Assessment of Interphase Drag Correlations in the RELAP5/MOD2 and TRAC-PF1/MOD2 Codes (NUREG/IA-0015)
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Date Published: July 1989
K.H. Ardron, A.J. Clare
Central Electricity Generating Board
Prepared as part of:
The Agreement on Research Participation and Technical Exchange
under the International Thermal-Hydraulic Code Assessment
and Application Program (ICAP)
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555
An assessment is carried out of the interphase drag correlations used in modelling vertical two-phase flows in the advanced thermalhydraulic codes RELAN5/Mod2 and TRAC-MFI/Modl. The assessment is performed by using code models to calculate void fraction in fully developed steam-water flows, and comparing results wi:h predictions of standard correlations and test data. The study is restricted to the bubbly and slug flow regimes (void fractions below 0.75).
For upflows, at pressures of interest in PWR small break LOCA and transient analysis the performance of She code models is generally satisfactory. E:ccepcions are (i) small hydraulic diameter channels at low pressure (p ≤ 4 MPa) (ii) large pipe diameters at void fractions exceeding 0.5; in these cases void fraction errors are cuCsLde normal uncertainty ranges.
For downflows, the code models give good agreement with limited available void fraction data.
The numerical results given in this paper allow a rapid estimate co be made of void fraction errors likely to arise in a particular code application due to deficiencies in interphase drag modelling.