Critical heat flux prediction in rod bundles under upward low mass flux densities

• Published in: Nuclear engineering and design Vol. 183; no. 3; pp. 249 - 259
• Main Authors: Khabensky, V.B, Malkin, S.D, Shalia, V.V, Nigmatulin, B.I
• Format: Journal Article
• Language: English
• Published: Elsevier B.V 01.07.1998
• Subjects: CHF, critical heat flux; WWER, water–water energy reactor (PWR-type); RBMK, uran–graphite channel reactor (Chernobyl-type)
• ISSN: 0029-5493; 1872-759X
• DOI: 10.1016/S0029-5493(98)00156-3

The analysis of experimental data and results of calculations for heat transfer crisis in heated channels under low upward coolant mass flux densities is presented. This analysis allows the determination of the basic features of the boiling crisis phenomenon. It is shown that the methods currently used for critical heat flux (CHF) prediction have insufficient accuracy in the given range of parameters. A new relationship for the CHF calculation is presented. It should be used for the water–water energy reactor (WWER) and uran–graphite channel reactor—Chernobyl-type (RBMK) rod bundles, and is verified by the test data. The comparison of results obtained by a new CHF correlation and the relationship used in RELAP5/MOD3.1 Code is presented. It is shown that the latter overpredicts the CHF values at atmospheric pressure and for x cr>0.4 and does not provide conservative estimations for the RBMK fuel bundles.