Abstract The present paper is a continuation of Part I: “Recalculation of single-phase and two-phase pressure loss measurements”. It deals with recalculations of void distribution measurements with the advanced two-phase, three-field sub-channel code F-COBRA-TF. Again, experimental data of both the OECD/NRC BFBT benchmark and in-house tests in AREVA NP's KATHY loop are used. The results of the recalculations of the measurements especially demonstrate the capability of a three-field code to predict void fractions with good accuracy, whereas the code is not based on a conventional void correlation which derives void fraction from quality according to an empirical function. In fact, the code relies on interfacial friction correlations for each flow regime. The quantities volume fraction of continuous liquid, volume fraction of entrained liquid and volume fraction of vapor are variables in the basic transport equations, which it is directly solved for. Thereby, the F-COBRA-TF standard models – which are usually applied for all sorts of calculations (pressure loss, void distribution, lateral mixing, critical heat flux, etc.) – were used. As already in Part I of the present paper, it was not necessary to do special code tuning with respect to certain experiments.
[1]
O. G. Martynenko,et al.
Handbook of hydraulic resistance
,
1986
.
[2]
L. Hochreiter,et al.
Analysis of FLECHT-SEASET 163-rod blocked bundle data using COBRA-TF
,
1985
.
[3]
M. Ishii,et al.
Flow regime transition criteria for upward two-phase flow in vertical tubes
,
1984
.
[4]
Lawrence E. Hochreiter,et al.
NEA NUCLEAR SCIENCE COMMITTEE NEA COMMITTEE ON SAFETY OF NUCLEAR INSTALLATIONS NUPEC BWR FULL-SIZE FINE-MESH BUNDLE TEST (BFBT) BENCHMARK Volume I: Specifications
,
2005
.
[5]
Thomas J. Hanratty,et al.
The interfacial drag and the height of the wall layer in annular flows
,
1976
.
[6]
M. Glück,et al.
Sub-channel analysis with F-COBRA-TF – Code validation and approaches to CHF prediction
,
2007
.
[7]
F. J. Moody,et al.
The thermal hydraulics of a boiling water nuclear reactor
,
1977
.
[8]
A. E. Dukler,et al.
Modelling flow pattern transitions for steady upward gas‐liquid flow in vertical tubes
,
1980
.