RELAP5/MOD3.3 CODE MANUAL VOLUME III: DEVELOPMENTAL ASSESSMENT PROBLEMS

Transcription

1 RELAP5/MOD3.3 CODE MANUAL VOLUME III: DEVELOPMENTAL ASSESSMENT PROBLEMS Nuclear Systems Analysis Division December 21 Information Systems Laboratories, Inc. Rockville, Maryland Idaho Falls, Idaho Prepared for the Division of Systems Research Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission Washington, DC 2555

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3 ABSTRACT The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. In this volume the results of a wide range of test applications of RELAP5/MOD3.3 are presented for the purpose of demonstrating the applicability of the code. The applications include results from ten phenomenological tests, nineteen separate effects tests, and five integral experiments. The phenomenological problems are primarily used to determine whether the code produces qualitatively correct results; however two of the problems have analytical solutions that serve as quantitative tests as well. The separate effects problems all provide quantitative as well as qualitative tests for correct simulation. Each problem is selected to emphasize a particular physical effect that provides a test for correct functioning of a particular model or group of models. The integral problems provide tests for qualitative and, if data is available, quantitative correctness of all the code models working in concert. The primary objective of the work reported in this edition of Volume III is to provide a comparison of results obtained with the current version of RELAP5, MOD3.3 to results obtained using the previous version MOD3.2. However, for those problems having analytical solutions or data, the MOD3.3 results are also compared to those data as well. iii NUREG/CR-5535/Rev 1-Vol III

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5 EXECUTIVE SUMMARY EXECUTIVE SUMMARY The light water reactor (LWR) transient analysis code, RELAP5, was originally developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC). Code uses include analyses required to support rulemaking, licensing audit calculations, evaluation of accident mitigation strategies, evaluation of operator guidelines, and experiment planning analysis. RELAP5 has also been used as the basis for a nuclear plant analyzer. Specific applications have included simulations of transients in LWR systems such as loss of coolant, anticipated transients without scram (ATWS), and operational transients such as loss of feedwater, loss of offsite power, station blackout, and turbine trip. RELAP5 is a highly generic code that, in addition to calculating the behavior of a reactor coolant system during a transient, can be used for simulation of a wide variety of hydraulic and thermal transients in both nuclear and nonnuclear systems involving mixtures of steam, water, noncondensable, and solute. The MOD3 version of RELAP5 has been developed jointly by the NRC and a consortium consisting of several foreign and domestic organizations that were members of the International Code Assessment and Applications Program (ICAP) and its successor organization, the Code Applications and Maintenance Program (CAMP). Credit also needs to be given to various Department of Energy sponsors, including the INEL laboratory-directed discretionary funding program. The mission of the RELAP5/MOD3 development program was to develop a code version suitable for the analysis of all transients and postulated accidents in LWR systems, including both large- and small-break loss-of-coolant accidents (LOCAs) as well as the full range of operational transients. The RELAP5/MOD3 code is based on a nonhomogeneous and nonequilibrium model for the twophase system that is solved by a fast, semi-implicit numerical scheme to permit economical calculation of system transients. The objective of the RELAP5 development effort from the outset was to produce a code that included important first-order effects necessary for accurate prediction of system transients but that was sufficiently simple and cost effective so that parametric or sensitivity studies were possible. The code includes many generic component models from which general systems can be simulated. The component models include pumps, valves, pipes, heat releasing or absorbing structures, reactor point kinetics, electric heaters, jet pumps, turbines, separators, accumulators, and control system components. In addition, special process models are included for effects such as form loss, flow at an abrupt area change, branching, choked flow, boron tracking, and noncondensable gas transport. The system mathematical models are coupled into an efficient code structure. The code includes extensive input checking capability to help the user discover input errors and inconsistencies. Also included are free-format input, restart, renodalization, and variable output edit features. These user conveniences were developed in recognition that generally the major cost associated with the use of a system transient code is in the engineering labor and time involved in accumulating system data and developing system models, while the computer cost associated with generation of the final result is usually small. The development of the models and code versions that constitute RELAP5 has spanned approximately 23 years from the early stages of RELAP5 numerical scheme development to the present. RELAP5 represents the aggregate accumulation of experience in modeling reactor core behavior during v NUREG/CR-5535/Rev 1-Vol III

6 EXECUTIVE SUMMARY accidents, two-phase flow processes, and LWR systems. The code development has benefited from extensive application and comparison to experimental data in the LOFT, PBF, Semiscale, NRU, and other experimental programs. Several new models, improvements to existing models, and user conveniences were added to RELAP5/MOD2 to produce RELAP5/MOD3.. Additional models, model improvements, and corrections have been incorporated with the MOD3.2 and MOD3.3 releases. The primary focus of this document will be the performance of the RELAP5/MOD3.3 version of the code in comparison to previous versions. Most comparisons will be to the MOD3.2 results and to the data or analytical solutions where they exist. The new models that have been added since the conclusion of the MOD2 development include: Several counter-current flow limiting correlations that can be activated by the user at each junction in the system model. The ECCMIX component for modeling of the mixing of subcooled emergency core cooling system (ECCS) liquid and the resulting interfacial condensation. A zirconium-water reaction model to model the exothermic energy production on the surface of zirconium cladding material at high temperature. A surface-to-surface radiation heat transfer model with multiple thermal radiation enclosures defined through user input. A level tracking model. A thermal stratification model. Improvements to existing models include: New correlations for interfacial friction for all types of geometry in the bubbly-slug flow regime in vertical flow passages. Use of junction-based interphase drag. An improved model for vapor pullthrough and liquid entrainment in horizontal pipes to obtain correct computation of the fluid state convected through a break. A new critical heat flux correlation for rod bundles based on tabular data. An improved horizontal stratification inception criterion for predicting the flow regime transition between horizontally stratified and dispersed flow. A modified reflood heat transfer model. NUREG/CR-5535/Rev 1-Vol III vi

7 EXECUTIVE SUMMARY Improved logic for vertical stratification inception to avoid excessive activation of the water packing model. An improved boron transport model. A mechanistic separator/dryer model. An improved crossflow model. An improved form loss model. The addition of a simple plastic strain model with a clad burst criterion to the fuel mechanical model. The addition of a radiation heat transfer term to the gap conductance model. Modifications to the noncondensable gas model to eliminate erratic code behavior and failure. Improvements to the downcomer penetration, ECCS bypass, and upper plenum deentrainment capabilities. An improved equation of state that includes the meta-stable regions and uses thermodynamically consistent interpolation. Additional user conveniences include: Modifications that place both the vertical stratification and water packing models under user control so they can be deactivated. Removal of bit packing and vectorization to improve portability and readability. Computer portability through the conversion of the FORTRAN coding to adhere to the FORTRAN 77 standard. Code execution and validation on a variety of systems. The code should be easily installed (i.e., the installation script is supplied with the transmittal) on the CRAY X-MP (UNICOS), DECstation 5 (ULTRIX), DEC Alpha Workstation (OSF/1), IBM Workstation 6 (UNIX), SUN Workstation (UNIX), SGI Workstation (UNIX), and HP Workstation (UNIX). The code has been installed (although the installation script is not supplied with the transmittal) on the IBM 39 (MVS) and IBM-PC (DOS). The code can be installed easily on all 64-bit machines (integer and floating point operands) and any 32- bit machine that provides for 64-bit floating point. vii NUREG/CR-5535/Rev 1-Vol III

8 EXECUTIVE SUMMARY The RELAP5/MOD3 code manual consists of seven separate volumes. The modeling theory and associated numerical schemes are described in Volume I, to acquaint the user with the modeling base and thus aid in effective use of the code. Volume II contains more detailed instructions for code application and specific instructions for input data preparation. Both Volumes I and II are expanded and revised versions of the RELAP5/MOD2 code manual a and Volumes I and III of the SCDAP/RELAP5/MOD2 code manual. b This volume, Volume III, presents the results of developmental assessment cases run with RELAP5/ MOD3.3 to demonstrate and validate the models used in the code. The assessment matrix contains phenomenological problems, separate-effects tests, and integral systems tests. Volume IV contains a detailed discussion of the models and correlations used in RELAP5/MOD3. It presents the user with the underlying assumptions and simplifications used to generate and implement the base equations into the code so that an intelligent assessment of the applicability and accuracy of the resulting calculations can be made. Thus, the user can determine whether RELAP5/MOD3 is capable of modeling his or her particular application, whether the calculated results will be directly comparable to measurement or whether they must be interpreted in an average sense, and whether the results can be used to make quantitative decisions. Volume V provides guidelines for users that have evolved over the past several years from applications of the RELAP5 code at the Idaho National Engineering Laboratory, at other national laboratories, and by users throughout the world. Volume VI discusses the numerical scheme in RELAP5/MOD3, and Volume VII is a collection of independent assessment calculations. a. V. H. Ransom et al., RELAP5/MOD2 Code Manual, Volumes I and II, NUREG/CR-4312, EGG-2396, Idaho National Engineering Laboratory, August 1995 and December 1985, revised March b. C. M. Allison and E. C. Johnson, Eds., SCDAP/RELAP5/MOD2 Code Manual, Volume I: RELAP5 Code Structure, System Models, and Solution Methods, and Volume III: User s Guide and Input Requirements, NUREG/CR-5273, EGG-2555, Idaho National Engineering Laboratory, June NUREG/CR-5535/Rev 1-Vol III viii

9 ACKNOWLEDGMENTS ACKNOWLEDGMENTS Development of a complex computer code such as RELAP5 is the result of team effort and requires the diverse talents of a large number of people. Special acknowledgment is given to those who pioneered and continue to contribute to the RELAP5 code, in particular, V. H. Ransom, J. A. Trapp, and R. J. Wagner. A number of other people have made and continue to make significant contributions to the continuing development of the RELAP5 code. Recognition and gratitude is given to the members of the INEL RELAP5 team: V. T. Berta C. E. Lenglade R. A. Riemke K. E. Carlson M. A. Lintner R. R. Schultz C. D. Fletcher C. C. McKenzie A. S-L. Shieh E. E. Jenkins G. L. Mesina R. W. Shumway E. C. Johnsen C. S. Miller C. E. Slater G. W. Johnsen G. A. Mortensen S. M. Sloan J. M. Kelly P. E. Murray M. Warnick H-H. Kuo R. B. Nielson W. L. Weaver N. S. Larson S. Paik G. E. Wilson The list of contributors is incomplete, as many others have made significant contributions in the past. Rather than attempt to list them all and risk unknowingly omitting some who have contributed, we acknowledge them as a group and express our appreciation for their contributions to the success of the RELAP5 effort. The list of contributors while the RELAP5 Program was at SCIENTECH, Inc., included: Bill Arcieri Doug Barber Robert Beaton Robert Copp Byron Hansen Scott Lucas Glen Mortensen Dan Prelewicz Rex Shumway Randy Tompot Weidong Wang The current list of contributors to the RELAP5 Program at Information Systems Laboratories, Inc., include: Bill Arcieri Doug Barber Robert Beaton Mark Bolander Don Fletcher Glen Mortensen Dan Prelewicz Rex Shumway Victor Ransom The RELAP5 Program is indebted to the technical monitors from the U. S. Nuclear Regulatory Commission and the Department of Energy-Idaho Operations Office for giving direction and management to the overall program. Those from the NRC include W. Lyon, Y. Chen, R. Lee, R. Landry, H. Scott, M. Rubin, D. E. Solberg, D. Ebert, S. Smith, T. Lee, V. Mousseau, and Weidong Wang. Monitors from ix NUREG/CR-5535/Rev 1-Vol III

10 ACKNOWLEDGMENTS DOE-ID when the RELAP5 program was at the INEL include N. Bonicelli, C. Noble, W. Rettig, and D. Majumdar. The technical editing of the RELAP5 manuals when the RELAP5 program was at the INEL was done by D. Pack and E. May. Finally, acknowledgment is made of all the code users who have been very helpful in stimulating timely correction of code deficiencies and suggesting improvements. NUREG/CR-5535/Rev 1-Vol III x

11 CONTENTS 1 INTRODUCTION ADDED CAPABILITY OF RELAP5/MOD INTENDED APPLICATIONS OF RELAP5/MOD KNOWN LIMITATIONS OF RELAP5/MOD Developmental Assessment Objective DEVELOPMENTAL ASSESSMENT PROBLEMS PHENOMENOLOGICAL PROBLEMS Nine-Volume Water Over Steam Nitrogen-Water Manometer Problem Branch Reentrant Tee Problem Cross-Flow Tee Problem Cross Tank Problem Three-Stage Turbine Workshop Problem Workshop Problem Horizontally Stratified Countercurrent Flow Pryor s Pipe Problem References SEPARATE - EFFECTS PROBLEMS Edwards Pipe Problem Dukler Air-Water Flooding Tests Marviken Test Marviken Test LOFT Test L3-1 Accumulator Blowdown Bennett s Heated Tube Experiments Royal Institute of Technology Tube Test ORNL Bundle Tests Christensen Subcooled Boiling Test Shoukri Subcooled Flow Boiling and Condensation Test MIT Pressurizer Test ST FLECHT-SEASET Forced Reflood Tests FLECHT-SEASET Boil Off Test Summary of Separate Effects Assessment References INTEGRAL TEST PROBLEMS LOFT Small-Break Test L LOFT Large-Break Test L Semiscale Natural Circulation Tests S-NC-2 and S-NC Zion-1 PWR Small Break References Page xi NUREG/CR-5535/Rev 1-Vol III

12 3 CONCLUSION Phenomenological Problems Separate Effects Problems INTEGRAL TEST PROBLEMS NUREG/CR-5535/Rev 1-Vol III xii

13 FIGURES Page RELAP5 Nodalization Diagram of the Nine-Volume Water Over Steam Problem The History of Void Distribution for the Nine-Volume Problem (Volume 1) The History of Void Distribution for the Nine-Volume Problem (Volume 3) The History of Void Distribution for the Nine-Volume Problem (Volume 5) The History of Void Distribution for the Nine-Volume Problem (Volume 7) The History of Void Distribution for the Nine-Volume Problem (Volume 9) Void Distribution in the Vertical Pipe at Various Times as Calculated by MOD Void Distribution in the Vertical Pipe at Various Times as Calculated by MOD RELAP Nitrogen-Water Manometer Nodalization Liquid Velocity at the Bottom of the Manometer (junction 77711) MOD3.2 and MOD3.3 Calculated Water Level Comparison for the Manometer Problem Nodalization of the Branch Tee Problem Using Two Non-Sink Junctions Nodalization Diagram for the Cross Flow Tee Problem Calculated Mass Error Comparison for the Crossflow Tee Problem RELAP5 Cross Tank Nodalization Diagram Mass Flow Rate Comparison: Pipe 2, Volume 14 (Liquid Region) Mass Flow Rate Comparison: Pipe 4, Volume 14 (Liquid Region) Mass Flow Rate: Cross Flow Junction 315 (Connecting Pipe 2 to Pipe 4 in the Liquid Region) Mass Flow Rate Comparison: Pipe 2, Volume 18 (Gas Region) Mass Flow Rate Comparison: Pipe 4, Volume 18 (Gas Region) Mass Flow Rate: Cross Flow Junction 319 (Connecting Pipe 2 to Pipe 4 in the Gas Region) Nodalization Used for the Three-stage Group Turbine Problem Mass Error Comparison for the Three-stage Turbine Problem Relap5 Nodalization Diagram for Workshop Problems 2 and Pressurizer Pressure History of Workshop Problem Core Outlet Pressure History of Workshop Problem Steam Generator Steam Dome Pressure History of Workshop Problem Steam Generator Secondary Side Liquid Level for Workshop Problem Mass Flow in the Primary Loop Hot Leg for Workshop Problem Mass Flow in the Primary Side of Steam Generator for Workshop Problem Reactor Core Pressure Response for Workshop Problem xiii NUREG/CR-5535/Rev 1-Vol III

14 Pressurizer Pressure Response for Workshop Problem Mass Flow in the Reactor Core for Workshop Problem Mass Flow in the Pressurizer for Workshop Problem Mass Flow in the Steam Generator Secondary Side (Riser) for Workshop Problem Mass Flow in the Steam Discharge Line for Workshop Problem Steam Generator Steam Dome Pressure Response for Workshop Problem Relap5 Nodalization Diagram for a Horizontally Stratified Countercurrent Flow Problem Relap5-Calculated Junction Liquid Velocity at the Left End Relap5-Calculated Junction Liquid Velocity at the Mid-section Relap5-Calculated Junction Liquid Velocity at the Right End Relap5-Calculated Junction Vapor Velocity at the Left End Relap5-Calculated Junction Vapor Velocity at the Mid-Section Relap5-Calculated Junction Vapor Velocity at the Right End Relap5 Nodalization Diagram for Pryor s Pipe Problem Void Fraction Response for Volume 2 of the Pryor s Pipe Problem Void Fraction Response for Volume 4 of the Pryor s Pipe Problem Void Fraction Response for Volume 6 of the Pryor s Pipe Problem Void Fraction Response for Volume 8 of the Pryor s Pipe Problem Void Fraction Response for Volume 1 of the Pryor s Pipe Problem Volume Pressure Response for Volume 2 of the Pryor s Pipe Problem Volume Pressure Response for Volume 4 of the Pryor s Pipe Problem Volume Pressure Response for Volume 6 of the Pryor s Pipe Problem Volume Pressure Response for Volume 8 of the Pryor s Pipe Problem Volume Pressure Response for Volume 1 of the Pryor s Pipe Problem Relap5 Hydrodynamic Nodalization for Edwards Pipe Experiment Pressure Comparison at Left Section of Edwards Pipe Blowdown Experiment Vapor Void Fraction Comparison at Left Section of Edwards Pipe Blowdown Experiment Pressure Comparison at Left Section of Edwards Pipe Blowdown Experiment, Heavy Water Medium Vapor Void Fraction Comparison at Left Section of Edwards Pipe Blowdown Experiment, Heavy Water Medium Pressure Comparison at Left Section of Edwards Pipe Blowdown Experiment, Nearly Implicit Advancement Scheme Vapor Void Fraction Comparison at Left Section of Edwards Pipe Blowdown Experiment, Nearly Implicit Advancement Scheme Edwards Pipe Blowdown Pipe Transient - Restart at.1 Second, Mod Schematic of the Dukler Air/water Test Facility...54 NUREG/CR-5535/Rev 1-Vol III xiv

15 Relap5 Nodalization for Dukler s Air/water Test Problem Data for the Liquid Downflow Rate Versus Air Flow Injection Rate for Dukler s Air/Water Problem Measured and Mod3.2 Calculated Liquid Downflow Comparison for Dukler s Air/Water Problem Measured and Mod3.3 Calculated Liquid Downflow Comparison for Dukler s Air/Water Problem Measured and Mod3.3 Calculated Liquid Downflow Comparison for Dukler s Air/Water Problem With Gas Constant and Slope Fitted to Data Schematic, Nodalization and Initial Temperature Profile for Marviken Test Measured and Calculated Pressure in the Top of the Vessel for Marviken Test Measured and Calculated Mass Flow Rate at the Nozzle for Marviken Test Measured and Calculated Density in the Middle of the Discharge Pipe for Marviken Test Measured and Calculated Pressure (Including Phase Slip at Choked Conditions) in the Top of the Vessel for Marviken Test Measured and Calculated Mass Flow Rate (Including Phase Slip at Choked Conditions) at the Nozzle for Marviken Test Measured and Calculated Density (Including Phase Slip at Choked Conditions) in the Middle of the Discharge Pipe for Marviken Test Schematic, Nodalization and Initial Temperature Profile for Marviken Test Measured and Calculated Pressure in the Top of the Vessel for Marviken Test Measured and Calculated Mass Flow Rate at the Nozzle for Marviken Test LOFT L3-1 Accumulator A and Surgeline Schematic RELAP5 LOFT L3-1 Accumulator Model Schematic Measured and Calculated Accumulator Gas Dome Pressure Versus Volume, Loft Test L Measured and Calculated Accumulator Gas Dome Pressure, Loft Test L Measured and Calculated Accumulator Liquid Level, LOFT Test L Measured and Calculated Accumulator Gas Temperature, LOFT Test L RELAP5 Nodalization Diagram for Bennett s Heated Tube Experiment Measured and Calculated Axial Wall Temperature Profiles for Bennett s Heated Tube Low Mass Flux Experiment - Test Measured and Calculated Axial Wall Temperature Profiles for Bennett s Heated Tube Intermediate Mass Flux Experiment - Test Measured and Calculated Axial Wall Temperature Profiles for Bennett s Heated Tube High Mass Flux Experiment - Test xv NUREG/CR-5535/Rev 1-Vol III

16 Measured and Calculated Critical Heat Flux Position for the Royal Institute of Technology Tube Test Measured and Calculated Surface Temperature for Ornl Bundle CHF Test 3.7.9B Measured and Calculated Surface Temperature for Ornl Bundle CHF Test 3.7.9N Measured and Calculated Surface Temperature for Ornl Bundle CHF Test 3.7.9W Measured and Calculated Axial Void Fractions for the Ornl Void Profile Test Measured and Calculated Rod Temperature for the Ornl Void Profile Test Measured and Calculated Steam Temperature for the Ornl Void Profile Test Measured and Calculated Axial Void Fractions for the Christensen Subcooled Boiling Test Measured and Calculated Axial Void Fractions for the Shoukri 3c Experiment Schematic of the Experimental Apparatus for the Mit Pressurizer Test Measured and Calculated Rate of Pressure Rise for the Mit Pressurizer Test Calculated Vapor Generation Rate for Volume 38, MIT Test ST Tank fluid and Inside Wall Temperature at 35s into the MIT Pressurizer Test Tank Fluid and Inside Wall Temperature at 35s into the Transient During the MIT Pressurizer Test with the Thermal Front Tracking Model Active Measured and Calculated Rate of Pressure Rise for the MIT Pressurizer Test, Thermal Front Tracking Model Active RELAP5 Nodalization for the FLECHT-SEASET Forced Reflood Tests Measured and Calculated Rod Surface Temperature Histories for FLECHT-SEASET Forced Reflood Run 3154 at the.61 m (2 ft) Elevation Measured and Calculated Rod Surface Temperature Histories for FLECHT-SEASET Forced Reflood Run 3154 at the 1.22 m (4 ft) Elevation Measured and Calculated Rod Surface Temperature Histories for FLECHT-SEASET Forced Reflood Run 3154 at the 1.83 m (6 ft) Elevation Measured and Calculated Rod Surface Temperature Histories for FLECHT-SEASET Forced Reflood Run 3154 at the 2.46 m (8 ft) Elevation Measured and Calculated Rod Surface Temperature Histories for FLECHT-SEASET Forced Reflood Run 3154 at the 2.85 m (9.25 ft) Elevation Measured and Calculated Rod Surface Temperature Histories for FLECHT-SEASET Forced Reflood Run3154 at the 3.8 m (1 ft) NUREG/CR-5535/Rev 1-Vol III xvi

17 Elevation Measured and Calculated Rod Surface Temperature Histories for FLECHT-SEASET Forced Reflood Run3154 at the 3.38 m (11 ft) Elevation Measured and Calculated Steam Temperatures for FLECHT-SEASET Forced Reflood Run 3154 at 1.23 m (4 ft) Elevation Measured and Calculated Steam Temperatures for FLECHT-SEASET Forced Reflood Run 3154 at 1.85 m (6 ft) Elevation Measured and Calculated Steam Temperatures for FLECHT-SEASET Forced Reflood Run 3154 at 2.46 m (8 ft) Elevation Measured and Calculated Steam Temperatures for FLECHT-SEASET Forced Reflood Run 3154 at 2.85 m (9.25 ft) Elevation Measured and Calculated Steam Temperatures for FLECHT-SEASET Forced Reflood Run 3154 at 3.8 m (1 ft) Elevation Measured and Calculated Steam Temperatures for FLECHT-SEASET Forced Reflood Run 3154 at 3.54 m (11 ft) Elevation Measured and Calculated Total Bundle Mass Inventory for FLECHT-SEASET Forced Reflood Run Measured and Calculated Void Fractions at.92 to 1.23 m (3 to 4 ft) for FLECHT-SEASET Forced Reflood Run Measured and Calculated Void Fractions at 1.23 to 1.54 m (4 to 5 ft) for FLECHT-SEASET Forced Reflood Run Measured and Calculated Void Fractions at 1.54 to 1.85 m (5 to 6 ft) for FLECHT-SEASET Forced Reflood Run Measured and Calculated Void Fractions at 1.85 to 2.15 m (6 to 7 ft) for FLECHT-SEASET Forced Reflood Run Measured and Calculated Void Fractions at 2.15 to 2.46 m (7 to 8 ft) for FLECHT-SEASET Forced Reflood Run Measured and Calculated Axial Void Profile at 1s for FLECHT-SEASET Forced Reflood Run Measured and Calculated Axial Void Profile at 2s for FLECHT-SEASET Forced Reflood Run Measured and Calculated Axial Void Profile at 3s for FLECHT-SEASET Forced Reflood Run Measured and Calculated Rod Surface Temperatures for FLECHT-SEASET Forced Reflood Run m (2 ft) Elevation Measured and Calculated Rod Surface Temperatures for FLECHT-SEASET Forced Reflood Run m (4 ft) Elevation Measured and Calculated Rod Surface Temperatures for FLECHT-SEASET Forced Reflood Run 3171 at the 1.83 m (6 ft) Elevation Measured and Calculated Rod Surface Temperatures for FLECHT-SEASET Forced Reflood Run 3171 at the 2.46 m (8 ft) Elevation...12 xvii NUREG/CR-5535/Rev 1-Vol III

18 Measured and Calculated Rod Surface Temperatures for FLECHT-SEASET Forced Reflood Run 3171 at the 3.8 m (1 ft) Elevation Measured and Calculated Total Bundle Mass Inventory for FLECHT-SEASET Forced Reflood Run Measured and Calculated Void Fraction History at the to 1ft Level for Test Measured and Calculated Void Fraction History at the to 1ft Level for Test Measured and Calculated Void Fraction History at the 1 to 2ft Level for Test Measured and Calculated Void Fraction History at the 2 to 3ft Level for Test Measured and Calculated Void Fraction History at the 3 to 4ft level for Test Measured and Calculated Void Fraction History at the 4 to 5ft Level for Test Measured and Calculated void fraction history at the 5 to 6ft level for Test Schematic of LOFT Test Facility RELAP5 Nodalization for LOFT Test L3-7: Vessel and Broken Loop RELAP5 Nodalization for LOFT Test L3-7: Intact Loop CPU Time Versus Simulated Time for LOFT Test L Measured and Calculated Primary System Pressure for LOFT Test L Measured and Calculated Secondary System Pressure for LOFT Test L Measured and Calculated Liquid Velocity in the Intact Loop Hot Leg for LOFT Test L Measured and Calculated Vapor Velocity in the Intact Loop Hot Leg for LOFT Test L Measured and Calculated Liquid Temperature at the Core Inlet for LOFT Test L Measured and Calculated Liquid Temperature at the Core Outlet for LOFT Test L Measured and Calculated Mass Flow Rate at the Break for LOFT Test L Measured and Calculated Density in the Intact Loop Hot Leg for LOFT Test L RELAP5 Nodalization for LOFT Test L2-5: Vessel and Broken Loop RELAP5 Nodalization for LOFT Test L2-5: Intact Loop Measured and Calculated Primary System Pressure for LOFT Test L Measured and Calculated Secondary System Pressure for LOFT Test L Measured and Calculated Mass Flow Rate in the Broken Loop Cold Leg for LOFT Test L NUREG/CR-5535/Rev 1-Vol III xviii

19 Measured and Calculated Mass Flow Rate in the Broken Loop Hot Leg for LOFT Test L Measured and Calculated Mass Flow Rate in the Intact Loop Cold Leg for LOFT Test L Measured and Calculated Mass Flow Rate in the Intact Loop Hot Leg for LOFT Test L Measured Intact Loop Hot Leg Differential Pressure; LOFT Test L Absolute Value of Measured and Calculated Mass Flow Rate in the Intact Loop Hot Leg for LOFT Test L Measured and Calculated Density in the Intact Loop Hot Leg for LOFT Test L Measured and Calculated Accumulator Level for LOFT Test L Calculated Mass Flow Rate from the Accumulator for LOFT Test L Measured and Calculated Pump Speed for Primary Coolant Pump 2 (Hydrodynamic Volume 165) for LOFT Test L Measured and Calculated Upper Plenum Temperature Below the Nozzle for LOFT Test L Measured and Calculated Lower Plenum Temperature for LOFT Test L Measured and Calculated Fuel Centerline Temperature.69m (27in.) Above the Bottom of the Core for LOFT Test L Measured and Calculated Fuel Cladding Temperature.13m (5in.) Above the Bottom of the Core for LOFT Test L Measured and Calculated Fuel Cladding Temperature.53m (21in.) Above the Bottom of the Core for LOFT Test L Measured and Calculated Fuel Cladding Temperature.69m (27in.) Above the Bottom of the Core for LOFT Test L Measured and Calculated Fuel Cladding Temperature.99m (39in.) Above the Bottom of the Core for LOFT Test L Measured and Calculated Fuel Cladding Temperature 1.37m (54in.) Above the Bottom of the Core for LOFT Test L Measured and Calculated Fuel Cladding Temperature 1.47m (58in.) Above the Bottom of the Core for LOFT Test L Semiscale Mod-2A Single-loop Configuration Schematic of RELAP5/MOD3 Natural Circulation Test Model Measured and Calculated Primary System Mass Flow Rate at the 6kW Core Power for Test S-NC Measured and Calculated Primary System Hot Leg Fluid Temperature at the 6kW Core Power for Test S-NC Measured and Calculated Primary Side Steam Generator Outlet Fluid Temperature at the 6kW Core Power for Test S-NC Measured and Calculated Primary System Pressure at the 6kW Core Power for Test S-NC xix NUREG/CR-5535/Rev 1-Vol III

20 Measured and Calculated Primary System Mass Flow Rate Versus Steam Generator Secondary Side Heat Transfer Area for Test S-NC Measured and Calculated Primary Side Hot Leg Fluid Temperature Versus Steam Generator Secondary Side Heat Transfer Area for Test S-NC Measured and Calculated Primary Side Steam Generator Outlet Temperature Versus Steam Generator Secondary Side Heat Transfer Area for Test S-NC Measured and Calculated Primary System Pressure Versus Steam Generator Secondary Side Heat Transfer Area for Test S-NC RELAP5 Nodalization for the Zion-1 PWR: Vessel Model RELAP5 Nodalization for the Zion-1 PWR: Loop Model MOD3.2 and MOD3.3 Primary System Pressure (Core Outlet) Comparison for a Small Break Transient in a Typical PWR MOD3.2 and MOD3.3 Break Mass Flow Rate Comparison for a Small Break Transient in a Typical PWR MOD3.2 and MOD3.3 Intact Loop Accumulator Liquid Volume Comparison for a Small Break Transient in a Typical PWR MOD3.2 and MOD3.3 Broken Loop Accumulator Liquid Volume Comparison for a Small Break Transient in a Typical PWR MOD3.2 and MOD3.3 Core Outlet Void Fraction Comparison for a Small Break Transient in a Typical PWR MOD3.2 and MOD3.3 Intact Loop Steam Generator System Pressure (Steam Dome) Comparison for a Small Break Transient in a Typical PWR MOD3.2 and MOD3.3 Mass Error Comparison for a Small Break Transient in a Typical PWR NUREG/CR-5535/Rev 1-Vol III xx

21 TABLES Page Developmental Assessment Matrix Results for the Branch Tee Problem Results for the Crossflow Tee Problem Turbine Parameters for the Assessment Problem RELAP5/MOD3 Nonequilibrium Model Developmental Assessment Matrix..71 xxi NUREG/CR-5535/Rev 1-Vol III

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23 INTRODUCTION 1 INTRODUCTION The RELAP5 computer code is a light water reactor transient analysis code developed for the U.S. Nuclear Regulatory Commission (NRC) for use in rulemaking, licensing audit calculations, evaluation of operator guidelines, and as a basis for a nuclear plant analyzer. Specific applications of this capability have included simulations of transients in LWR systems, such as loss of coolant, anticipated transients without scram (ATWS), and operational transients such as loss of feedwater, loss of offsite power, station blackout, and turbine trip. RELAP5 is a highly generic code that, in addition to calculating the behavior of a reactor coolant system during a transient, can be used for simulation of a wide variety of hydraulic and thermal transients in both nuclear and nonnuclear systems involving steam-water noncondensable solute fluid mixtures. The MOD3. version of RELAP5 was developed jointly by the NRC and a consortium consisting of several of the countries and domestic organizations that are members of the International Code Assessment and Applications Program (ICAP). The mission of the RELAP5/MOD3. development program was to develop a code version suitable for the analysis of all transients and postulated accidents in PWR systems, including both large- and small-break loss-of-coolant accidents (LOCAs), as well as the full range of operational transients. Although the emphasis of the RELAP5/MOD3. development was on large-break LOCAs, improvements were made to existing code models based on the results of assessments against small-break LOCAs and operational transient test data. Since the completion of the RELAP5/MOD3. development, error correction, refinement, and improvement of model has continued. These improvements have been incrementally included in the MOD3.x releases of the code. RELAP5/MOD3.3 is the latest in this series and this version is the focus of this developmental assessment. RELAP5/MOD3.3 contains new and improved modeling capability as well as additional user conveniences compared to RELAP5/MOD2. The purpose of this volume is to document the developmental assessment problems performed using RELAP5/MOD3.3. The remainder of this section briefly describes the newly developed features and improvements in RELAP5/MOD3.3, the intended application of the code, and its known limitations. Section 2 presents the developmental assessment problems. The problems are divided into three categories: phenomenological problems (Section 2.1); separate effects-problems (Section 2.2); and integral problems (Section 2.3). Finally, Section 3 presents conclusions drawn from the assessment. 1.1 ADDED CAPABILITY OF RELAP5/MOD3.3 The added capability in RELAP5/MOD3.3 since the release of RELAP5/MOD2 includes new modeling, improvements to existing models, and new user conveniences. A detailed description of the MOD3.3 capabilities can be found in Volumes I and II of the RELAP5/MOD3.3 code manual. The new models that have been added since the initiation of the RELAP5/MOD3 development include: A counter-current flow limiting model that uses correlations, which are based on actual geometry and can be activated by the user at each junction in the system model. 1 NUREG/CR-5535/Rev 1-Vol III

24 INTRODUCTION The ECCMIX component for modeling of the mixing of subcooled emergency core cooling system (ECCS) liquid and the resulting interfacial condensation. A zirconium-water reaction model to model the exothermic energy production on the surface of zirconium cladding material at high temperature. A surface to surface radiation heat transfer model with multiple radiation enclosures defined through user input. A level tracking model. A thermal stratification model. Improvements to existing models include: New correlations for interfacial friction for all types of geometry in the bubbly-slug flow regime in vertical flow passages. Use of junction based interphase drag. An improved model for vapor pull-through and liquid entrainment in horizontal pipes to obtain correct computation of the fluid state convected through a break. A new critical heat flux correlation for rod bundles based on tabular data. An improved horizontal stratification inception criterion for predicting the flow regime transition between horizontally stratified and dispersed flow. A modified reflood heat transfer model. Improved logic for vertical stratification inception to avoid excessive activation of the water packing model. An improved boron transport model. A mechanistic separator/dryer model. An improved crossflow model. An improved form loss model. The extension of water packing logic to horizontal volumes. A new default critical flow model (Henry-Fauske). The addition of a simple plastic strain model with clad burst criterion to the fuel mechanical model. The addition of a radiation heat transfer term to the gap conductance model. Modifications to the noncondensable gas model to eliminate erratic code behavior and failure. NUREG/CR-5535/Rev 1-Vol III 2

25 INTRODUCTION Improvements to the downcomer penetration, ECCS bypass, and upper plenum deentrainment capabilities. An improved equation of state that includes the meta-stable regions and uses thermodynamically consistent interpolation. Additional user conveniences include: Modifications that place both the vertical stratification and water packing models under user control so they can be deactivated. Removal of bit packing and vectorization to improve portability and readability. Computer portability through the conversion of the FORTRAN coding to adhere to the FORTRAN 77 standard. Code execution and validation on a variety of systems. The code should be easily installed (i.e., the installation script is supplied with the transmittal) on the CRAY X-MP (UNICOS), DEC station 5 (ULTRIX), DEC Alpha Workstation (OSF/1), IBM Workstation 6 (UNIX), SUN Workstation (UNIX), SGI Workstation (UNIX), and HP Workstation (UNIX). The code has been installed (although the installation script is not supplied with the transmittal) on the IBM 39 (MVS) and IBM-PC (DOS). The code can be installed easily on all 64-bit machines (integer and floating point operands) and any 32-bit machine that provides for 64-bit floating point. 1.2 INTENDED APPLICATIONS OF RELAP5/MOD3.3 RELAP5/MOD3.3 is designed for use in the analysis of pressurized water reactor transients resulting from large- and small-break loss-of-coolant accidents and operational transients such as anticipated transients without scram, loss of feedwater, and turbine trip. Both primary and secondary systems, including balance of plant components, can be modeled. The code has generic modeling capability so that separate effects experiments can also be modeled for use in the assessment of code capability and for extrapolation of separate effects results to integral system behavior. The modeling philosophy to be followed in using RELAP5/MOD3.3 is the same as for MOD2 except where new modeling capability requires special treatment. In general, all modeling capability present in the MOD2 version of the code has been retained in the MOD3 versions, and it has been an objective to keep input decks which have been developed for MOD2 compatible with MOD3.3. With a few exceptions, this has been achieved. 1.3 KNOWN LIMITATIONS OF RELAP5/MOD3.3 This discussion of known limitations of RELAP5/MOD3.3 is based on experiences with previous versions, the developmental assessment work reported herein, and known or suspected limitations of the models used in the code. The Henry-Fauske critical flow model does not consider slip between the phases as it is currently coded. This can lead to an over-prediction of the mass flow rate under critical flow conditions. In addition, the Henry-Fauske model is not coded in the nearly implicit numerical scheme. 3 NUREG/CR-5535/Rev 1-Vol III

26 INTRODUCTION The reflood model developed for RELAP5/MOD3 has shown good agreement with nonuniformly heated rod bundle data with respect to time to maximum temperature. The liquid entrainment appears to be about right, since the liquid inventory in the core is predicted well. However, the axial drag distribution is not correct because an inverted void profile develops at low reflood rates. This results in quench front velocity stalling at about core midplane. The code tends to over-calculate the interfacial heat transfer in mist flow. In the case where the droplet size is small, the surface area is large and the heat transfer between the two phases is over-stated. The energy dissapation due to form loss is not a default in RELAP5. It was included in earlier versions of the code, but was removed because of several problems it caused. It is now a users option (option 41). The pump model does not have an integral model to represent cavitation. The pump model does allow the user the option of accounting for the cavitation or two-phase degradation effects on pump performance through the input of homologous, two-phase curves for head and torque, curves that are in the form of difference curves. This approach is not generally acceptable, particularly when correlations for a given pump are available for determining the cavitating pump head as a function of available net positive suction head. Users have resorted to using the RELAP5 control system, or hardwired updates for a particular pump in order to base the pump head on the net positive suction head. The code has only approximate two- and three-dimensional capability, which must be invoked by using cross-flow junctions to cross connect a matrix of volumes. This approach appears to be adequate for friction-dominated cases. This method is approximate, in that all convected momentum terms are neglected at the cross flow junctions and the primary flow direction only includes the axially convected momentum terms. For horizontal flow, the plug flow regime is not generally available in the code. It is only present in the ECC mixer component. The reactor kinetics model uses point kinetics. There are situations where one- and threedimensional capability is required but is not available in the code. To handle these situations, the PARCS code has been coupled to RELAP5/MOD3.3 to provide 1- and 3-D neutronic effects. 1.4 Developmental Assessment Objective The objective of the developmental assessment is to determine the qualitative and quantitative accuracy of the code for problems that are consistent with the intended application of the code. This is accomplished using three types of problems: phenomenological problems, modeling of separate effects experiments, and modeling of integral experiments. The phenomenological problems are used to demonstrate that the code is in qualitative agreement with the physics of the problem and in cases where analytical solutions exist, the qualitative accuracy of the code can be judged as well. The separate effects tests are designed to provide data on a primary physical effect. These problems are selected to test a key model or models of the code. Qualitative agreement with the data is the first NUREG/CR-5535/Rev 1-Vol III 4

27 INTRODUCTION criteria that must be satisfied, i.e. the correct trends are predicted. Given this, then the code results can be quantitatively compared to the data. The integral problems provide evidence that the collection of models in the code function in concert. Code predictions of intergral system parameters, such as pressure, clad temperature, and mass inventory, are used to assess overall accuracy of the code. The scope of this developmental assessment is limited to the suite of test problems that were previously used to assess the RELAP5 MOD2 and MOD3. versions of the code, and that are not proprietary. This suite of problems provides tests of the major code models. While beyond the scope of this effort, a PIRT analysis to select a suite of problems covering all significant models of the code would improve the level of confidence relative to the general accuracy of the code. 5 NUREG/CR-5535/Rev 1-Vol III

28 INTRODUCTION NUREG/CR-5535/Rev 1-Vol III 6

29 DEVELOPMENTAL ASSESSMENT PROBLEMS 2 DEVELOPMENTAL ASSESSMENT PROBLEMS A total of 34 developmental assessment calculations were performed using RELAP5/MOD3.3. Table 2.-1 shows the developmental assessment matrix, which includes a brief description of the objective of each problem. The matrix contains 1 phenomenological problems, 19 separate-effects problems, and 5 integral test problems. The phenomenological problems are presented in Section 2.1, the separate-effects problems are presented in Section 2.2, and the integral problems are presented in Section 2.3. A code to code or code to data assessment is made for each of the problem types in the development assessment matrix. The codes used in the assessment are RELAP5/MOD3.2 and RELAP5/MOD3.3 (these code versions are referred to throughout the text as MOD3.2 and MOD3.3 respectively). Table 2.-1 Developmental Assessment Matrix Problem Type Nine-Volume Water Over Steam Nitrogen-Water Manometer Problem Assessment Objective(s) Phenomenological Problems Gravitation head effect, Two fluid kinematics Noncondensable state, Oscillatory flow Input Deck Name(s)* ninevol.i manom3.i Branch Reentrant Tee Problem Tee model using branch component brtee.i Crossflow Tee Problem Tee model using crossflow feature crfltee.i Cross Tank problem Crossflow feature, Recirculating flow crosstank.i Three-Stage Turbine Turbine component turbine.i Workshop Problem 2 Workshop Problem 3 Horizontal Stratified Countercurrent Flow Hypothetical two-loop PWR, System Modeling, Control system, Steady-state option Hypothetical two-loop PWR, System Modeling, Control system, Transient option Countercurrent flow model wrkshp_prob2.i wrkshp_prob3.i stwave.i Pryor s Pipe Problem Water packing pryors.i Separate-Effects Problems 7 NUREG/CR-5535/Rev 1-Vol III

30 DEVELOPMENTAL ASSESSMENT PROBLEMS Table 2.-1 Developmental Assessment Matrix (Continued) Problem Type Assessment Objective(s) Input Deck Name(s)* Edwards Pipe Problem Vapor generation model edhtrk.i edhtrkd.i edhtrkn.i Dukler Air-Water Flooding Tests Noncondensable, Interphase drag mode, Countercurrent flow model dukler1.i dukler25.i dukler5.i dukler1.i Marviken Test 24 Subcooled choking model marv24.i Marviken Test 22 Subcooled choking model marv22.i LOFT Test L3-1 Accumulator Blowdown Bennett s Heated Tube Experiments Experiment 5358 Experiment 5294 Experiment 5394 Royal Institute of Technology Tube Test 261 ORNL Bundle CHF Tests Test 3.7.9B Test 3.7.9N Test 3.7.9W ORNL Void Profile Test Test 3.9.1i Christensen Subcooled Boiling Test 15 Shoukri Subcooled Boiling Experiment at Low Pressure MIT Pressurizer Test FLECHT-SEASET Forced Reflood Tests Test 3154 Test 3171 Accumulator model Nonequilibrium heat transfer, Vapor generation model, CHF correlation Nonequilibrium heat transfer, Vapor generation model, CHF correlation Nonequilibrium heat transfer, Vapor generation model, CHF correlation Nonequilibrium heat transfer, Vapor generation model Subcooled boiling model Subcooled boiling model Wall condensation model, Stratified interfacial heat transfer Reflood model l31acc.i ben5294.i ben5358.i ben5394.i rit261.i r261-pg.i 379B.i 379N.i 379W.i or391i.i chris15.i sh3c.i mitst4base.i fs3154.i fs3171.i NUREG/CR-5535/Rev 1-Vol III 8