics of core during reflood process for the safety analysis of break accident. Based on the subchannel analysis code CTF, the transient analysis and calculation are carried out for the reflood process of Xi'an pulsed reactor in typical large and small break loss of coolant accidents. In the analysis, the core is divided into eight channels based on power. The spatiotemporal distribution characteristics of the quenching front and core temperature are obtained, and the temperature changes of the cladding and fuel rod center at different heights of fuel rods at different positions are analyzed. The results show that the temperature of the outer fuel rods is lower and completely cooled earlier than that of the hottest rod at the center. Under both large and small break conditions, the coolant can completely submerge the core, allowing it to cool completely. Under large break conditions, due to higher decay power, it takes longer for the core to be completely cooled. The risk of cladding failure is relatively low under large break conditions, while the risk of cladding failure in small break accidents is higher compared to large break conditions. Fuel pellet melting will not occur under both large and small break conditions.
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Key words:
- Pulsed reactor /
- Reflood /
- Core temperature /
- Subchannel code /
- Safety features
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图 1 西安脉冲反应堆堆芯结构和子通道划分示意图
D1~D5—稳态控制棒；E1—脉冲控制棒；E2、E3—不锈钢棒；I1、I2—跑兔管；A1—中子源；C1~C12—12个位置的燃料棒，C1为最热棒
Figure 1. Schematic Diagram of Core Structure and Subchannel Division of Xi'an Pulsed Reactor
图 5 不同破口失水事故工况剩余堆芯功率变化示意图
Figure 5. Schematic Diagram of Residual Power Changes under Different Operating Conditions
图 6 大破口失水事故工况下z=0.41 m处不同燃料棒包壳温度对比
Figure 6. Comparison of Different Fuel Rod Cladding Temperatures at z=0.41 m under Large Break Condition
图 7 小破口失水事故工况下z=0.372 m处不同燃料棒包壳温度对比
Figure 7. Comparison of Different Fuel Rod Cladding Temperatures at z=0.372 m under Small Break Condition
图 8 不同破口失水事故工况下骤冷前沿推进曲线
Figure 8. Advancement of Quench Front under Different Break Conditions
图 9 不同破口失水事故工况下最热棒温度峰值
Figure 9. Peak Temperature of the Hottest Rod under Different Break Conditions
图 10 大破口失水事故工况下燃料棒温度瞬态响应曲线
Figure 10. Transient Response Curves of Fuel Rod Temperature under Large Break Condition
图 11 小破口失水事故工况下燃料棒温度瞬态响应曲线
Figure 11. Transient Response Curves of Fuel Rod Temperature under Small Break Condition
图 12 大破口失水事故工况下最热通道中z=0.41 m处气体温度变化
Figure 12. Changes in Gas Temperature at z=0.41 m in the Hottest Channel under Large Break Condition
图 13 小破口失水事故工况下最热通道中z=0.372 m处气体温度变化
Figure 13. Change in Gas Temperature at z=0.372 m in the Hottest Channel under Small Break Condition
表 1 堆芯几何参数
Table 1. Core Geometry Parameters
参数 数值 水池筒体内径/m 2.5 堆芯筒体内径/m 0.694 燃料棒间距/m 0.043 燃料棒长度/m 0.71 活性区长度/m 0.39 
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表 2 边界条件
Table 2. Boundary Conditions
参数 大破口失水事故 小破口失水事故 再淹没时间/s 5901 70374 再淹没压力/MPa 0.101 0.101 再淹没质量流量/(kg·s−1) 1.67 1.67 初始气体焓值/(kJ·kg−1) 682 1655 棒束表面平均
线功率密度/(kW·m−1)7.9 1.7 入口冷却剂温度/K 298 298
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