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Development of a steady thermal-hydraulic analysis code for the China Advanced Research Reactor |
| TIAN Wenxi1, QIU Suizheng1, GUO Yun1, SU Guanghui1, JIA Dounan1, LIU Tiancai2, ZHANG Jianwei2 |
| 1.Department of Nuclear Engineering, Xi′an Jiaotong University, Xi′an 710049, China; 2.China Institute of Atomic Energy, Beijing 102413, China |
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Abstract A multi-channel model steady-state thermal-hydraulic analysis code was developed for the China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed mass flow distribution in the core was obtained. The result shows that structure size plays the most important role in mass flow distribution, and the influence of core power could be neglected under single-phase flow. The temperature field of the fuel element under unsymmetrical cooling condition was also obtained, which is necessary for further study such as stress analysis, etc. of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of the mean and hot channel was carried out and it is proved that all thermal-hydraulic parameters satisfy the Safety design regulation of CARR .
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Issue Date: 05 June 2007
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