Experimental Tests and Qualification of Analytical Methods to Address Thermohydraulic Phenomena in Advanced Water Cooled Reactors
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XA0055000 IAEA-TECDOC-1149 Experimental tests and qualification of analytical methods to address thermohydraulic phenomena in advanced water cooled reactors Proceedings of a Technical Committee meeting held in Villigen, Switzerland, 14-17 September 1998 31/27 INTERNATIONAL ATOMIC ENERGY AGENCY May 2000 IAEA SAFETY RELATED PUBLICATIONS IAEA SAFETY STANDARDS Under the terms of Article III of its Statute, the IAEA is authorized to establish standards of safety for protection against ionizing radiation and to provide for the application of these standards to peaceful nuclear activities. The regulatory related publications by means of which the IAEA establishes safety standards and measures are issued in the IAEA Safety Standards Series. 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IAEA-TECDOC-1149 Experimental tests and qualification of analytical methods to address thermohydraulic phenomena in advanced water cooled reactors Proceedings of a Technical Committee meeting held in Villigen, Switzerland, 14-17 September 1998 INTERNATIONAL ATOMIC ENERGY AGENCY May 2000 The originating Section of this publication in the IAEA was: Nuclear Power Technology Development Section International Atomic Energy Agency Wagramer Strasse 5 P.O. Box 100 A-1400 Vienna, Austria EXPERIMENTAL TESTS AND QUALIFICATION OF ANALYTICAL METHODS TO ADDRESS THERMOHYDRAULIC PHENOMENA IN ADVANCED WATER COOLED REACTORS IAEA, VIENNA, 2000 IAEA-TECDOC-1149 ISSN 1011-4289 © IAEA, 2000 Printed by the IAEA in Austria May 2000 FOREWORD Worldwide there is considerable experience in nuclear power technology, especially in water cooled reactor technology. Of the operating plants, in September 1998, 346 were light water reactors (LWRs) totalling 306 GW(e) and 29 were heavy water reactors (HWRs) totalling 15 GW(e). The accumulated experience and lessons learned from these plants are being incorporated into new advanced reactor designs. Utility requirements documents have been formulated to guide these design activities by incorporating this experience, and results from research and development programmes, with the aim of reducing costs and licensing uncertainties by establishing the technical bases for the new designs. Common goals for advanced designs are high availability, user-friendly features, competitive economics and compliance with internationally recognized safety objectives. Large water cooled reactors with power outputs of 1300 MW(e) and above, which possess inherent safety characteristics (e.g. negative Doppler moderator temperature coefficients, and negative moderator void coefficient) and incorporate proven, active engineered systems to accomplish safety functions are being developed. Other designs with power outputs from, for example, 220 MW(e) up to about 1300 MW(e) which also possess inherent safety characteristics and which place more emphasis on utilization of passive safety systems are being developed. Passive systems are based on natural forces and phenomena such as natural convection and gravity, making safety functions less dependent on active systems and components like pumps and diesel generators. In some cases, further experimental tests for the thermohydraulic conditions of interest in advanced designs can provide improved understanding of the phenomena. Further, analytical methods to predict reactor thermohydraulic behaviour can be qualified for use by comparison with the experimental results. These activities should ultimately result in more economical designs. The Technical Committee Meeting on Experimental Tests and Qualification of Analytical Methods to Address Thermohydraulic Phenomena in Advanced Water Cooled Reactors was convened to review the current status and the remaining needs in this area. The meeting was hosted by the Paul Scherrer Institute, Villigen, 14-17 September 1998. The IAEA officer responsible for this publication was J. Cleveland of the Division of Nuclear Power. EDITORIAL NOTE This publication has been prepared from the original material as submitted by the authors. The views expressed do not necessarily reflect those of the IAEA, the governments of the nominating Member States or the nominating organizations. The use of particular designations of countries or territories does not imply any judgement by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries. The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement or recommendation on the part of the IAEA. The authors are responsible for having obtained the necessary permission for the IAEA to reproduce, translate or use material from sources already protected by copyrights. CONTENTS SUMMARY 1 CONDENSATION TESTS AND QUALIFICATION OF MODELS (Session 1) Effect of pool turbulence on direct contact condensation at a steam/water interface 13 J.D. Jackson, C.L. Zhao, S. Doerffer, J.E. Byrne, H. Falaki Characterization of direct contact condensation of steam jets discharging into a subcooled water •; 21 Chul-Hwa Song, SeokCho, Hwan-Yeol Kim, Yoon-Young Bae, Moon-Ki Chung Experimental research on in-tube condensation in the presence of air 39 A. Tanrikut, O. Yesin Effects of non-condensible gas on the condensation of steam 53 J.D. Jackson, P. An, A. Reinert, M. Ahmadinejad Evaluation of the condensation phenomenon in presence of non-condensible gases for PCCS related geometry 83 V. Faluomi Modelling containment passive safety systems in advanced water cooled reactors 97 L.E. Herranz, M.H. Anderson, M.L. Corradini, J.L. Munoz-Cobo DATA AND MODELS FOR CRITICAL HEAT FLUX (CHF) AND POST-CHF HEAT TRANSFER (Session 2) Experiments on critical heat flux for CAREM reactor 109 CM. Mazufri Enhancing the moderator effectiveness as a heat sink during loss of coolant accidents in CANDU-PHW reactors using glass-peened surfaces 121 T. Nitheanandan, R.W. Tiede, D.B. Sanderson, R.W.L. Fong, C.E. Coleman Uncertainty of evaluations of flow film boiling heat transfer coefficient 139 A.A. Ivashkevitch, P.L. Kirillov, V. V. Sergeev, IP. Smogalev, V.N. Vinogradov Dispersed flow film boiling heat transfer of flowing water in vertical tubes — CIAE steady state data and prediction methods 155 Chen Yuzhou, Chen Haiyan PASSIVE SYSTEMS FOR FLUID AND HEAT TRANSPORT (Session 3) Effectiveness of passive systems tested in the NOKO facility 167 E.F. Hicken, H. Jaegers, M. Fethke, L. Rossner The DIVA programme: General presentation and first results 179 P. Dumaz, B. Due A flashing driven moderator cooling system for CANDU reactors: Experimental and computational results 189 H.F. Khartabil Natural circulation in an integral CANDU test facility 201 P.J. Ingham, J.C. Luxat, A.J. Melnyk, T.V. Sanderson Natural circulation performance in nuclear power plants 213 F. D 'Auria, G.M. Galassi, M. Frogheri Boiling induced mixed convection in cooling loops 221 J. U. Knebel, G. Janssens-Maenhout, U. Muller Thermal hydraulic research on next generation PWRs using ROSA/LSTF 233 T. Yonomoto,