1 Copyright © 2011 by ASME Proceedings of the ASME 2011 International Mechanical Engineering Congress & Exposition IMECE2011 November 11-17, Denver Colorado, USA IMECE2011-65893 APPLICABILITY OF CHOKING FLOW MODELS FOR SUBCOOLED FLASHING FLOW THROUGH TUBE CRACKS Brian Wolf Purdue University West Lafayette, IN, USA Shripad T. Revankar Pohang University of Science and Technology and Purdue University West Lafayette, IN, USA Jovica R. Riznic Canadian Nuclear Safety Commission Affiliation Ottawa, Ontario, Canada ABSTRACT Recently there is some database available on choking flow through cracks relevant to steam generator (SG) tubes to model the critical flow. These data are used in assessing the key choking flow models. Based on this assessment a mechanistic choking model is developed. The model is used to predict the choking flow rates for various experimental conditions for subcooled flashing flow through narrow slits with L/D varying from small values (~5) to large values (100). Results are presented on the effects of thermal and mechanical non- equilibrium on the choking flow for small L/D channels. A mechanistic model was developed to model two-phase choking flow through slits. A comparison of model results to experimental data shows that the homogeneous equilibrium based models markedly under predict choking flow rates in such geometries. As subcooling increases, and channel length decreases the non-equilibrium effects play a greater role in the choking phenomenon, therefore the difference in model predictions and experimental results increases. INTRODUCTION Choking flow is a phenomenon which occurs in a wide range of industrial systems. It is especially important in a nuclear reactor, where high pressure subcooled water in the example of a PWR is used to generate steam. In the case of a loss of coolant accident (LOCA), choking flow determines the water inventory of the reactor vessel. If choking were not to occur, the reactor water inventory would be depleted rapidly. This is not the case however because as the pressurized subcooled water nears the break, it flashes to vapor which limits the mass flow rate due to choking. Therefore, the integrity of the core during a LOCA is dependent upon this choking phenomenon. Most experimental studies of the past and data have been focused on large breaks, such as in a main steam line. In fact many current models have been fully or partially developed and benchmarked based on the infamous data produced by the first full scale choking flow tests for large break loss of coolant analysis referred to as the Marviken tests (Marviken,1979). While full scale data is invaluable in most avenues of research, it is produced in order to minimize or eliminate scaling error. For the same reason, the Marviken data sets should not necessarily be used to develop choking flow models for geometries related to SG tubes. A LOCA from a small break or large break in the main steam line or valve is not the only place in a reactor where coolant could possibly escape from the primary side of the reactor. Steam generator tubes have a history of small cracks and even ruptures, which lead to a loss of coolant from the primary side to the secondary side. Some of the main tube degradation mechanisms that have been identified are due to corrosion, mechanical wear, and fatigue. Some examples of such mechanisms are intergranular attack and outside diameter stress corrosion cracking, primary water stress corrosion cracking, tube fretting and wear, foreign material wear, pitting, high cycle fatigue, and wastage or thinning (Fuller et al., 2006). Currently, steam generators operate under a leak-before-break approach. This approach is used in a variety of industrial systems by demonstrating that a crack can grow through-wall resulting in leakage, and that this through-wall flaw will be detected well before the flaw becomes unstable and the tube ruptures (IAEA,1993). A rupture then signifies the loss of the integrity of the tube itself. Therefore, choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant, but also to everyday operation. In the case of leakage through Proceedings of the ASME 2011 International Mechanical Engineering Congress & Exposition IMECE2011 November 11-17, 2011, Denver, Colorado, USA IMECE2011-65893 Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/15/2014 Terms of Use: http://asme.org/terms