Neutronic thermal-hydraulic coupling analysis for PT-SCWR-reactor core
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摘要: 针对压力管式超临界水堆(PT-SCWR)新型62棒设计,其功率密度与燃料温度、冷却剂密度/温度紧密耦合,利用中子物理分析程序(WIMS-AECL)和子通道分析程序(ATHAS),对该设计堆芯进行核热耦合分析,并进行优化,结果表明该耦合方法是有效的。分析结果指出新型62棒燃料组件设计包壳最高温度和冷却剂出口温度都低于设计限值,满足设计目标；并且可以通过调整内外圈燃料富集度至5.5%和4.6%、调整燃料组件内圈棒束节圆由5.30 cm到5.175 cm,进行优化来获取一个均匀的温度分布；通过对比不同栅距下的慢化剂温度系数和空泡系数,得到一个最佳栅距为21 cm。Abstract: According to the pressure tube supercritical water reactor (PT-SCWR) new 62-element design, the power density and the fuel temperature, coolant density/temperature are coupled. Neutron physics analysis code (WIMS-AECL) and sub-channel analysis code (ATHAS) are used to optimize the design. The results show that the coupling method is effective. The results indicate that the maximum cladding surface temperature of the bundle and the coolant outlet temperature are lower than the design limits. So the scheme meets the design objectives. We can adjust the fuel enrichment from 5% to 5.5% and 4.6%, adjust the fuel assembly pitch circle of the inner bundle from 5.30 cm to 5.175 cm to obtain a uniformity temperature distribution. By comparing the moderator temperature coefficient and void coefficient under different pitch, we obtained an optimum pitch of 21cm.
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Key words:
- PT-SCWR /
- neutronic thermal-hydraulic coupling analysis /
- 62-element bundle design /
- WIMS-AECL /
- ATHAS /
- CANDU
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