NUREG/IA-0002 Intemational Agreement Report Heat Transfer Processes During Intermediate and Large Break Loss-of-Coolant Accidents (LOCAs) Prepared by 1. Vojtek Reactor Safety Corporation Gesellschaft fuer Reaktorsicherheit Porschungsgefaende, 8046 Garching, The Federal Republic of Germany Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, D.C. 20555 September 1986 Prepared as part of The Agreement on Research Participation and Technical Exchange * under the International Thermal- Hydraulic Code Assessment and Application Program (ICAP) Published by U.S. Nuclear Regulatory Commission NOTICE This report was prepared under an international cooperative agreement for the exchange of technical information. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus pro- duct or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights. Available from Superintendent of Documents U.S. Government Printing Office P.O. Box 37082 Washington, D.C. 20013-7082 and National Technical Information Service Springfield, VA 22161 NUREG/IA-0002 International Agreement Report Heat Transfer Processes During Intermediate and.-Large Break Loss-of-Coolant Accidents (LOCAs) Prepared by 1. Voitek Reactor Safety Corporation Gesellschaft fuer Reaktorsicherheit Porschungsgelaende, 8046 Garching, The Federal Republic of Germany Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, D.C. 20555 September 1986 Prepared as pert of The Agrement on Research Participation and Technical Exchange under the International Thermal-Hydraulic Coda Assessment end Application Program 4ICAP) Published by U.S. Nuclear Regulatory Commission Translated by: Fischer Translation Service 1928 Catoctin Terrace Silver Spring, MD 20906 NOTICE This report documents work performed under the sponsorship of the Kraftwerk Union in the Federal Republic of Germany. The information in this report has been provided to the USNRC under the terms of an information exchange agreement between the United States and the Federal Republic of Germany (Technical Exchange and Cooperation Arrangement Between the United States Nuclear Regulatory Commission and the Bundesminister Fuer Forschung und Technologie of the Federal Republic of Germany in the field of reactor safety research and development, April 30, 1981). The Kraftwerk Union has consented to the publication of this report as a USNRC document in order that it may receive the widest possible circulation among the reactor safety community. Neither the United States Government nor the Kraftwerk Union or any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability of responsibility for any third party's use, or the results of such use, of any information, apparatus, product or process disclos'e-d inf this report, or represents that its use by such third party would not infringe privately owned rights. SUMMARY Within the framework of this research project there were examined the heat transfer ranges, as they occur during the high pressure phase of a LOCA with intermediate and large breaks. Special attention was given to the phenomena important for the prediction of the highest clad tube temperatures of the maximum and minimum critical heat flux and to the steam-droplets cooling. The experimental results of the 25-rod bundle tests, conducted at the KWU facility in Karistein, were used as a data- base for the verification of the assembled models and correlations. The values of the heat flux and of the heat transfer coefficients, ob- tained from these measurements, were used for the comparison with the calculated results and allowed the evaluation of the used correlations and models. The local values of the important thermal and fluid dynamic parameters, required for this comparison, were calculated with the aid of the computer code BRUDI-VA. -Inp articular, the following correlations were evaluated on hand of these experimental results: Maximum critical heat flux: W-3 correlation - B-W-2 correlation - Macbeth correlation - Zuber-Griffith correlation - Biasi correlation - CISE correlation - Slifer-GE correlation - Smalin correlation - Tong correlation - Thorgenson correlation - Monde-Katto correlation - Minimum heat flux: Berenson modification of the Zuber correlation - i Minimum temperature difference in case of rewetting: Berenson correlatio'n - Henry correlation - Ilceje correlation - Correlations for the calculations of the heat transfer coefficients in the sphere of steam-droplets cooling: modified Dougall-Rohsenow correlation - Groeneveld-5.7 correlation - Condie-Bengston-IV correlation - Groeneveld-Delorme correlation - Chen-Ozkaynak-Sunderam correlation - The verification of the correlations for the calculations of the maximum critical heat flux made apparent the limitation of the spheres of applica- tion of the individual correlations and showed, that none of these correla- tions can be recommended for the entire range of the test parameters for a safe prediction of the DNB moment and location. The verification of the correlations for the calculation of the heat transfer coefficients after the exceeding of the maximum critical heat flux showed, that these correlations also lead only in certain ranges of the test parameters to a good consistency between the measurement and the calculated results. The use of the chosen correlations for the calculation of the minimum temperature difference between wall and coolant and the minimum critical heat flux showed, that none of these correlations, at least in the para- meter combinations that resulted from the 25-rod bundle tests, can be used for the prediction of the rewetting phenomenon. On hand of the results of the verification of the correlations for the calculation of the heat transfer coefficients in the case of steam-droplets cooling there was developed a new "two components correlation." The application of this correlation within the entire range of the test parameters of the 25-rod bundle measurements (pressure 2 to 12 MPa, mass flow density 300 to 1400 kg/m'.s, steam quality 0.3 to 1, and wall temperature 300 to 700'C) led to a very good consistency between measure- ment and calculated results. ii ABSTRACT The general purpose of this project was the- investigation of the heat trans- fer regimes during the high pressure portion of blowdown. The main at- tention has been focussed on the evaluation of those phenomena which are most important in reactor safety, such as maximum and minimum critical heat flux and forced convection film boiling heat transfer. The experimen- tal results of the 2S-rad bundle blowdown heat transfer tests, which were performed at the KWU heat transfer test facility in Karistein, were used as a database for the verification of different correlations which are used or were developed for the analysis of reactor safety problems. The compu- ter code BRUDI-VA was used for the calculation of local values of Impor- tant thermohydraulic parameters in the bundle. In particular the following correlations have been evaluated in this study: Maximum critical heat flux: correlation -W-3 correlation -B-W-2 correlation -Macbeth correlation -Zuber-GriffIth correlation -Biasi correlation -CISE - Slifer-GE correlation - Smolin correlation ng correlation - Thorgerson correlation - Monde-Katto correlation - Minimum critical heat flux and minimum film boiling temperature: - Berenson modification of the Zuber correlation - Berenson correlation - Henry correlation - lioeje correlation Heat transfer coefficients in flow film boiling: - modified Dougal l-Rohsenow correlation - Groeneveld-5.7 correlation iii Candie-Bengston-lV correlation - - Groeneveld-Delorme correlation - Chen -Ozkayna k-Sundaram correlation. The evaluation of correlations for the prediction of critical heat flux, film boiling heat transfer coefficients and minimum film boiling temperature showed that none of the correlations should be used over the entire range of test parameters investigated. Using results of this investigations a new equilibrium correlation for the cal- culat!on of forced film boiling heat transfer coefficients has been developed. This correlation is shown to agree well. with the experimental data over the following range of testparameter: Mass flow rate 300 to 1400 kg/M2.s, pres- sure 2 to 12 MPa and quality 0.3 to 1.0. iv TABLE OF CONTENTS 1. Introduction~ 1 2. Foundations and definitions 2 2.1 Forced convection with water as a coolant 3 2.2 Subcooled nucleate boiling 4 2.3 Nucleate boiling 4 2.4 Transfer from nucleate boiling ti ofilm boiling, resp. steam-dropl ets flow 4 2.4.1 Transfer from nucleate boiling ti0 film boiling 4 2.4.2 Transfer from nucleate boiling ti steam-droplet cool ing 5 2.5 Fil~m boiling 5 2.6 Steam-droplets cooling 6 2.7 Transfer fromfilm boiling, resp. steam-dropl ets cool ing to nucleate boiling 6 2.8 *Forced steam convection 6 3. Experimental in vestogations 7 3.1 The test device 7 3.2 Instrumentation 8 3.3 Test program 9 3.3.1 DNB tests 9 3.3.2 Post DNB tests 11 4. Calculation of the heat transfer coefficients from the test data and the evaluation of the results 12 4.1 Calculation of the heat transfer coefficients from- the test data 12 4.2 Compilation of the important parameters from the test 12 5. Calculation of the local thermal and fluid dynamic parameters 14 5.1 The computer code BRUDI-VA 14 5.1.1 Basic equations of the homogenous point of equilibrium model 14 5.1.2 Calculation of the two-phase pressure loss 15 V 5.1.3 Heat conduction model 1 5.1.4 Heat transfer model 16 5.2 Calculation of the chosen tests and determination of the local thermal and fluid dynamic parameters at the DNB, dry-out and RND moment 16 6. Compilation and comparison of the correlations for the calculation of the heat transfer coefficients in the region of the forced convection with one-phase coolant 18 6.1 Forced water convection 18 6.2 Forced steam convection 19 6.3 Comparison of the correlations 20 7. Compilation and comparison of the correlations for the calculation of the heat transfer coefficients (HTC) of the subcooled and saturated nucleate boiling 21 7.1 Calculation of the necessary temperature difference for bubbles formation with subcooled fluid 21 7.2 Calculation of the heat transfer coefficients in the region of the subcooled nucleate boiling 21 7.3 Calculation of the heat transfer coefficients in the region of nucleate boiling 22 7.4 Comparison of the correlations 23 8. Transfer from nucleate boiling to film boiling, resp. stea~m-droplets cooling (maximum critical heat flux) 25 - 8.1 Vessel film boiling 25 8.2 Heat transfer crisis with forced convection, high heat flux and low steam-cOntent (DNB) 27 - 8.2.1 Semi-empirical models 27 8.2.2 Empirical models 31 8.3 Heat transfer crisis with forced convection, low to intermediate heat flux, and intermediate to high steam content (dry-out) 33 8.3.1 Analytical models 33 8.3.2 Semi-empirical and empirical correlations 36 8.4 Verification of the correlations for the calculation of the maximum cri tical heat flux (KHB) based on the experimental results 38 vi