Coupled 3-D Neutronics/Thermal-Hydraulics Analysis for SCWR Core Typical Transients
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摘要: 采用超临界水堆堆芯三维核热耦合瞬态性能分析方法,研究中国百万千瓦级超临界水堆(CSR1000)在控制棒弹出堆芯、控制棒失控抽出等典型瞬态过程中堆芯的瞬态性能。堆芯三维瞬态分析表明:控制棒弹出堆芯事故过程中燃料最大包壳壁面温度峰值低于事故安全限值(1260℃),控制棒失控抽出瞬态过程中燃料最大包壳壁面温度峰值低于瞬态安全限值(850℃)。燃料温度和水密度的显著反应性反馈以及必要的保护停堆措施,能够保证CSR1000堆芯在典型瞬态过程中的安全性能。
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关键词:
- 超临界水堆(SCWR) /
- 三维瞬态分析 /
- 控制棒弹出 /
- 控制棒失控抽出 /
- 最大包壳壁面温度
Abstract: Transient performance of CSR1000 core during some typical transients, such as CR ejection and uncontrolled CR withdrawal is analyzed and evaluated with the coupled three dimensional neutronics/thermal-hydraulics SCWR transient analysis code. The 3-D transient analysis shows that the maximum cladding surface temperature retains lower than safety criteria 1260℃ during the process of CR ejection accident, and the maximum cladding surface temperature retains lower than safety criteria 850℃ during the process of uncontrolled CR withdrawal transient. The safety of CSR1000 core can be ensured during the typical transients under the salient fuel temperature and water density reactivity feedback and the essential reactor protection system. -
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