Validation of VIPRE-W sub-channel void predictions using NUPEC/BFBT measurements Brynjell-Rahkola, Karolina [Engineering Physics, Uppsala University, Uppsala (Sweden)]; Le Corre, Jean-Marie; Adamsson, Carl [Westinghouse Electric Sweden AB, Vaesteras (Sweden)] 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; BENCHMARKS; BWR TYPE REACTORS; FUEL ASSEMBLIES; MOCKUP; POWER DISTRIBUTION; PRESSURE DROP; PRESSURE RANGE MEGA PA; REACTOR SAFETY; SUBCOOLED BOILING; THERMAL HYDRAULICS; TWO-PHASE FLOW; V CODES; VALIDATION; VOIDS The validation of a sub-channel analysis code using detailed void measurements from full-scale fuel assembly mock up experiments, at the operating conditions of interest, is an essential step to perform before considering the code for applications to reactor safety analysis. Recently, NUPEC has released a large database (392 data points) of detailed steady-state void measurements from a 8x8 fuel assembly design under typical BWR conditions and an international benchmark effort has been organized. The VIPRE-W sub-channel analysis code (Westinghouse version of VIPRE-01) was used in this benchmark exercise where various void models were investigated, including the EPRI drift flux void correlation and the four-equation drift flux model of two-phase flow. Detailed and statistical analyses of the code results were performed using the entire BFBT sub-channel void database and it was demonstrated that the VIPRE-W sub-channel void predictions are generally reliable over a wide range of conditions, including where the radial power gradient is very large. However, in the case of low pressure (< 1.4 MPa) and for bundle boundary sub-channels, in particular at low pressure and low flow, a void over-prediction was observed. The entire steady-state cross-section averaged exit void database, restricted to outlet pressure > 1.4 MPa (351 data points), was predicted with an absolute mean error (M-P) of -0.016 and a standard deviation of 0.021. (author) Atomic Energy Society of Japan, Tokyo (Japan) Available from Atomic Energy Society of Japan, 2-3-7, Shimbashi, Minato, Tokyo, 105-0004 JAPAN Japan 2009-07-01 English Conference Conference Conference: NURETH-13: 13. international topical meeting on nuclear reactor thermal hydraulics, Kanazawa, Ishikawa (Japan), 27 Sep - 2 Oct 2009; Other Information: Available as CD-ROM Data in PDF format, Folder Name: FullPaper, Paper ID: N13P1080.pdf; 11 refs., 15 figs., 7 tabs.; Related Information: In: NURETH-13: Proceedings of the 13th international topical meeting on nuclear reactor thermal hydraulics| [4617 p.] Medium: X; Size: [17 page(s)] https://doi.org/ [] TRN: JP1203240093544 INIS 2013-03-14 22003465