한국원자력학회:학술대회논문집 (Proceedings of the Korean Nuclear Society Conference)
- 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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- Pages.386-391
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- 1997
Critical Heat Flux for Low Flow in Vertical Annulus under Various Pressure Conditions
- Chun, Se-Young (Korea Atomic Energy Research Institute) ;
- Jun, Hyung-Gil (Korea Atomic Energy Research Institute) ;
- Chung, Heung-June (Korea Atomic Energy Research Institute) ;
- Moon, Sang-Ki (Korea Atomic Energy Research Institute) ;
- Chung, Moon-Ki (Korea Atomic Energy Research Institute)
- 발행 : 1997.05.01
초록
It is important to understand correctly a CHF under low flow condition for the purpose of enhancing the reactor safety and performance in the LWRs. The CHF experiments have been carried out for an internally heated vertical annulus in RCS loop facility. The experimental conditions cover ranges of pressure from 1.82 to 12.08 MPa, mass flux from 300 to 550kg/
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