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Dynamics of Two-Phase Flows

ISBN Druckformat: 0-8493-9925-4



It is important for the design of nuclear reactors to clarify, from various viewpoints, the thermal hydraulic behavior in a fuel rod bundle. This paper reports an analytical and experimental study on the critical power characteristics in a fuel rod bundle with a narrow gap under Boiling Water Reactor (BWR) conditions.
Some correlations have already been developed to predict critical heat flux (CHF), the local heat flux at which a rapid reduction in heat transfer coefficient occurs, for a tight lattice. These correlations are based on the so-called local condition hypothesis. This hypothesis is valid to describe high-pressure, high-flow phenomena. In BWR conditions, on the other hand, an integral approach is employed, since upstream history in a nonuniform heat flux profile is quite important. The upstream history strongly depends on flow regime and, thus, quality.
Therefore, a critical-quality-type correlation for a tight-spaced triangular lattice, based on the Biasi correlation has been studied. This correlation is a critical quality-boiling length (XC−LB) type correlation, based on the experimental boiling transition data derived from triangular arrays of rod clusters at the Bettis Atomic Power Laboratory of Westinghouse Corporation.
Further, the lattice tightening effect on the critical power characteristics was studied, using a simple-shaped experimental apparatus. Comparisons with the critical power data obtained are also presented for steady state conditions and transient conditions. From these comparisons, it is considered that the present correlation is applicable to the critical power prediction in a tight-spaced, triangular fuel lattice.
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