M amoru Ishii Thermal-Hydraulics and Reactor Safety Laboratory School of Nuclear Engineering Purdue University West Laf

1. Behavior of Containment Emergency Systems:A Review of State of the Art Mamoru Ishii Thermal-Hydraulics and Reactor Safety Laboratory School of Nuclear Engineering Purdue University West Lafayette, IN 47907 Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

2. Contents • Introduction • PUMA-PCCS Separate Effect Test • PANDA ISP-42 Test • JAEA’s Research Project on Horizontal Heat Exchanger for PCCS • Code Evaluations for PUMA Integral Test Including PCCS Performance • Code Capabilities • Summary Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

3. IntroductionPhenomenon 4: Behavior of Containment Emergency Systems • Focus of Phenomenon 4 • Natural Circulation Cooling and Heat Transfer in Various Containment Passive Cooling Systems under Accident Conditions to Remove the Energy out of the Containment by Natural Circulation and Condensation Heat Transfer • Typical Systems are the Tube Condensers Such As the PCCS in SBWR/ESBWR and External Air Cooling Systems in AP600/1000 • Purpose of the Containment Safety Systems • To Protect the Containment under Both DBAs and Severe Accidents • To Prevent the Significant Release of Radioactive Materials to the Atmosphere • Requirements for the Containment Safety Systems • To Remove the Load on the Containment under Accidents • Most of the Load Comes from the Released Steam from the Reactor Primary Coolant System • Noncondensable Gases will Affect the Condenser Efficiency Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

4. IntroductionDefinition of Phenomenon 4 Nuclear power reactor containments are equipped with safety systems which protect the containment integrity under various accident conditions. The focus of Phenomenon 4 is the natural circulation cooling and heat transfer in various containment passive cooling systems under accident conditions to remove the energy out of the containment by natural circulation and condensation heat transfer. Typical systems are the tube condensers such as the Passive Containment Cooling System (PCCS) and external air cooling system or external liquid film cooling and internal condensation of steam in the containment by natural circulation. The major purpose of these containment systems is to protect the containment under both Design Basis Accidents (DBAs) and severe accidents involving serious core damages and to prevent the significant over pressurization and release of radioactive materials to the atmosphere. These systems are required to remove the load on the containment from the Loss of Coolant Accidents (LOCAs) and other accidents by removing the heat but containing the mass within the structure. Most of the load comes from the released steam from the primary coolant system due to the LOCA or venting of the pressure relief valves. The major part of the noncondensable (NC) gases consists of the original containment atmosphere such as air or nitrogen, however with the core damage, hydrogen or fission gases can also be released into the containment atmosphere. The thermal-hydraulic phenomena of importance are tube surface condensation with NC gases, natural circulation of steam and NC gases, degradation of condensation by the accumulation of NC gases and purging of NC gases from condenser systems. The PCCS can be vertical or horizontal tube condensers in external water pool, exposed condenser tube system in the containment cooled by natural circulation water through the tubes from the external pool or by external air circulation and others. Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

5. IntroductionContainment • The Final Barrier Against the Release of Radioactive Materials into the Environment ESBWR Containment Figure Comes from http://www.ne.doe.gov/np2010/pdfs/esbwrOverview.pdf Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

6. IntroductionPassive Safety Systems for Containment • Passive Safety System for Containment Utilize Natural Circulation and Condensation Heat Transfer to Suppress the Pressure and Temperature of the Containment Atmosphere. • Examples: • Vertical PCCS Condenser (GE’s SBWR/ESBWR); Leonardi et al., 2006 • Horizontal PCCS Condenser (Japan); Kondo et al., 2006 • External Air Cooling System (Westinghouse’s AP600/1000); Sha et al., 2004 Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

7. IntroductionTest Facilities • GIRAFFE Full height, 1/400 volume ratio test facility simulates GE’s SBWR; Performed separate effects and system response tests; • PANDA Full height, 1/25 volume ratio test facility simulates GE’s SBWR; Performed tests to check the effects of the NC concentration and pool inventory; Performed tests to investigate the PCCS start-up and long-term capabilities; • PANTHERS Full size prototypic heat exchangers; Performed tests under the same thermal hydraulic conditions as GIRAFFE and PANDA; • PUMA 1/4 height ratio, 1/400 volume ratio test facility simulates GE’s SBWR/ESBWR; Performed separate effects tests to check the effects of the NC concentration; • LSTF Full height, 1/48 volume ratio test facility simulates Westinghouse PWR; • Horizontal Heat Exchanger Test Facility in JAEA Halved full height, prototypical-scale bundle to represent one of four HEXs in ABWR-II; Single-tube test and tube bundle test have been performed; Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

8. IntroductionContributions from the CRP Participants • PUMA-PCCS Separate Effect Test Purdue University, USA; 3rd RCM • Overview on PANDA Test Facility and ISP-42 PANDA Tests Data Base Paul Scherrer Institute, Switzerland; 2nd RCM • JAEA’s Research Project on Horizontal Heat Exchanger for PCCS Japan Atomic Energy Agency, Japan; 2nd RCM • Code Evaluations for PUMA Integral Test Including PCCS Performance Purdue University, USA; 2nd RCM Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

9. PUMA-PCCS Separate Effect TestObjectives • To study the effect of noncondensable gas concentration on PCCS performance using PUMA for three operational modes: • Bypass Mode • Cyclic Venting Mode • Long-Term Cooling Mode • To study the effect of PCCS inlet (Drywell) pressure • To study the effect of PCCS inlet flow rate • To study the effect of PCCS pool water level • To compare the test results with other data Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

10. PUMA-PCCS Separate Effect TestTest Facility Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

11. z PCCS Condenser Venting Frequency Decreases Vent Line Drain Line from DW Horizontal Vent SP SP SP PUMA-PCCS Separate Effect TestPCCS Operational Modes • Blowdown period before GDCS water injection to RPV • Continuous flow through NC gas vent line • After GDCS water injection and restart boiling in RPV • NC gas accumulation in the PCCS • Final phase of LOCA • Low NC gas fraction (< 1%) • Venting frequency approaches zero (ideal case) Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

12. PUMA-PCCS Separate Effect TestTest Matrix Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

13. Pure Steam Steam-Air Mixture PUMA-PCCS Separate Effect TestEnergy Balance • Energy Balance • NC Gas Effect on Energy Balance • As NC gas fraction increases, NC gas purging with steam increase increase of mass and energy discrepancies between PCCS inlet and outlet Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

14. Cyclic Venting Bypass Vierow PUMA-PCCS Separate Effect TestNC Effect on Heat Transfer Coefficient • NC Gas Effect on Average HTC • As NC gas fraction increases, HTC decreases due to the increase of thermal resistance in the gas phase. • As NC gas fraction increases, the discrepancies between PUMA, Kuhn and Vierow’s data decrease, because the NC venting condition is getting closer to a single-tube experiment (flow-through mode) as NC gas fraction increases. Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

15. PUMA-PCCS Separate Effect TestCombined Correlated Test Results • Combined correlated data compared with PUMA-PCCS test data • Both cyclic venting and bypass with continuous venting follow the same trends as other research results: • PUMA and PANTHERS data fall on the same heat transfer coefficient trend line. Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

16. PANDA ISP-42 Test Review Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

17. JAEA’s Horizontal Heat Exchanger Test Review Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

18. Code Evaluation for PUMA Integral TestPUMA Facility Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

19. Code Evaluation for PUMA Integral TestRELAP5 Nodalization Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

20. Code Evaluation for PUMA Integral TestMSLB Test Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

21. Code Evaluation for PUMA Integral TestTest Data vs. RELAP5 Simulation Downcomer Collapsed Water Level Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

22. Code Evaluation for PUMA Integral TestTest Data vs. RELAP5 Simulation GDCS Loop A Injection Flow Rate Drywell Pressure Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

23. Code Evaluation for PUMA Integral TestTest Data vs. RELAP5 Simulation Decay Heat Removal, Test Data Decay Heat Removal, RELAP5 Prediction Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

24. Code Evaluation for PUMA Integral TestSummary of Code Evaluation • RELAP5 was Evaluated PUMA Model of RELAP5 • Code Modeling Problems • Suppression Pool Condensation Model 1-D Model Insufficient Recirculation Pass  Instability Artificial Flow Restriction Required • Suppression Pool Over Stratified 1-D Model Insufficient Containment Pressure Affected • Film Condensation Model (PCCS & ICS) Insufficient Modeling Accuracy Effect of Noncondensable Gas not Certain Containment Pressure Affected • Overall Performance All Major Events and Trends Predicted Facility and Code Scaling Capability are Good PCCS Condensation Rate Improvement Desired RELAP-5 appears Adequate for DBA Analysis Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

25. Code CapabilitiesHeat Transfer Models in Thermal Hydraulic System Code • TRACE Mode 4.0 • Wall Vapor HTC: Nusselt and empirical model • Wall Liquid convention HTC: flow factor F:Chen correlation • Interfacial Heat Transfer: Empirical model of Sklover & Rodivilin Developed for cross-flow of gas-vapor mixtures on liquid jets • RELAP Mode 3.3 Beta • Default Model: Saha, Nusselt • Alternative model: UCB model Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

26. Code CapabilitiesRELAP5 Calculation Comparing with PUMA Experimental Data • Default Model: Under predict • UCB model: over predict and fluctuation • Kuhn Correlation: Fluctuation • One of main reason of disagreement is caused by limitation of NC gas venting phenomenon in SP water in the code calculation Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

27. Code CapabilitiesCode Calculation Comparing with PUMA Experimental Data Table. Cooling Capability Comparison (inlet steam flow/condensate water flow) • When steam with NC flow more discrepancy between experimental data and code calculation in terms of condensate water flow and temperatures inside condenser tubes; It may be caused by condensation model limitation in calculating NC gas profile in the condenser tubes and venting of gas to suppression pool water. • Saturation temperatures of inlet, outlet and condenser tubes are over-predicted compared to experimental data when the NC gas exiting. • Code couldn’t simulate cyclic venting of the NC gas. Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

28. Code CapabilitiesCode Calculation Comparing with PANDA ISP-42 Data • Overall Best Results were Obtained by the Lumped Parameter Code SPECTRA; • System Codes like CATHARE and RELAP5 Produced Acceptable Results; • Containment Code COCOSYS Produced Acceptable Results; • CATHARE and RELAP5 have the Flexibility to Simulate Special Components, like the PCCS; • Strict QA Procedure for Nodalization and Input Deck Generation should be Followed; • Appropriate Input Parameters should be Given in the Code Analysis; • Lumped Parameter Approach should be Chosen for Simple Physical Situations (i.e., PCCS Start-up Phase); • Further Assessment of the 3-D Models and Advanced Modeling Features are Necessary; • No CFD Codes Calculation has been Submitted. Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

29. Code CapabilitiesCode Calculation and Development in JAEA’s Research Project • Code Development • RELAP5 for Primary Side Calculation • ACE-3D was Developed for Secondary Side Calculation • RELAP5 and ACE-3D Coupled at the Condenser Tube Surface • Calculation Successfully Predicted the Distribution of the Quality in the Condenser Tubes and the Void Fractions in the Bundle Side • RELAP5 MOD3 Modification • Heat Transfer Package was Developed for Modeling the Horizontal Heat Exchanger • Model Predicted Well the Total Heat Removal Rate and the Pressure Drop Across the Heat Exchanger Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

30. Summary • Introduction to Phenomenon 4 • Review of the Recent Research Results on PCCS • PUMA Separate Effect Test (Vertical PCCS for SBWR/ESBWR) • PANDA ISP-42 Test (Vertical PCCS for SBWR/ESBWR) • Horizontal Heat Exchanger Test in JAEA (Horizontal PCCS for ABWR-II) • Review of the Code Capabilities • System Code (RELAP5, TRACE) • System Code Coupled with 3-D CFD Code (RELAP5/ACE-3D) Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University