한국원자력학회:학술대회논문집 (Proceedings of the Korean Nuclear Society Conference)
- 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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- Pages.75-81
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- 1996
Assessment of COBRA-TF for Critical Heat Flux
- Chun, Tae-Hyun (Korea Atomic Energy Research Institute) ;
- Lim, Jong-Sun (Korea Atomic Energy Research Institute) ;
- Motoaki Okazaki (Japan Atomic Energy Research Institute)
- 발행 : 1996.05.01
초록
COBRA-TF is a two fluid, three field subchannel code. Three fields are continuous vapor, continuous liquid and droplet. Some assessments are conducted to validate the related models and to estimate a code ability through dryout and post-CHF experiment in a tube and DNB test in rod bundles. It turned out form dryout and post-CHF experiment that the predicted dryout locations and wall temperature profiles are in close agreement with the experiments. On the other hand, DNB prediction of COBRA-TF are performed for two kinds of rod bundles along with EPRI CHF correlation. To estimate its performance COBRA-IV of homogeneous model is also run for the same data. The results say that COBRA-TF/EPRI is better in DNB prediction than COBRA-IV/EPRI. In addition the thermal-hydraulic behaviors due to the different two-phase flow models are presented at the condition of CHF.
키워드