Critical Heat Flux Experiments on the Reactor Vessel Wall Using 2-D Slice Test Section
Author:
Affiliation:
1. Korea Advanced Institute of Science and Technology Department of Nuclear & Quantum Engineering, 373-1 Guseong-dong Yuseong-gu, Daejeon, 305-701, Korea
2. Korea Atomic Energy Research Institute P.O. Box 105, Yuseong-gu, Daejeon, 305-600, Korea
Publisher
Informa UK Limited
Subject
Condensed Matter Physics,Nuclear Energy and Engineering,Nuclear and High Energy Physics
Link
https://www.tandfonline.com/doi/pdf/10.13182/NT05-A3667
Reference4 articles.
1. In-vessel coolability and retention of a core melt
2. The coolability limits of a reactor pressure vessel lower head
3. T.N. DINH, J.P. TU, T. SALMASSI, and T.G. THEOFANOUS, “Limits of Coolability in the AP1000-Related ULPU-2400 Configuration V Facility,” CRSS-03/06, Center for Risk Studies and Safety, University of California, Santa Barbara (2003).
4. A hydrodynamic critical heat flux model for saturated pool boiling on a downward facing curved heating surface
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