Steady-State Thermal-Hydraulic Analysis of the Oregon State University TRIGA Reactor Using RELAP5-3D
Nuclear Science and Engineering(2009)
摘要
Oregon State University has recently conducted a complete core conversion analysis as part of the Reduced Enrichment for Research and Test Reactors Program. The goals of the thermal-hydraulic steady-state analysis were to calculate natural-circulation flow rates, coolant temperatures, and fuel temperatures as a function of core power, as well as peak values of fuel temperature, cladding temperature, surface heat flux, critical heat flux ratio, and temperature profiles in the hot channel for both the highly enriched uranium and low-enriched uranium cores. RELAP5-3D Version 2.4.2 was used for all computational modeling during the thermal-hydraulic analysis. This is a lumped parameter code forcing engineering assumptions to be made during the analysis. A single-hot-channel model's results are compared to results produced from more refined two- and eight-channel models in order to identify variations in thermal-hydraulic characteristics as a function of spatial refinement.
更多查看译文
关键词
Thermal-Hydraulic Analysis,Lead-Cooled Fast Reactor,High Temperature Behavior,Reliability Evaluation
AI 理解论文
溯源树
样例
![](https://originalfileserver.aminer.cn/sys/aminer/pubs/mrt_preview.jpeg)
生成溯源树,研究论文发展脉络
Chat Paper
正在生成论文摘要