Text

Nonproprietary AP600 LOFTRAN-AP & LOFTTR2-AP Final Verification & Validation Rept
	ML20086M916
Person / Time
Site:	05200003
Issue date:	07/31/1995
From:	; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	;
Shared Package
ML20086M898	List: ML20086M916; NRC-95-4512, Forwards Nonproprietary WCAP-14308, AP600 LOFTRAN-AP & LOFTTR2-AP Final Verification & Validation Rept;
References
	WCAP-14308, WCAP-14308-R, WCAP-14308-R00, NUDOCS 9507250050
	Download: ML20086M916 (180)
	v • d • e

_ . _ _ _ .

IAE PeormIETARY CIAN 3

' WCAP-14308 f

I AP600 LOFTRAN AP AND LOFTTR2-AP FINAL VERIFICATION AND VALIDATION REPORT July,1995 C 1995 Westinghouse Electric Corporation All Rights Reserved 9507250050 950719 PDR ADOCK 05200003 A PDR m:W2061-oon\2%1 w.gu.e:lt>470795 REVISION: 0

TABLE OF CONTENTS Section Title Pm S UMM ARY . . . . . ... ....... ....... ... ..... ..... . .. ........I

1.0 INTRODUCTION

. . . . . . . . ... .............. . .................... 1-1 1.1 Use of LOFTRAN in AP600 Safety Analysis ............. ........... 1-1 1.2 Verification and Validation of LOFTRAN-AP and LOFITR2-AP . . .. .... . 1-2 1.2.1 Summary of Key Phenomena for Validation . . . . . . . . . . . . . . . . . .. 1-2 ,

1.2.2 Tests Used for Validation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -3 ,

2.0 LOFTRAN-AP AND LOFITR2-AP CODE DESCRIFTION . . . . . .............. 2-1 2.1 O ve rview . . . . . . . . . . . . . . . . . . . . . . . .. ........ .... . .. .... 2-1 2.2 Code Modifications . . . . . . . . . . . . . . . . . . . . . . ........ . . ..... 2-2 2.3 LOFTRAN-AP Code Modifications for Test Simulations . . . . .... .... .. 2-5 3.0 ROLES OF TESTS IN LOFTRAN-AP AND LOFITR2-AP CODE VALIDATION . . . . ....... ........ .............. ..... . 3-1 3.1 Overview . . . . . . . .. .............. ....... ........ .... . 3-1 3.2 Role of CMT Component Tests in LOFTRAN-AP Validation . .......... 3-2 3.2.1 CMT Component Tests Description . . . . . . . .. ............. . 3-2 3.2.2 Role of the CMT in AP600 Safety Analysis with LOFTRAN-AP . . . . . 3-3 3.2.3 CMT Component Test Results Used . . . . . . . . ..... ... ..... 3-3 3.3 Role of SPES-2 Tests in LOFTRAN-AP Validation . ... .. . ...... .. 3-4 3.3.1 SPES-2 Tests Description . . . . . . . . . . . . . . ............... . 3-4 3.3.2 SPES-2 Test Results Used ... .... .... .................. 3-4 4.0 LOFTRAN-AP CMT MODEL AND VALIDATION . . . ............ ... ... 4-1 4.1 Validation Approach ..................... ..,....... ........ 4-1 4.2 Key Phenomena . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ....... ..... 4-1 4.3 LOFTRAN-AP CMT Component Test Facility Model . . . ............... 4-2 4.3.1 Boundary Conditions . . . . . . . . . . . . . . . . ...... ............ 4-3 4.3.2 Input Deck .................... ................ .... 4-3 4.3.2.1 Geometrical Data . . . . . . . . . .... ............... 4-3 4.3.2.2 Friction Factors of the Lines . . . . . . ................ 4-4 4.3.2.3 Metal Parameters . . . . . . . . ......................44 4.4 Analytical Simulations . . . . ........... .......................... 4-6 4.4.1 Cold Inlet Balance Line (Cases A and B) . . . . . . . . . . . . . ....... 4-6 1 4.4.2 liot Inlet Balance Line . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... . 4-7 )

4.4.3 Iland Calculations of Momentum and Energy Balance . . . . . . . . . . . . . 4-8 1 4.4.4 Conclusions - Analytical Simulations . . ...... . . . . . . . . . . . 4- 10 4.5 500-Series Tests Simulations ... ........ .... .... ........ . . 4-10 4.5.1 C064506 Test ... ......... .. . ........ ......... 4-10 4.5.1.1 Description of the Runs . . . . . ......... . . . . . . . . 4-10 4.5.1.2 Calculation Results . . . .. .. .... ... . . . . . . . . . 4- 1 1 4.5.2 C072509 Test ..... . ...... . . ..... . ... ...... 4-13 4.5.3 500-Series Tests Conclusk,as ... .... ........ . .... . 4-14 4.6 Assessment of CMT Component Test Cimuldon Results . ........... 4-15 m:\a;+00C061 -non\2061 w.non: I Mno795 iji REVIsloN: 0 s

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TABLE OF CONTEN'IS (Cont.)

Section Title Page ;

r I

5.0 LOFTRAN SPES-2 MODEL AND INTEGRAL SYS' EMS VALIDATION . . . . . . . . . 5-1 i 5.1 Validation Approach . . . . . . . . . . . . . . . . . . ............... ........ 5-1 ,

5.1.1 Steam Generator Tube Rupture . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 .

5.1.2 Steam Line Break ...................................... 5-2 5.2 Key Phenomena . . . . . . . . . . . . . . . . . . ........................... 5-2 5.2.1 Steam Generator hbe Rupture . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-2 5.2.2 Main Steam Line Break . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 5.3 LOFTRAN-AP SPES-2 Model Description ........................... 5-5 j 5.3.1 Primary System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-5 :

5.3.1.1 Power Channel Pressure Vessel and Reactor Coolant !

Loop Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-5 l 5.3.1.2 Power Channel Rod Bundle . . . ..................... 5-6 :

5.3.1.3 Reactor Coolant Pumps and Loop Flow Model . . . . . . . . . . 5-6 5.3.1.4 Pressurizer and Surge Line . . . . . . . . . . . . . . . . . . . . . . . . 5-7 5.3.1.5 Vessel Head Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-8 5.3.2 Secondary Coolant System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-9 5.3.2.1 Steam Generators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-9 5.3.2.2 Steam Generator hbe Rupture Break Flow Model . . . . . . . 5-10 '

5.3.2.3 Steam Pipe Break Flow Model . . . . . . . . . . . . . . . . . . . . . 5-10 5.3.3 Passive Safety Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-11 5.3.3.1 CMT System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 1 1 ;

5.3.3.2 Passive Residual Heat Removal System Heat Exchanger . . . 5-12 ;

5.3.4 Heat Losses Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-12 i r

5.3.4.1 Primary System Heat Losses Model . . . . . . . . . . . . . . . . . 5-13 l 5.3.4.2 Pressurizer Heat Losses . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-13 5.3.4.3 SGs Secondary-Side Heat Losses Model . . . . . . . . . . . . . . 5-14 l 5.3.4.4 CMTs Heat Losses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-14 !

5.4 Test-Specific LOFTRAN-AP Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-17 I 5.4.1 Steam Generator hbc Rupture Test-Specific Input . . . . . . . . . . . . . . . 5-17 ;

5.4.2 Main Steam Line Break LOFTRAN-AP Test-Specific Input . . . . . . . . . 5 19 5.5 Test Simulation Results: Test 9,10, and 11 - Steam Generator Tube Rupture . 5-22 5.5.1 O vervi ew . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-22 5.5.2 Matrix Test 10 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-24 5.5.2.1 Descripdon of the Runs . . . . . . . . . . . . . . . . . . . . . . . . . . 5-24 '

5.5.2.2 Results Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-25 5.5.2.2.1 Initial Case (Run 1) . . . . . . . . . . . . . . . . . . . . . 5-25 5.5.2.2.2 Sensitivity Studies . . . . . . . . . . . . . . . . . . . . . . 5-30 5.5.2.3 Conclusions Concerning Test 10 . . . . . . . . . . . . . . . . . . . 5-33 5.5.3 Matrix Test 9 ........................................5-117 5.5.3.1 Description of the Runs . . . . . . . . . . . . . . . . . . . . . . . . . . 5-117 ,

5.5.3.2 Results Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-117 5.5.3.2.1 Base Case (Run 1) . . . . . . . . . . . . . . . . . . . . . . 5-117 5.5.3.2.2 Sensitivity Study . . . . . . . . . . . . . . . . . . . . . . . 5-120 5.5.3.3 Conclusion Concerning Test 9 . . . . . . . . . . . . . . . . . . . . . 5-120 m Amp 6000061.non\2061 w. non:Ib-07(7795 jy REVISION: 0 ,

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Section Title Page 5.5.4 Matri x Test 11 . . . . . . . . . . . . . . . . . . . . . . ..... .. . . . . . 5-175 5.5.4.1 Description of the Runs . . . . . . . . ...... .... ... . 5-175 5.5.4.2 Results Analyses . ........................... 5-176 5.5.4.2.1 Blind Simulations . . . . . .... . . . . . . . . . . . . 5- 17 6 5.5.4.2.2 Blind Simulation After Updating of the Input Data (Run 1) . .............. ........ 5-177 5.5.4.3 Conclusion Concerning Test 11 . . . . . . . . . . . . . . . . . . 5- 181 5.6 Test Simulation Results: Test 12 - Main Steam Line Break . . ........... 5-234 5.6.1 Pre-Data Release Simulations . . . . . . ... ........ . ...... . 5-234 5.6.2 Post-Data Release Simulation . . . .... .............. ..... 5-234 5.7 Assessment of SPES-2 Simulation Results . .. .. ............ . . . 5-268 5.7.1 Steam Generator 'Ibbe Rupture Test Simulations . . ... . .... . . 5-268 5.7.2 Main Steam Line Break Test Simulation . . . . . . . ..... . . . . 5-269 6.0

SUMMARY

OF THE LOFTRAN CODE VALIDATION EFFORT . . . . . . ........ 6-1 6.1 Role of LOFTRAN in Safety Analysis . ...... ..................... 6-1 6.2 Adaptation of LOFTRAN and LOFTRAN Based Safety Methodology to Advanced Passive Plant Designs . . . . . . . . . . . . . . . . . . . . ........... . 6-1 6.3 Code Validation Tests . . . . . . . . . . . . . . . . . . . . ....... . ........ 6-2 6.3.1 SPES-1 Natural Circulation Tests . . . . . . . ........ ........ . 6-2 6.3.2 PRHR Tests .. ..... .. ........ ........... ....... 6-3 6.3.3 CMT Tests . . . . . . . . . . . . . . . . . . . . ...... ............... 6-3 6.3.4 SPES-2 Tests . . .. ... ... . ........ .............. 6-4 6.3.4.1 SGTR Test Simulations . . . . ... .... . . . . . . . . . . . . 6-4 6.3.4.2 MSLB Test Simulation ........... ...... .... .. 6-4 6.4 LOFIRAN Application Envelope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-5 6.5 Conclusions . . . . ........ ...... .. ......... ............. 6-5

7.0 REFERENCES

. . . . . . . . .......... .................. ..... ....... 7-1 APPENDIX A CMT COMPONENT TESTS A.1 CMT Component Test Facility Description ..... ........... .........A-1 A.2 CMT Component Tests Used for LOFTRAN Code Validation . . . . . . . . . . . . . . A-2 A.3 Waltz Mill CMT Test Results and Data Used . . . . . . . . . . . . . . . . . . . . . . . . . . A-2 APPENDIX B SPES-2 TESTS B.1 SPES-2 Test Facility Description . . ..... ..... ..................B-1 B.1.1 Introduction . . . . . ...... .......... ........ ...... . . B-1 B.I.2 Facility Scaling Summary . . ....................... . . . B-1 B.I.3 Facility Description . . . . . . . . .... . ................ . . . . . B-2 i B.1.4 Instrumentation Data Acquisition System . ... .... ...... . . . . B-3 i B.I.5 Controt Loops . . . . . . . . . . . ............... .. . . . . . . . . . . . B -4 B.2 SPES-2 Tests Used for LOFTRAN Code Validation . . . . . . . . . . . . . . . . . . . . . B s l

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$10 TABLE OF CONTENTS (Cont.)

Section Title ._Page B.3 Test Descriptions ............................................. B-6 B.3.1 Design-Basis Steam Generator Tube Rupture with Nonsafety Systems Operational and Operator Action for Mitigation (SPES.2 Matrix Test S01309) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B -6 B.3.2 Design Basis Steam Generator Tube Rupture Without Nonsafety Systems and No Operator Action to isolate the Steam Generator (SPES-2 Matrix Test S01110) . . . . . . . . . . . . . . . .. . . . .. . . . . ' B-8 B.3.3 SGTR With Inadvertent ADS Actuation (S01211) . . . . . . . . . . . . . . . . . . B-9 B.3.4 Large Steam Line Break at Hot Standby Conditions With Passive Safety Systems (S01512) .....................................B-10 m:\ap60N061-mon \2061w. mon:Ib 070795 vi REVI$loN: 0

, --s - - _

,g LIST OF TABLES

.T.Eh!t Il'l!! .Pagt 1-1 Phenomena Identification Ranking Table for AP600 Non-LOCA and Steam Generator Tbbe Rupture Design Basis Analyses . . . . . . . . . . . . . . . . . . .

1-5 3-1 500 CMT Test Series - Tests Selected for Simulation . . . . . . . . . . . . . . . . . . 3-6 4-1 Phemmena Identification for the AP600 CMT . . . . . . . . . . . . . . . . . . . . ... 4-17 4-2 CM'E Water Node Sizes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-18 .

'4-3 CMT Steel Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-19 4-4 Metal Parameters - Variable Volume Noding . . . . . . . . . . . . . . . . . . . . . . . . 4-20 4-5 Metal Parameters - Equal Volume Noding . . . . . . . . . . . . . . . . . . . . . . . . . . 4-21

, 4-6 Analytical Simulations - Run Descriptions . . . . . ..................... 4-22 4-7 Test C064506 and C072509 - Run Parameters . . . . . . . . . . . . . . . . . . . . . . . 4-23 5.3-1 RCS Heat Losses at 605'F . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-16 5.3-2 SG Secondary-Side Heat Losses, Including Steel Inertia . . . . . . . . . . . . . . . . 5-16 5.5.1-1 SGTR Matrix Tests 9,10,11 - Refinements Between Preliminary and Final Validation Report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-23 5.5.2-1 Sensitivity Studies for Matrix Test 10 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-35 5.5.2-2 Comparison of Test and LOFITR2.AP Initial Conditions for Matrix Test 10 . . 5-36 5.5.2-3 Sequences of Events for Matrix Test 10 - Initial Case (Run 1) . . ........ 5-37 5.5.3-1 Comparison of Test and LOFTIR2-AP Initial Conditions for Matrix ') .,st 9 . . . 5-122 5.5.3-2 Manual SG PORV and ADS Valve Actuation Sequence . . . . . . . . . . . . . . . . 5-123 5.5.3-3 Sequence of Events for Matrix Test 9 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-126 5.5.3-4 Operator Actions for Matrix Test 9 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-127 5.5.4-1 Comparison of Test and LOF1TR2-AP Initial Conditions for Matrix Test 11 .. 5-182 5.5.4-2 Sequence of Events for Matrix Test 11 - Blind Simulation . . . . . . . . . . . . . . . 5-183 5.5.4-3 Sequence of Events for Matrix Test 11 - Run 1 . . . . . . . . .. . . . . . . . . . . . . . 5-184 5.5.4-4 Sequence of Events for Matrix Test 11 - Run 2 . . . . . . . . . . . . . . . . . . . . . . 5-185 5.5.4-5 .SGTR Matrix Test 11 - Evolution Between the Blind Simulation, Run 1 and R un 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 18 6 5.6-1 Comparison of Test LOFTRA-AP Conditions for Matrix Test 12 . . . . . . . . . . 5-237 5.6-2 Sequence of Events for Test S01512 (Matrix Test 12) . . . . . . . . . . . . . . . . . . 5-238 5.7-1 Assessment of SPES-2 Simulation Results . . . . . . . . . . . . . . . . . . . . . . . . . . 5-272 A1 CMT 500-Series Tests Used for LOF TRAN-AP Validation . . . . . . . . . . . . . . . . A-3 B-1 Comparison of Specified and Actual Test Conditions for S01309 (Matrix Test 9) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-13 B-2 Sequence of Events for Test S01309 (Matrix Test 9) . . . . . . . . . . . . . . . . . . . . B-15 B-3 Comparison of Specified and Actual Test Conditions for S01110 (Matrix Test 10) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B -16 B-4 Sequence of Events for Test S01110 (Matrix Test 10) . . . . . . . . . . . . . . . . . . . B-18 B-5 Comparison of Specified and Actual Test Conditions for S01211 (Matrix Test 11) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-19 B-6 Sequence of Events for Test S01211 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-21 B7 Comparison of Specified and Actual Test Conditions for S01512 . . . . . . . . . . . B-22 B-8 Sequence of Events for Test S01512 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-24 m:\ap6002061 mon \2061w. son:1W70795 yjj REVistoN: 0

LIST OF FIGURES Firure Title Page 1-1 AP600 Passive Cooling System ... ....... ........ .... ... .. . 1-6 3-1 CMT Test Facility and AP600 Plant . . .. ......................... 3-7 3-2 AP600 Core Makeup Tank Test Piping and Instrumentation Diagram . . .... . 3-8 3-3 SPES-2 Test Facility Schematic . . . .......... .......... ....... 3-9 4-1 Deviation of Boundary Conditions and other Variables for CMT Component Test Facility Model ............................ 4-24 4-2 CMT Component Test Facility LOFTRANCMT Noding . . . . . . . . . . . . . . . . 4-25 4-3 Analytical Simulation, Cold Inlet Balance Line Injection Line Flow Rate . . . . . 4-26 4-4 Analytical Simulation, Cold inlet Balance Line CMT Fluid Temperature, 4.9 in. from the Top of the CMT . . . . ...... ........... ....... 4-27 4-5 Analytical Simulation, Cold Inlet Balance Line CMT Fluid Temperature, 11.6 in. from the Top of the CMT . . . ... . ............. 4-28 4-6 Analytical Simulation, Hot Inlet Balance Line injection Line Flow Rate . .... 4-29 4-7 Analytical Simulation, Hot Inlet Balance Line CMT Fluid Temperature, 11.6 in. from the Top of the CMT . . . .... .. .............. . . . . 4-30 4-8 Analytical Simulation, Hot Inlet Balance Line CMT Fluid Temperature, 101.6 in. from the Top of the CMT . . . .. ......................... 4-31 4-9 Analytical Simulation, Hot Inlet Balance Line Water to CMT Wall Heat Flux . 4-32 4-10 Analytical Simulation, Hot Inlet Balance Line CMT Wall-to-Air Heat Flux . . . 4-33 4-11 C064506 Test Injection Line Flow Rate . . .. . . ........ ... ..... 4-34 '

4-12 C064506 Test Pressure at the Reservoir Top . . . . . . . . . . ......... .. 4-35 4-13 C064506 Test C'TP Inlet Fluid Temperature . ... . . . . . . . . . . . . . . . . . . 4- 3 6 4-14 C064506 Tes'. Chir Outlet Fluid Temperature . . .................... 4-37 4-15 C064506 Test CMT Fluid Temperature 4.9 in. from the Top of the CMT . . . . 4-38 4-16 C064506 Test CMT Fluid Temperature 53.1 in. from the Top of the CMT ... 4-39 4-17 C064506 Test CMT Fluid Temperature,101.6 in. from the Top of the CMT . . 4-40 4-18 C064506 Test Injection Line Flow Rate . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-41 4-19 C064506 Test Pressure at the Reservoir Top . . . . . ............. .... 4-42 4-20 C064506 Test CMT Wall Teat Transfer . . . . . . . . . ........ ....... .. 4-43 4-21 C064506 Test CMT Water-to-Wall Heat Transfer . . . . . . . . . . . . . . . . . . . . 4-44 4-22 C064506 Test CMT Inlet Fluid Temperamre . . . . . . . . . . . . . . . . . . . . . . 4 -45 4-23 C064506 Test CMT Outlet Fluid Temperature . . . . . . . . . . . . ... . ... . 4-46 4-24 C064506 Test CMT Fluid Temperature,4.9 in. from the Top of the CMT . . . . 4-47 4-25 C064506 Test CMT Fluid Temperature,53.1 in. from the Top of the CMT . . . 4-48 4-26 C064506 Test CMT Fluid Temperature,101.6 in. from the Top of the ChfT , 4-49 4-27 C064506 Test Injection Line Flow Rate . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-5 0 4-28 C064506 Test Pressure at the Reservoir Top . .... .................4-51 4-29 C064506 Test CMT Wall Heat Transfer . . . . . . . . . . . . . . . . . . . .. . . . . 4-52 4-30 C064506 Test CMT Water-to-Wall Heat Transfer . . . . . . . . . . . . . . . ... . 4-53 4-31 C064506 Test CMT Inlet Fluid Temperature . . . . . . . . . . . . . . . .... 4-54 4-32 C064506 Test CMT Outlet Fluid Temperature . . . . . . . . . . . . . . . . . .... 4-55 4-33 C064506 Test CMT Fluid Temperature,4.9 in. from the Top of the CMT . . . 4-56 :

4-34 C064506 Test CMT Fluid Temperature,53.1 in. from the top of the CMT . . . 4-57 {

4-35 COMSM Test CMT Fluid Temperature,101.6 in. from the Top of the CMT .. 4-58 1 4-36 C064506 Test Injection Line Flow Rate . . . . . . . . . . ....... .... .... 4-59 4-37 C064506 Test CMT Water-to-Wall Heat Transfer . .. .. ..... .. .. 4-60 4

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4 LIST OF FIGURES (Cont.)

Fleure Title _P_ age 4-38 C054506 Test CMT Outlet Fluid Temperature . . . ......... ..... ... 4-61 4-39 C0645% Test CMT Fluid Temperature,4.9 in. from the Top of the CMT . . . . 4-62 4-40 C064506 Test injection Line Flow Rate . . . . . . . ......... ......... 4-63 4-41 C064506 Test CMT Water-to-Wall Heat Transfer . . . . . . . . . . . . . . . . . . . . . 4-64 4-42 C064506 Test CMT Outlet Fluid Temperature . . . . ....... .......... 4-65 4-43 C064506 Test CMT Fluid Temperature,4.9 in from the Top of the CMT . . . . 4-66 4-44 C072509 Test injection Line Flow Rate . . . . . . . . . . . . . ........ ..... 4-67 4-45 C072509 Test Pressure at the Top . . . . . . . . . . . . . . . . . . . . . . . .. .. .. 4-68 4-46 C072509 Test CMT Wall Heat Transfers . . . . . . . . . . . . . . . . . . . . . ..... 4-69 4-47 C072509 Test CMT Water-to-Wall Heat Transfer . . . . . . .............. 4-70 4-48 C072509 Test CMT Inlet Fluid Temperature . . . . . . . ................ 4-71 4-49 C072509 Test CMT Outlet Fluid Temperature . . . . . . . . . . . . . . . . . . . . . . 4-72 4-50 C072509 Test CMT Fluid Temperature,4.9 in. from the Top of the Chit . . . . 4-73 4-51 C072509 Test CMT Fluid Temperature,53.1 in. from the Top of the CMT . . . 4-74 4-52 C072509 Test CMT Fluid Temperature,101.6 in. from the Top of the CMT . . 4-75 4-53 C072509 Test hijection Line Flow Rate . . . . . . . . . . . . . . . . ..... .. ... 4-76 4-54 C072509 Test CMT Water-to-Wall Heat Transfer . . . . . . . . . . . . . . . . . . . . . 4-77 4-55 C072509 Test CMT Outlet Fluid Temperature . . . . . . . . . . . . . . . . . . . . . . 4-7 8 4-56 C072509 Test CMT Fluid Temperature,4.9 in. from the Top of the CMT . . . . 4-79 5.5.2-1 Test S01110 - Core Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 -3 8 5.5.2-2 Test S01110 - Pressurizer Pressure . . . . . . . . . . . .. .......... . . . . 5-39 5.5.2 3 Test S01110 - SG-A Pressure . . . .......... . . . . . . . . . . . . . . . . . . . 5-40 5.5.2-4 Test S01110 - SG-B Pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-41 5.5.2-5 Test S01110 - Tube Rupture Break Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-42 5.5.2-6 Test S01110 - Primary Side Hot Leg Temperature . . . . . . . . . . . . . . . . . . 5-4 3 5.5.2-7 Test S01110 - Primary Side Inlet Temperature . . . . . . . . . . . . . . . . . . . . . . 5-44 5.5.2-8 Test S01110 - Pressurizer Liquid Level . . . . . . .. . . . . . . . . . . . . . . . . . . . . . 5-45 5.5.2-9 Test 501110 - PRHR Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-4 6 5.5.2-10 Test S01110 - PRHR Inlet Temperature . . . . . . . . . . . ................ 5-47 S.5.2-11 Test 501110 - PRHR Outlet Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-48 5.5.2-12 Test S01110 - CMT Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 -4 9 I 5.5.2-13 Test S01110 - Upper Head Mass and Level . . . . . . . . . . . . ..... ..... . 5-50 )

l 5.5.2-14 Test S01110 - CMTs Fluid Mass . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-51 l l 5.5.2-15 Test S01110 - RCS Steel Heat Transfer . . . . . . . . . . . . . . . . . . . . ....... 5-52 l l 5.5.2-16 Test S01110 - SGs Fluid to Steel Heat Transfer . . . . . . . ....... . . . . . . 5-53

]

5.5.2-17 Test S01110 - Integrated Break Flow ............................5-54 5.5.2-18 Test S01110 - Pressurizer Pressure . . . . . . .......... . . . . . . . . . . . . . 5-55 5.5.2-19 Test S01110 - SG-A Pressure . . . . . . . . . . . . . . . . . . .... . .... . . . . 5-56 j 5.5.2-20 Test S01110 - SG-B Pressure . . . . ......... . . . . . . . . . . . . . . . . . . . 5 -5 7 I 5.5.2-21 Test S01110 - Tube Rupture Break Flow . . . ............... . . . . . . . 5-5 8 5.5.2-22 Test S01110 - Primary Side Hot leg A Temperature . . . . . . . . . . . . . . . . . . . 5-59 5.5.2-23 Test S01110 - Primary Side Hot Leg B Temperature . . . . . . . . . . . . . . .... 5-60 i l

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I mAap6000061 -nonC061 w. mon:l b-070795 ix REVIs10N: 0 l

l

LIST OF FIGURES (Cont.)

Figure Title ,P, age, 5.5.2-24 Test S01110 - Primary Side Inlet Temperature . . . . . . . . . . . . . . . ....... 5-61 5.5.2-25 Test 501110 - Pressurizer Liquid Level . . . . . . . . . . . . . . . . . . . .... . . 5-62 5.5.2-26 Test S01110 - PRHR Flow . . . . . . . . . ....... .......... . . . . . . . 5 -6 3 5.5.2-27 Test S01110 - CMT Flow ..... ...............................564 5.5.2-28 Test S01110 - Upper Head Mass . . . . . . . . ... ............... . 5-65 5.5.2-29 Test S01110 - Integrated Break Flow ..................... . ..... 5-66 5.5.2-30 Test S01110 - Pressurizer Pressure . . . . . . . . . . . . . . ....... . ...... 5-67 5.5.2-31 Test S01110 - SG-A Pressure . . ................................ 5-68 5.5.2-32 Test 501110 - SG-B Pressure . . . . . . . . . . . . ...... ...... ... .. . 5-69 5.5.2-33 Test 501110 - Tube Rupture Break Flow . .. ......................5-70 5.5.2-34 Test S01110 - Primary Side Hot Leg A Temperature . . . . . ........ .... 5-71 5.5.2-35 Test S01110 - Primary Side Hot Leg B Temperature . . . . ..... ........ 5-72 5.5.2-36 Test S01110 - Primary Side Inlet Temperature .... ................. 5-73 5.5.2-37 Test Soll10 - Pressurizer Liquid Level . . . . . ..... . .......... . . 5-74 5.5.2-38 Test S01110 - PRHR . . . . . . . . . . . . . . ........ ......... . . . . . . . 5-75 5.5.2-39 Test S01110 - CMT Flow ..... .... ... ... . .. ...... .... 5-76 5.5.2-40 Test S01110 - Upper Head Mass . . . .... . . ... . . . . . . . . . . . . . . 5 -77 5.5.2-41 Test S01110 - Integrated Break Flow . .... ...................... 5-78 ;

5.5.2-42 Test S01110 - Pressurizer Pressure . . . . . .... . .. ........ . . . . . . 5-79 1

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5.5.2-43 Test 501110 - SG-A Pressure . . . . . . . . ... .... ............. . 5-80 i 5.5.2-44 Test S01110 - SG-B Pressure .......... .. .. ........... ..... 5-81 5.5.2-45 Test S01110 - Tube Rupture Break Flow . . . . . ..... ... ........ . 5-82 5.5.2 do Test S01110 - Primary Side Hot Leg A Temperature . . . . . . . . . . . . . ..... 5-83 5.5.2 r.' Test S01110 - Primary Side Hot Leg B Temperature . . . . . . . . . . ........ 5-84 l 5.5.*A 8 Test S01110 - Primary Side Inlet Temperature . . . . . . . . . . . . . . ....... 5-85 5.5).-49 Test S01110 - Pressurizer Liquid Level . . . . . . . ............ . ..... 5-86 l

5.5.2-50 Test S01110 - PRHR Flow . . . ................... ......... .... 5-87 5.5.2-51 Te.st S01110 - CMT Flow . ........ . ............. . . . . . . . . . 5 -8 8 5.5.2-52 Test S01110 - Upper Head Mass . . . . . . . . . ... ............ ..... 5-89 i

5.5.2-53 Test S01110 - Integrated Break Flow ..... . .. ............ .... 5-90 l 5.5.2-54 Test S01110 - Pressurizer Pressure . . . . . . . .... .......... ........ 5-91

( 5.5.2-55 Test S01110 - SG-A Pressure . . . . . . . . . . . . . . .. ............ .... 5-92 5.5.2-56 Test S01110 - SG-B Pressure . . ................... . . . . . . . . . . . 5-9 3 5.5.2-57 Test S01110 - Tube Rupture Break Flow . . . . . . . . . . ...... . . . . . . . . 5-94 5.5.2-58 Test S01110 - Primary Side Hot Leg A Ternperature . . . .............. . 5-95 5.5.2-59 Test 501110 - Primary Side Hot Leg B Temperature . . . . . . . ...... ,. . 5-96 5.5.2-60 Test S01110 - Primary Side Inlet Temperature . . . . . . . . . . ... . . . . . 5-97 5.5.2-61 Test S01110 - Pressurizer Liquid Level . . . ................. ... . . 5-98 5.5.2-62 Test 501110 - PRHR Flow . . . . . . . . . . . . .. ... .. ....... ... . 5-99 5.5.2-63 Test S01110 - CMT Flow ............... .... .... . ... .. 5-100 l

5.5.2-64 Test S01110 - Upper Head Mass . . . . . . . . . .. ... . ...... .... . 5-101 1 5.5.2-65 Test S01110 - Integrated Break Flow ....... ........ ... . . . . 5-102 5.5.2-66 Test S01110 - Pressurizer Pressure . . . . ..... ... .. ........ . 5-103

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1 LIST OF FIGURES (Cont.) I l

Firure Title _Page 5.5.2-67 Test S01110 - SG-A Pressure . ....... ... .. ...... .. .... .. 5-104 5.5.2-68 Test S01110 - SG-B Pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 105 5.5.2-69 Test S01110 - Tube Rupture Break Row . . . . . ............. . . . . . . . 5-106 5.5.2-70 Test S01110 - Primary Side Hot Leg A Temperature . . . . . . . . . . . . . . . . . . 5 - 107 5.5.2-71 Test S01110 - Primary Side Hot Leg B Temperature ...... . . . . . . . . . 5-108 5.5.2-72 Test S01110 - Primary Side Inlet Temperature . ......... ........ .. 5 109 5.5.2-73 Test S01110 - Pressurizer Liquid Level . . . . ....... ... .. .. . . . . 5-110 5.5.2-74 Test S01110 - PRHR Flow . . . . . . . . . . . . . . . . ... . . . . . . 5-111 5.5.2-75 Test S01110 - CMT Flow ........ . .. ........ . . ..... . 5-112 5.5.2-76 Test S01110 - Upper Head Mass . . . . . . . . . .. ............. . . . . . 5-113 5.5.2-77 Test 501110 - Integrated Break Flow ............. .......... .... 5-114 5.5.2-78 Test S01110 - RCS Water to Steel Heat Transfer . . . . . . . . . . . . . . . . . . . 5 - 1 15 5.5.2-79 Test 501110 - RCS Steel to Air Heat Transfer ... ....... . . . . . . . . 5-116 5.5.3-1 Test S01309 - Core Power . . . . . . . . . . . . .. .......... ... ..... 5-128 5.5.3-2 Test SO1309 - Pressurizer Pressures . . . . . . . . . . . . . . .... . . . . . . . . 5-129 5.5.3-3 Test S01309 - SG-A Pressure . . . ..... ...... .. ..... . .. . 5-130 5.5.3-4 Test S01309 - SG-B Pressure . .. ..... . .. . . . . . . . . . . . . . . . . . . 5- 131 5.5.3-5 Test S01309 - hbe Rupture Break Flow . ..... .... ... .... . 5-132 5.5.3-6 Test S01309 - Primary Side Hot Leg Temperature . .. . ............. 5-133 5.5.3-7 Test S01309 - Primary Side Inlet Temperature . . . . . . . . . . . . . . . . . . . . 5-134 5.5.3-8 Test SO1309 - Pressurizer Liquid Level . . . ... .. .......... . 5-135 5.5.3-9 Test SO1309 - PRHR Flow . . . . ... .................... . . . . . 5-136 5.5.3-10 Test S01309 - PRHR Inlet Temperature . . . .. ..... ..... ...... 5-137 5.5.3-11 Test S01W - PRHR Outlet Temperature . . . . ......... ... ...... 5-138 5.5.3-12 Test sol 309 - CMT Flow . . . . . . . . ........... ... ..... . . . . . 5-139 5.5.3-13 Test S01309 - Upper Head Mass and Level . . . .............. . . . . . . 5-140 5.5.3-14 Test S01309 - CMTs Fluid Mass .. ............. ......... . .5-141 5.5.3-15 Test SO1309 - RCS Steel Heat Transfer ................ ..... . . . 5-142 5.5.3-16 Test S01309 - SGs Fluid to Steel Heat Transfer . . . . . . . . . . . . . . . . . . . . . . 5-143 5.5.3-17 Test sol 309 - Integrated Break Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-144 j 5.5.3-18 Test S01309 - Starting Feedwater Flow - Loop A .................... 5-145 5.5.3-19 Test SO)309 - Starting Feedwater Flow - Loop B . . . . . . . ....... . . . . . 5-146 5.5.3-20 Test S01309 - PORV Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-147 5.5.3-21 Test S01309 - Integrated PORV Flow . . . . . . . ... . . . . . . . . . . . . . . . . . 5 - 14 8 5.5.3-22 Test S01309 - ADS Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-149 5.5.3-23 Test sol 309 - Integrated ADS Flow . . . . . .......................5-150 :

5.5.3-24 Test S01309 - Pressurizer Pressure . . . . . . . . . . . . ............... . . 5-151 5.5.3-25 Test S01309 - SG-A Pressure . . . . . . . .... .... ........... . . . . 5-152 I 5.5.3-26 Test S01309 - SG-B Pressure . . . . . . . . . . . . . . ...... ............ . 5-153 1

5.5.3-27 Test S01309 - Abe Rupture Break Flow ..... ................ ... 5-154 5.5.3-28 Test S01309 - Primary Side Hot Leg A Temperature . . . . . . . . . . . . . . . . . 5-155 l 5.5.3-29 Test S01309 - Primary Side Hot Leg B Temperature . . . . . . . . . . . . . . . . 5- 15 6 l l

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LIST OF FIGURES (Cont.)

FE' ur_e Title _Pgg -

5.5.3-30 Test S01309 - Primary Side Inlet Temperature . . . . . . . . . . . . . . . . . . . . . .. . 5-157 5.5.3-31 Test S01309 - Pressurizer Liquid Level . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-158 !

5.5.3-32 Test S01309 - PRHR Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-159 5.5.3-33 Test S01309 - CMT Flow ............ ..... ....... . . . . . . . . . . 5- 160 !

5.5.3-34 Test S01309 - Upper Head Mass . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 161 5.5.3-35 Test S01309 - Integrated Break Flow ............... .. . . . . . . . . . . 5-162 5.5.3-36 Test S01309 - Pr, ssurizer Pressures . . . . . . . . . . . .. . . . . . . . . . . . . . . . . 5- 163 7 5.5.3-37 Test S01309 - SG-A Pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 164 e t

5.5.3-38 Test S01309 - SG-B Pressure . . . .................. . . . . . . . . . . . . 5- 165 '

5.5.3-39 Test S01309 - Tube Rupture Break Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-166 !

5.5.3-40 Test S01309 - Primary Side Hot Leg A Temperature . . . . . . . . . . . . . . . . . . 5-167 ,

5.5.3-41 Test S01309 - Primary Side Hot Leg B Temperature . ... ... ..... .. 5-168 '

5.5.3-42 Test S01309 - Primary Side Inlet Temperature . . . . . . . . . . . . . . . . . . . . . . 5-169 ;

5.5.3-43 Test S01309 - Pressurizer Liquid Level . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-170 5.5.3-44 Test S01309 - PRHR Flow . . . . . . . . . . ......................... 5-171 l 5.5.3-45 Test S01309 - CMT Flow ......... .................. . . . . . . . . 5-172 .

5.5.3-46 Test S01309 - Upper Head Mass . . . . . . . . . . . . . . . ....... . . . . . . . . . 5-173 5.5.3-47 Test S01309 - Integrated Break Flow ....... ..................... 5-174 5.5.4-1 Test S01211 - Core Power . . . . . . . . . . . . . . . . . . ................ . 5-187 5.5.4-2 Test 501211 - Pressurizer Pressure . . . . . . . . . .....................5-188 5.5.4-3 Test S01211 - SG-A Pressure . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . 5 189 5.5.4-4 Test S01211 - SG-B Pressure ......... ....... . . . . . . . . . . . . . . . . 5 - 190 5.5.4-5 Test S01211 - Tube Rupture Break Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-191 5.5.4-6 Test S01211 - Primary Side Hot Leg Temperature . . . . . . . . . . . . . . . . . . . . 5-192 5.5.4-7 Test S01211 - Primary Side Inlet Temperature ....... .............. 5-193 5.5.4-8 Test S01211 - Pressurizer Liquid Level . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 194 5.5.4-9 Test S01211 - PRHR Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-195 5.5.4-10 Test S01211 - PRHR Inlet Temperature . . . . ...... . . . . . . . . . . . . . . . . 5- 196 5.5.4-11 Test S01211 - PRHR Outlet Temperature . . . . . . . . . . . . . . . . . . . . . . . . . 5-197 5.5.4-12 Test S01211 - CMT Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 198 5.5.4-13 Test S01211 - RCS Steel Heat Transfer . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-199 5.5.4-14 Test S01211 - SGs Fluid to Steel Heat Transfer . . . . . . . . . . . . . . . . . . . . . . 5-200 5.5.4-15 Test S01211 - Integrated Break Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-201 5.5.4-16 Test S01211 - Core Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-202 5.5.4-17 Test S01211 - Pressurizer Pressure ..............................5-203 5.5.4-18 Test S01211 - SG-A Pressure ........... ...................... 5-2 M 5.5.4-19 Test S01211 - SG-B Pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-205 5.5.4-20 Test S01211 'Ibbe Rupture Break Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-206 5.5.4-21 Test S01211 - Primary Side Hot Leg Temperature . . . . . . . . . . . . . . . . . . . . 5-207 5.5.4-22 Test S01211 - Primary Side Inlet Temperature . . . . . . . . . . . . . . . . . . . . . . 5 -208 5.5.4-23 Test S01211 - Pressurizer Liquid Level . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-209 5.5.4-24 Test S01211 - PRHR Flow . . . . . . . . . . ............. . . . . . . . . . . . . 5-210 m:\ap60m2061 -non\2061 w.non:Ib-070795 xii REVislON: 0

LIST OF FIGURES (Cont.)

Firure Title _P_ age 5.5.4-25 Test S01211 - PRHR Inlet Temperature . . . . .. . . . . . . . . . . . . . . . . 5 -21 1 5.5.4-26 Test S01211 - PRHR Outlet Temperature . . . .......... . ....... . 5-212 5.5.4-27 Test S01211 - CMT Flow ..... . .... .. ......... ... . . . . . 5-213 5.5.4-28 Test S01211 - RCS Steel Heat Transfer . . . . . . . . . . . . . . . . . . .. ..... . 5-214 5.5.4-29 Test S01211 - SGs Fluid to Steel Heat Transfer . . . . . . . . . . . . . . . . . . . . . 5-215 5.5.4-30 Test S01211 - Integrated Break Flow . ..... .. .. . . ... . . . . 5-216 5.5.4-31 Test S01211 - Core Power . . . . . . . ... ........ . . ....... . . 5-217 5.5.4-32 Test S01211 - Pressurizer Pressure . .. .......... ..... ... .. 5-218 5.5.4-33 Test S01211 - SG-A Pressure . . . ..................... . ... . 5-219 5.5.4-34 Test S01211 - SG-B Pressure . . . . . . . . . . .... ... . . . . ..... 5-220 5.5.4-35 Test S01211 - Tube Rupture Break Flow . . . . . ... ........... .. . 5-221 5.5.4-36 Test S01211 - Primary Side Hot Leg Temperature ... ............ .5-222 5.5.4-37 Test S01211 - Primary Side inlet Temperature . . . . . . . . .. ......... . 5-223 5.5.4-38 Test S01211 - Pressurizer Liquid Level . . . . ... .. .. ... .... ... 5-224 5.5.4-39 Test S01211 - PRHR Flow . . . . . . . . . . . . . ... . . . ..... . .... 5-225 5.5.4-40 Test S01211 - PRHR Inlet Temperature . . .. .... .. . . . . . . . . . . . . 5 -226 5.5.4-41 Test S01211 - PRHR Outlet Temperature . . . .... .... . . ...... 5-227 5.5.4-42 Test S01211 - CMT Flow . ..... ... . ... .. ....... ...... . 5-228 5.5.4-43 Test S01211 - RCS Steel Heat Transfer . . ...... . .............. 5-229 5.5.4-44 Test S01211 - SGs Fluid to Steel Heat Transfer . . . . . . .. . . . . . . . . . . . 5 -2 30 5.5.4-45 Test S01211 - Integrated Break Flow ... ...... . . . . . . . . . . . . . . . 5 -2 31 5.5.4-46 Test S01211 - ADS Flow . . ............ ........ . . . . . . . . . . . . 5-232 5.5.4-47 Test S01211 - Integrated ADS Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-2 3 3 5.6-1 Test 501512 - Pressurizer Pressure . . . . . . . . . ... ....... . . . . . . . . . 5-2 39 5.6-2 Test S01512 - SG-A Pressure . ................................5-240 5.6-3 Test S01512 - SG-B Pressure . . . . . . . . . . . . . . .............. . . . . 5-241 5.6-4 Test S01512 - PRHR Flow . ...................................5-242 5.6-5 Test S01512 - CMT Flow . . . . . . . . . . . . ..................... .. 5-243 5.6-6 Test S01512 - Pressurizer Pressure . . . . . . . . . . . . . . . . . . . ... ... . . . 5-244 5.6-7 Test S01512 - SG-A Pressure . . . . . . ... ...... .. .......... .. 5-245 5.6-8 Test S01512 - SG-B Pressure . . . .. . ..... ..... ........ . . . 5-246 5.6-9 Test S01512 - PRHR . . . . . . . . . . . . . . . . .......... ............. 5-247 5.6-10 Test S01512 - CMT Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-248 5.6-11 Test S01512 - Integrated Break Flow .............................5-249 5.6-12 Test S01512 - SG Inlet Header Temperature - Faulted Loop . . . . . . . . . . . .5-250 5.6-13 Test S01512 - SG Outlet Header Temperature - Faulted Loop . . . . . . . ... .5-251 5.6-14 Test S01512 - Pressurizer Pressure . . . . . . . ................... ... 5-252 5.6-15 Test S01512 - SG-A Pressure . . ... . .. ........ . .......... .5-253 5.6-16 Test S01512 - SG-B Pressure . . . . . . . . .. ... .. .. ... . . . . . . 5-254 5.6-17 Test S01512 - PRHR Flow . . . ... ........... ................ 5-255 5.6-18 Test S01512 - CMT Flow . ........... ........ ..... ..... . 5-256 5.6-19 Test S01512 - integrated Break Flow ........................ ... 5-257 in Aapb00\2061.non\2061 w.non:ltr070795 xiji REVislON: 0

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LIST OF FIGURES (Cont.) ;

r Firure Title h 5.6-20 Test S01512 - SG Inlet Header Temperature - Faulted Loop . . . . . . . . . . . . . 5-258 ;

5.6-21 Test S01512 - SG Outlet Header Temperature - Faulted Loop . . . . . . . . . . . . 5-259 l 5.6-22 Test S01512 - Pressurizer Pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-260 5.6-23 Test S01512 - SG-A Pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-261 ,

5.6-24 Test S01512 - SG-B Pressure . . . . . . . . . . . . . . . . . . . ..... . . . . . . . . . 5-262 i 5.6-25 Test S01512 - PRH R Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . S-263 '

5.6-26 Test S01512 - CMT Flow .....................................5-264 5.6-27 Test S01512 - Integrated Break Flow .............................5-265 ,

5.6-28 Test S01512 - SG Inlet Header Temperature - Faulted Loop . . . . . . . . . . . . . 5-266 {

5.6-29 Test S01512 - SG Outlet Header Temperature - Faulted Loop . . . . . . . . . . . . 5-267

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SUMMARY

nis report provides details of the LOFTRAN-AP simulations of the CMT and SPES-2 tests used to validate the LOFTRAN CMT model and integral plant response during transient situations. he CMT model validation is based on separate-effects tests conducted at the CMT test facility. This report also presents a summary of related efforts for the PRHR component and SPES-1 natural-circulation tests, which support code validation.

While not a primary purpose of this report, the validation exercises presented in the report support existing LOFTRAN models.

The LOFTRAN code is used to calculate NSSS transients given a set of boundary conditions and a transient-forcing function. The code simulates the transient based on user-supplied input. By specifying minimum- or maximum-initial conditions, safety system setpoints, relief and safety valve capacities, core kinetics' parameters, and safeguards system, thermal-hydraulic performances, the code supplies conservative and bounding analysis results. The transient forcing functions, such as the steam break model, also contain conservative modeling assumptions or are supplied with conservative input parameters to achieve a conservative system response.

Code inputs are based on design data and where applicable, uncertainties are included and applied in the direction, which provides conservative response relative to acceptance criteria or safety-analysis limit. The safety-analysis limit includes margin to design limits. The overall approach then includes conservative models, minimum- or maximum-code input values, and margin in the acceptance criteria giving an overall conservative result. The LOFTRAN code is not intended for use where significant, two-phase flow occurs. Transients which employ LOFTRAN, in general, do not exhibit significant, two-phase flow conditions. If there is potential for two-phase flow, acceptance criteria are established based on prohibiting large-scale RCS boiling. Additionally, LOFTRAN is not used where CMT draindown could occur, for post-trip ADS, during IRWST injection phase, or for long-term cooling phases.

LOFIRAN-AP code simulations of SPES-2 SGTR test data show that LOFTRAN accurately predicts the operational behavior of the passive systems as well as the integrated plant operation under transient conditions. Comparison to test data shows good agreement with all key parameters identifled in the PIRT. In particular, the code accurately predicts CMT and PRHR behavior. Combined with analysis assumptions, which maximize break flow; it is apparent that LOFTTR2-AP provides a good model for conservative design-basis SGTR calculations.

Comparison of LOFTRAN-AP simulations with the SPES-2 MSLB test data demonstrated that i

LOFIRAN-AP accurately predicts the overali transient trends and provides conservative results I suitable for design-basis safety analyses. LOFTRAN provided valid predictions of CMT and PRHR f

behavior for the SPES-2 MSLB test. l an:\apb00\2061 -non\2061 w.non :l t>-071 195 1 REVisloN: 0 l l

. . _ . . - - . . _ . - _ - . - - _ . _ . _ _ . . - . _ - - . . - . _ . . ~ . . .

. 1 l

Based on code-validation efforts documented in this report and the related efforts documented in i Appendix ISB, Revision 0 of the SSAR, it is concluded that the LOFTRAN-AP code provides an l accurate model of the AP600 plant over the range of conditions required for the analysis of design- l basis non-LOCA and SGTR events. In conjunction with conservative input parameters based on established safety ar.alysis methodologies, LOFTRAN provides an excellent tool for design-basis safety ;

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1.0 INTRODUCTION

This report presents the final results of the LOFTRAN-AP and LOFTTR2-AP code verification and validation effort, supporting analyses contained in the AP600 Standard Safety Analysis Report (SSAR).

The documentation extends and supersedes the LOFTRAN-AP code validation effort presented in the preliminary Core Makeup Tank (Reference 1) and SPES-2 (Reference 2) Validation reports. Included l

in this document are the final LOFTRAN-AP simulations of the CMT tests performed at the Waltz l Mill CMT test facility, the steam generator tube ruptnre (SGTR), and the main steam line break (MSLB). The simulations are compared with the test data (including blind test data) and the results demonstrate the ability of LOFTRAN-AP and LOFTTR2-AP to model AP600 non-LOCA and SGTR transients.

The report also provides background information about the LOFTRAN-AP and LOFTTR2-AP codes and other verification and validation work.

1.1 Use of LOFTRAN in AP600 Safety Analysis LOFTRAN* is a digital computer code that was developed to simulate transient behavior in a multiloop pressurized water reactor system. The code simulates a multiloop system by modeling the reactor core and vessel, hot and cold leg piping, steam generator (SG) tube and shell sides, pressurizer, and reactor coolant pumps (RCPs) with up-to-four coolant loops. The code has an extensive history in performing design- and licensing-basis non-loss-of-coolant accident (non-LOCA) analyses. The code has been reviewed and approved for use in the non-LOCA analyses by the United States Nuclear Regulatory Commission (US NRC).*

LOFTTR2 is a specialized version of the LOFTRAN code, modified for the analysis of steam generator tube rupture (SGTR) events. The main reactor coolant system (RCS) models of LOFITR2 are the same as those specified in the LOFTRAN code. Additionally, LOFTTR2 includes an enhanced SG secondary-side model, a tube rupture break flow model, and improvements to allow simulation of operator actions. 'Ihe code is documented and has been reviewed and approved by the US NRC for SG tube rupture analyses."^')

A key feature of the AP600 plant is the addition of passive safeguards systems. The passive residual heat removal (PRHR) heat exchanger (HX) functions as a passive alternative to the auxiliary feedwater system. The core makeup tanks (CMTs) function as a passive emergency coolant injection and boration system (See Figure 1-1).

The application of LOFTRAN and LOFITR2 to the AP600 design is described in detail in the LOFTRAN and LOITTR2 AP600 Code Applicability Document.* The AP600-specific versions of these codes are designated LOFfRAN-AP and LOFITR2-AP. Modifications to the codes include the addition of the PRHR, core makeup tank (CMT), and reactor vessel head vent models. Due to the major commonalities of the LOFFRAN-AP and LOFTTR2-AP codes, whenever LOFFRAN-AP maap600C061 -nonu061 w.noo :lb-070795 11 REV!sloN: 0

models are referred to in this report, except as otherwise noted, the applicability of the LOFITR2-AP models is implied. A full history of LOFTRAN-AP and LOFITR2-AP is provided in Section 2 of this repon.

1.2 Verification and Validation of LOFTRAN AP and LOFITR2-AP Verification of LOFTRAN-AP and LOFITR2-AP comprises a series of checks desigr.ed to provide confidence that the computer codes are correctly solving and applying the equations and correlations within them. In particular, the checks are aimed at confirming correct implementation of the code modifications associated with the AP600 design and the validation effort.

I De verification process is controlled by internal Westinghouse procedures. Compliance with these procedures demonstrates adequate verification of LOFTRAN-AP and LOFITR2-AP. Verification is not discussed funher in this report. l l

Validation of LOFTRAN-AP and LOFITR2-AP is accomplished by comparing code simulations with scale model test data. He comparisons are aimed at showing that the key phenomena for AP600 transient analyses are correctly modeled by the codes. The phenomena and the tests are briefly described below. .

1.2.1 Summary of Key Phenomena for Validation Table 1-1 presents the phenomena identification ranking table (PIRT) for the AP600 non-LOCA events that can be analyzed by the LOFTRAN-based code. Table 1-1 ranks the importance of various component or system phenomena to specific events; phenomena that show an H indicate high importance, those with an M are of moderate importance, and those marked with an L are oflow importance. In some cases, a phenomena may not be applicable to a transient, and this is indicated as N/A.

He importance rankings of Table 1-1 are based on the analysis time frame as presented in Chapter 15 of the AP600 SSAR. De design-basis analyses results of Chapter 15 Eenerally cover the transient until a safe state is reached. For many ANSI 18-2 Condition 11 events, the initiating fault may be quickly terminated by an automatic protection system action. For example, a fault that causes inadvertent rod cluster control assembly (RCCA) withdrawal from at-power condition will cause reactor power to increase until an overpower reactor trip occurs. De reactor trip causes an immediate reduction in power and also terminates the inadvertent RCCA withdrawal. Immediately following a reactor trip, the plant will be in a safe state and the plant may be maintained in a stable state or cooled down funher using normal plant shutdown procedures. He analytical results presented in Chapter 15 for this event only cover the event from initiation until shonly after reactor trip, and only the phenomena considered applicable over this time frame are considered in the PIRT.

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Inspection of the AP600 phenomena identifled in Table 1-1 indicates that many are the same as those I 1

for conventional pressurized water reactors (PWRs); however, a key difference between AP600 and J conventional PWRs is the increased importance of natural circulation flow and related phenomena. )

l De PIRT identifies the key phenomena of interest for the non-LOCA and SGTR events. The key phenomena relevant to the MSLB and SGTR tests and simulations include:

. Natural circulation flow and heat transfer

. Break flow from the ruptured tube (SGTR events)

Break flow from the faulted SG (MSLB events)

= Steam generator secondary-side conditions j

. Decay power of the core !

. Mixing in the reactor vessel l

. Pressurizer response Two other important phenomena associated with the AP600 passive systems are:

. CMT recirculation !

. PRHR recirculation and heat transfer l l

1.2.2 Tests Used for Validation Validation of the LOFTRAN-AP PRHR model, CMT model, and integral AP600 plant response with these passive safeguards systems,is based on the following tests:

. SPES-1 natural circulation tests

PRIIR component tests

. CMT component tests

. SPES-2 steam generator tube rupture and steam line break tests Comparisons of the SPES-1 natural circulation and PRHR tests to LOFTRAN-AP simulations have been completed and are presented in Appendix 15B, Revision 0 of the SSAR.

The CMT component tests provide data on CMT recirculation behavior, which can be used to validate i the LOFTRAN-AP CMT model.

De SPES-2 SGTR and MSLB tests provide integral-systems-effects data, which can be used to validate the LOFTRAN-AP code for the key phenomena identified in the PIRT, as well as to confirm the behavior of the passive safety system models in these events.

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M roles of the SPES-1, PRHR CMT and SPES-2 tests in LOFTRAN-AP and LOFITR2-AP

, validation effort are further discussed in Section 3, and the CMT and SPES-2 validation results are i presented in Sections 4 and 5.

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il TABLE 1-1 g PIIENOMENA IDENTIFICATION RANKING TABLE FOR AP600 NON-LOCA AND I STEAM GENERATOR TUBE RUlrrURE DESIGN BASIS ANALYSES 8

h Component & System i%ima % (1) (2) 0) (4) (5) (6) (7) (8) (9) (10) (11) (12) (13) (14)

{ FW III SLB Inad- LOL Loss ILB LOSS LR SUIL RWAP Inad- RCS SGTR g Malf vertent ac d &

a vertent Dep, u PRHR & RCS BS CMT LONF How a g CVS e

Critical How N/A N/A H N/A N/A N/A N/A N/A N/A N/A 11 N/A M ll Vessel 11 L II Il L M M L L H L M L M Mixing Hashing in Urger llead N/A N/A M L N/A L L N/A N/A N/A N/A L L L Core M H M L M 11 11 M M II M L L L Reactivity Feedback Renaw Trip H L II II II II

{ 11 11 H 11 H II H II Decay Heat L L L L II II L L 11 L L H L 11 Forced Convection II il 11 H H 11 11 II H 11 II M 11 L Natwal Circulation How and M L 11 11 L 11 H L L L L H L M Heat Transfer RCP Coastdown Performance L N/A L L L L L H 11 N/A N/A L L L Presswizer L L M M L M L L L L L M L M Pressurizer Huid Level Surge Line Pressure Drop L L L L II L L M H L L L L L Steam Generator (SG) Heat Transfer II 11 11 L H H !! L L L M L L M Secondary Conditions M L 11 L L M M L L L L L L H Is M

C

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9 TABLE 1-1 (Cont.)

g PHENOMENA IDENTIFICATION RANKING TABLE FOR AP600 NON-LOCA AND s STEAM GENERATOR TURE RUPTURE DESIGN BASIS ANALYSES S

y Component & System Phenomenon (1) G) (3) (4) (5) (6) G) (8) (9) (10) (11) (12) (13) (14)

FW ELI SLB Inad- LOL less H.B LOSS LR SUIL RWAP Inad- RCS SGTR f

n Malf vertent PRIIR ac d

RCS

& vertent Dep.

~ BS ChR LONF Flow w CVS S

S RCS Wall Stored Heat L L L L N/A L L N/A N/A L N/A L L M CMT N/A N/A H 11 N/A H M N/A N/A N/A N/A H N/A L Recircidation Injection Gravity Draining injection N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Vapor Condensation Rate N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Balance une Presstre Drop N/A N/A 11 11 N/A II M N/A N/A N/A N/A H N/A L Balance Line Initial Temperature N/A N/A H II N/A 11 M N/A N/A N/A N/A H N/A L Dist.

Accumulators N/A N/A M N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Injection Flow Rate PRIIR N/A N/A L H N/A 11 11 N/A N/A N/A N/A II N/A 11 Flow Rate and Ifeat Transfer

,1) tw Maia -

Feedwater Malunction that Kesults in a peacase in t cedwater h-r ature or an inuense as teedwater How Q) ELI -

Excessive Increau in Secondary Steam Flow (3) SLB -

Steamline Break (4) Inadvertent PRilR - Inadvertent Operatio of the PRHR (5) LOL -

1 mss d Secondary SWe Lead Events ,

(6) Loss ac & LONF Less of ac Power and Loss of Normal Feedwater G) 11.B -

Feed Une Break (8) Loss of RCS Flow - Los, d Forced RCS Flov (9) LR & BS - lect;ed RCP Rotor and Bioken RCP Shaft (10) SUIL -

Startup of an Inactive Ren ser Coolant Pump at an Incorrect Temperature l

(1I) RWAP -

RCCA Withdrawal at Pow tr (12) Inad-vertent CMT or CVS - Inadvertent Operation of ti e ChR w Chemical end Volume Control System (13) RCS Dep. -

Inadvertent RCS Depresstnization (14) SGTR Steam Generator Tube Rug ture ,

$ H - High Inpostance M - Moderate Importance L - low Importance N/A - Not Agplicable 3

o 4

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V Pressurizer r' 3 4$ RHR HX O O/ e W Sparger yt g S

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3 I

2.0 LOFTRAN-AP AND LOFTTR2-AP CODE DESCRIPTION

)

2.1 Background '

LOFTRAN is used for non-LOCA analyses presented in Chapter 15 of SSAR, simulations of anticipated transients without trip, equipment sizing studies, and calculating mass and energy releases front secondary-side breaks. ,

Development of the LOFTRAN code began in the late 1960s. Initially the code contained a single l RCS loop and was used for the analysis of symmetric design-basis non-LOCA transients. Analyses of only asymmetric system transients was performed with other computer codes.

In 1976, LOFTRAN was modified to explicitly simulate four RCS loops. 'Ihe modification resulted in use of LOFTRAN for the RCS transient response to symmetric and asymmetric non-LOCA design '

basis transient analyses. Selected Chapter 15 safety analysis events analyzed with LOFTRAN include:

Feedwater system malfunctions resulting in a decrease in feedwater temperature or an increase in feedwater flow

Excessive increase in steam flow Inadvertent opening of a steam generator relief or safety valve

Steam system piping failure '

Inadvestent operation of the passive residual heat removal heat exchanger

Loss of external electrical load

Tbrbine trip '

Inadvertent closure of main steam isolation valves

Loss of ac power to the plant auxiliatics

Loss of normal feedwater flow

Feedwater system pipe break

Partial and complete loss of forced reactor coolant flow

= Reactor coolant pump shaft seizure (locked rotor) and shaft break ,

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Startup of an inactive reactor coolant pump at an incorrect temperature

. Inadvertent operation of the core makeup tanks (CMrs) during power operation

Chemical and volume control system malfunction that increases reactor coolant inventory

Inadvertent opening of a pressurizer safety or inadvertent operation of the ADS

Steam generator tube rupture ne original validation program for LOFFRAN included comparisons of LOFTRAN results to actual plant data and to other similar thermal-hydraulic codes (Reference 2). Rese comparisons consisted of fourteen transients that included simulations of actual plant loss of loads, reactor trips and load step changes. Rese comparisons were used to demonstrate the ability of LOFTRAN to analyze non-LOCA transients.

As part of the original validation process an analysis of the R. E. GINNA SG~lR event of January 25, 1982, was performed and submitted to the NRC in Reference 23. Shortcomings of LOFTRAN were identifled by this comparison and demonstrated that LOFTRAN was able to model the GINNA tube rupture event prior to the failure of the PORV to close.

To better analyze tube rupture events, a specialized version of LOFTRAN called LOFITR2 was developed. LOFTTR2 was developed by modifying LOFFRAN in two stages. He LOFITR2 program is an updated version of the LOFITRI program, which was developed from the LOFTRAN program for SGTR analysis, and was used for the generic SGTR evaluation of previous Westinghouse PWR designs. The original LOFITRI program was subsequently modified to model SG overfill and was designated as LOFITR2, which was then used for the evaluation of the consequences of overfill in Reference 7.

He LOFITR2 program is identical to the LOFITRI program, with one exception. De LOFITR2 program has the additional capability to represent the transition from two regions (steam and water) on the secondary side to a single water region if overflll occurs, and to model the transition back to two regions again, if indicated by the calculated secondary conditions.

2.2 Code Modifications for the AP600 The AP600 deMgn is similar in many respects to previous PWR designs. The analysis of many of the design basis events is unimpacted by AP600 features and the safety analysis methods used on previous PWRs remain applicable. 'Ihe principle new features that impact the non-LOCA safety analyses are the passive residual heat removal heat exchanger and the core makeup tanks. These two components provide the safety related methods for decay heat removal and boration of the RCS. The LOFTRAN and LOFITR2 codes were modified to incorporate models of those AP600-specific features.

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Core Makeup Tank Model The core makeup tank model is a multi-node model that simulates the tank, the balance line connecting the reactor coolant system cold leg with the top of the CMT, and the injection line connecting the bottom of the CMT with the reactor vessel. The thermal-hydraulics model simulates the flow in the CMT lines and tracks mass, energy, and boron concentration in the CMT. The CMT model calculations are performed explicitly from the RCS thermal-hydraulic conditions. A single CMT is used to simulate two CMTs by doubling the flow rates into and out of the CMT model.

Fluid noding in the CMT model is as follows:

[

]"

Ileat transfer from the tank fluid through the walls of the tank is simulated and [

]" are used.

Boron concentration is tracked on a node basis in the cold leg balance line and the injection line. In the CMT, boron is tracked on a tank average basis, which effectively assumes perfect mixing of the boron within the tank with fluid entering from the cold leg balance line. This assumption conservatively underpredicts the boron concentration of the CMT injection. More details of the CMT model can be found in Reference 8.

PRHR and IRWST Model

'the passive residual heat removal (PRHR) heat exchanger model is divided into the following regions:

I l

]"

Up to [ ]" nodes can be simulated in these five regions. The inlet and outlet piping regions are ,

l simulated as [ ]" and the inlet and outlet header and channel head regions are simulated as [ ]" The heat exchanger region is set up to model either vertical or C-tube type heat exchangers. User input allows specification of whether a heat exchanger node is vertical or horizontal, but [

]" Depending upon the orientation of the PRHR nodes, different heat transfer correlations are l used. No heat transfer is simulated in the inlet and outlet regions. ,

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E- ,

t Heat removed from the PRHR is transferred to the in-containment refueling water storage tank

' (IRWST) which is modeled as [ ]'* Initial IRWST conditions, such as temperature and fluid mass, are input to the model, as well as pressure as a function of time. Energy and mass are .

tracked in the IRWST node. Fluid in the node is assumed to be a homogeneous mixture (i.e., perfect mixing is assumed in the IRWST tank). Steaming from the poolis accounted for if saturation -

- temperature is reached in the IRWST.

Minor changes or enhancements were made to other existing models in LOFTRAN and LOFITR2.

'Ihe changes to existing models consisted of the following:

Reactor vessel head vent model Following extended operation of the core makeup tanks or the CVCS overfilling of the RCS

, may occur under some postulated assumptions. The AP600 includes a safety related reactor vessel heat vent which may be to opened bleed excess fluid injected into the RCS. 'Ihis feature gives the operator the flexibility to remove excess fluid injected in the RCS rather than L terminating the CMT injection. An option has been added to LOFTRAN and LOFITR2 to simulate this fluid relief path from the RCS. The option allows the fluid relief flow rate to be

. controlled as a function of time or calculated by the code as a function of the local fluid conditions based on the Fauske/ HEM critical flow model. Details of the head vent model can be found in Reference 8.

Protection system actuation logic

'Ihe AP600 protection system includes new functions for automatic actuation of the PRHR and CMT and additional automatic functions for other safeguards features. New automatic.

protection system actuation logic was added to LOFTRAN as needed to model design-basis accident analyses.

. Modifications to the pressurizer safety valve n.odel to. allow simulation of slower ADS valve opening The AP600 ADS is used to depressurize the RCS in a controlled manner following small-break LOCA transients. It is not used in the mitigation of non-LOCA events and is not expected to be actuated. However, an inadvertent opening of an ADS stage is addressed as a design basis event using LOFTRAN. The ADS valves are designed to open slowly. 'Ihe existing ,

pressurizer relief valve model was modified, so that a slow opening valve could be simulated.

. Addition of user input to model the elevation difference between reactor vessel inlet and outlet -

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2.3 LOFTRAN AP Code Modifications for Test Simula'. sons Modifications for CMT Component Test Simulations Part of the validation of the LOFTRAN-AP core makeup tank model is performed by comparing LOFTRAN calculated CMT performance data to CMT component test data. The Waltz Mill CMT component tests were used for this purpose. He component CMT test facility consisted of an instrumented tank used to simulate the CMT and a steam / water reservoir that simulated the rest of the RCS. Connecting lines to supply steam and/or liquid to the top of the CMT were provided, as well as a drain line to allow flow out of the bottom of the CMT. A source of saturated steam from a boiler was attached to the reservoir. A more detailed of description of the CMT component test facility is given in Appendix A.

To simplify comparison of the LOFTRAN-AP CMT model to the CMT component tests, a stand-alone CMT simulation code called LOFTRANCMT was setup. LOFTRANCMT uses CMT thermal-hydraulic coding identical to LOFTRAN-AP. De code was developed by combining the LOFTRAN-AP CMT coding with input, output, and driver routines compatible with the test instrumentation. This approach simplified supplying the test boundary conditions to the LOFFRAN-AP CMT numerics. Further details of the validation method are given in Section 4.0.

Modifications for SPES-2 Test Simulations he LOFFRAN code as described in Reference 3 contains a simple metal heat capacity model for the RCS. A lumped metal heat capacity model is used and the RCS is divided into the following metal heat capacity regions:

[

]"

For each of the [ ]" regions a constant metal heat capacity [ ], and constant RCS fluid to j metal heat transfer coefficient [ ] are input. De model is used to account for heat additions l to the RCS fluid from the metal during SGTR analyses and steam line break mass and energy release analyses. The model is turned off during other non-LOCA analyses because the metal heat transfer and heat capacity would have no impact or would make the event less severe.

Heat losses from the SPES-2 facility were approximately [ ]" percent of the power generated at scaled AP600 full power and 610*F. Following reactor trip, rod power was reduced to values scaled to m:\ap600\2061 -non\2061 w.non: lt>-070795 25 REVIsloN: 0

AP600 decay hat levels, of about several hundrec kW, In this condition, heat losses from the SPES-2 RCS were about the same as the simulated decay heat. It was obvious the heat losses would !

have a significant impact on the test results. Because the heat losses are significant at the SPES-2 facility, [ 1" added to the test post trip decay power to attempt to compensate for heat loss.

The metal heat capacity model of LOFTRAN was modified to take account of the heat losses and improve the simulations of the SPES-2 facility. For each of the sesen metal heat regions, an external heat transfer component was added.

The change in energy from the RCS fluid to the metal and from the metal to the containment air is calculatn.1 using following:

AEi = UA, (Tu - T,) dt AEs = UAs U, - T int) dt where AE 1

= Energy transferred between the primary fluid and the metal over a time step, Btu

= Energy transferred betweert the metal and the containment air over a time step, Btu AEs UA, = Fixed input value for primary fluid to metal heat transfer coefficient multiplied by interfacial area, Btu /sec. *F UAs = Fixed input value for metal to metal containment atmosphere heat transfer coefficient multiplied by interfacial area, Btu /sec. 'F Tp = Primary fluid temperature, 'F Tu = Metal temperature, 'F TmA

= Fixed input value for containment air temperature, *F dt = time step size, secorxis The change in metal temperature in each region over the code time step is computed using the following:

Tu(t) = Tu(t-dt) - (AE + AEs) / MCp 1

where Tu(t) = Metal temperature at current time step, 'F Tu(t-dt) = Metal temperature from previous time step, 'F m:\ap6000061 -non\2061 w.sno :ll>.070795 2-6 REVisloN: 0

MC, = Metal specific heat capacity. Fixed input value equal to metal specific heat multiplied by metal mass, Btu /*F Further details of the SPES-2 heat loss modeling are provided in Subsection 5.3.4 t

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3.0 ROLES OF TESTS IN LOFTRAN-AP AND LOFTTR2-AP CODE VALIDATION 3.1 Overview ;

Validation of the LOFTRAN-AP PRHR model, CMT model, and the integral AP600 plant response with these passive safeguards systems, is based on the following tests:

. SPES-1 natural circulation tests l I

PRHR component tests

. CMT component tests

. SPES-2 steam generator tube rupture and steam line break tests Each of these test programs and how they relate to the LOFTRAN-AP code validation effort are ;

briefly described below. More detailed discussions of the CMT component test facility and SPES-2 ,

tests' roles are provided in Subsections 3.2 and 3.3.

SPES-1 Natural Circulation Tests he SPES-1 facility was a three loop full-height facility scaled in the ratio of 1/427 with respect to a standard Westinghouse PWR three-loop plant. Scaling criteria are aimed toward natural circulation I and small-break LOCA.

He LOITRAN-AP RCS natural circulation capability is validated by comparison of simulations of tests performed at the SPES-1 facility, with test data. In particular, the validation is based on simulations of test SPNC-01, which focuses on single-phase natural circulation. The test can be ;

accessed in the report listed in Reference 9. His comparison has been completed and is summarized in Appendix 15B of the SSAR.

PRHR Component Tests 1

The Westinghouse PRHR test facility is a full-height simulation of three PRHR tubes in the IRWST.

The heat tmnsfer mechanisms used in the LOFTRAN-AP PRHR model are verifled by comparisons of simulations from a LOFTRAN-AP model of the three-tube arrangement of the test facility with the test data. His comparison has been completed, and the results are presented in Appendix 15B of the SSAR. ;

CMT Component Tests i The Westinghouse CMT component test facility comprises a scale CMT tank, a steam / water reservoir, I instrumentation, and piping. l l

t m:Wo61-noeuo61w.non: tw70795 3-1 REVis10N: 0 f

i e

The LOFTRAN-AP CMT model is verified primarily by comparison of simulations of CMT component tests with test data. During design-basis non-LOCA and SGTR events, the CMTs exhibit the recirculation mode ofinjection instead of the drain-down mode of injection. De verification uses the CMT 500-series tests, which are natural circulation tests (followed by drain-down and depressurization).

SPES-2 Steam Generator Tube Rupture and Steam Line Break Tests he SPES-2 test facility is a 1/395-scale full-height, high-pressure test facility. He facility includes the reactor vessel loops, the pressurizer, the SGs, the PRHR heat exchanger, and the CMTs.

To validate LOFTRAN-AP, four full-system transient SGTR and MSLB tests are simulated. Two of the tests are blind tests.

Comparison of the test simulations with the data validates the integrated behavior of the LOFTRAN-AP reactor coolant loop models and the new passive safeguards system models of LOFTRAN-AP.

3.2 Role of CMT Component Tests in LOFTRAN-AP Validation 3.2.1 CMT Component Tests Description ne Westinghouse CMT test facility (Reference 10) consists of an instrumented test vessel that simulates the CMT and a steam / water reservoir that simulates the remainder of the RCS. Connecting lines to supply steam and/or liquid to the top of the CMT are provided, as well as a drain line to allow flow out of the bottom of the CMT. A comparison of the CMT test facility and AP600 layout is shown in Figure 3-1. A source of saturated steam from a boiler is attached to a steam reservoir and is connected to the liquid / steam reservoir. De test apparatus is shown schematically in Figure 3-2, and a '

detailed description is given in Reference 10. A data acquisition system (DAS) is provided to record l signals from thermocouples, pressure sensors, and flow meters. He test matrix is presented in l

Reference 11. !

The CMT test program has been developed to perform scaled, separate effects tests in which the boundary conditions are controlled over a wide range to produce thermal-hydraulic conditions of interest for computer code validation. De test facility CMT is 1/2-scale in height and 1/7.7-scale in l diameter. De scaling logic that supports the application of the data for code assessment is described in detail in Reference 12. His report shows that the key thermal-hydraulic phenomena ofinterest were reproduced in the test facility and that the test facility can be operated and controlled over a sufficiently broad range that captures all CMT modes of operation relevant to non-LOCA/SGTR analysis. )

i I

i m:\ap6002061 - non\2061 w.non:1 b-070795 3-2 Revision: 0

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i

.' i 4

j 3.2.2 Role of the CMT in AP600 Safety and Analysis with LOFTRAN-AP l Be AP600 passive core cooling system includes two CMTs, located above the cold legs of the AP600 RCS. Each tank stores 2000 ft.' of cold borated water at RCS pressure that is gravity-injected into the i RCS to provide reactivity control and core cooling. I J

The AP600 plant system design is shown in Figure 1-1. The AP600 plant system design includes a I normally open pressure balance line from the RCS cold leg to the top of the CMT. De CMTr are ;

also connected to the RCS via a discharge line from the bottom of each CMT to the reactor vessel.

De CMTs provide the same function as the high-pressure safety injection system in existing PWRs, ;

with the difference being that current plants require the availability of ac power to perform their safety function, whereas the CMTs perform this function using only gravity-driven flows. During accidents, the role of the CMT is to provide coolant and boron to the RCS.

When a safeguards signal ("S" signal) occurs (typically low pressurizer pressure), the reactor coolant '

pumps (RCPs) trip and the CMT isolation valves open. De cold leg balance line to the CMT line is ;

initially warmer than the CMT and the injection line. The water density difference between the balance !

line and the CMT is sufficient to create a positive gravitational head that initiates flow between the I cold leg, CMT, and direct vessel injection line, initiating the recirculation mode of the CMT.

During the transients analyzed with LOFrRAN-AP, subcooling exists in the reactor cold leg, thus the !

CMTs work most of the time in single-phase natural circulation. Moderate void generation can occur

{

for some transients when the RCS pressure drops very low, leading to a decrease in the water r

subcooling at the top of the CMT (e.g., steam line break, steam generator tube rupture). The ;

1 LOFTRAN homogeneous-equilibrium slug flow model is capable of handling such situations. '

i Rus, there are two CMT operational modes that are applicable to non-LOCA transients analyzed with .

LOFTRAN-AP:

l

Single-phase natural recirculation

Two-phase natural circulation with moderate void generation l 3.2.3 CMT Component Test Results Used i

Referring to the CMT operational modes identified in Section 3.2.2 and comparing to the test matrix, j the natural circulation phases (phase 1) of the 500-series tests are selected for validation of the !

LOFTRAN CMT module. De draindown phases (Phase 2) of these tests are outside'the scope of l

LOFTRAN-AP since they simulate a plant configuration with a water level inside the RCS loops.

Only tests with the CMT completely heated are simulated because they essentially repeated the tests with a partially heated CMT (see Table 3-1).

J nanp6aos2061.nonuo61..non:lta70795 3-3 REVIsloN: 0 l

In addition, the cold pre-operational test results are used to determine the friction factor of each line in preparation of the input deck for the CMT component test facility simulations. .

l The CMT tests used for the LOFTRAN-AP code validation effort are summarized in Appendix A of .

]

this report.

3.3 Role of SPES-2 Tests la LOFTRAN AP Validation - -!

3.3.1 SPES-2 Tests Description :

'Ihe SPES-2 test facility is a full-height, full-pressure,1/395-volume scale model of the AP600 plant. !

SPES-2 has a two-loop primary circuit, a secondary system (up to the main steam isolation valves), the passive safety systems, a normal residual heat removal system (NRHR), a chemical and volume -l control system (CVCS), and a startup feedwater system (SFWS). The primary coolant system consists of a pressure vessel with electrically heated rods, two RCPs, two SGs, a pressurizer, and coolant loop piping. The passive safety systems consist of two accumulators, two CMTs, an in-containment refueling water storage tank (IRWST), PRHR heat exchanger (HX), and an ADS. A detailed Ll description of the SPES-2 facility is provided in Reference 13. Figure 3-3 gives the general system l

layout and identifies its key components. The test matrix is presented in Reference 24. !

t k

The overall objective of the SPES-2 tests is to provide experimental data for validation of the [

computer codes used in the safety analysis to obtain design certification for the AP600 plant. The l SPES-2 layout has been designed to duplicate, as close as possible, the thermal-hydraulic phenomena E that would occur in the AP600 during transients. j i

t 3.3.2 SPES-2 Test Results Used l Most non-LOCA analyses do not employ the advanced plant features for event mitigation. Simulation f of the SPES-2 SGTR and MSLB tests with LOFTRAN-AP provides validation of the code for l modeling the CMT and PRHR systems, in addition to the component test efforts. Simulations of the l SPES-2 SGTR tests provide code validation for the integral plant response over a range of SGTR

{

scenarios, which exceed those encountered in the SGTR SSAR analysis. Simulations of the SPES-2

{

MSLB test validates integral plant behavior at conditions, which extend beyond those normally [

encountered in non-LOCA design basis calculations. ;

i

'The SPES-2 tests used for LOFTRAN-AP validation are- !

Matrix Test S01009 - Design-basis steam generator tube rupture with nonsafety systems on and l operator action to isolate steam generator f

Matrix Test S01110 - Design-basis steam generator tube rupture with nonsafety systems on and l

no operator action :

mAap6002061 -oonUO61 w.non:l t>.070795 3-4 REVIsloN: 0 I

_ . . . . - . . _ - . _ _ _ _ _. . __ _ ~.- - -.__

)

Matrix Test S01211 - Design-basis steam generator tube rupture with manual ADS (blind test)

Matrix Test S01312 - Large steam line break (blind test)

In addition, data from hot and cold pre-operational SPES-2 tests were used to develop heat loss and .

line resistance modeling. '

'Ihe SGTR and MSLB SPES-2 tests used for the LOFFRAN-AP code validation effort are summarized in Appendix B of this report.

f F

I t

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m.\ap6000061 -nonuG61 w.non :l t>.070795 3-5 RWisioN: 0 ;

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TABLE 3-1 500 CMT TEST SERIES - TESTS SELECTED FOR SIMUI.ATION Length Pressure beated Test ( psi ) (-) Comments Selected C066501 1085 1/5 Phase 1 included in C064506 No C059502 1085 1/5 Same phase I as C066501 No C068503 1085 1/2 Phase 1 included in C064506 No C0615G4 1085 1/2 Same phase I as C068503 No C070505 1085 1/1 Same phase I as C064506 No C064506 1085 1/1 Yes C076507 1835 1/5 Phase 1 included in C072509 No C074508 1835 1/2 Phase 1 includes 1 in C072509 No C072509 1835 1/1 Yes 1

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1 4.0 LOFTRAN-AP CMT MODEL AND VALIDATION l

1 4.1 Validation Approach l l

The validation of the LOFTRAN-AP CMT model complements the validation of the PRHR and natural circulation models presented in Appendix ISB, Revision 0 of the SSAR. These exercises validate LOFTRAN-AP modeling of AP600-specific features. The methodology used for the LOFTRAN-AP CMT module validation includes three complementary steps:

= First, the LOFTRAN-AP CMT module (in the stand-alone LOITRANCMT code) is used to simulate simple transients with results that can be verified by hand calculations. These simulations are called analytical simulations in the following sections.

Second, the LOFTRAN-AP CMT module (in the stand-alone LOFTRANCMT code) is compared with the results of the CMT component tests described in Section 3.2.3 and Appendix A. This step includes sensitivity studies.

=

Given the independently quallfled LOFTRAN-AP CMT module, the last step is to verify that the CMT module is correctly coupled with other LOFTRAN-AP modules. This is achieved using simulations of the integral full-pressure SPES-2 facility test results.

The first two steps are presented in Subsections 4.4 and 4.5. The third step is presented in Section 5.0.

4.2 Key Phenomena As discussed in Subsection 3.2.2, there are two CMT operational modes that are applicable to non-LOCA transients analyzed with LOFTRAN:

= Single-phase natural recirculation Two-phase natural circulation with moderate void generation

'Ihis section identifies and discusses the thermal-hydraulic phenomena that are involved in these modes and were simulated in the CMT component tests.

CMT recirculation results from the density difference between the cold CMT water and the hotter cold leg balance line water. Recirculation then provides colder, denser, and highly borated water to the reactor vessel from the CMT, which is replaced by hotter, less dense, and lower-borated cold leg water from the balance line. There is a net mass transfer of water and boron from the CMT to the RCS due to the density and concentration differences, as well as a net energy transfer from the RCS back to the CMT. The rates of energy and mass transfer depend on the buoyancy differences and the hydraulic resistances in the flow path. Recirculation flow continuously diminishes with time as the CMT heats m:\ap6002061 -oon\2061 w. non : I b-o70795 4.] REVISION: 0 i

up and the resulting buoyancy head decreases. As the cold leg piping and cold leg balance line void, the buoyancy head increases, which increases the discharge flow. Single-phase recirculation predominates for most non-LOCA transients with a potential for two-phase recirculation for some steam line breaks and SGTRs.

There is CMT wall heat transfer during recirculation; the hot fluid from the cold leg balance line will transfer heat to the initially cold CMT walls.

Fluid mixing also occurs to a limited extent at the top of the CMT during the recirculation phase of the transient. The hot liquid from the cold leg balance line is injected into the CMT through the nozzle and mixes with the initially colder water in the CMT. The geometrical configuration of the CMT with hotter, less dense water arriving at the top of the CMT, leads to stable thermal I stratification.

During a long recirculation transient, it is possible that the water at the top of the CMT remains hotter than the RCS water, if the RCS pressure drops dramatically, boiling inside the cold leg balance line and inside the CMT may occur. Boiling inside the cold leg balance lines increases the buoyancy head and the recirculation flow. Boiling at the top of the CMT reduces the recirculation flow. The LOFTRAN-AP homogeneous slug flow model takes that into account if there is no steam stratification in the CMT circuit. Stratification may occur only if the water velocity is very low. In that situation, steam may accumulate at the top of the CMT or more precisely inside the vertical CMT inlet pipe.

The LOFTRAN-AP CMT model is not sophisticated enough to precisely compute this phenomena.

Conservative calculations are made, using a buoyancy head penalty, as soon as the water subcooling is lower than an input user data. The buoyancy head penalty calculation is made assuming that an input user length of pipe is full of steam.

l Table 4-1 summarizes the phenomena identified, giving a short description of the LOFTRAN-AP model principle and the methodology used to validate each phenomena separately. It is noted that i boron transport was not simulated in the CMT component test and is therefore not pertinent to this l validation exercise.

1 1

4.3 LOFTRAN AP CMT Component Test Facility Model For the validation of the CMT model, the LOFTRAN-AP CMT module was isolated from the LOFFRAN-AP code and built into a stand-alone code, LOFTRANCMT, which runs in conjunction with a main program that supplies boundary conditions. This section describes the module and the input used to model the Chfr component test facility. For a description of the test facility see Appendix A.

i mAmp6000061-non\2061w.non:Ib 070795 4-2 REVISION: 0

4.3.1 Houndary Conditions

'Ihe water reservoir of the CMT component test facility is not simulated as a component with the LOFTP.ANCMT code. Only the tank and its inlet and outline lines are modeled. In the facility, the CMT lines operate with a very low driving pressure difference (=10 psi), compared to the design pressure of the CMT (2250 psia). The simulations are performed using one test pressure measurement and one level measurement (PTl and PDT7) and calculated loop pressure drops. Multiple pressure measurements were not used because point-to-point measurements, combined to obtain pressure drops, were too noisy.

For the 500-series tests, steam line number one was closed during the natural circulation phase.

Referring to Figures 4-1 and 4-2, LOITRANCMT code needs three boundary conditions, computed as follows: !

PPBL = PTl + (PDT7 - IIEIT)

p,/144 PPVESS = PPBL + HEIT

p,/144 IIHilli = enthalpy corresponding to TC75 where:

PPBL: pressure at the inlet to steam line 2 (psia)

PPVESS: pressure at the injection line outlet (psia) 11111111: fluid enthalpy at the inlet to steam line 2 (Btu /lbm)

IIEIT: distance from steam line 2 entrance to bottom of water reservoir (ft.)

ITI: pressure at the top of the water reservoir (psia); measured data PDT7: water reservoir level (ft. of water); measured data TC75: water temperature inside the water reservoir ('F) p.,: water density above the balance line inlet (Ibm /ft.')

p.,: water density below the balance line inlet (Ibm /ft.3) 1 4.3.2 Input Deck l l

4.3.2.1 Geometrical Data The CMT facility as built drawings were used without any modification to prepare the CMT model.

Geometrical data are obtained from Reference 10. The measured volume of the CMT (18.7 ft.')is obtained from the A-01 cold pre-operational test results presented in Reference 14. Simulation of the l CMT was performed using a [ ]" node model. Two types of CMT node sizes are investigated:

. Variable node sizes (small at the top, large at the bottom)

. Equal node volume sizes I

m:\apMXA2061-non\2061 w.non:ltW70795 43 REVISION: 0

l l

Table 4-2 gives the water node volumes for each configuration, and Figure 4-2 shows the noding of the facility.

43.2.2 Friction Factors of the Lines ne total friction factors of the lines are input to the LOFTRAN-AP code based on the total hydraulic resistance of the lines, f 11D + K.

where:

f: friction factor used to compute the regular hydraulic resistance UD: equivalent length of a resistance to flow, in pipe diameters '

K: resistance coefficient for hydraulic singularities l

Injection Line De resistance of the injection line at maximum flow was measured during the A04R1 and A04R2 tests presented in Reference 14. The Reynolds number during these tests was higher than that for the 500-series tests [

}"

Cold Leg to CMT Balance Line l

)

'Ihe friction factor of line 2 was measured during the A06 test presented in Reference 14. [

1 i

Ju 43.23 Metal Parameters l

For each node, three input parameters are used to simulate metal-to-fluid heat transfer:

i UACMT: Heat transfer coefficient between the CMT fluid and the steel multiplied by the surface area for each node, Bru/sec. 'F )

Note: the UACMT input accounts for the steel conductivity as well as steel-to-water film coefficient. 'Ihe conductivity effect is limiting.

]

1 I

i r

ar\ap600C061 - non\2061 w. non; 1 b-070795 4-4 Revision: 0

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UACMTE: Heat transfer coefficient between the steel and the containment atmosphere for each node multiplied by the surface area, Btu /sec. 'F Note: the UACMTE input accounts for the steel conductivity as well as the steel-to-air coefficient. The air transfer is limiting.

XMCCMT: 'Ihick metal heat capacity for each node, Btu /*F The input parameters are computed using the following:

XMCCMT = Ms Cp

= P. V Cp where:

Ms: mass of steel (Ibm)

V: volume of steel (ft.')

Cp: specific heat capacity of steel (0.110 Btu /lbm *F) p,: density of steel (480 lbm/ft.')

UACMT = 2 Ss A/e where:

Ss: steel-to-water contact surface area (ft.2)

A,: thermal conductivity of the steel (65.4E-3 Btu /sec-ft2 ,.py e: average thickness of the steel UACMTE = Se he where:

Se: steel-to-air contact surface area (ft.2) he: heat transfer coefficient, including convective and radiative heat transfer (Btu /sec.-ft.')

he = [ ]'b" he = [ ]'6d Geomettical data concerning volume and steel surfaces are obtained from Reference 10. Table 4-3 summarizes the data used.

Table 4-4 and 4-5 give the metal parameters for each water node configuration.

mAsp6000061 -non\2061 w. mon: l t@0795 45 REVislON: 0 l

i

)

l 4.4 Analytical Simulations Four separate analytical simulations were performed using the LOFTRANCMT model to investigate model behavior under different constant boundary conditions. In each case, [

ju Two configurations were simulated:

Cold water (CMT temperature) at the cold leg to CMT balance line inlet - Natural circulation is expected to stop when essentially all the balance line hot water is replaced by the cold inlet water (Cases A and B).

Hot water maintained at the cold leg to CMT balance line inlet - Natural circulation is expected to stop only when all the CMT cold water is replaced by hot water (Cases C and D).

Each simulation is made with heat transfer (Cases B and D) and without (Cases A and C) heat transfer between water and CMT wall. Table 4-6 gives the run descriptions.

'Ihe results of these simulations are discussed below.

4.4.1 Cold Inlet Balance Line (Cases A and B)

These results are presented in Figures 4-3 to 4-5.

Case A (No heat transfer between CMT water and wall)

After the opening of the CMT discharge valve (t-O seconds) the discharge flow mte increased rapidly (Figure 4-3). 'Dx: maximum discharge flow rate was approximately 0.7 lbm/sec. and was obtained at two seconds. The flow rate decreased repidly as the cold water replaced the hot water in the balance line and stopped completely at 760 seconds, when a new momentum equilibrium was reached; the low density of the water at the top of the CMT was compensated by some hot water that remained in the last balance line node.

Case B (With heat transfer between CMT water and wall) jo.c m:pt-no uostw. on:15070795 4-6 REV!sloN: 0

s t

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4.4.2 Hot Inlet Balance Line .

These results are presented in Figures 4-6 to 4-10.

Case C (No beat transfer between CMT water and wall) -

i .

After the opening of the CMT discharge valve, the discharge flow rate increased rapidly and reached {

0.83 lbm/sec. (Figure 4-6). Because the inlet balance line water was hot, the injection flow rate

{

decreased only when a significant part of the CMT water was warmed, more than 1000 seconds later. l 1

The injection flow rate continuously decreased to a very small value at 20,000 seconds (Figure 4-6). .i It is to be noted that the simulation was made with very small nodes at the top of the CMT and larger nodes at the bottom of the CMT (variable volume noding). This noding led to an overestimate of the

{

- long-term flow rate (see 500-series test simulations, Subsection 4.5). 4 i

Case D (With heat transfer between CMT water and wall) 4 At the beginning of the transient (i.e., before time 1000 seconds), the injection flow rate was essentially the same as Case C (Figure 4-6). This was because the wall heat transfer affected only the l

warmed area, which is small at the beginning of the transient. On the other hand, when a large i portion of the CMT water was warmed, the heat losses exactly compensated for the convective heat ~

transfer, and a new stable steady-state was reached. The injection flow rate remains high, around '

40 percent of the initial value (See Figure 4-6). The temperature at the bottom of the CMT and inside l

the injection line stabilized at 478'F (Figure 4-8). l l

[ !

l l

ja.c 1

This transient confirmed that the heat losses should not significantly affect the injection flow rate I during the first 1000 seconds of a transient. On the other hand, the long term behavior was affected -

by heat transfer to the CMT wall.

1 i

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J m:WO61-nonUO61w.non:lt>-07M95 47 REVIsloN: 0 )

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1 4.4.3 Hand Calculations of Momentum and Energy Balance !

' To verify the CMT model momentum and the energy balance, hand calculations are performed for. j Case D.

t Momenttun Balance !

t The momentum balance is checked at the beginning of the transient (T=5 seconds). At this time, the j buoyancy head hand calculation is simplified since the entire CMT is still cold. j The following condition should be verified: f BH = Kg Ww :/p + Kw Ww: j pw where: ;

i 5

BH: buoyancy head at the beginning of the transient ;

BH = (HEIT + TOPCMT ) * (pq - pw) /144 (HEIT + TOPCMT) = 34.57 ft. (see Figure 4-1)

C p,: CMT and injection line water density: 62.43 lbm/ft.8 i pw: balance line water density: 47.73 lbm/ft.8 I

K,: friction factor of the injection line ;

K, = (fIJD + K) / (144.

2

g

S 2) f IJD + K: total resistance of the line. For Case D, the injection line is valve resistance set to provide 12.2 gpm. 'Ihis leads to f IJD + K = 277.8. !

S: cross-sectional area of the injection line (0.01003 ft.2) ,

g: 32.2 ft/sec.2 ;

'Ihen, Kw = 297.7 psi-sec.2/lbm-ft.8 i

Kw: friction factor for the balance line : 21.4 psi-sec.2/lbm ft.8 I W: % injection line flow rate calculated by LOFTRANCMT: 0.833 lbm/sec. !

B I

m:\ap6000061 -non\2061 w.non: l b.o70795 48 REVISION: 0 ,

i l

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I

Ww: balance line flow rate calculated by LOFTRANCMT: 0.684 lbm/sec.

The balance line flow rate is lower than the injection flow rate, due to the density .

difference.

I The following numerical application shows that the momentum balance is maintained:

BH [ ]" [

~i

-K,,W ,2fp g ju j E

2

-Kw Ww fpw g ja.

-i TOTAL [ ] '

Energy Balance :

Case D is also used to check the energy balance at time 15,000 seconds. This time is chosen since a !

thermal steady-state was reached. I The following condition should be met:

Q, = Q., + Q,a !

where : l l

t Q,: convective energy flow at the CMT inlet .

Q, = W

H, !

i Wm: inlet flow rate computed by LOFTRANCMT: 0.3285 lbm/sec.

l H: inlet enthalpy: 530 Bru/lbm

-l Q,,: convective energy flow at the CMT outlet . '

I Q,, = W,,

H. i W.,: outlet flow rate computed by LOFTRANCMT: 0.3285 lbm/sec.' ~j t

H.: outlet enthalpy computed by LOFTRANCMT: 461.8 Btu /sec. :

1 I

I adap6000061 -moe\2061 w. mon:I b470795 49 REVIsloN: 0 ;

i

Q,g external heat losses between CMT wall and air: 22.39 Btu /sec. '

The order of magnitude of the value computed by LOFTRANCMT may be checked using data of Table 4-5 and an average fluid temperature of the CMT of 500*F (See Figures 4-7 and 4-8). The external air temprature is 97'F.

'Ihe following numerical application shows that the energy balance is maintained.

Qin = [ ]" ;

1

-Q , = [ ]"

-Qwall = [ ]"

TOTAL [ ]"

4.4.4 Conclusions - Analytical Simulations These basic simulations show that LOFTRAN-AP CMT modeling leads to credible simulations that validate the CMT component test facility modeling and the method used to simulate the water reservoir tank.

4.5 500-Series Tests Simulations Matrix Tests 501 to 509 simulated the heating of the CMT water by natural circulation with subsequent dramdown and depressurization. Only the natural circulation phase of the tests is simulated with LOFTRANCMT (Subsection 3.2.3).

During the natural circulation phase, the water reservoir contains water (close to saturation) and steam; the level is above the line 2 inlet. Line I was closed and natural circulation was initiated by fully opening the injection line (valve V3). During the transient, the reservoir pressure was kept constant as much as possible.

Only tests with the CMT completely heated are simulated because they essentially repeat the tests with a partially heated CMT (see Subsection 3.2.3). These tests are C064506 and C072509.

4.5.1 Test C064506 4.5.1.1 Description of the Runs Five runs were made to analyze LOFTRAN-AP CMT model behavior. These runs included sensitivity studies to investigate the effect of varying the time steps and noding used for these calculations.

m Amp 6000061 -non2061 w.nos : WO70795 4-10 REVISION: 0

e Run 1: Run with the LOFITIAN-AP CMT default parameters / variable volume CMT nodes /

no heat transfer between the water and the CMT steel wall

Run 2: Same as nm 1, but with heat transfer between the water and the CMT steel wall taken into account / data used are described in Section 4.3.2

Run 3: Same as run 2, but with equal volume CMT nodes Run 4: Sensitivity study on the time step /same as run 3, but the time step is divided by 4

Run5: Sensitivity study on the water reservoir enthalpy to simulate boiling inside the CMT/same as run 3, but boiling conditions are simulated inside the loop (This configuration is credible because the test is initiated with the water temperature inside the reservoir very close to the saturation. Boiling is obtained by increasing j the balance line inlet enthalpy of 5 Btu /lbm during the transient. A buoyancy head penalty is used assuming that the vertical inlet pipe is full of saturated steam when the loop subcooling is lower than 10*F. This pipe is 1.6 ft. long.)

Table 4-7 summarizes the run parameters.

l 4.5.1.2 Calculation Results !

l The calculation results are presente.d in Figures 4-11 to 4-43. The pressure at the water reservoir j (Irrl-ABS) and the temperature at the CMT inlet (TC76-COR) are used to verify the accuracy of the boundary conditions used.

Run1: Variable Node Sizes and no Water-to-Steel Heat Transfer. l l

The general behavior of the injection flow rate plot is similar to the test results. [

]a.b.c l

Both calculated and experimental plots show two points where the injection line flow rate slope changes dramatically; these are described as follows:

[

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n-l ;

ik ._ -

i I

l I

. j=u - :

l '[ , j i

1 j=u .;

Run 2: Simulation of the Water-to-Steel Heat Transfer i: [

Run 2 is a repeat of Run I with the addition of water-to-steel heat transfer. The [ f i

.}'b5 i Figure 4-20 shows the calculated heat transfer. [

J'* 1he calculated water-to-steel heat transfer matche's the experimental values (Figure 4-21). !

At the end of the transient (after 1500 seconds) the injection flow rate is increasingly overestimated 1 because [ ]'6" as confirmed by '

l Run 3 (see Figures 4-23 to 4-26).

. 1 Run 3: CMT Nodes of Equal Size and Water-to-Steel Heat Transfier 1 i

Run 3 is an extension of Run 2, except that equal volume fluid nodes are used in the CMT. As shown in Figure 4-27, [

]" A small overestimation of the injected flow rate is observed all along the transient. The shapes of the calculated and experimental injection flow rate plots are very similar. Fluid temperature

- evolution (Figure 4-33 to 4-35) and CMT water-to-wall heat transfer (Figure 4-30) are also in good i agreement.

During the complete transient, the calculated injection flow rate is higher (5 to 10 percent) than the experimental value. This explains why the hot water reaches the CMT outlet earlier (see TC77-COR Figure 4-32). It should induce a faster decrease of the buoyancy and also a faster decrease of the' injection flow rate. The probable explanation is that the friction factor of the line is underestimated at low flow; only one value is used, and the usual increase of the friction factor when the Reynolds number decreases, is not taken into account. .!

)

I arpl-mosuG61w. mon:ll>C70795 4 12 REVISloN: 0 1

't ',

Run 4: Time Step Influence Run 3 uses time steps of [ ]" afterwards. He time step size was decreased by a factor of four for Run 4 t

Run 3 and 4 results are compared in Figures 4-36 to 4-39 and are shown to be identical. .his -

demonstrates that [

]" .

Run 5: Simulation With Boiling Inside the CMT And the Balance Line i Run 5 is the same as Run 3 except that the reservoir water enthalpy is increased by 5 Btu /lbm, and a .

calculational penalty is applied to conservatively account for the potential accumulation of steam in the '!

piping inlet. He initial injection flow rate is lower than that for Run 3 because the buoyancy penalty -

applies (Figure 4-40). As the pressure of the loop decreases from 1120 psi to 1090 psi between l times 0 and 100 seconds (Figure 4-12), boiling occurs for the LOITRANCMT calculation; this induces an initial increase of the injection flow to 1.9 lbm/sec., compared to 1.8 lbm/sec. for the l simulation with no boiling. He reason for this increase is that the balance line density decrease has a I bigger effect than the penalty used (1.62 ft.) to simulate the potential accumulation of steam at the - -

CMT inlet pipe.

l l

[ 'j

]" l B

I

]" t i

When subcooling of the water reaches 10*F (conservative input data), the buoyancy head penalty is set l to zero; then, the injection flow rate increases again (Figure 4-40, at time 1960 seconds),

f 4.5.2 C072509 Test Run 6 and Run 7 are executed for C072509 test. His test was performed at a system pressure of approximately 1350 psia. He run parameters are those described in Table 4-7. Calculation results are presented in Figures 4-44 to 4-56. !

Run 6: Water CMT Nodes of Equals Sizes and Water-to-Steel Heat Transfer Run 6 is similar to Run 3, except the initial and boundary conditions of test C072509 are used. All ;

the comments made for the C064506 test are still applicable. Test results and calculations are in good f

ndap6000061 -non\2061 w. mon:lb 470795 '

4 13 REVISloN: 0

, . _ , . . . _ _ , , - _ . - . ~ .. - - - . . - -. - -

j agreement. The calculation overestimates the injection flow rate by approximately seven percent during the transient (Figure 4-44).

During the test, the loop pressure increases considerably during several periods [ ]'**

psia - Figure 4-45). No major experimental injection flow changes are observed during these periods.

This observation indicates that there is no boiling inside the loop for this test. If boiling were occuning, the pressure peaks would have collapsed the steam bubbles and significant flow rate fluctuations would have been observed in Figure 4-44. The pressure peaks may be caused by the reservoir pressure control valve.

Run 7: Simulation With Boiling Inside the CMT and the Halance Line This high is a repeat of Run 5, except that the boundary and initial conditions for test C072509 are used. As for test C064506, the inlet balance line enthalpy is increased by [

Ja.c

[

j a.,

[

F ja.,

4.5.3 500-Series Tests Conclusions The 500-series tests simulations presented in subsections 4.5.1 and 4.5.2 show:

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ju 4.6 Assessment of CMT Component Test Simulation Results The ability of the LOFTRAN-AP CMT module to predict the thermal-hydraulic behavior of the AP600 CMT has been investigated. 'Ihe separate thermal-hydraulic phenomena identified and simulated in the CMT Component Test in Table 4-1 are addressed as follows:

= Convective heat transfer

[

ju

= CMT wall heat transfer

[

ju

= Single-phase natural circulation 1

3u

= Two-phase natural circulation

[

ju i

Two sensitivity studies were performed: l I

I i jo.e In the range of the time steps used, the prediction is independent of the time step.

The LOFTRAN-AP CMT hydraulic model is not intentionally biased in either the conservative or nonconservative direction. Depending on the transients simulated, conservatism is introduced with the input data to provide minimal or maximal flow.

I i

m:\ap600N2061 -non\2061 w.non:1 M70795 4 15 REVIs!ON: 0 l

l I

Simulations have been made with the stand-alone LOFTRANCMT code. The SPES-2 simulation results (Section 5.0) show that the LOFIRAN-AP CMT module is still accurate when coupled with the other LOFTRAN-AP modules.

l 1

-)

m:WO61-ooeUO61w.non:1b 070795 4 16 REVISION: 0

'l

_ _ _ _ _ .~_______..._.I

5-i 15 h TABLE 4-1 g PHENOMENA IDENTIFICATION FOR THE APtee CMT

{

S Phenomena LOFTRAN-AP Modeling Validatkm S Single-Phase Natural Circulation Momentum balance solved using buoyancy Flow comparison using the 500-series tests. '

head calculation and hydraulic resistances in the flow path.

Two-Phase Natural Circulation Momentum equation solved using buoyancy No available test.

(Flashing of the Hot CMT Liquid Layer) head calculation and hydraulic resistances in Addressed using buoyancy head penalty.

the flow path. Applicable only for situation p with moderate void generation and high flow G rate.

Convective Heat Transfer Homogeneous SLUG flow model. Transient temperature profile comparisons (Mixing at CMT Top) using the 500-series tests.

CMT Wall Heat Transfer 1

Thermal balance between each water and Heat flux comparison using the 500-series CMT steel node. tests.

Boron Transport

Global point boron balance for the CMT. Not timulated in CMT Comptment Test

-=

SLUG model for each node of the lines.

Is M

o

. . . _ . - ~.- . _ _ _ - . - - _ _ . ~ ~ _ - .., . , . . , . . - , - - . . - . , . ,- -- - . - -- -- . - . - . . . . . . . - - . . ._,-_ _ . _ . . _ - -

. j TABLE 4-2 CMT WATER NODI SIZES Node Number Variable Volume Equal Volume (1 = Top) Noding (ft.') Noding (ft.')

u 4

+

a l

t 5

atWo61-oonuo61w.non:1&o70795 4 18 REVISION: 0

i

)

J TABLE 4-3 CMT STEEL DATA i i

Water volume Steel Volunne Steel to Water Steel to Air i Region ( ft.' ) ( ft.' ) ~ Surface ( ft') Surface ( ft.2 ) l Upper Cap 0.545 Cylindrical Part 17.61 1 Lower Cap 0.545 l r

Total 18.70 l i

t l

e l

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i l

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1 TABLE 4-4 ,

METAL PARAMETERS - VARIABLE VOLUME NODING Water XMCCMT UACMT UACMTE ,

Node Volume ( ft.8 ) ( BturF ) ( Btu /sec 'F ) ( Blu/sec/F ) i u l t

f l

l l

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l h

TABLE 4 5 -

METAL PARAMETERS - EQUAL VOLUME NODING Water XMCCMT UACMT UACMTE UACMTE Volume 1985 psi 1835 psi Node ( ft.' ) ( Btu / F ) ( Btu /sec. *F ) ( Btu /sec. *F ) ( Btu /sec. 'F) _u r

l i

I i

i t

i i

i f

I I

l

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j l

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l TABLE 4-6 ANALYTICAL SLMULATIONS - RUN DESCRIPTIONS !

Water to Steel Default Node Description Case Number Heit Transfer . ;mes A No Yes i

~

Cold inlet Balance Line B Yes Yes C No Yes Hot Inlet Balance Line D Yes Yes i

I l

l l

l i

l l

m:\ap6000061 -non\2061 w.non:lt@0795 4 22 REVISION: 0

.=

TABLE 4-7 TEST C064506 AND C072509 - RUN PARAMETERS Run Node Wall Heat Reduced Time Increased Reservoir Test Number Sizes Transfer Step Enthalpy 1 Variable No No No 2 Variable Yes No No 3 Equal Yes No No C064506 4 Equal Yes Yes No 5 Equal Yes No Yes C072509 6 Equal Yes No No 7 Equal Yes No Yes m.wuo2061 -monco61 w. mon:1 b-070795 4 23 REVISION: 0

i Connection line [ Steam Line 1) ppi t

l Balance line [ Steam line 2)

CMT TOPCMT WATER TANK I

PPTK (Waterlevel]

V9 Valve a, PPBL ETK HEIT injection line #

i i

Figure 41 Derivation of Boundary Conditions and other Variables for CMT !

Component Test Facility Model mAa;W)oc061 oon\2061w.oon::b 070795 4 24 REVISION: 0

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i FIGURES 4-2 THROUGH 4 56 ARE NOT INCLUDED IN THIS NONPROPRIETARY DOCUMENT. !

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i 5.0 LOFTRAN SPES-2 MODEL AND INTEGRAL SYSTEM VALIDATION 5.1 Validation Approach The LOFTRAN-AP simulations of the full-pressure SPES-2 integral effects tests are the last stage in the validation process. He simulations demonstrate the ability of the code to perform conservative design-basis analyses for the AP600 plant In particular, the simulations are used to confirm that the AP600-specific models in the LOFTRAN-AP code are interacting correctly and capture the key phenomena of the plant transients (see Subsection 5.2). He simulations also provide further validation of the individual passive system models.

The simulations are performed using the same code versions as used in the final SSAR Chapter 15 design-basis analyses. A SPES-2-specific input deck (modified from test-to-test and from run-to-run) is used to model the facility. See Subsections 5.3 and 5.4 for descriptions of the SPES-2 model and test-specific input.

He validation approach that was used to simulate the SPES-2 SGTR tests (Tests 9,10 and 11) and the steam line break test (Test 12) is outlined in the following.

5.1.1 Steam Generator Tube Rupture A preliminary set of SPES-2 SGTR test simulations was presented in Reference 2, including the blind simulation of Test 11. De preliminary simulations of Tests 9 and 10 showed a need for further refinement ofinputs and assumptions related to tube rupture break flow, primary- and secondary-side heat loss modeling, metal masses, line resistances, and other parameters.

Ileat loss and metal mass modeling is critical in SPES-2 simulations because of the facility's large system heat, surface-to-volume ratio and metal-to-fluid mass ratio. System heat losses and metal mass heat capacity have a significant impact on the test behavior. Note that these factors are not important in analysis of the AP600 plant.

The final simulations presented here include the necessary refinements. In addition, several i simulations of Matrix Test 10 are run to test sensitivity to particular heat loss treatments, time step sizes, and noding.

New nonblind simulations of Test 11 are run to correct an error in the blind simulation input and incorporate the refinements discussed above.

The pertinent parameters in each simulation are compared with the test data (described in Appendix B) to demonstrate the code's ability to capture the key phenomena in AP600 design-basis analyses. The ability to simulate CMT and PRHR performance is also examined.

maap6000061-oonC061w-5.non It470795 5-1 REVISION: 0

he SGTR test simulations are presented in Subsection 5.5. The assessment of the simulation results is presented in Subsection 5.7.1.

5.1.2 Steam Line Break The SPES-2 MSLB test was a blind test. Both pre-data release and post-data release simulations are included in this report.

Prior to performance of the blind simulation, it was recognized that because of the conservative nature of LOFfRAN-AP's steam line break model and the code's limitations in modeling heat losses and thick metal heat capacity, the simulations would over-predict the speed and extent of the test facility cooldown. The approach taken for the pre-data release simulation was therefore to use the same conservative assumptions as the SSAR transient analyses. These assumptions are:

Break flow is saturated steam

. Break flow model assumes fIlD = 0

No steam line friction In addition, another pre-data release simulation was performed to confirm the sensitivity of the prediction to assumed blowdown quality. This simulation uses a quality profile constructed for design-basis containment mass and energy-release calculations.

Post-data release simulations were perfonned to assist in the examination of differences between the test data and the blind simulation. A blowdown quality profile was constructed based on the blowdown mass accumulation rate. The break flow model was modified to match the blowdown from each SG to the test data. Heat capacity is added to the SG tube metal mass to show the sensitivity of the simulation accuracy to metal mass effects.

De simulations were assessed to show that LOFTRAN-AP is capable of performing conservative AP600 safety analysis calculations.

He MSLB test simulations are presented in Subsection 5.6. The assessment is presented in Subsection 5.7.2.

5.2 Key Phenomena 5.2.1 Steam Generator Tube Rupture His subsection describes the SGTR transient and the primary LOFTRAN-AP model interactions that were validated by comparison to the SPES-2 tests. Key to these simulations is the operation of the CMT and PRHR during natural circulation flow conditions. Successful modeling of these systems during the SGTR tests validates LOFTRAN-AP for modeling integral plant response during the naap60cc061 nonu061w-5.non:lb-070795 $-2 REVislON: 0

.k [

3 1

'I

- -)

Loperation of these key passive safety features,in addition to the separate CMT and PRHR test i programs. Prior to CMT and PRHR actuation, plant behavior is similar to a standard Westinghouse ,j PWR. De phenomena of interest are identified for this transient in the PIRT as given in j Subsection 1.2 and Table 1-1. !

i I

Matrix Tests 9,10, and 11 are SGTR experiments. Each of these tests simulated a single-tube rupture. !

Tests 9 and 10 differ in that Test 9 employed both operator action to control the tube rupture and 'l employed nonsafety (control) systems. Test II, the blind SGTR test, was similar to Test 10 until ADS l actuation occurred. De LOFITR2-AP simulation of Test 11 was terminated shortly after ADS j actuation (150 seconds after pressurizer low-low level was reached). De following is a brief I description of the SGTR event and the system interactions validated by this series of tests. -l AP600 design-basis SGTR analyses are performed in part to verify that the SG will not become i t^

liquid. solid (overfill) and that the primary and secondary pressures are brought into equilibrium, terminating flow through the break. De key parameters for SGTR analyses are primary pressure, !

secondary pressure, flow through the break and PRHR heat removal. The code-calculated values for i t

these parameters should compare closely to the test. CMT injection is not critical to the transient, T provided the level in the CMT does not drop. De most important parameter is the primary-to-secondary break flow. Break flow is dependent upon the relative values of the primary and secondary :

pressure, which in turn is highly dependent upon the cooling provided by the PRHR.

l i

i Important automatic protection system functions in the SGTR tests include: safeguards actuation of ;

the CMT, PRH.R, RCP trip, and reactor trip on low-pressurizer level. Dese are closely tied to the f

magnitude of the break flow. He' code should predict the occurrence of safeguards actuation at

[

essentially the same time as the test.

l I

Based on previous discussions, key parameters for the LOFITR2-AP SGTR simulations include: i i

e Break flow

Pressurizer pressure and level !

SG pressure (faulted and intact loops) l

=

Primary-side SG inlet and outlet temperatures (both loops)

=

l RCS temperature upstream and downstream of the PRHR injection point

. CMT flow and level

= PRHR flow and heat removal rate t

5.2.2 Main Steam Line Break j

Matrix Test S01512 simulated an MSLB similar to a design-basis double-ended pipe rupture. De test !

was initiated by opening a power-operated relief valve on SG-A with a flow orifice scaled to simulate l a comparable AP600 MSLB cvent. Steam line check valves were removed; thus initially, prior to i steam line isolation, both SGs contributed to break flow. De test was conducted with the SPES-2

\

m:wr600co61-mo uo61w.5 om:ib-070795 5-3 REVIsloN: 0 j

i l

facility at conditions comparable to hot-standby for the AP600 design. Control systems were not operable, and the feedwater system was isolated.

The steam line break event is characterized by a rapid RCS cooldown. He rate of system cooldown can be observed by noting the evolution of the pressurizer pressure. Initially the rate of cooldown is severe because high quality steam discharges from both SGs and the primary side RCS is at full flow conditions. %e rate of system depressurization decreases slightly after steam line isolation and decreases more after reactor coolant pump (RCP) trip as the RCS flow quickly decreases. After the pressurizer empties, the rate of system cooldown further decreases reducing primary-to-secondary heat transfer. The rate of system cooldown at this point is governed by the SG blowdown, the PRHR heat removal, and cool CMT inventory. This rate of system cooldown and pressure decrease continues until the upper head saturation pressure is reached. When system pressure reaches the saturation pressure of the upper head, the upper head acts as a pressurizer and system cooldown is controlled by the PRHR, CMT, and system heat losses. All of these break points should be predicted well by LOFTRAN-AP. Discussion and comparisons of simulation to test data are provided in Subsection 5.6.

The test and simulation assume a safety signal at event initiation. After specified delays, various automatic protection functions occur. Protection system response includes steam line isolation, RCP trip, CMT actuation, and PRHR actuation. Passive accumulator injection occurs when the primary-side pressure drops below 700 psia.

Based on this discussion and the parameters identified in Table 1-1 the key criteria for steam line break event are as follows:

Pressurizer pressure

SG pressure

Primary-side RCS temperature

RCS flow a CMT flow and level

PRHR flow Rese parameters are presented for the LOFTRAN-AP MSLB simulations with comparisons to the SPES-2 test data.

It should be noted that reactivity feedback effects are not included in the listed parameters. The test was conducted with the intent of maximizing the RCS cooldown rate. System heat loss compensation (important to the SPES-2 facility) was terminated at break initiation. His test did not model core decay heat or any core reactivity feedback effects.

m:\ap6002061 -oon\2061 w-5.non: Ib-070795 5-4 REVIsloN: 0

5.3 LOFTRAN-AP SPES-2 Model Description De same version of LOFTRAN-AP used for AP600 analyses models the SPES-2 test facility by calculating SPES-2 specific LOFTRAN-AP input. Some inputs include SPES-2 specific component elevations and volumes, line friction factors, and R.CS pressure drops. De application of LOFTRAN-l AP to the SPES-2 test facility is described in Subsection 53. Deviations from the basic LOFTRAN-

! AP SPES-2 model required the modeling of individual tests, these tests are described in

(

1 Subsection 5.4.

l Consistent with the particular test being simulated, initial conditions were input into the SPES-2 LOFTRAN-AP model. Heater rods controlled heat generation in the SPES-2 facility. In i LOITRAN-AP test simulations, core power was input as a function of time based on test data. For I

the SGTR tests, core-decay heat was simulated with channel power increased to compensate for system heat losses.

Although the same version of LOITRAN-AP was used for the SPES-2 test simulations and the AP600

, SSAR, some modifications were made to the code specifically addressing phenomena related to the i

1895 scale SPES-2 facility. Rese changes were primarily required to model the heat losses in the SPES-2 facility, which have increased significance compared to the AP600 facility. He IS95-volume scaling results in a much larger, surface-area-to-volume ratio than the actual AP600 Although validation of LOFTRAN-AP heat loss models is not a goal of this program, adding heat loss models was necessary to adequately model the SPES-2 system transient response.

He SPES-2 facility is briefly described in Appendix B. This Subsection describes the significant features from a code modeling aspect and indicates how the major LOFTRAN models were specified to simulate the SPES-2 facility tests.

53.1 Primary System The SPES-2 primary system consists of a pressure vessel with electric heating rods, loop piping, pressurizer, SG U-tubes, and coolant pumps. Each of these components, as well as the LOFTRAN-AP model considerations, is described in the following subsection.

53.1.1 Power Channel Pressure Vessel and Reactor Coolant Loop Piping ne SPES-2 power channel pressure vessel includes a lower plenum, riser, upper plenum, upper head, and downcomer. De SPES-2 downcomer consists of an annular downcomer around the upper plenum and a tubular downcomer that connects the upper and lower plenum regions. To better simulate the AP600 riser inlet conditions, the tubular downcomer inlet to the lower plenum enters an annular section.

m:WJ061-nonUO61w-5.aon:1M)70795 5-5 Revision: 0

With respect to the AP600 reference design, the total RCS volume and system component volumes (lower plenum, riser, upper plenum, downcomer, and upper head) are volume-scaled. Vertical elevations are preserved with the exception of the lower plenum and upper head, which have no influence on the natural circulation.

The SPES-2 facility has two reactor coolant loops and each loop has one hot leg and two cold legs, an SG, and an RCP. The cold leg split occurs downstream of the RCP The cold legs enter the power channel annulus downcomer through separate nozzles. The primary-system loop piping maintains vertical elevations and preserves the Froude number. The horizontal portion of the hot leg has the same length-to-diameter ratio as the AP600.

The power channel pressure vessel and RCPs are simulated in LOFTRAN-AP with separate volume inputs for the power channel, power channel inlet, upper plenum, hot and cold leg piping, SG inlet and outlet regions, and SG tube region. The power channel inlet volume includes the downcomer, lower plenum, bottom of the riser, and bypass volumes. A separate volume input is included for the upper power channel, which includes the area above the active fuel and below the upper core plate.

The reactor coolant power channel and loop volume input is divided into control volumes or nodes.

The number of nodes can be specified by the user. LOFTRAN-AP can have up to 160 core sections, ten hot leg sections per loop, eight cold leg sections per loop, and ten SG sections per loop. The loop model reproduces the layout of a standard Westinghouse PWR. This general layout correlates to the AP600 design and SPES-2 facility.

53.1.2 Power Channel Rod Bundle The SPES-2 facility power channel consists of 97 electrical heating rods, which have the same heated length and geometry (rod pitch, rod diameter) as the AP600 fuel rod design. Rod heat flux is preserved. Overall core power was selected to maintain the AP600-design fluid thermodynamic conditions, power-to-volume ratio, and power-to-flow ratio. Rod-bundle power flux was increased to compensate for system heat losses. For the SPES-2 SGTR and steam line break simulations, rod-bundle power is not an independent variable and was input to LOFTRAN-AP as a function of time.

5.3.13 Reactor Coolant Pumps and Loop Flow Model he SPES-2 facility has one RCP per loop. The AP600 nominal pump head and fluid loop transient time was preserved. Pump and flow coastdown was simulated at the facility. The SPES-2 nominal flow rate was selected to achieve the same power-to-flow ratio as the AP600 design. Component elevations were preserved to accurately simulate natural-circulation flow conditions.

The LOFTRAN-AP code solves the basic equation of motion including the effects of friction pressure losses, elevation, pump head, and fluid momentum. Ec code computes pump head and torque based en:\ap6002061-non\206Iw 5.non:lt470795 5-6 REVIsloN: 0

e on homologous RCP curves. For test simulations, pump speed was adjusted to match the initial flow conditions. Pump windage was adjusted to match the experimental flow coastdown in SGTR Test 10.

For two-phase flow conditions, LOFTRAN uses a homogenous slug flow model; thus, the code will handle void generation. The steam and liquid phases are always in equilibrium and there is no slip.

While this model is adequate for cases with moderate void generation, LOFFRAN-AP is not intended for transients where primary-side, two-phase flow effects are important.

Specific SPES-2 input to the LOFTRAN-AP code that is important to RCS flow includes: component elevations, lengths, and frictional pressure drop under full flow conditions for the RCS loop.

Reoretical pressure drops were calculated and compared to experimental values. Where available, experimental pressure drop values were used. He input-pressure-drop data are normalized to match the pump head. Test simulations were conducted with a code option, which allow friction factors to increase as flow decreases based on data in Reference 15.

The fractional volume of the upper plenum (where mixing occurs)is specified by input to the code.

Power-channel bypass flow is also specified in the code. The fraction of bypass flow was given for each test. Due to the small size of the inlet and outlet plenna in relation to coolant flow rate, perfect mixing was assumed for test simulations.

5.3.1.4 Pressurizer and Surge Line The SPES-2 pressurizer consists of a cylindrical vessel with flanged ends and contains submersible electrical heaters. The surge line is connected from the bottom flange to the primary-circuit hot leg.

Six external electrical heaters compensate for heat losses. The pressurizer has one safety valve located at the top of the vessel. When necessary, pressure is :elieved manually through the ADS.

With respect to the AP600 design, the SPES-2 pressurizer is volume-scaled, the bottom elevation is preserved, and the level swelling phenomena was reproduced. The level swelling phenomenon can occur in the pressurizer due to flashing of the contained liquid or an insurge of steam from the primary circuit. This swelling can significantly affect the quality of the fluid discharging through the ADS valves located at the top of the pressurizer; thus, the SPES-2 pressurizer diameter was selected using the Wilson bubble rise model (Reference 16) to match the average void fraction in the AP600 for similar thermal-hydraulic conditions.

The LOFTRAN-AP model consists of a two-region (water and steam) pressurizer model. Since the water level is expected to change during a transient, a variable control volume model is used.

Condensation or superheating is allowed in the steam region, and evaporation and subcooling is allowed in the water region. Water drops are uniformly distributed in the steam region and fall at a constant rate, while steam bubbles are uniformly distributed in the water region and rise with a constant velocity. The pressurizer model also has the capability to consider the effects of pressurizer heaters, spray, relief, and safety valves.

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Specific SPES-2 test facility input to LOFTRAN-AP includes the overall pressurizer volume, pressurizer surge line volume and frictional pressure drop, pressurizer heater output, pressurizer relief valve flow rate, initial pressurizer pressure, and initial pressurizer water volume. Initial pressurizer water volume and level are test-specific parameters. Pressurizer pressure control system sprays and relief valves can be automatically actuated based on pressurizer pressure or be manually actuated. For the SPES-2 test, these systems were manually actuated as a function of time, simulating operator action when required.

5.3.I.5 Vessel IIcad Model The LOFTRAN-AP model of the vessel head consists of a single volume. The code simulates the vessel head bypass flow from the reactor vessel inlet (cold leg) to the upper plenum. The behavior differs depending on the RCP operation:

When the RCPs operate, flow is by forced circulation from the top of the downcomer to the upper head and returns to the upper plenum. For the SPES-2 facility and the AP600, the bypass flow is large enough to maintain the upper heml temperature below the hot leg temperature.

After the RCPs trip, natural circulation governs the flow direction. Since the upper plenum is hatter than the downcomer, flow can occur in the opposite direction (from the upper plenum to ;

the vessel head and to the top of the downcomer into the cold legs). )

When the RCS begins cooling, flow can stop because the geometrical configuration leads to a stable thermal stratification. The hotter (less dense) water is located at the top of the circuit.

The temperature of the water in the upper head volume will vary depending upon the vessel )

head heat losses. If a depressurization of the RCS occurs up to a point where saturation is reached in the upper head, boiling occurs and induces draining of the upper head.

This general behavior of the vessel head model has been observed during most of the SPES-2 tests.

LOFTRAN-AP uses a simple model based on the following user-supplied input:

Upper-head bypass flow coefficient expressed as the fraction of the total RCS loop flow (this coefficient is constant during the transient)

Initial water enthalpy in the vessel head To simulate the SPES-2 behavior during the SGTR tests, the upper head bypass flow was set to zero, and the initial enthalpy of the upper head water to the saturation enthalpy corresponding to the RCS mAap600C061-non\2061w-5 non:lt470795 58 REVisloN: 0

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pressure at the time boiling occurs. This simulation reproduces the real phenomena only after the pumps trip. Before the pump trip, the impact is small because core flow is modified by less than one percent.

l

} 53.2 Secondary Coolant System f 53.2.1 Steam Generators The SPES-2 plant has two identical SGs that allow the transfer of thermal power from the primary-to-I secondary circuits. He SGs consist of a pressure vessel, tube bundle, steam separator, and dryers.

He SG vessel consists of a bottom plena (primary side), an intermediate section containing the tube bundle and two extemal dryers, and an upper section that includes the separator and the dryers. The intermediate section consists of a tall cylindrical vessel that houses the U-tube bundle. A dividing j plate prevents the mixing of water from the two external downcomer pipes and feedwater, separating the riser into a cold and hot zone. The two external downcomers connect the upper SG shell to the j bottom of the tube bundle. The SG section consists of a cylindrical vessel, flanged to the intermediate vessel with a large dome at the top. This section contains the feedwater nozzle and the main steam line nozzle.

The SPES-2 SGs are also volume-scaled to the AP600 design. De number and length of the U-tubes is not as well proportioned, leading to a 2.5 percent less than scaled heat transfer surface area.

Vertical elevations are preserved in the secondary side at the top of the steam separator. The steam dome elevation has not been maintained since it has no influence on natural circulation.

The LOFIRAN-AP SG secondary-side model is represented by a single volume node. In the standard LOFTRAN-AP version, this node consists of a homogeneous saturated mixture of steam and water.

l LOFITR2-AP differs from the standard version m that the node is divided into two regions. The i lower region may be saturated or subcooled and the upper region may be saturated or superheated.

On the primary side, the SG model contains multiple (up to [ J'**) tube section nodes. Primary-to-secondary heat transfer is simulated with a log-mean temperature difference (LMTD) type response.

He overall heat transfer coefficient (UA) is initialized by the code to match the nominal input conditions. The code uses the primary mass flow rate, heat flux, and secondary-side pressure to compute changes in the heat transfer coefficient due to changes that can affect the primary- and secondary-side film resistance.

The ability of the SG to transfer heat depends upon four factors: the primary-fluid convective film coefficient, the SG U-tube conductive resistance, the secondary-fluid convective film coefficient, and the extent to which the SG tube bundle is covered with secondary fluid. LOFTRAN-AP has the capacity to model each of these parameters. LOFTRAN-AP decreases the heat transfer area linearly as the water mass decreases when the SG water mass reaches a value corresponding to the level at the top of the tube bundle.

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Specific SPES-2 test facility input to LOFTRAN for the secondary side includes feedwater temperature, overall SG volume, and initial SG water volume height from the top of the tube sheet to the top of the tubes, height from the top of the tubes to the riser region, SG heat transfer area, tube bundle volume, minimum water volume required to cover the tubes, tube material, and tube metal heat capacity. He initial SG water level is a test specific parameter. Default code parameters were used for the primary- and secondary-side film resistance. The SG model also has the capability to model relief and safety valves. Steam and feed line isolation are simulated and check valves can also be specified.

SG PORVs can be automatically actuated on pressure or be manually actuated. For the SPES-2 SGTR test 9 simulations, the PORVs were opened as a function of time sirnulating operator actions. No secondary-side control or safety valve operations are modeled in the steam line break simulation.

53.2.2 Steam Generator Tube Rupture Break Flow Model LOITTR2-AP contains a detailed and flexible failed SG tube model capable of modeling any number broken or ruptured tubes. The model considers tube frictional loses and entrance and exit form losses.

%e code also models critical flow using the modified Zouledek correlation (Reference 17).

For the SPES-2 SGTR simulations, the break location and area was provided. Break flow versus time data were available for Tests 9 and 10. 'Ihis data were used to calculate an overall friction factor.

The coefficient calculated from Tests 9 and 10 data were used for the blind test simulation (Test 11).

53.23 Steam Pipe Break Flow Model Steam flow is determined in the LOFTRAN-AP based user-selected options. Steam flow to the turbine, through the safety valves, steam dump, and break flow can be modeled. The effects of isolation valve closure and check valves can be modeled as desired. The steam header is modeled as a zero-volume node and requires a mass balance of the flow to and from the header.

Steam line breaks or inadvertent steam system valve openings are modeled in LOFTRAN-AP via a user-specified break area. A steam line break per SG and a header break can be modeled simultaneously. Break flow is calculated based on the Moody critical flow correlation with fZ/D=0.

For the SPES-2 MSLB test (Matrix Test S01512), the break was modeled on SG-A. Check valves were not modeled allowing steam flow from SG-B until steam line isolation occurred. The code accounts for the pressure drop from the intact SG to the header by specifying the pressure drop at nominal steam flow. A SPES-2 specific value was used for the nominal pressure drop.

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5.33 Passive Safety Systems Passive safety systems important to the MSLB and SGTR tests include the CMTs, the PRHR HX, and the two accumulator tanks.

i 5.33.1 Core Makeup Tank System he CMT system is a passive gravity-driven system that provides reactor coolant makeup and ;

emergency boration. The CMT system has two operating modes: water recirculation and draindown with steam entering the CMT, and water displacement. He MSLB and SGTR events should only experience the CMT recirculation mode. l l

l The SPES-2 facility has two CMTs that deliver coolant directly to the reactor vessel annular downcomer through separate lines. The balance lines, which take coolant from the cold leg, are connected to each of the cold legs in reactor coolant loop B (the loop opposite the pressurizer). ;

i With respect to the AP600 reference design, the CMTs are 1/395-volume-scaled. Elevations of the CMTs and connecting lines are preserved. The metal mass is scaled to obtain the same overall condensation of steam on the CMT walls and the same water / metal temperatures during transient simulations. This condition required placing the CMTs within a secondary pressurized tank. ;

I The resistances of the CMT lines are computed using the cold pre-operational test results. Two sets of hypotheses are used in this report, which consider the uncertainty of the pressure drop and flow j measurements. To better match actual test results, maximum line resistances are used for all calculations, except for Run 4 of Test 10. Rese values are obtained using the maximum pressure !

drop and minimum flow measurement, according to the uncertainty of the test data. Minimum line resistances are the result of opposite assumptions.

I The LOFTRAN-AP code simulates the dynamics of a single CMT and assumes the performance of a second CMT as identical. The resistance of the injection line of the two CMTs differs by 20 percent (Reference 19) in the facility. An equivalent single injection line has been simulated. To compare LOFTRAN-AP results with the SPES-2 tests, the flow from each CMT should be added. l A separate LOFTRAN-AP model validation effort was performed for the CMT system and is presented in Section 4.0. The SPES-2 tests provide additional data to validate the LOFPRAN-AP i CMT model during transient situations similar to the design-basis MSLB and SGTR events. Details of )

the LOFTRAN-AP model can be found in Subsection 3.1 in Reference 8. l l

l 533.2 Passive Resi<!ual Heat Removal Heat Exchanger The PRHR HX is located on the primary-side of the system and is designed to remove core decay heat during emergency situations. De PRHR HX connects to loop A, which also contains the pressurizer.

m:\ap600C061 -non\2061 w.5.non: I b-O'r0795 5-11 Revision: 0

i l

I l

The SPES-2 facility includes a C-shaped PRHR HX submerged in the IRWST. With respect to the AP600 design, the total heat transfer surface area, tube diameter, tube thickness and pressure drops, l

and elevations are preserved to the extent possible. The HX contains three tubes. The proper scaling for the AP600 is 1.7 tubes. For the SGTR tests, one tube was used to minimize event mitigation effects. For the MSLB (test 12), three tubes were used to maximize the severity of transient system cooldown.

The SPES-2 PRHR supply line runs from the top of the loop A hot leg. A two meter pipe elevation l was added to preserve the AP600 piping elevation. This line is heated prior to tests to match the ;

expected AP600 fluid conditions. 'Ihe return line is routed to the suction line of RCP-A.

LOFTRAN-AP contains a multi-node PRHR model with inlet piping connected to the hot leg of the ;

pressurizer loop and outlet piping connected to the SG outlet plenum. The heat transfer mechanisms used in the LOFTRAN-AP PRHR model were verified by a separate test performed at the Westinghouse PRHR test facility. The results of those test comparisons are presented in Appendix ISB, Revision 0 of the AP600 SSAR. A detailed description of the LOITRAN-AP PRHR !

model is provided in Subsection 3.2 of Reference 8. '

l 5.3A Heat Losses Models f

The LOFITtAN-AP code includes models to simulate heat losses at principal locations such as: the i primary system, the secondary side of the SGs, and the CMTs.

I The SPES-2 heat losses were measured during the hot pre-operational tests. The results of Test H-01 presented in Reference 19 are used. The total heat losses of the loop (RCS and SG) have been ;

estimated to be about 150 kW at a stable temperature of 605'F. Test H-01 also provides an l

approximate split of the heat losses: [ ]'6" kW for the hot legs and power channel, [ ]'6#kW ;

for the cold legs, and [ ]'6# kW for the two SGs.

As stated in Subsection 5.1, the validation of the heat losses is not a goal for this report; however, the representation is required for these specific tests because the water at the PRHR and CMT inlet pipe.is close to saturation. Subcooling is lower than [ ]'6#'F for Test 10 (Reference 19). Results from the I i

I f

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simulation of Test 10 also show that when boiling occurred at the inlet of the PRHR, the average flow increased and oscillations occurred due to the condensation of steam in the PRHR.

l 5.3.4.1 Primary System IIeat Losses Model l

The RCS metal mass temperature evolutions were modeled by defining RCS water-to-steel and steel-to-air heat transfer coefficients. The coefficients accounted for relative convective and conduction heat transfer resistances and are based on experimental data from the H-01 test. Based on relative fluid and steel contact area, experimental data, and comparison of simulation results to the experimental data; l the following assumptions were made relative to partitioning of component heat losses:

[ ]'6# percent of the SGs heat losses ([ ]'6# kW per SG) occur in the inlet and outlet headers on the primary side, with [ ]'6# percent of the SG losses occurring on secondary side.

[ ]'6# percent of the hot legs and power channel heat losses are applied to the power channel, with [ ]'6" percent on the hot legs.

During initial simulations, unexpected boiling occurred in the inlet PRHR pipe due to an overestimation of the hot leg water temperature by [ ]'6#*F. Investigation proved subcooling during the test was probably caused by heat transfer between the upper plenum and the annular downcomer of the facility. These two volumes are separated by a thin tube ([ ]'6# m Shss Mme 13 Approximate evaluations show that the power exchanged between the upper plenum and the annular downcomer could have been in the range of [ ]'6# kW. Since the major objective of the SPES-2 simulation is to validate specific features of the passive core cooling system (and not the heat losses of the SPES-2 loop), it is assumed that a heat transfer of [ ]'6# kW occurs between the legs at the power channel inlet and outlet. This heat transfer gives a more accurate inlet temperature at the PRHR inlet and it also gives an adequate flow configuration (single phase during a significant period of each transient).

Table 5.3-1 summarizes the RCS heat losses used for all the transients presented in this document.

This table does not account for the heat source and sink of [ ]'6# kW used to simulate the heat transfer between the upper plenum and the annular downcomer.

5.3.4.2 Pressurizer IIeat Losses LOFTRAN-AP can simulate pressurizer heat losses as a heat sink applied to the pressurizer steam l phase. SIET estimated it as [ ]'** kW at nominal conditions. The tests simulated in this report L

confirm this value.

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Test 9 used external pressurizer heaters to compensate for pressurizer heat losses. Two of six external heaters remained on throughout this transient after the S signal. He power of two external heaters is

[ ]'6" kW, with [ ]* percent efficiency. The output power of the two external heaters is more than pressurizer heat losses at stable and saturated conditions, and explains the reason for the presence of superheated steam in the pressurizer during Test 9.

LOFITR2-AP does not simulate the SPES-2 pressurizer with external heaters. It is assumed that the external heaters exactly compensate for the heat losses.

5.3.43 SGs Secondary-Side Heat Losses Model As presented in Table 53-1, [ ]* kW of heat loss was applied to each secondary side of the SG sid:s. This value is representative of conditions with a stable temperature ([ ]**F). LOFTRAN-AP does not take the steel mass on the SG secondary sides into account. Due to the scale of the SPES-2 facility, energy stored inside the steel of the secondary side of the SGs increases considerably in relation to the energy stored in the liquid. His phenomenon was taken into account by correcting the permanent heat losses, assuming that the steel mass temperature followed the fluid temperature with a typical time constant of [ ]* seconds (Estimate based on the SG main steel walls).

Because the SG pressure increased from [ ]* psi for Test 10, this aspect of the SPES-2 facility induced a high transfer after the MSIV isolation. De corresponding energy needed to warm the SG secondary steel mass is around [ ]* Btu per SG, or an average power of [ ]'*" Btu /sec.

Based on the sensitivity calculation, the initial SG pressure response was adequately simulated assuming that this power varies linearly from [ ]* the average power to zero during the period starting with MSLIV isolation ([ ]* sec. for Test 10) and the time when pressure was stable before decreasing again ([ ]* sec. for Test 10).

The final SG secondary-side heat loss values versus time are provided in Table 5.3-2. Initial heat loss is smaller for Test 9 because startup feedwater slowed the secondary-side pressure increase after the MSLI valve closure. Although the integrated heat loss evolution with time reflects the global energy balance (permanent heat losses corrected by the steel mass thermal inertia),it was difficult to duplicate the full results of this test. However, experimental SG pressure evolution was well matched and this is the condition required to validate the other modules of LOFTRAN-AP. !

5.3.4.4 CMTs Heat Losses LOFTRAN-AP simulates the heat transfer between water, steel, and air. He interpretation of the tests performed on the CMT component test facility showed that this specific model is pertinent and l provides an adequate thermal profile for CMTs with one wall (Section 4.0).

The metal mass of the SPES-2 CMTs is scaled to obtain the same overall condensation of steam on the CMT walls and the same water / metal temperatures during fast transient simulations. His required m Aap600,2061-mon \2061 w-5.non:1 b-070795 5-14 REVIsloN: 0

placing the CMTs within a secondary subatmospheric tank. The transients simulated in this report are long-term transients and the external heat losses to be applied to the external face of the internal tank are also important. The global thermal response of the two SPES-2 CMTs is not simulated by LOFTRAN-AP.

The transients presented in this report used an equivalent external heat transfer coefficient applied to the external wall of the internal tank. Since the transients simulated in the preliminary report (Reference 2) showed a high sensitivity on the net mass injected by the CMTs in the RCS, a sensitivity study was made to investigate the importance of these parameters.

The following hypotheses for the simulations were investigated:

Simulation with low-heat transfer - The external heat transfer was computed assuming that a perfect thermal steady-state exists between the two CMT walls. The heat transfer between the two tanks and outside the external tank accounted for convective and radiative heat transfer.

Simulation with increased-heat transfer - The heat transfer was computed assuming that the external tank remains cold at the initial temperature. This hypothesis should have been appropriate at CMT actuation, but became inaccurate for long-tenn simulations.

The final assumptions used for all the calculations presented in this report were a combination of the two assumptions presented. The top CMT nodes use high-heat transfer (CMT is cold at the beginning of the transient), and nodes at the CMT bottom use low-heat transfer (CMT is warmed when hot water comes to the bottom of the CMT). The number of nodes with high-heat-transfer was adjusted to almost match the water temperature profile inside the CMT.

l l

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TABLE 5.31 RCS HEAT LOSSES AT 605'F Test H-01 LOFTRAN-AP Simulation i

Power Power Cornponents l ] Location [ ]

Two Hot Legs + PC Two SGs t

Four Cold legs tow - - -

TABLE 5.3 2 SG SECONDARY-SIDE HEAT LOSSES, INCLUDING STEEL INERTIA Heat Loss I ]

Time Test 9 - Run 1 Test 10 - Run 1 Test 11 - Run 1 10 = MSLIV Closure - - **

t0 + 300 see 10 + 400 sec i

3000 sec _ _,,

j l

l 4

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5.4 Test-Specific LOFTRAN AP Input As discussed in Subsection 5.3, SPES-2 specific LOFTRAN-AP input was developed for modeling the SPES-2 SGTR and steam line break tests. General model input is discussed in that section. This subsection describes the LOFTRAN-AP input for the individual simulations.

5.4.1 Steam Generator Tube Ilupture Test-Specific Input Initial Conditions Test-specific initial conditions were input for the following parameters:

+

Pressurizer pressure ;

Power channel inlet temperature !

+

Upper vessel head temperature

+ Core flow rate I

+

Downcomer bypass flow Pressurizer level 1 Accumulator volume, temperature, and pressure

+ IRWST temperature j

+

PRIIR IIX and supply line temperatures

+

CMT and CMT balance line temperatures l

+ SG water mass

+

Main feedwater temperature

+

Containment air temperature Additionally, power-channel rod power was provided for each test, including the blind test, as a function of time. Specific input parameters used in SGTR Test Simulations 9,10, and 11 are listed in Tables 5.5.2-2,5.5.3-1 and 5.5.4-1 respectively.

Break Flow Input Key to simulation of the SGTR tests is modeling the break flow. Tube break flow is determined by the break area, SG tube friction and form pressure losses, and the primary-to-secondary side pressure difference. The SPES-2 SGTR break size was scaled to 1.2 times the area of a single AP600 SG tube with an inner diameter of [ ]'6# inches. Using this information, a break area of [

]'6# ft.2 was calculated.

In LOFITR2-AP, an SGTR results in break flow from both ends of the ruptured tube providing flow from both the hot legs and cold legs of the primary system. The SPES-2 SGTR tests simulated a break from the primary side of RCP-B suction to the secondary side of SG-B. To model the SPES-2 test break, the friction factor for the hot leg tube break was set to a very large value, prohibiting flow ra%su2061-oo \ 061,-5..on:ib.070795 5-17 REVISION: 0

.l from the hot leg. It should be noted that although LOFTRAN allows input of separate friction factors j or loss coefficients for the tube entrance and exit, inlet plenum, and tube length, the LOFITR2-AP ' i SPES-2 friction losses were simulated with a single overall loss coefficient. 'Ihe loss coefficient was )

~ based on Test 10 break flow data and was calculated so that break flow matched the test data for the !

given primary- to secondary-side pressure difference.

System Heat Losses LOFIRAN-AP was modified for the SPES-2 test comparisons to model heat losses in the primary- . .

and secondary-side systems. Prior to this effort, LOFTRAN-AP had a metal mass heat capacity model l on the primary side and for the SG tube mass. Initially LOFTRAN-AP was modified to allow heat loss from the primary side to the containment. It was determmed that additional heat loss ,

compensation was required for the SGs and pressurizer (see Subsection 5.3.4).

The SG heat loss and metal-mass-heat capacity had a significant effect on secondary-side pressure. To j compensate for this effect LOFTRAN-AP was modified to allow heat loss versus transient time to be .

input for the individual SGs. This input was calculated based on Test 9 and Test 10 data. A similar profile was used for the blind SGTR test.

Heat losses in the pressurizer resulted in the pressurizer filling after the upper head boiled lu Test 10. l To demonstrate this phenomena, a simple heat loss versus time model was also added for the pressurizer (see Subsection 5.3.4).

Automatic Protection System Actuation i

Test-specific actuation times for the reactor and the RCP trip, steam line and feed line isolation, and CMT and PRHR valve opening were provided. Matrix Tests i and 10 were simulated with the actual time of all the test events as input (including trip, passive systent actuations, operator actions, etc.). j Test i1 (blind test with calculation results previously presented .n Reference 2) was simulated with '

both the time of the events calculated by LOFITR AP or with the time of the events as input. ,

l Protection system response times used for SGTR Tests 9,10, and 11 simulations are recorded in Tables 5.5.2-3,5.5.3-3 and 5.5.4-2 respectively. In all cases, action is assumed to occur at the specified time. No allowances have been made for gradual valve closures. !

Operator Action SGTR Test 9 used operator actions to help mitigate the effect of the tube rupture. These actions were input in a simplified form, as described in Subsection 5.5.3.

I I

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i 5.4.2 Main Steam Line Break LOFTRAN-AP Test Specific Input -j With few exceptions, the base LOFTRAN-AP input deck used for the MSLB simulation was identical to that used in the SGTR simulations. Test-specific input includes initial conditions, protection system timing, and input required to model the steam line break.

Initial conditions For the MSLB test simulations, test-specific initial conditions were input for the following parameters:

i

Pressurizer pressure

Power channel inlet temperature !

Upper head vessel temperature j

Core flow rate 1

Downcomer bypass flow rate l 1

Pressurizer level

Accumulator volume temperature and pressure

PRHR system and supply line temperature

CMT and CMT balance line temperatures

SG water mass

Containment air temperature

. IRWST temperature ,

i Heater rod power was [ ]'A' kW prior to the steam line break simulation. This value modified I the RCS metal to containment air heat transfer coefficients and was validated by a period of steady- !

state code simulation prior to break initiation. Specific input parameters used in the LOFTRAN-AP MSLB simulations are provided in Table 5.6-1.

Automatic Protection System Actuation Times Protection system actuation times were based on a pre-test prediction that showed that a safety signal would occur close to break initiation. 'Ihe safety signal isolates the intact SG-B, opens the CMT and PRHR valves, and trips the RCPs. The actuation times used in the simulation were supplied in the test data and are provided in Table 5.6-2. Since LOFTRAN has a single CMT, the average actuation time of 4.25 seconds was used. Similarly for main steam line isolation the average (10.5 seconds) of the two valves was used.

I l

Operator Action and Control System Operation l

Consistent with Matrix Test S01512, no operator actions or control system operation were modeled.

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Feedwater and Safety Feedwater Assumptions The test simulated a main steam line break with no load conditions and no feedwater flow. Safety feedwater was inadvertently added to SG-A for several minutes during the event. Although the overall safety feedwater contribution was only [ ]"# lbm, which is small compared to the [ ]'6# lbm mass initially in SG-A, the safety feedwater flow was modeled in the LOITRAN simulations.

To model the safety feedwater, [ ]'6# lbm of water at [ ]'6#*F was added to SG-A using a constant flow rate from 33 to [ ]"# seconds.

Component Heat Loss and Ileat Capacity The SPES-2 facility is a full height,1/395 volume scale of the AP600 plant. There is approximately 20 times more surface area per unit volume for the SPES-2 components than for the AP600 plant.

Consequently, component heat losses and stored energy in the component metal masses are more important to the SPES-2 facility than for the AP600 (except for the pressurizer, LOFTRAN-AP models the primary system component metal mass heat capacities including SG tubes). Prior to the SPES-2 test simulation efforts, there was no way to model primary- or secondary-side heat losses in LOITRAN-AP.

IIcat losses in the pressurizer and SGs had a significant effect on the SGTR tests. To simulate the SPES-2 SGTR tests, LOFTTR2-AP and LOFTRAN-AP were modified with a very simple heat loss / addition table for these components. A more sophisticated model was not developed since heat losses from these components during SGTR and non-LOCA events are not as significant for the j AP600 plant. Modeling metal mass heat capacities, and not modeling heat losses, provides !

conservative results for design-basis analyses.

For the MSLB test conditions,150 kW heater rod power was required to keep the system at hot-standby conditions. RCS to containment, air heat-transfer coefficients were adjusted so that the system remained in equilibrium with a heater rod power of [ ]'6# kW. Based on heat-loss partition data ,

contained in the hot pre-operational test data (Reference 19), [ ]'6# kW of heat loss was placed in the SGs. At steady-state conditions, the primary side of the reactor coolant system lost [ ]"#kW. j For the MSLB simulations, heat losses were not modeled in the pressurizer. The pressurizer heater was on at the test facility prior to the break to maintain a system pressure of 2240 psia. This is :

comparable to the way LOFTRAN-AP is initialized. Once the break occurs, heat loss in the pressurizer is not very important to the MSLB simulations since the pressurizer drains relatively quickly (within [ ]'6# seconds).

Heat addition from the metal mass of the SGs significantly influences the SPES-2 MSLB test results.

As described above, the SPES-2 facility has a much larger component metal-mass-heat capacity compared to its volume than the AP600 plant. After blowdown begins, the SG metal mass is hotter mwae2061..o co6:w-5..wibmo 95 5-20 REVISION: 0

b l

f than SG water inventory. The simplified SG heat loss / gain model that was added for the SPES-2 SGTR simulations is inadequate for MSLB simulations due to the speed and severity of the cooldown.

l The effect of SG metal mass and associated heat input on the test results is discussed in the results j section. A sensitivity run is presented, which simulates the effect of the large SG metal mass on the j test results by adding metal mass to the SG tubes and water mass to the intact SG. '

i Steam Line Break Flow Model *

'Ihe steam line break was simulated in LOITRAN-AP by modeling a break area of [ ]'" ft. on ,

SG-A. "Ihis corresponds to a single-ended steam line break area of [ ]'* ft. in the AP600 plant. !

LOFTRAN-AP does not predict break flow quality, so quality as a function of time was input lastead.

Prior to release of the blind data, two cases were run: a steam quality of one, and a typical quality i used in calculating steam line break mass and energy releases for design-basis containment analyses. l Upon release of the blind data, a test-specific quality was derived by matching a quality profile that i replicated the experimental break flow accumulation and total mass blowdown from SG-B.

For design-basis calculations, a conservative quality profile was used. For example, the main steam ;

line break analyses were performed to determine core response assuming a steam quality of one.

When a more realistic quality profile is used, the profile is based on NOTRUMP calculations and the ;

values are skewed to generate conservative results relative to regulatory limits. j i

l l

n I

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i 5.5 Test Simulation Results: Tests 9,10, and 11 - Steam Generator Tube Rupture 5.5.1 Overview :

'Ihis section contains a description of the LOFITR2-AP test simulations, including sensitivity studies.

Matrix Test 10, which was a SGTR with no Operator actions, was selected to present most of the l sensitivity studies. This transient is the least complicated and allows an easier analysis of the separate phenomena.

The input deck simulating the test facility and the events of each transient used experimental data. No conservatism was introduced intentionally by adjusting the input data.

Almost all the data used for the simulations presented in the preliminary validation report ;

(Reference 2) were used, but refinements were made using the final data of Reference 10. Table 5.5.1-1 summarizes the evolution of the modeling between the preliminary anxi the final validation ;

report. ;

i Matrix Tests 9 and 10 were simulated with the actual time of all the test events as input (including trip, passive system actuations, operator actions, etc.). Test 11 (blind test with (alculation results '

previously presented in Reference 2) was simulated with both the time of th9 events calculated by LOFITR2-AP or with the time of the events as input. ;

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5.5.2 Matrix Test 10 :

i Matrix Test 10 (Test S01110) was an SGTR test at full-power conditions. Nonsafety systems including the CVCS, the NRHR, and the SFWS were not activated during the test. Test analyses were done using LOFITR2-AP. Actual test results show boiling in the vessel head and in the upper plenum at [ J'6# seconds. Situations with significant boiling in the upper plenum are out of the scope of LOFTRAN-AP. The test simulations were continued, however, until code instabilities were detected.

5.5.2.1 Description of the Runs Test 10 was selected to perform most of the sensitivity studies because there are no Operator Actions and the behavior is less complicated to analyze. The following aspects have been investigated :

Run 1 - Initial Case The assumptions used for the initial case of Test 10 are provided in Subsection 5.3.

Run 2 - SGs Secondary Side Heat Losses Using the initial case as a basis, the SG heat losses were modified as follows:

=

Initial case: [ ] percent on the SGs secondary side, [ l'*# percent on the RCS side Run 2: [ ]'6' percent on the SGs secondary side, [ ]'6" percent on the RCS side l This sensitivity case explains the reason for the SGTR back flow calculated in the initial calculation.

Because this simulation is closer to the test data, it was used as the base case.

Run 3 - SGs Secondary-Side Heat Losses i

1 Using Run 2 as a basis, SG heat losses were modified neglecting the thermal inertia of the steel masses of the SG secondary-side. This sensitivity case illustr .tes the influence of SG metal mass's thermal inertia on the SG pressure evolutions.

Run 4 - Minimum Resistance of the CMT Lines l

Using Run 2 as a basis, minimum resistances for the CMT line were used. 'Ihe assumptions for computing the friction coefficients are provided in Subsection 5.3.3.1.

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Run 5 - PRHR Resistance Sensitivity Study There are some uncertainties regarding the various systems that contribute to the energy balance of the facility. To investigate the effect ofincreased energy removal, a sensitivity study was performed by simulating an increase in the PRHR flowrate. Using Run 2 as a basis, the friction factor of the PRHR heat exchanger was decreased by [ ]'6' percent. Core power, heat losses SGs, CMTs and the PRHR contribute to the energy balance. Since the influence of the other systems has been tested, the PRHR .

l resistance sensitivity study simulates an increase of extracted power from the RCS.

Run 6 and 7 - Time Step Influence A sensitivity simulation was performed on the time step by varying the timestep after reactor trip.

Runs 6 and 7 duplicated the initial case and Run 2 respectively, with the time step size decreased by half during the stable period after the trip.

( Run 8 - Noding Influence i

Run 8 duplicated Run 2 with about twice as many nodes simulating the RCS. The noding in the CMT (fixed in the code) and in the PRHR was not modified.

Run 9 - Heat Transfer Between RCS Water and Steel l

l Run 9 duplicated Run 2 with heat transfer coefficients between the RCS water and steel divided by two. Although there is no rigorous justification for this sensitivity study, the intent was to try to explain why an overestimation of RCS pressure and temperature between [ ] seconds occurs. One reason may be that some RCS steel masses, like the flanges that represent about

[ ]** percent of the steel masses, are initialized at too high a temperature. After reactor trip, heat transfer between the water and steel then decreases too fast because the hot leg water temperature rapidly drops by [ ]***F. Decreasing the heat transfer coefficients will initialize the flanges with lower steel temperatures and cause a lower drop in heat transfer after the trip.

Table 5.5.2-1 summarizes the parameters of each run.

5.5.2.2 Results Analysis 5.5.2.2.1 Initial Case (Run 1)

A comparison of code-calculated results and actual test results is shown in Figures 5.5.2-1 through 1 5.5.2-17. Table 5.5.2-3 provides the sequence of the events. As mentioned previously, this test was I simulated using the actual times of the events. Minor variations in the results in Table 5.5.2-3 were caused by the instantaneous closing and opening of valves in the LOFITR2-AP simulations (e.g., as in !

the main steam line isolation valve closure). l m:\ap607.2061. son \2061 w-5. con: lb-070795 5 25 REVISION: 0

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RCS Parameter Evolutions ,

i The opening of the SGTR break at zero seconds induced a loss of mass out of the RCS and a I decreased pressurizer level. When th, pressurizer internal heaters switched on at [ ]*# kW, the i pressurizer-pressure decreased slowly. At [ ]'*# seconds, the pressurizer level reached [ ]** ft., !

and the pressurizer heaters automatically turned off by the plant computer, inducing faster ;

depressurization.

In the test, reactor trip and safeguards actuation occurred when the low-pressurizer level setpoint .

([ ]*#) was reached. This setpoint was reached in the test at [ ]** seconds. At ;

[ ]"# seconds, code-calculated, pressurizer liquid level was consistent with the test and showed a level of about [ J# ft. in the pressurizer (Figure 5.5.2-8). The reactor trip induced a fast drop in core power (Figure 5.5.2-1) consequently, the core outlet temperature was reduced by approximately ,

[ ]'6#*F. This reduction resulted in the contraction of the RCS water and in complete pressurizer !

draindown by the surge line (Figure 5.5.2-8).

i Various safety features actuated in the test when the low-level setpoint was exceeded. These included:

closure of main steam isolation valves, closure of main feedwater isolation valves, PRHR actuation, i CMT actuation, and tripping of the RCPs. After these events, the facility operated in free evolution, f

with no active systems. RCS pressure and temperature evolutions resulted from mass and heat !

transfers by natural circulation in the RCS. SGs, PRHR, CMT, break, and external heat losses. [

Core power in the test was input to LOFITR2-AP as a boundary condition. The SPES-2 core power l' was increased by [ ]"# kW above the scaled AP600 decay heat (Figure 5.5.2-1), starting

[ ]** seconds after the trip, to compensate for higher-than-scaled facility heat losses. For a brief period after the RCPs tripped [ ,

]*# seconds, the passive systems extracted less power than the core power, and the global energy balance was positive. Natural circulation in the RCS resulted in l 1

the increase of the outlet RCS temperature (Figure 5.5.2-6) and also the RCS pressure. After 550 seconds, when the core power decreased to 400 kW, the RCS temperatures and prer,sure started decreasing slowly.

Overall, the RCS pressure and temperature evolutions are predicted well by LOFITR2-AP. The main differences occurred between [ ]'6* seconds. RCS pressure and temperatures were overestimated by about [ ]*' psi and [ ]'***F. These minor differences between the actual test results and the calculations can easily be explained by phenomenon like heat transfer between the RCS fluid and steel heat transfer.

SGs Secondary Side Pressure The SG-A and -B pressure evolutions are shown in Figures 5.5.2-3 and 5.5.2-4. The SG pressures ,

were essentially constant prior to reactor trip because SG pressure was regulated by the control system )

of the facility. I manp6 coco 61.nonuo61 5.non:tb-070795 5-26 REVIs10N: 0 I

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After the SG steam and feedwater valves were isolated, no mass transfer occurred in the SGs, except for in the SGTR break flow in SG-B. SG pressure was then driven by the energy balance, including heat transfer with the RCS and the SGs secondary-side steel masses. De temperatures of the SG steel masses depend on the steel thermal inertia and on the heat transfer with the SG water and external air.

After steam line isolation valve closure, SG pressure increased from [ ]'*' psi as a result of the heat transfer between the RCS and the SGs. No secondary side relief valves were opened. He reactor trip induced a fast decrease in core power and also in the hot leg temperature. As a result of core power decrease and of heat transfer with the SG steel masses (relatively cold, in thennal equilibrium with the initial SG pressures), the SG pressure stabilized at [ ]*** seconds and decreased slowly. In addition, some subcooled water in the SGs may have contributed to stabilizing SG pressure, shortly after the trip.

After the RCP trip occurred at [ ]'*# seconds, the RCS natural circulation developed with a temporary increase in the hot leg temperature to approximately [ ]'A'*F. At the same time, heat transfer decreased in the SG secondary-side steel because the steel became warmer. The combination of these phenomena was the reason for the SG pressure increase between [ ]'** seconds.

After [ ]'** seconds, SG pressure was mainly governed by SG secondary-side heat losses, including the thermal inertia of the steel. He RCS temperature changed slowly, inducing the same slow temperature variation in the SGs water.

He global behavior of SG pressure was correctly simulated, but the SG pressure was overestimated by

[ ]'** psi after [ ]'** seconds. Sensitivity studies presented in Section 5.5.2.2.2 show that heat losses in the SG secondary-sides and the heat stored in the steel masses are key parameters. As stated earlier, the SPES-2 facility is a 1/395-volume scale of the AP600 reference design. Volume scaling results in a much larger surface-area to volume-ratio for the SPES-2 facility than the AP600 facility. His significantly increased the effects of phenomena associated with system heat losses and the heat energy stored in the system metal masses.

SGTR Break Flow Measured test break flow and code-calculated break flow are shown in Figure 5.5.2-5. The SGTR break flow is a function of the difference between the primary- and secondary-side pressures. He !

water temperature at the break location also had an effect. He SPES-2 SGTR was simulated with a i pipe connected between the pump B suction line and the SG-B secondary-side; this may have induced some minor bias.

At time zero seconds when the SGTR break opened, calculated-break flow was [ ]'** lbm/sec.

compared to [ ]** lbm/sec. for the actual break flow. This [ ]'** percent underestimation disappeared rapidly probably because of cold water in the pipe that simulated the SGTR. Before I reactor trip, the differences between the measured-test and code-calculated, primary-to-secondary-side mAapMKM061 oon\2061w-5.non:lt*070795 5-27 REVislON: 0 I

l pressure drop are very small and the code-calculated tube rupture flow was close to the measured flow. l The integrated calculated and measured flow (Figure 5.5.2-17) were in excellent agreement, and !

validated the break flow model used in the code-calculation. l l

i Between the trip ([ ]'6# seconds) and [ ]"# seconds, break flow was underestimated or overestimated, depending on the prediction of the pressure difference between the RCS and SG-B.

After [ ]'6# seconds, the SG-B pressure was overestimated (up to [ ]'6# psia). Since the RCS pressure was predicted well, ts.e calculated SG-B pressure was slightly higher than RCS calculated pressure, and led to SGTR back flow [ ]"# lbm/sec. This behavior was observed in the test later in the transient at [ ]"# seconds. The sensitivity study (Run 2) presented in Subsection 5.5.2.2.2 showed that a small modification of the SGs* heat losses was sufficient to better simulate the SGs*

pressure, and consequently break flow.

The integrated break flow (Figure 5.5.2-17) was predicted very well up to [ ]'6# seconds. At the end of the simulation ([ ]** seconds), the integrated, calculated-break flow was [ ]"# lbm,

[ ]'6# percent lower than the experimental value.

In the design SGTR analyses for plant licensing purposes, control system functions would be assumed so that the primary-side pressure would be maximized. Using the control system to maximize primary-pressure increases, the tube rupture break flow and the duration of the break flow, which would increase the severity of the transient. The CVCS and pressurizer heaters would be assumed operational in a licensing-basis calculation. Matrix Test 10 represents a less severe transient in that the primary-side depressurization rate was maximized. As a result of this, the primary-side pressure decreased to the point where boiling occurred. The LOFITR2 code assumes homogenous fluid conditions in the primary side. Cases involving stratification of steam and liquid phenomena are outside the range of LOFITR2. This is why the code calculation of Matrix Test 10 terminated at about [ ]'6#

Vessel llead and Pressurizer Levels In the calculation, vessel head boiling occurred at [ ]'6* seconds (Figure 5.5.2-13), [ ]"#

seconds later than for the test as a result of the modeling (see Subsection 5.3.1.5) and of the overestimation of the RCS pressure at this time.

The calet. lated pressurizer level was in good agreement throughout the test (Figure 5.5.2-8) and no external-pressurizer heaters were used during this test. 'Ihe accurate prediction of the pressurizer level evolution confirms the validity of the pressurizer heat loss simulation ([ ]"# kW at [ ]"#*F).

After [ ]"# seconds, RCS pressure decreased slowly and the pressurizer level increased. Since the i

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RCS water was close to saturation, the only physical way to explain this behavior is that pressurizer heat losses caused steam to condense. The accurate pressurizer-level prediction proves that pressurizer heat losses were correctly simulated.

CMT Flow The CMT actuated [ ]** seconds after the start of the transient. During the test, the CMT operated only in the water recirculation mode. As expected during non-LOCA events, the CMTs did not drain; therefore, the ADS was not actuated during this event. The CMT flow rate is shown in Figure 5.5.2-12. In the SPES-2 facility, there are two CMTs. The total code-calculated CMT flow rate is consistent with the total measured test CMT flow rate. The decrease in the total CMT flow due to the warming of the CMT during the transient, was also well predicted.

At [ ]** seconds, the experimental CMT flow started to decrease faster because the hot water reached the injection line. This trend was not predicted by LOFTTR2-AP, due to numerical diffusion that slightly modified the water thermal profile in the CMT. This is typical behavior for the SPES-2 CMT. Full experimental results (Reference 19) showed that the CMT flow never stopped and remained stable at [ ] percent of the initial value because the CMT heat losses continuously cooled the CMT water. Overall, the LOFTRAN-AP CMT model combined with the heat losses simulation as described in Subsection 53.4.4 predicted the CMT flow well.

Figure 5.5.2-12 shows the measured flow from each of the two CMTs in the SPES-2 test facility.

1here were no significant differences in the flow rates of the CMTs. CMT flow rate is a function of cold leg fluid conditions (pressure and temperature) and pressure in the vessel downcomer where the injection line is connected. The slight variation in CMT flow rates was due to the difference of resistance in the injection and balance lines of each CMT, indicating that symmetrical conditions !

occurred at CMT connection points during SGTR events. l The LOFITR2-AP code simulates only one CMT and assumes that performance of two CMTs are identical. In licensing-basis calculations, variation between the two CMT layouts is handled by using the conservative configuration for defining CMT model input. The test confirms the assumption that a single CMT model can be used in the code with appropriate input conservatism for steam tube ruptures or any transients where symmetrical conditions are expected at the CMT-to-RCS connection points.

PRIIR Flow and Temperatures l

Matrix Test 10 was performed with one tube in the PRIIR. The PRIIR lines' friction factors I (Test C-09, Reference 18) have been updated using the final results of the cold pre +perational tests, since the preliminary calculations were made (Reference 2). Using these factors results in a decrease of the PRIIR flow by approximately [10]** percent.

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Comparisons between the test data and the code-predicted PRHR flow rate, inlet temperature, and outlet temperature are shown in Figures 5.5.2-9 to 5.5.2-11. The PRHR was initiated [ ]"# seconds after the start of the transient. De RCPs continued to run until [ ]"# seconds. Because the RCPs continued running, PRHR actual flow developed rapidly and peaked at approximately [ ]"# lbm/sec.

and after the pumps were tripped, decreased to stable natural circulation flow of [ ]"# lbm/sec.

i PRHR flow in the test remained stable at about [ ]'6" Ibm /sec. until around [ ]"# seconds when steam began entering the PRHR from the vessel. After the RCP tripped, the LOFTTR2-AP calculation matched the test closely.

LOF1TR2-AP predicts the initial surge of flow, but probably overestimates it by a factor of [ ]"#.

From detailed investigations, it appears that the LOFITR2-AP calculation is correct. SIET has confirmed that the PRHR flow meter (F_A80E) was out of range for this test (Maximum span =

[ ]"# lbm/sec.). The range of the meter was increased after the SGTR tests. For illustration, Test S01613 indicated a PRHR flow of [ ]""lbm/sec., when the RCS pumps were operating (see Reference 18 and Subsection 4.2.6). Test S01613 had three tubes in the PRHR. One tube in the PRHR represents approximately [ ]"# percent of the fluid resistance in the PRHR loop. With one tube, the PPJ1R flow should be approximately [ ]"# percent of [ ]"# lbm/sec., which is close to the value calculated by LOF1TR2-AP ([ ]"# lbm/sec.).

The calculated temperature drop across the PRHR HX is in good agreement with the test data, indicating the heat transfer models selected are appropriate.

S.5.2.2.2 Sensitivity Studies This subsection presents the results of the sensitivity studies performed, illustrating the influence of the parameters where uncertainty exists.

Run 2 - SG Steady Heat Loss Sensitivity Study Run 2 duplicates Run 1 with a modified or distribution partition of the SGs steMy-state heat losses.

[ ]'6# percent of the steady-state heat losses are accounted for in the SG secondary side, while Run 1 uses [ ]*# percent. Key parameters of Run 2 are presented in Figures 5.5.2-18 through 5.5.2-29.

The decrease by [ ]"# kW ([ ]'6# percent of [ ]"# kW) of the steady-state heat losses in the SGs ,

induced a faster decrease in SG pressure. For example, at [ ]"# seconds, the SG-B pressure (Figure 5.5.2 20) was [ ]** psi, [ ]'6# psi lower than for Run 1. The main impact is that the calculated SGTR break flow (Figure 5.5.2-21) is closer to the experimental value. The SGTR back i flow observed after [ ]"" seconds for Run I disappeared and the integrated calculated break flow became higher than the experimental value at [ ]** seconds ([ ]"# lbm for the calculation,

[ ]"# lbm for the test).

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De increased SGTR break flow induced a minor decrease in RCS pressure (Figure 5.5.2-18).

Extensive boiling in the upper plenum and in the PRHR began at 3000 seconds, both in the calculation and in the test. The simulation was stopped at this time.

The CMT and PRHR flows are not significantly affected by this sensitivity study.

Run 3 - SG Secondary Side Steel Mass Sensitivity Study Run 3 duplicates Run 2 without the SGs' steel mass thermal inertia factor. Key parameters of Run 3 are presented in Figures 5.5.2-30 through 5.5.2-41.

Bis calculation clearly shows that the steel masses on the SGs' secondary sides have a significant impact on the SGs' pressure evolutions (Figures 5.5.2-31 and 5.5.2-32). After reactor trip, the simulation with no SGs steel masses shows a pressure peak at [ ]'6# psi, [ ]'6# psi higher than the experimental value. After [ ]'6" seconds, when the SGs pressure starts decreasing, the l

calculation shows a faster rate of decrease than the experimental result because the steel masses tend to i minimize the temperature (and therefore pressure) evolutions. Once again, the importance of the steel masses for SPES-2 is the result of the 1/395 volume-scale of the facility.

The SG steel masses are an important factor in the total heat capacity of the loop (approximately

[ J'6# percent for each SG). These relatively cold-temperature steel masses contributed significantly to the cooling of the SG water just after the trip. His factor explains why the RCS temperatures in l Run 3 (Figures 5.5.2-34 through 5.5.2-36) increased by approximately [ J F before

[ ]'6# seconds. As a result, the RCS fluid density decreases by approximately [ ]'6# percent.

Fluid expansion was equivalent to an RCS injection of [ ]'6# Ibm of water, and was the reason for l break flow increases between [ ]'6# seconds (Figure 5.5.2-33). RCS pressure responded in such a way that the total integrated break flow increased by approximately [ ]' 6" lbm (Figure 5.5.2-41).

l Excess break flow was the reason for the later recovery of the pressurizer level in Run 3 (Figure 5.5.2-37).

I The CMT and PRHR flows are not significantly affected by this sensitivity study. Run 3 was stopped at 3000 seconds, when boiling in the upper plenum developed.

Run 4 - CMT Line Resistance Sensitivity Study ;

l 1

Run 4 duplicates Run 2 with CMT line resistances at their minimum value (line resistances of Run 2 I minus [ ] percent). based upon the experimental uncertainty resulting from the flow and pressure in Aap6000061-oonC061 w-5.non: n-070795 5-31 REVIsloN: 0 W

drop measurements during the cold pre-operational tests. Key parameters of Run 4 are presented in Figures 5.5.2-42 through 5.5.2-53.

As expected, the reduction of the CMT line resistances by [ ]'6" percent induced the increase of flow injected by the CMTs in the RCS by approximately eight percent (Figure 5.5.2-51). Excess CMT flow induced a faster cooling of the RCS, equivalent to an increase of RCS power extraction of approximately [ ]'** kW. Consequently, RCS and SG pressures and temperatures decreased slightly faster.

Overall, the impact on key parameters of the transient is small. The PRHR flow is not significantly affected by this sensitivity study.

Run 5 - PRHR Line Resistance Sensitivity Study Run 5 duplicates Run 2 with the PRHR lines resistances decreased by [ ]** percent. As mentioned in Subsection 5.5.2.1, the sensitivity study is not based on PRHR data uncertainty, but tests the influence of an excess of power extraction. Increasing the PRHR flow will have the same effect as decreasing the core power or increasing the total loop heat losses. Key parameters of Run 5 are presented in Figures 5.5.2-54 through 5.5.2-65.

As expected, the PRHR flow was increased by approximately [ ]'6* percent, and led to the increase of the power extracted by the PRHR by approximately [ ]'6* kW and faster cooling of the facility.

The integrated break flow is not significantly affected (Figure 5.5.2-65) because of the excess cooling in the RCS (fluid contraction) is compensated for by faster draining of the vessel head (Figure 5.5.2-64),

which is due to the faster decrease in RCS pressure (Figure 5.5.2-54).

Runs 6 and 7 - Time Step Sensitivity Study Runs 6 and 7 showed a very low sensitivity to the time step in the period before boiling occurred in !

the PRIIR. (No figures are provided for these runs.) l l

Run 8 - Noding Sensitivity Study Run 8 duplicates Run 2 with more fluid nodes in the RCS. There are about twice the number of fluid nodes in the RCS. The results show no significant differences between Run 8 and Run 2 validating the noding. (No figures are provided.)

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Hun 9 - RCS Water to Steel lleat Transfer Sensitivity Study Run 9 duplicates Run 2 with heat transfer coefficients between water and steel divided by two. Key parameters of Run 9 are presented in Figures 5.5.2-66 through 5.5.2-79.

The intent of this calculation is to examine the reason for overestimation of RCS pressure and temperature between [ ]** seconds. Decreasing the heat transfer coefficients will result in initialization with lower steel temperatures, and therefore cause a lower drop in the heat transfer after the trip.

As expected, the heat transfer between the RCS water and the steel in Run 9 was higher after the trip (Figure 5.5.2-78). His transfer induced a faster cooldown of the RCS (Figures 5.5.2-70 and 5.5.2-71) and consequently decreased RCS pressure (Figure 5.5.2-66) and SG pressures (Figures 5.5.2-67 and 5.5.2-68).

Since the RCS and the SG pressures are both affected by this sensitivity study in the same way, the SGTR break flow is not really modified (Figure 5.5.2-69). At the end of the simulation, the integrated break flow of Run 2 and Run 9 is identical (Figure 5.5.2-77).

Overall, the transient is not very sensitive to the heat transfer between the RCS water and steel. De CMT and PRHR flows are also not significantly affected in this study.

5.5.2.3 Conclusions Concerning Test 10 he simulation of SPES-2 Matrix Test 10 shows that LOFITR2-AP correcdy simulated the overall trend of the transient.

Based on the results of the sensitivity studies performed,it is apparent that a key parameter is heat loss from the secondary side of the SG. Heat losses caused the faulted SG to depressurize, and therefore have a strong effect on the long-term break flow. The sensitivity study performed in Run 2 shows that a modification of only 0.9 kW or 5 percent of the total-estimated heat loss induced a 40 percent variation in the integrated break flow. Run 1 (base case) shows SGTR back flow after 1600 seconds, while in Run 2 back flow occurred after 4000 seconds.

Since the uncertainty on secondary-side heat loss is higher than 1 kW, it is difficult to make definitive conclusions on the SGTR break flow model during the long-term phase of the transient. Nevertheless, based on Test 10 simulation results, the SGTR break flow model behaved correctly according to the RCS and SG pressure calculations. His behavior proved that the LOFITR2-AP model will conservatively calculate break flow in the SSAR calculations, when maximum pressure drop between the RCS and the faulted SG is assumed.

i I

l mAap600%2061-non\2061w 5.non:lb-070795 5-33 Revision: 0 l

1 The CMT and PRHR flows and temperatures were well predicted during the transient. All sensitivity studies performed showed that they are not overly dependant on RCS parameters when single-phase flow conditions exist.

In regard to time steps and noding, the simulations proved to be independent of these factors.

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Table 5.5.2 2 COMPARISON OF TEST AND LOFTTR2-AP INITIAL CONDITIONS FOR MATRIX TEST 10 LOFTfR2-AP Condition Test Simulation Rod Power , kW Pressurizer pressure, psia Average Hot leg Temperature, 'F Reactor Vessel (Core) Inlet Temperature, F Cold Leg Flow Rate, Ibm /sec.

DC-UH Bypass Flow Rate, Ibm /sec.

Pressurizer lesel, ft CMT level, ft.

CMT Temperature, 'F Initial SG Water Level, ft.

SG MFW Temperature, 'F '

SG Pressure, psia Ambient air Temperature, 'F - -

1 l

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TABLE 5.5.2-3 SEQUENCE OF EVENTS FOR MATRIX TEST 10 - INITIAL CASE (Run 1)

Time (seconds)

Simulation with Event Specified Test LOfTTR2-AP*

Break Opens 0 Pressurizer beaters turned off Pressurizer Low level PZR = [ ]' 6' m Setpoint Reached MSLIV Closure PZR LL [ ]'*' seconds MFWIV Closure PZR LL [ ]'** seconds CMT Initiation PZR LL [ ]*6" seconds PRHR Actuation PZR LL [ ]'** seconds SCRAM simulated PZR LL [ ]'** seconds RCPs Tripped PZR LL [ ]'** seconds Break Flow Terminates Pressurizer Empties Note:

Time of the events are not computed by LOFITR2-AP for this test. Experimental times of the events are used as input data.

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f 5.53 Matrix Test 9 Matrix Test 9 (Test S01309) is a single tube rupture test with the plant at full power. Nonsafety I systems (CVCS, SG valves, and PORVs) were activated during this test. LOFITR2-AP was used to analyze this test. A comparison of the significant sequence-of-events in the simulation and the actual I test results is shown in Table 5.53-3. l Since the PORV and ADS valves 1 and 3 were opened and closed frequently (Table 5.53-2), it was i not possible to simulate the exact sequence of operator actions (LOFITR2-AP can simulate up to l 50 operator actions). Sequencing was grouped according to the total opening time and the integrated !

mass discharge by the valves. Table 5.53-4 gives the exact opening and closing sequence used for the !

PORV and the ADS valves in the simulation. l 5.53.1 Description of the Runs a

De runs performed for Test 9 are as follows:

Run 1 - Base Case

'Ihe assumptions used for the base case of Test 9 are identical to Test 10, Run 1, and are provided in i Subsection 53. l:

Run 2 - SG Secondary Side Heat Losses Starting from Run 1, the SG heat losses are decreased by [ ]'6# between the trip and l

[ ]"#. After [ ]'6# the heat losses of the broken SG are tuned to match the !

actual SG pressure evolution. This case is like a simulation where the SG secondary-side pressure of .l the faulted SG is a known boundary condition. j Run 3 - Initial Pressurizer Level f The preliminary calculations presented in Reference 12 show that initial pressurizer level has a }

significant impact on the pressurizer pressure evolution after the trip. His sensitivity study was l performed to confirm this point for final calculations. i i

5.5.3.2 Results Analysis ;

5.53.2.1 Base Case (Run 1) !

A compadson of the code-calculated results and the actual-test results is shown in Figures 5.53-1 through 5.53 23. As stated previously, this test is simulated according to the actual times of the .;

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events. Minor variations in Table 5.5.3-3 are due to the instantaneous valve closures and openings with LOFITR2-AP simulations (eg., main steam line isolation valve closure).

RCS Parameter Evolution The initial phase of the transient, up to the time when the operator actions were carried out, is similar to Test 10 and is not described in detail for Test 9. In the test, the trip occurred at [ ]"#

when the low-pressurizer level setpoint ([ ]'6#) was reached, [ ]'6# sooner than for Test 10. Several parameters contributed to the earlier trip in Test 9 such as, higher power in the pressurizer heaters, which maintained a higher pressurizer pressure and break flow, and a higher setpoint for the low-pressurizer level ([ ]"# compared to [ ]*6#). However, the CVCS flow at [ ]"# contributed to the reactor trip delay.

The pressurizer-pressure calculation (Figure 5.5.3-2) is closer to the experimental evolution. When the pressurizer emptied (around [ ]""), the rate of depressurization was similar in the LOFITR2-AP simulation and the test (approximately { ]"#). The opening of one ADS-1 valve for ten seconds at [ ] induced a pressure drop close to [ ]"# in both the test and in the simulation. At [ ]"#, RCS pressure was overestimated by [ ]'6#. After

[ ]'6#, the calculation was closer to the test result Calculations presented in the preliminary report indicated that pressurizer-pressure evolution after the trip, was sensitive to the initial pressurizer level. Run 3, presented in Subsection 5.5.3.2.2 of this report, confirms this point.

"Ihe pressurizer behavior during the repeated opening of one ADS-3 valve between [ ]"# and

[ ]'6# (Figure 5.5.3-2) was not predicted very well by the code. In the calculation, repeated opening of the valves induced a larger, pressurizer pressure drop. In the test, the pressurizer pressure increased after each ADS-3 opening. Poor, pressurizer-pressure calculation during this period was probably due to an underestimate of the pressurizer level (Figure 5.5.3-8), which induced less water flashing. Also, the external pressurizer heaters of the pressurizer and the pressurizer wall are not precisely simulated by LOFTTR2-AP (see Subsection 5.3.4.2).

Between [ ]"# and [ ]"#, the code becomes unstable due to boiling in the upper plenum. However, this configuration will not be encountered during the SSAR calculation.

SG Pressure Evolution The SG pressure evolution (Figures 5.5.3-3 and 5.5.3-4) was in relatively good agreement with the test results. The maximum difference during the simulated transient was in the range of +/- [ ]' 6#,

and around [ ]"' most of the time. Compared to Test 10 (no operator action and no startup feedwater), the average rate of depressurization is increased ([ ]"# compared to [

]"#) due to the startup feedwater flow (Figures 5.5.3-18 and 5.5.3-19) that reduced the rate of steaming in the SGs during depressurization.

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4 During the PORV opening phases of SG-A, pressure evolution was qualitatively well predicted. At

[ ]'**, the first opening of the PORV (for 10.5 seconds) induced in the LOFITR2-AP simulation and in the test, a pressure drop of [ ]' 6#. Between [ ]'^*, SG-A pressure dropped by approximately [ J'6# in the calculation and in the test. Total water mass discharged by the PORV during this period was in good agreement with the test results (Figure 5.5.3-21).

Both SG-A and SG-B pressure were underestimated during the transient. This behavior was probably a result of the secondary-side modeling, which uses only one water node at average temperature. In the test, the water was colder in the SG downcomer than in the riser because the startup feed water was cold. The modeling minimized potential flashing of the hotter water in the SGs when the pressure decreased, and led to a conservative calculation of the SGTR break flow.

SGTR Hreak Flow The break flow evolution (Figure 5.5.3-5) was consistent with the pressure difference evolution between the RCS and the faulted SG. At [ ]'b* (end-of-simulation), integrated break flow (Figure 5.5.3-17) was overestimated by about [ ]'6 '.

Vessel IIcad and Pressurizer Level in the calculation. vessel head boiling occurred at [ l'*# seconds (Figure 5.5.3-13),

[ ]'** later than in the test. This effect is related to the modeling (see Subsection 5.3.1.5) ,

I and an overestimate of the RCS pressure at this time.

'Ihe calculated-pressurizer level was in good agreement before the trip (Figure 5.5.3-8). Because the SGTR break flow was overestimated, pressurizer-level recovery happened later in the calculation and the pressurizer level emptied again after [ ]'**.

CMT Flow l The CMT injection flow (Figure 5.5.3-12) was close to experimental flow. At the end of the simulation, the calculated mass lost by each CMT added to the RCS, was approximately [ J' 6 *

(Figure 5.5.3-14).

PRIIR Flow and Temperature The PRIIR flow and temperature (Figures 5.5.3-9 through 5.5.3-11) are well predicted by the code.

Once again, the PRHR flow measured by the flow meter (F_.A80E) is not correct, when the RCS pumps are operating. The experimental value should be closer to [ ]#, as explained in Subsection 5.5.2.2.1. l m:\ap600\2%1-non\2061 w-6c.noo:Itwo71295 5-119 REVIsloN: 0

l 5.5.3.2.2 Sensitivity Study )

l Two sensitivity studies were performed to simulate the influence of the SGs secondary-side heat losses and the initial pressurizer level. '

Run 2 - Steam Generator Heat Losses Run 2 duplicates Run I with SG heat losses decreased by [ ]"# up to [ ]"# and by about [ ]** after this time. These values were selected because they result in a better match with the test SG-B pressure evolution. A comparison of the results of Runs 1 and 2 is presented in Figures 5.5.3-24 through 5.5.3"35.

The main impact of this sensitivity study was that the faulted SG pressure decreased slowly (Figure 5.5.3 26) and was closer to the experimental value. This induced a better prediction of break flow (Figure 5.5.3-27) after the repeated opening of the ADS-3 valve. At [ ]"#,

overestimation of the integrated SGTR break flow is reduced by [ ]"# (Figure 5.5.3-35). For the same reason, draining the vessel head was slower (Figure 5.5.3-34).

i The CMT and the PRIIR flows were not significantly affected by this sensitivity study.

Run 3 -Initial Pressurizer Level Sensitivity Study Run 3 duplicates Run 2 with the initial pressurizer level decreased by [ ]"# so that the calculated-pressurizer level matched the actual-test pressurizer level at the time when the trip was simulated. The comparison of results for Runs 2 and 3 is presented in Figures 5.5.3-36 through 5.5.3-47.

Figure 5.5.3-36 shows that pressurizer-pressure evolution is improved after the trip but remains higher than the actual results (between [ ]"#). Since SG pressure was underestimated during this period (Figures 5.5.3-37 and 5.5.3-38), break flow was still overestimated (Figures 5.5.3-39 and 5.5.3-47).

Overall, the transient was sensitive to the initial pressurizer level, but was less than initially thought during preliminary calculations (Reference 2).

The CMT and PRHR flows were not significantly affected by this sensitivity study.

5.53.3 Conclusion Concerning Test 9 The comments made in the conclusion concerning Test 10 (Subsection 5.5.2.3) apply for Test 9. The three simulations presented have a general tendency to overestimate RCS pressure after the trip and before the ADS opening (at [ ]*#). This tendency causes overestimation of the integrated mAap60(h2061 aon'2061w-6c.non:lb-M1295 5-120 REVisloN: 0

O break flow by about [ ]"#. His behavior was probably due to the modeling of the external pressurizer heaters.

he simulation of the passive safety systems and the nonsafety systems, such as CVCS and SG valves, was accurately performed by LOFITR2-AP.

l l

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TABLE 5.5.3-1 COMPARISON OF TEST AND LOFTTR2-AP INITIAL CONDITIONS FOR MATRIX TEST 9 LOFTTR2 AP l Condition Test Simulation Rod Power , kW -

- a.be Pressurizer pressure, psia Average Hot Leg Temperature, *F Reactor Vessel (Core) Inlet Temperature, F Cold Leg Flow Rate, Ibadsec.

DC-UH Bypass Flow Rate, Ibm /sec.

Pressunzer level, ft.

CMT Level, ft CMT Temperature, 'F Initial SG Water 1.evel, ft.

SG MFW Temperature, 'F SG Pressure, psia Ambient Air Temperature, 'F __ _

)

I 1

l mAarne2061-noeuo61 4c.non:Ib-071295 5-122 REVISION: 0

TABLE 5.5.3-2 MANUAL SG PORY AND ADS VALVE ALTUATION SEQUENCE Steam Generator-A PORY Opened Closed at (sec.) At (sec.)

(sec.) (sec.) Opened Closed

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1 TABLE 5.53-2 (Cont.) 1 MANUAL SG PORY AND ADS VALVE ACTUATION SEQUENCE ADS 1 Valve Opened Closed At (sec.) At (sec.)

(sec.) (sec.) Opened Closed

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TABLE 5.5.3-2(Cont.)

MANUAL SG PORY AND ADS VALVE ACTUATION SEQUENCE ADS 3 Valve Opened Closed At (sec.) At (sec.)

(sec.) (sec.) Opened Closed a b.c a

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TABLE 5.5.3-3 SEQUENCE OF EVENTS FOR MATRIX TEST 9 )

Time (seconds)

Simulation with ,

Event Specified Test LOFITR2-AP* l Break Opens 0 CVCS on PZR L = 3.1 m l

Pressurizer heaters ;

turned off Pressurizer Low Level PZR = 0.676 m Setpoint Reached MSLIV Closure PZR LL +[ 2 seconds]"'

MFWIV Closure PZR LL + [2 seconds]'*' ,

CMT Initiation PZR LL + [2 seconds] ,

PRHR Actuation PZR LL + [2 seconds]'*'

SCRAM simulated PZR LL + [5.7 semnds]"'

SFW Actuation PZR LL = [.676 m]'*'

RCPs Tripped PZR LL + [16.2 seconds] !

Break Flow Terminates Pressurizer Emptics ._.

Note:

Time of events are not computed by LOFTIR2.AP for this test. Experimental times of the events are [

used as input data. !

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TABLE 5.5.3-4 OPERATOR ACTIONS FOR MATRIX TEST 9 Operator Action time (Test Tirne Sec) Operator Action CVCS Start PZR lleaters turned of PORV SG-A Open ADS Opened PORV SG-A close ADS Closed ADS Opened ADS Closed ADS Opened ADS Closed ADS Opened ADS Closed ADS Opened ADS Closed ADS Opened AD5 Closed ADS Opened CVCS turned off ADS Closed ADS Opened ADS Closed PORV SG-A Opened PORY SG-A closed PORV SG-A Opened PORY SG-A closed PORV SG-A Opened PORV SG-A closed IORV SG-A Opened PORV SG-A closed 1 PORV SG-A Opened I IORV SG-A closed I IORV SG-A Opened PORV SG-A closed FORV SG-A Opened IORV SG-A closed PORV SG-A Opened IORY SG-A closed IORV SG-A Opened l

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5.5.4 Matrix Test 1I Matrix Test 11 (Test S01211) is an SGTR test with the plant operating at full power. Nonsafety systems (CVCS, NRIIR, and SFWS) were not activated during the test. Test analyses were done using LOFTTR2-AP. This simulation was a blind test and the blind simulator was presented in Reference 2.

For Matrix Test 11, the ADS actuated [ ]"* after the low-pressurizer water level signal was j reached. This simulation only considered the period of interest to the design-basis SGTR event and I

was terminated shortly after ADS actuation occurred. LOFITR2-AP cannot simulate a situation with a large void generation in the RCS, ADS actuation does not occur for the design-basis SGTR event.

5.5.4.1 Description of the Runs The following three runs explain the differences between the blind simulation and test results.

l Illind Simulation l

The results of the blind simulation, based on the preliminary test initial conditions presented in the Quick look Report in Reference 20, are compared to the actual test results.

Run 1: Updated Input Data

' Die blind simulation needs to be improved because the Guick look Report, from which the input deck was generated, has two errors:

. Initial pressurizer level was at [ ]"# not [ ]'*#

. The pressurizer heaters were on during the first 260 seconds of transient, as described in Subsection 5.5.4.2-1.

Hun 2 : Refinements Since the preliminary report has been completed (Reference 2), refinements have been made. Run 2 of Test 11 uses mainly the same refinements as Test 10 (Table 5.5.1-1). For this run, the time of the events was considered as data.

Table 5.5.4-5 summarizes the evolution of the data among the blind simulation, Run 1, and Run 2.

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5.5.4.2 Resuks Analysis 5.5.4.2.1 Blind Simulation This subsection compares the blind simulation results and the actual test results. Figures 5.5.4-1 through 5.5.4-15 present the code-calculated results for key parameters. Table 5.5.4-2 presents the sequence-of-the events.

In the test simulation, reactor trip and safeguard actuation were assumed to occur when the low-pressurizer level setpoint ([ ]"") was reached. The setpoint was reached in the simulation at

[ ]"". Various safeguards were actuated based on when the low-level setpoint was reached. These included: closure of MSIVs, closure of MFIVs, PRHR actuation, CMT actuation, and <

tripping of the RCPs. These features were performed following prototypical delays after the setpoint was reached. The actuation times listed in Table 5.5.4-2 were used in the LOFTTR2-AP simulation.

l l In the test, the reactor trip occurred at [ ]"#, [ ]"" carlier than in the calculation. As can be seen in Figure 5.5.4-8, the later trip in the calculation was mainly due to the initial pressurizer level being [ ]"", not [ ]'6#, as mentioned above (Reference 20).

I Some other incorrect preliminary input data (Reference 20) were also used with the blind test simulation: j i

Test 11 used pressurizer heaters up to [ ]"# at [ ]'6'. 'Ihe blind simulation l

was performed with no pressurizer heaters, which explains why pressurizer pressure decreased too fast in the blind calculation (Figure 5.5.4-2).

i

The power was increased by [ ]"# to compensate for heat losses that were turned off l

when ADS-1 actuated (Figure 5.5.4-1). In reality, this happened when ADS-2 actuated and is not simulated in the calculation (Reference 19). l

Temperatures in the PRHR lines (Figures 5.5.4-10 and 57.4-11) were initiated with the default !

option of the code, which is not appropriate for the SPES-2 facility. At this time, it was known that the initialization was not correct, according to the actual results of Test 10. This point was not modified because it has very little impact. The RCS pumps operated when the PRHR was actuated. Flow in the PRHR was by forced convection and the water in the inlet and outlet pipes was rapidly replaced, typically in 10 seconds.

Due to the imperfection of the data, the blind simulation is not discussed further. Table 5.5.4-5 summarizes the errors in the input deck used for the blind simulation.

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5.5.4.2.2 Blind Simulation After Updating of the Input Data (Run 1)

Run 1 duplicates the blind simulation after the input data was updated (Table 5.5.4-5). Key parameters of Run 1 are presented in Figures 5.5.4-16 through 5.5.4-30. Table 5.5.4-3 provides the sequence-of-events.

He reactor trip, based on low-pressurizer level, occurs at [ ]'6# earlier than in f Test S01211 (Figure 5.5.4-23). Pressurizer-level evolution is perfectly predicted up to [

l

]'6#. After this time, the calculated-pressurizer level decreased faster than in the test. The reason why the level decreased faster than the test is because the RCS pressure was overestimated (Figure 5.5.4-17) by [ ]'6#, leading to a larger break flow between [ ]'6* and the trip (Figure 5.5.4-20). The overestimation of RCS pressure also induced a higher water density in the l

RCS* and hence, a smaller volume.

As seen in Figures 5.5.4-16 through 5.5.4-30, the general trend of the actual test results is well predicted. A detailed description of this test is provided in the discussion of Run 2 in the following subsection.

5.5.4.2.3 Simulation With Refinements (Run 2)

Run 2 duplicates Run 1 and includes refinements and correction of the PRHR line temperature initialization (Table 5.5.4-5). A comparison of the code-calculated results and the actual-test results is shown in Figures 5.5.4-31 through 5.5.4-47. Table 5.5.4-4 provides the sequence-of-events. Run 2 j simulated the actual times of the events. I RCS Parameter Evolutions The opening of the SGTR break at time zero seconds induced a loss of mass in the RCS and consequently, decreased the pressurizer level. Since the pressurizer's heaters are turned on (at

[ J'6#) as soon as the pressurizer-pressure decreases, pressurizer pressure decreased slowly. At

[ ]'6#, when the pressurizer level reached [ ]'6#, the pressurizer heaters were automatically turned off by the plant computer, inducing a faster rate of depressurization.

In Test S01211, Run 2, the reactor trip and safeguards actuation occurred when low-pressurizer level setpoint ([ ]'6#) was reached. When the setpoint was reached, the pressurizer had a level of

[ ]'6#. This setpoint was reached in the test at [ ]'6#, At[ ]'6#,

code-calculated pressurizer liquid level was [ ]'6# (Figure 5.5.4-38). As explained in Subsection 5.5.4.2.2, a faster decrease in the calculated-pressurizer level was due to overestimating pressurizer-The water density variation between 2000 and 1930 psi at 600 F is 0.2 percent, equivalent to 0.26 ft, of water in the pressurizer.

=V0N061-nonuo61*-7.non:ttat 195 5-177 REVIsloN: 0

pressure. De reactor trip induced a sharp drop in core power, (Figure 5.5.4-31) consequently, the core outlet temperature dropped sharply by about [ ]'6# (Figure 5.5.4-36). At the same time, the reduction in core temperature also caused contraction of the RCS water, and consequently complete pressurizer draindown by the surge line (Figure 5.5.4-38).

Various safety features were actuated in the test when the low-level setpoint was exceeded. These included closure of main steam isolation valves, closure of main feedwater isolation valves, PRHR actuation, CMT actuation, and tripping the RCPs. After these events, the facility operated in free evolution, with no active systems. RCS pressure and temperature evolutions resulted from mass and heat transfers by natural circulation in the RCS, SGs, PRIIR, CMT, break, and external heat losses.

De core power from the test was input to LOFTTR2-AP as a boundary condition. SPES-2 core power was increased by [ ]'*# above the scaled AP600 decay heat (Figure 5.5.4-31), beginning at [ ]'*# after the trip, compensated for higher-than-scaled facility heat losses.

After the RCPs tripped, and during [ ]'*# to [ ]'*#, the passive systems extracted less power than the core power and the global energy balance was positive. Natural circulation in the RCS ,

resulted in increased RCS outlet temperature (Figure 5.5.4-36) and RCS pressure. After

[ ]'*#, when core power decreased to [ ]'*#, the RCS temperature and pressure decreased slowly.

After ADS actuation, at [ ]'*#, primary pressure dropped at a fast rate. The calculation follows the actual test results closely (Figure 5.5.4-32). He simulation was terminated shortly after the ADS valve opening.

Overall, RCS pressure and temperature evolutions are well predicted by LOFITR2-AP, but RCS ]

pressure is overestimated by about 50 psi after the pressurizer heaters are turned off. His is likely l

due to not simulating the heat transfer between the pressurizer steam and the wall during this period. l SGs Secondary-Side Pressure SG-A and -B pressure evolutions are shown in Figures 5.5.4-33 and 5.5.4-34. The SGs pressure is essentially constant prior reactor trip because SG pressure is regulated by the control system of the i facility.

After the SGs steam and feedwater valves are isolated, there is no mass transfer in the SGs except for the SGTR break flow in SG-B, SGs pressure is then driven by the energy balance including heat I transfer with the RCS and the SGs secondary-side steel masses. The temperatures of the SG steel masses depend on the steel thermal inertia and heat transfer with the SG water and the external air. !

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After the steam valve's isolation, the SG pressures initially increased from [ ]'6"asa result of the heat transfer between the RCS and the SGs. No secondary-side relief valves were opened. The trip induced a fast decrease in core power, and also in the hot leg temperature.

Combined with the heat transfer of the SG steel masses (relatively cold,in thermal equilibrium with the initial SG pressures), the SG pressures stabilized at [ ]# and decreased slowly.

Eventually, some subcooled water in the SGs contributed to stabilize SG pressure, shortly after the ,

reactor trip.

After the RCP trip occurred at 496 seconds, RCS natural circulation developed with a temporary increase in the hot leg temperature to about [ ]'6" During the same period, heat transfer with the SG secondary-side steel decreased because the steel became warmer. The combination of these phenomena is the reason for the SG pressure increase after 550 seconds.

Global behavior of the SG pressures is qualitatively well simulated.

SGTR Break Flow The measured-test break flow and the code-calculated break flow are shown in Figure 5.5.4-35. The SGTR break flow is a function of the difference between the primary and secondary-side pressures.

Water temperature at the break location also has an effect. SPES-2 SGTR is simulated with a pipe connecting the pump-B suction line and the SG-B secondary-side; this may induce some minor bias.

At zero seconds, when the SGTR break opened, calculated-break flow was 0.122 lbm/sec., compared to 0.130 lbm/sec. for the actual break flow. This six percent initial underestimate disappeared rapidly, probably due to the presence of cold water in the pipe that simulated the SGTR. Prior to reactor trip, the differences between measured-test and code-calculated, primary-to-secondary side pressure drop was small and the code-calculated tube rupture flow was close to the measured flow. When the reactor tripped, integrated calculated and measured flow (Figure 5.5.4-45) were in excellent agreement, and validated the break flow model used in code calculation.

Between reactor trip and [ ** break flow was underestimated by [

, ]'** probably because LOFTTR2-AP simulated L i:. break flow as a sink located in the SG outlet header. The calculated fluid temperature was higher than at the actual temperature at the inlet of the pipe that simulated the SGTR in the facility at this point.

Integrated break flow (Figure 5.5.4-45) was well predicted.

l mWxco61-nonc061w-7 non:lt,07:195 5-179 Revision: 0

l Vessel Head and Pressurizer Levels No boiling occurred in the vessel head during the period simulated.

l De calculated pressurizer level was in excellent agreement before 300 seconds (Figure 5.5.4-38). As j mentioned earlier, it decreased too fast after 300 seconds, due to the RCS pressure overestimation. !

l i

CMT Flow he CMT actuated [ ]*# after the start of the transient. The CMT flow rate is shown in l Figure 5.5.4-50. With two CMTs in the SPES-2 facility, the total code-calculated CMT flow rate was l

consistent with the total-measured test CMT flow rate. Duling fast depressurization, the calculation indicated some peaks not observed during the test. Since these peaks were very brief, diey had a little impact on the integrated injected flow. Mass balance was maintained during the peaks.

PRHR Flow and Temperatures Matrix Test 10 was performed with one tube in the PRHR. Since the preliminary calculations presented in Reference 2, the PRHR lines friction factors have been updated, using the final results of the cold pre-operational tests, (C-09, Reference 18). As a result, the PRHR flow was decreased by ;

approximately 10 percent. !

Comparisons between the test data and the code-predicted PRHR flow rate, inlet temperature, and outlet temperature are shown in Figures 5.5.4-39 to 5.5.4-41. De PRHR was initiated 479 seconds j after the start of the transienL he RCPs continued to run until [ J'"#. Because the RCPs j continued to run, the PRHR actual flow developed rapidly to a peak of approximately [ ]*# l and then after the pumps were tripped, decreased to stable natural circulation flow at [ ]**. l The initial surge of flow was predicted by LOFITR2-AP, but it appeared to be overestimated by a factor of two (point already mentioned in the preliminary validation report - Reference 2). After detailed investigations, it appears that the LOFITR2-AP calculation is correct. SIET has confirmed j that the PRHR flow meter (F_A80E) was out ofits range for this test (Maximum flow = l

[ ]'**). De range of the flow meter was increased after the SGTR tests. For illustration, Test S01613 indicated a PRHR flow of [ ]*#, when the RCS pumps were operating I (Reference 12). Test S01613 has three tubes in the PRHR. With one tube in the PRHR, the tube represents approximately 20 percent of the fluid resistance of the PRHR loop. With one tube, the !

PRHR flow should be approximately 90 percent of [ ]'6 *, which is close to the i LOFTIR2-AP calculated value.

Boiling in the PRHR occurred at [ ]"# in the calculation; the calculation is not valid after this time. His event will not occur during non-LOCA events simulated with LOFITR2-AP.

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. l i

The calculated temperature drop across the PRHR HX was in good agreement with the test data,

~ indicating the heat transfer models selected were appropriate.

5.5.43 Conclusion Concerning Test 11 The initiation phase of Test 11 is identical to Test 10, except that the initial conditions are a little different, especially the initial pressurizer level and the pressurizer low-level setpoint. l The calculations presented in this report show that most of the differences between the blind simulation and the actual test data can be explained by errors in the input data used for the blind simulation. With updated input data, the pressurizer low-level setpoint is reached at [

]'6#, [ J'"# sooner than in the test. j The short period simulated after the ADS opening is correctly simulated up to the point where boiling l begins to occur in the facility, f

l l

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TABLE 5.5.4-I COMPARISON OF TEST AND LOFITR2-AP INITIAL CONDITIONS FOR MATRIX TEST II LOFTTR2-AP Condition Test Simulation ~

Rod Power, kW

-a, I Pressurizer pressure, psia Average lil Temperature, "F Reactor Vessel (Cae) Inlet Temperature, *F Cold Leg Flow Rate, Ibm /sec.

DC-UH Bypass Flow P.ste, Ibm /sec.

Pressurizer Lvel, ft. .

CMT Level, ft.

CMT Temperature *F Initial SG Water Level, ft.

SG MFW Temperature. *F SG Pressure, psia Ambient air Temperature. *F - ._

Note:

Test results show an initial pressurizer level of [ ]*' comes from Reference 20)

The temperature measured by this thermocouple is likely overestimated.

m:\ap6072061 - non\2061 w.7.non: l t>C71 195 3 182 REVISION: 0

.*- 1

.I ll l

1 l

TABLE 5.5.4-2 SEQUENCE OF EVENTS FOR MATRIX TEST 11 - Blind Simulation l

Time (seconds)

Simulation with Event Specified Test LOITTR2-AP Break Opens 0 - "'

Pressurizer beaters turned off Pressurizer Low level 3PZR = 0.676 m Setpoint Reached MSLIV Closure PZR LL + 2 seconds MFWIV Closure PZR LL + 2 seconds CMT Initiation PZR LL + 2 seconds PRHR Actuation PZR LL + 2 seconds SCRAM simulated PZR LL + 5.7 seconds RCPs Tripped PZR LL + 16.2 seconds ADS-1 Opened PZR LL + 150 seconds I

Pressurizer Empties _ _

1 l

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l l

. i TABLE 5.5.4-3 SEQUENCE OF EVENTS FOR MATRIX TEST 11 Run 1 Time (seconds)

Simulation with Event Specified Test LOFTTR2-AP ,

Break Opens 0 -

Plessurizer beatas turned off

Pressurizer Low Level PZR = 0.676 m Setpoint Reached MSLIV Closure PZR LL + 2 seconds MFWIV Closure PZR LL + 2 seconds CMT Initiation PZR LL + 2 seconds PRHR Actuation PZR LL + 2 seconds SCRAM simulated PZR LL + 5.7 seconds RCPs Tripped PZR LL + 16.2 seconds ADS-1 Opened PZR LL + 150 seconds [

Pressurizer Emptics - -

t e'

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s TABLE 5.5.4-4 SEQUENCE OF EVENTS FOR MATRIX TEST 11 - Run 2 Time (seconds)

Simulation with Event Specified Test LOFTTR2-AP

Bred Opens 0 - **

Pressurizer beaters turned off Pressurizer Low Level PZR = 0.676 m Serpoint Reached ,

MSLIV Closure PZR LL + 2 seconds MFWIV Closure PZR LL + 2 seconds CMT Initiadon PZR LL + 2 seconds PRHR Actuation PZR LL + 2 seconds SCRAM simulated PZR LL + 5.7 seconds RCPs Tripped PZR LL + 16.2 seconds ADS-1 Opened PZR LL + 150 seconds Pressurizer Empues - -

Notes:

Time of events are not computed by LOFTTR2-AP for this test. Experimental times of the events are used as input data.

L i

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. . _ . - . . . . -. _- - - . . . = - - . - _ . . --_

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l TABLE 5.5.4-5 i SGTR MATRIX TEST 11 - EVOLUTION BETWEEN THE BLIND SIMULATION, RUN 1, AND RUN 2 ,

Blind 2 Data Simulation Run1 Run 2 Initial Pressurizer level, ft. - -

a,b.c Pressurizer Heaters,10.2 kW during 260 sec. I Core Power shutoff by 150 kW _ _

Initial PRHR Line Temperatures Tuned No No Yes Refinements as Described in Table 5.5.1-7 No No Yes l

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5.6 Test Simulation Results: Test 12 - Main Steam Line Break Results of Matrix Test S01512 LOFTRAN-AP simulations are divided into two subsections.

Subsection 5.6.1 presents the results of the pre-data release simulations. Subsections 5.6.2 present the results of the post-data release simulations. An assessment of the LOFTRAN-AP MSLB simulations is provided in Subsection 5.7.2. An overall assessment of the LOFFRAN-AP and LOFITR2-AP code for use in design basis non-LOCA and SGTR analyses is provided in Section 6.

The initial conditions and sequence of events for Test S01512 and the simulations are shown in Tables 5.6-1 and 5.6-2, respectively.

5.6.1 Pre-Data Release Simulations Matrix Test 501512 is a blind MSLB test. Prior to release of the blind data, two cases were simulated. These cases are referred to as the base and sensitivity cases. Due to the conservative nature of the LOFTRAN steam break model, it was known that the code would overpredict the severity of the system cooldown and break energy.

i The base case modeled a break on steam line-A with both SGs blowing down pure steam (with a l quality of one) until a specified time when automatic steam iine isolation occurred. After steam line I

isolation, SG-A continued to blowdown. The steam quality sensitivity case used a conservative steam quality profile consistent with a full-scale Westinghouse SG used in design-basis mass and energy release for containment integrity analyses. The intent of the sensitivity case was to show a move in the direction toward the test data. Without the test data, there was no way to know the adequacy of this profile. Also, prior to data release,it was known that the break flow from the intact SG would be overpredicted since line losses are not modeled. Although these factors contribute to the overprediction of system cooldown. these modeling techniques are used in LOFTRAN since they are conservative for design-basis calculations.

Figures 5.6-1 through 5.6-3 present comparisons of pressurizer pressure and SG pressure for the two LOFTRAN-AP simulations and the test data. Figures 5.6-4 and 5.6-5 present comparisons of CMT and PRIIR flow rate. An assessment of the pre-data release simulations is provided in Subsection 5.7.2.1. Based on the results of the pre-data release simulations, additional simulations were performed. The results of these additional simulations are documented in Subsection 5.6.2.

5.6.2 Post Data Release Simulations To illustrate the adequacy of LOFTRAN-AP to predict the key phenomena of the MSLB event, several variations of code input were used. Each variation is described and the results are presented in the following. Additional simulation cases are presented to clearly demonstrate deviations in LOFTRAN-AP simulation results from the test data are either due to phenomena exaggerated by the m:\ap600\2061 non\2061w-9.non:lt>-071195 5-234 REVisloN: 0

SPES-2 facility and are not important to the AP600 plant, or are due to intentional conservative code simplifications.

In each case, the code predicted the overall trends very well and the conservative nature of the break flow model and LOFfRAN-AP code can be observed. In all cases, LOFfRAN-AP simulation results are in excellent agreement with the PRHR flow rate. The CMT flow rate was at first underpredicted, but later reached perfect agreement. The CMT did not drain and remained in the recirculation phase throughout the simulation.

Run 1 - Post Release case A base case is defined to compare the effect of several input variations. The base case differs from the base case used to release the blind data in that the break flow quality input was manipulated to !

achieve a mass flow rate comparable to the mass accumulation rate at the test facility. The integral of steam line break flow is provided in Figure 5.6-11 for the base case as compared to the test data.

Note that the test data is approximately [ ]'** lbm more than the simulation. 'Ihis difference is ,

largely due to the mass in the steam lines, which is not considered in LOFTRAN-AP since it does not contribute to the plant cooldown.

Transient plots are provided for the key parameters described in Subsection 5.2.2. Figures 5.6-6 through 5.6-13 provide transient plots of LOFTRAN-AP simulation data with test data for pressurizer pressure, broken (SG-A) and intact (SG-B) SG pressure, RCS temperature, CMT flow, and PRHR 110w.

The pressurizer-pressure evolution, which describes the system cooldown, was much quicker and was more severe than the test data; however, note that the basic trends and break points are consistent with j the test data. As previously stated, CMT and PRHR flow are well predicted. The CMT did not drain l

and remained in recirculation mode.

Run 2 - SG Metal Heat Capacity Run 2 simulated the quality profile from the base case (Run 1), but it also simulated the effect of the large metal heat capacity of the SGs and the effect of reverse heat transfer from the large heat capacity of the intact SG. 'Ihe effect of the large metal mass heat capacity of the SGs was simulated by increasing the SG tube metal mass, which is an existing LOFTRAN-AP input parameter. Note that the entire metal mass of one SG is 6500 lbm with a heat capacity of approximately 780 Btu /'F.

Since much of this metal is in locations where there would not be interaction with the flashing water, the effect was simulated by increasing the tube mass heat capacity by an arnount equivalent to the annular downcomer metal mass heat capacity.

Transient plots are provided for the key parameters described in Subsection 5.2.2. Figures 5.6-14 through 5.6-21 provide transient plots of LOFTRAN-AP simulation data with test data for pressurizer m46000061-na=\2061--9 non:1t>-071195 5-235 REVislON: 0

o.

.- 1 1

I pressure, broken (SG-A) and intact (SG-B) SG pressure, RCS temperature, CMT flow, and PRHR flow. All parameters are consistent with the test data.

Run 2 was varied by increasing the tube metal heat capacity by 20 percent or to a value equivalent to about 15 percent of the total SG metal mass capacity. Tids case provides the best agreement with test data. Figures 5.6-22 through 5.6-29 provide transient plots for tids case.

E l

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. l TABLE 5.61 COMPARISON OF TEST LOFTRAN-AP CONDrrIONS FOR MATRIX TEST 12 Condition Test LOFTRAN-AP Simulation .

~ ~ * !

Rod Power. kW Pressurizer Pressure, psia Average HL Temperature, 'F Total CL Flowrate, Ibm /sec.

DC-UH Bypass Flowrate Pressurizer Vol., fL8 Accumulator Water Temperature, 'F Accumulator Pressure, psia l

Cold Leg Balance Line Temperature, 'F -

CMT level ;

CMT Temperature, 'F I

l SG Water Mass, Ibm l I

SG Pressure, psia DfEd

[ jw

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TABLE 5.6-2 SEQUENCE OF EVENTS FOR TEST S01512 (MATRIX TEST 12)

Event Actual Time (Sec.)")

1 Break Opens (SG PORV A) **e '

CMT-IV Opening f

a PRHR HX Actuation MSLIV Closure :

RCPs Tripped SFW-A Flow Began SFW-A Flow Ended '

ADS (4)

IRWST Injection (4)

Notes:

(1) Actual times used as LOFIRAN-AP input except where noted.

[ ]=6e s

[ ]=6e l (4) ADS did not actuate due to CMT level, and the IRWST did not inject throughout this transient.

4 f

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DOCUMENT. !

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l S.7 Assessment of SPES 2 Simulation Results 5.7.1 Steam Generator Tube Rupture Test Simulations This subsection summarizes how LOFIRAN-AP handles the key plenomena defined in Subsection 5.2.1 for the SGTR simulations, in reference to the SPES-2 test results.

Break Flow Simulations of the three SGTR tests showed that the LOFITR2-AP break flow model predicts an accurate flow evolution, when the RCS and faulted SG pressure are accurately predicted.

The model responds well in direct flow as well as in back flow.

The short-term, integrated break flow (during the first 2000 seconds of each transient)is predicted within ten percent. Due to the uncertainty on the SG secondary-side heat losses, and the high sensitivity of these parameters, the calculation results are less accurate for the long term simulation. The sensitivity studies show results in the range of 25 percent.

Pressurizer Pressure and Level Before reactor trip, the pressurizer level and pressure evolutions were always accurately predicted, inducing the prediction of the trip time based on pressurizer low-level pressure within five percent.

During the long-term evolution and after the trip, all simulations showed a qualitatively good prediction of pressurizer-level recovery, but the deviation in the integrated SGTR break flow resulted in some deviation of the pressurizer-level calculation at the end of the simulation of Test 9.

Steam Generator Pressure Simulating SG pressure using the simple thermal model, which factors in steady SG heat losses and the thermal inertia of the SG walls, resulted in prediction of the SG pressure evolution within t 60 psi, for most of the transients durations.

It is important to note that for the AP600 SSAR calculations, the faulted SG pressure will be limited by the PORV valve setpoint. Therefore, the SG pressure will be stable at this setpoint, well predicted by LOFITR2-AP.

nr\np600C061 -mon \2061 w.9.non: l tW 1 195 5-268 REvlsioN: 0

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. Primary Side Inlet and Outlet Temperature De primary-side temperature evolution was well predicted by LOFTRAN-AP, and the difference between the calculations and measurements was within 10*F, most of the time.

. PRHR Flow and Temperature Evolutions The PRHR flow and temperature evolutions were well simulated in single-phase. The modeling was not thoroughly investigated with boiling in the PRHR because in the tests, significant boiling occurred in the outlet plenum of the facility at the same time as boiling in the PRHR. Boiling in the outlet plenum is out of the scope of LOFTRAN-AP.

All the sensitivity studies performed on the RCS and SG parameters showed that PRHR behavior was not significantly affected and indicated that PRHR model is robust in single-phase.

. CMT Flow The CMT flows were always predicted well by LOFTRAN-AP. As for the PRHR, the CMT flow was not significantly affected by the sensitivity studies performed on the RCS and ,

SG parameter. Once again, the modeling of the passive systems was shown to be robust.

. Automatic Depressurization System The short period simulated after ADS valve opening during Test 11 was correctly simulated up to the point were boiling occurred in the facility. The pressurizer-pressure drop-off was precisely predicted.

1

. Nonsafety Systems Nonsafety systems such as CVCS, SG valves, and pressurizer heaters were used during Test 9. Using the nonsafety systems did not induce perturbation of the behavior of the safety systems. The simulations presented show that LOFTRAN-AP has the capability to simulate the nonsafety systems.

Table 5.71 summarizes this subsection and illustrates why the LOFTTR2-AP code is valid for AP600 SSAR SGTR calculations.

5.7.2 Main Steam Line Break Test Simulations This subsection summarizes how LOFTRAN-AP addresses the key phenomena defined in Subsection 5.2.2 for MSLB simulations.

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e

, 1 Due to the simplified and conservative break flow model and non-prototypical characteristics of the SPES-2 facility, several simulations demonstrated the adequacy of the LOFTRAN-AP code for conservative design-basis analyses.

Steam Line Break Modeling The LOFTRAN-AP steam generator and steam pipe break model is based on input derived from production codes, design data, and other more sophisticated multi-phase flow codes. For design-basis analyses sensitive input parameters are adjusted in the conservative direction. The LOFTRAN-AP code does not predict break flow quality. For design-basis calculations either a pure steam quality of one is assumed to maximize the primary side cooldown or conservative profiles are developed based on NOTRUMP code calculations and established safety analysis methodology to maximize the effect on the parameters being compared to the applicable regulatory criteria.

Steam Generator Pressure The pressure and pressme transient evolution is related to the steam break modeling described above.

Pressure in the broken SG (SG-A) decreased faster than in the test data. This variation was expected and is consistent with conservative analysis. Overall, the rate of pressure decrease is faster in LOTTRAN-AP. LOFTRAN does not model the heat capacity of the SG, which contributes to tie under-prediction, relative to test data. In both the LOFTRAN-AP simulations and the test, SG-A blew down to ambient pressure at relatively the same time.

For SG-B, the pressure decreased rapidly to a pressure comparable to the test data. Reverse heat transfer for SG-B to the primary side occurred giving the same general trend as observed in the test data. It should be noted that LOFTRAN-AP does not model heat losses from the SG.

Pressurizer Pressure The rate and extent of primary-side cooldown is illustrated well by observing the pressure evolution.

Comparison of LOFTRAN-AP with SPES-2 data shows a tendency of LOFTRAN-AP to overpredict the system cooldown. This overprediction is a result of the conservative models developed for safety analysis calculations.

LOFTRAN-AP predicts pressure break-points and overall trends very well. As demonstrated by the sensitivity cases, given a more sophisticated break flow model and secondary-side component metal mass model, nearly perfect agreement to the test data would occur.

Primary Side Temperature I Prediction of primary-side temperature evolution was consistent with pressurizer pressure.

LOFTRAN-AP tends to overpredict the primary-side cooldown response to the SG blowdown.

m Aapw0C061 -non\2061 w-9.non: ! b-071195 $.270 REVIslON: 0 l

Sensitivity runs showed that overprediction is related to the conservative break flow model and the effect of the SG metal mass.

RCS Flow LOFTRAN-AP adequately predicted RCS flow during the time span ofinterest for design-basis non- l LOCA analyses. LOFFRAN does not allow negative flow in the loop with the pressurizer. However,

{

since the pressurizer loop also contains the PRHR system, loop flow never approached reverse-flow or j stagnant-flow conditions.

CMT Behavior l

CMT behavior was adequately simulated by the LOFTRAN-AP code model. Initially, LOFTRAN-AP ;

under-predicted CMT flow, which is conservative for design basis analysis. As the transient progressed, the code and test data converged to the same value. In both the test and code simulations, the CMTs did not drain and remained in the recirculation phase. :

The behavior of the individual CMT flow test data shows that even for an asymmetric MSLB event, both CMTs behave alike. This behavior provides justification for using the single CMT model in I conservative design-basis analysis applications.

Overall, the LOFTRAN-AP code demonstrates the ability to conservatively model the CMT system for design-basis, steam line break (and other non-LOCA) events.

PRHR Behailor The LOFTRAN AP code precisely predicted PRHR flow during Test 12. The apparent difference between the test and simulations is the result of a slightly positive bias in the test data. Simulated PRHR flow rate was not sensitive to the rate of system cooldown. >

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_ _ _ _ _ _ _ _ _ ~

R TABLE 5.7-1 g

ASSESSMENT OF SPES-2 SIMULATION RFEULTS I LOFITR2-AP Simulation

y Component & System 7 Phenomenon SPES-2 AP600 Comments 4

Pressurizer Pressure & Level + +

g

{ Steam Generator Pressure +/- ++ SPES-2 simulation not excellent, due to the uncertainty on the actual test SG heat losses and the high sensitivity of the simulation on this l parameter.

AP600 SSAR calculations will be performed with SG pressure controlled by the PORV setpoint.

RCS Temp-rature + ++ AP600 temperatures should be better predicted b:cause the effect of phenomena associated with heat losses and the energy stored in the system metal masses are less important for AP600 than for SPES-2.

'? The LOFITR2-AP modelling of this phenomena is simple.

d N Break Flow +/- ++ Excellent conservative prediction expected for AP600 because the SG pressure is fixed by the PORV setpoint, and the RCS pressure is maximized.

PRHR Flow and Temperature + +

CMT Flow + +

ADS + + AP600 and SPES-2 simulation valid only during a limited period, before boiling in the RCS.

Non Safety Systems + +

Note:

Simulations are ranked as follows:

- Poor fs + Good

++ Excellent

.O.

O 9

e 6.0

SUMMARY

OF TIIE LOFTRAN CODE VALIDATION EFFORT This section summarizes 6e overall code validation effon and the results presented in this repon.

This report provides 6mib of the LOFTRAN simulations of the CMT and SPES-2 tests used to validate the LOFTRAN CMT model and integral plant response during transient situations. The CMT model validation is based on separate-effects tests conducted at the CMT test facility. Similar efforts for the PRIIR components and SPES-1 natural circulation tests, which also support code validation are presented in Appendix 15B, Revision 0 of the SSAR and are summarized below.

While not a primary purpose of this report, the validation exercises herein further validate LOFTRAN models.

6.1 Role of LOFTRAN in Safety Analysis The LOFTRAN code is used to calculate NSSS transients given a set of boundary conditions and a transient forcing function. The code simulates the transient based on user-supplied input. By specifying minimum or maximum initial conditions, safety-system setpoints, relief- and safety-valve capacities, core kinetics' parameters, and safeguards system, thermal-hydraulic performances, the code supplies conservative and bounding analysis results. The transient forcing functions, such as the steam break model, also contain conservative modeling assumptions or are supplied with conservative input parameters to achieve a conservative system response.

Code inputs are based on design data and where applicable, uncertainties are included and applied in the direction, which provides conservative response relative to acceptance criteria or safety analysis limit. The safety-analysis limit includes margin to design limits. The overall approach then includes conservative models, minimum or maximum code input values, and margin in the acceptance criteria giving an overall conservative result.

LOFTRAN is used in conjunction with other codes. For example FACTRAN, a detailed fuel rod model, is used for detailed, heat-flux calculations, and THINC or WESTAR is used for DNBR calculations. LOFTRAN provides conservative and bounding boundary conditions to these codes.

6.2 Adaptation of LOFTRAN and LOFTRAN Based Safety Methodology to Advanced Passive Plant Designs The primary role of the LOFTRAN code is to perform conservative simulations of non-LOCA and SGTR licensing-basis events for pressurized water reactors. 'lhe LOFTRAN code has a long history of use in conservative design basis analysis calculations. LOFTRAN has been extensively reviewed by the USNRC and is approved for use in the non-LOCA and SGTR event, design-basis analyses.

m:\ag e00(2061 - oon\2061 w-9.non: l t-070995 61 REVlsloN: 0

Specifically, the LOFTRAN code is used to analyze the following subset of transients presented in AP600 SSAR (* - denotes transients which use or are the result of passive AP600 features).

Feedwater system malfunctions resulting in cooldowns

- Excessive increase in secondary steam flow

- Loss of load / turbine trip

- Complete loss of forced RCS flow

- Uncontrolled RCCA bank withdrawal at power Partial loss of forced RCS flow

- Locked or broken RCP shaft Stanup of an inactive RCP Inadvenent RCS depressurization

- Inadvertent opening of a steam generator relief valve, steam system piping failure

- Loss of ac power, loss of normal feedwater

- Feedwater system pipe breaks -

- CVCS malfunction resulting in increase in RCS inventory

- Inadvertent PRIIR operation

- Inadvertent ope-ation of the CMTs

- Steam generator tube rupture

LOFTRAN-AP is a modified version of LOFTRAN, which includes models for the passive safety systems. Of the above transients, approximately half do not actuate any of the AP600 passive features 1 l

and are analyzed with previously licensed LOFTRAN models.

The methods used in con,iunction with LOFTRAN have been previously reviewed and are consistent with NOREG-0800 (Refere:2ce 25) . The methods used and applied to the AP600 plant are consistent 1 with methods used for operating PWRs with active safety systems. Adaptation of previously used safety analysis methods to the AP600 plant is described in the Code Applicability Docwnent (Reference 8) and Chapter 15 of the AP600 SSAR.

l 63 Code Validation Tests !

l 63.1 SPES-1 Natural Circulation Tests Validation of the LOFTRAN natural-circulation flow model based on SPES-1 tests is presented in Appendix 15B, Revision 0 of SSAR. The conclusions of that effort are summarized in this subsection.

LOFTRAN natural-circulation model verification is based on SPES-1 Test SPNC-01, which focused on single-phase, natural-circulation conditions. LOFTRAN input ws developed for the SPES-1 test facility, with power levels and heat losses adjusted to match the test conditions. Comparison of l

1 m:WO61-nonV061w-9.non:lt470995 62 REVIsloN: 0

s

(

6.0

SUMMARY

OF THE LOFTRAN CODE VALIDATION EFFORT This section summarizes the overall code validation effort and the results presented in this report.

His report provides details of the LOITRAN simulations of the CMT and SPES-2 tests used to ,

validate the LOFTRAN CMT model and integral plant response during transient situations. De Chfr j model validation is based on separate-effects tests conducted at the CMT test facility. Similar efforts '

for the PRHR components and SPES-1 natural circulation tests, which also support code validation are :

presented in Appendix ISB, Revision 0 of the SSAR and are summarized below. ;

While not a primary purpose of this repon, the validation exercises herein funher validate LOFFRAN .

models.

6.1 Role of LOITRAN in Safety Analysis ne LOFTRAN code is used to calculate NSSS transients given a set of boundary conditions and a trmsient forcing function. The code simulates the transient based on user-supplied input. By siecifying minimum or maximum initial conditions, safety-system setpoints, relief- and safety-valve c Apacities, Core kinetics' parameters, and safeguards system, thermal-hydraulic performances, the code supplies conservative and bounding analysis results. The transient forcing functions, such as the steam break model, also contain conservative modeling assumptions or are supplied with conservative input ;

parameters to achieve a conservative system response.

Code inputs are based on design data and where applicable, uncertainties are included and applied in the direction, which provides conservative response relative to acceptance criteria or safety analysis limit. The safety-analysis limit includes margin to design limits. The overall approach then includes conservative models, minimum or maximum code input values, and margin in the acceptance criteria [

giving an overall conservative result. l l

LOFTRAN is used in conjunction with other codes. For example FACTRAN, a detailed fuel rod model, is used for detailed, heat-flux calculations, and THINC or WESTAR is used for DNBR l calculations. LOFTRAN provides conservative and bounding boundary conditions to these codes.

6.2 Adaptation of LOFTRAN and LOFTRAN Based Safety Methodology to Advanced Passive l Plant Designs ,

De primary role of the LOFTRAN code is to perform conservative simulations of non-LOCA and SGTR licensing-basis events for pressurized water reactors. The LOFTRAN code has a long history of use in conservative design basis analysis calculations. LOFTRAN has been extensively reviewed by the USNRC and is approved for use in the non-LOCA and SGTR event, design-basis analyses.

mAap6000061 non0061whoon:1b-070995 61 Revision: 0 i

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[

)

Specifically, the LOFTRAN code is used to analyze the following subset of transients presented in f AP600 SSAR (* - denotes transients which use or are the result of passive AP600 features). !

- Feedwater system malfunctions resulting in cooldowns ,

- Excessive increase in secondary steam flow Loss of load / turbine trip Complete loss of forced RCS flow i Uncontrolled RCCA bank withdrawal at power f

Partial loss of forced RCS flow i

- Locked or broken RCP shaft !

Startup of an inactive RCP l

- Inadvertent RCS depressurization

- Inadvertent opening of a steam Eenerator relief valve, steam system piping failure

- Loss of ac power, loss of normal feedwater [

- Feedwater system pipe breaks !

- CVCS malfunction resulting in increase in RCS inventory {

- Inadvertent PRHR operation [

i

- Inadvertent operation of the CMTs

- Steam generator tube rupture LOFTRAN-AP is a modified version of LOFTRAN, which includes models for the passive safety !

systems. Of the above transients, approximately half do not actuate any of the AP600 passive features l and are analyzed with previously licensed LOFTRAN models. ;

The methods used in conjunction with LOFTRAN have been previously reviewed and are consistent with NUREG-0800 (Reference 25) . 'The methods used and applied to the AP600 plant are consistent j with methods used for operating PWRs with active safety systems. Adaptation of previously used j safety analysis methods to the AP600 plant is described in the Code Applicability Docwnent l (Reference 8) and Chapter 15 of the AP600 SSAR.

6.3 Code Validation Tests '

q 6.3.1 SPES 1 Natural Circulation Tests Validation of the LOFTRAN natural-circulation flow model based on SPES-1 tests is presented in Appendix ISB, Revision 0 of SSAR. The conclusions of that effort are summarized in this subsection.

LOFTRAN natural-circulation model verification is based on SPES-1 Test SPNC-01, which focused on single-phase, natural-circulation conditions. LOFFRAN input was developed for the SPES-1 test facility, with power levels and heat losses adjusted to match the test conditions. Comparison of ar\npMKh2061-non\2061w 9.non:ltr070995 6-2 REVIslON: 0 l

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. l LOFTRAN simulations to the test data show a very good capability of the code to predict natural- l circulation flow rates. The RCS flow rate was predicted by the code with less than a 4 percent error i for the various power levels and heat loss cases. Moreover, if needed, the code response can easily be adjusted to provide a conservative response in terms of flow rates and pressure drops.

Additional details are presented in Appendix ISB, Revision 0 of SSAR.

63.2 PRHR Tests Validation of the LOFTRAN PRHR model based on Westinghouse PRHR Test Facility test data is )

presented in Appendix ISB, Revision 0 of SSAR. The conclusions of that effort are summarized in l this subsection.

'Ihe LOFTRAN PRHR model was verified by comparing LOFTPAN predictions to PRHR test data obtained at the Westinghouse PRHR test facility. LOFTRAN input was developed to simulate the three-tube arrangement of the test facility and test conditions. Comparisons of LOFTRAN to the ten data show close agreement. More importantly, the analyst, by selecting the appropriate F2HR heat transfer option in LOFTRAN, can choose to overpredict or underpredict the PRHR heat transfer providing a conservative response for safety analysis calculations.

Additional details of the PRHR separate effects tests are presented in Appendix ISB, Revision 0 of SSAR. Comparisons of LOFTRAN simulations to SPES-2 data show that the LOFTRAN PRHR i module continues to provides an accurate representation, when coupled with the other LOFTRAN l

modules. i l

6.33 CMT Tests ;

i Valklation of the LOFTRAN CMT model is based on comparisons to Westinghouse CMT Test Facility data. Details of the CMT validation effort are presented in Section 4 of this report. Based on the results presented within this report it is concluded that the LOFIRAN CMT module provides an i accurate representation of the AP600 CMT system. For single-phase, natural-circulation, the code predicted injection flow rate in the range of 5 to 10 percent higher than the experimental results. For j two-phase flow, a buoyancy head penalty is used to provide conservative simulations. i The LOFIRAN CMT hydraulic model is not intentionally biased in either the conservative or nonconservative direction. The 5 to 10 percent overprediction is explained by differences between the flow conditions in CMT tests and the flow conditions in the cold pre-operational tests used to develop user-specified, friction factor input. Conservatism is introduced in two ways. First, the CMT model assumes perfect mixing of boron in the tank with the fluid entering from the cold leg balance line.

During the CMT recirculation phase, applicable to non-LOCA events, this has the effect of instantly diluting the relatively higher boron concentration in CMT water with the replacement RCS water, conservatively underpredicting the boron concentration of the CMT injection. Second, the user can m:\np600Q061 - non\2061 w-9.aon: l t,-070995 6-3 REVisloN: 0

.t.

adjust code input to provide nunimum or maximum flow as appropriate for a given analysis to proside l conservative analysis results. CMT modeling assumptions in relation to particular non-LOCA events or SGTR are provided in Chapter 15 the SSAR.

I I

Comparisons of LOITRAN simulations to SPES-2 data show that the LOFTRAN CMT module continues to provides an accurate representation, when coupled with the other LOFTRAN modules, he SPES-2 test data showed that individual CMT behavior is amenable to simulations with the single CMT LOFTRAN module. For both the SPES-2 SGTR and MSLB tests, which are asymmetric i secondary-side transients, the SPES-2 test data showed relatively symmetric CMT flow rates. .

6.3.4 SPES 2 Tests 6.3.4.1 SGTR Test Simulations ,

4 SGTR event cases were run with and without nonsafety systems. Comparison of LOFTRAN simulations to SPES-2 test data show good agreement with all key parameters identified in the PIRT.

In particular, the code did a good job of predicting the CMT and PRHR behavior. Combined with analyses assumptions which maximize SG tube break flow,it is apparent that LOFITR2-AP provides a good model for use in conservative, design-basis SGTR calculations.

LOITRAN predictions of CMT and PRIIR behavior during the SGTR event demonstrated good to l excellent agreement with test data. De code models showed little variation in the sensitivity case, indicating that the design of these passive systems is robust for single-phase flow conditions.

De individual CMT flow rates in the SGTR tests behave very much the same providing justification for the use of the LOFTRAN single CMT model in SGTR analyses.

6.3.4.2 MSLB Test Simulation Comparison of LOITRAN simulations with the SPES-2 MSLB test data demonstrated that LOITRAN-AP accurately predicts the overall transient trends and provides conservative results suitable for design-basis safety-analyses. Evaluation of the main steam line break code simulations and comparisons to test data determined the primary reasons for differences with the test data. Dese are: a conservative blowdown model developed to nuximize break flow, break quality uncertainty, and thick metal heat capacity and system heat losses effecu. De thick metal heat capacity of the l steam generator and system heat loss effects are exaggerated by the SPES-2 facility. Dese differences l are explainable and are related to phenomena not associated with the passive plant design.

Comparison of the SPES-2 MSLB test data for the FRHR and CMT passive systems showed good agreement with LOITRAN simulations. Consistent with the SGTR test data, the behavior of the individual CMTs was essentially the same throughout the MSLB providing justification for the use of ;

the single CMT LOFTRAN model in MSLB analyses.

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LOITRAN generally employs conservative, transient-forcing functions. Combined with the use of l minimum or maximum input value assumptions designed to minimize the margin to applicable !

regulatory limits, LOFTRAN-AP provides an accurate, but conservative analysis tool.

6A LOFTRAN Application Envelope The LOFTRAN code is not intended for use where significant two-phase flow occurs. Transients l which employ LOITRAN,in general, do not exhibit significant two-phase flow conditions. If two phase flow could occur, acceptance criteria are established based on prohibiting large scale RCS boiling. Additionally, LOFTRAN is not used: where CMT drain down could occur, for post-trip ADS, during IRWST injecticn phase, or for long term cooling phases. I LOFTRAN models the two CMTs as a single CMT. For transients in which asymmetric cold leg transients in conjunction with CMT injection occurs, analysis assumptions are selected to conservatively bound actual plant response.

6.5 Conclusions j 'Ihis report presented the final phase of the LOFTRAN-AP code validation effort. Key to the i validation of the LOITRAN-AP code version is demonstration that the code adequately predicts the behavior of passive plant features as identified in the PIRT, such as the PRHR and CMTs and the intcgral plant response under conditions in which the passive features are required. Validation of the code with test data in conjunction with conservative input assumptions and methodologies described in the Code Applicability Document provides a sound basis for using the LOITRAN-AP code in design-basis safety-analyses.

Natural-circulation capability was validated by comparing LOFTRAN simulations to SPES-1 test data.

The results show a very good capability of the code to predict natural circulation in single-phase flow.

Additional details are available in Appendix ISB, Revision 0 of SSAR.

The LOFTRAN PRHR model was verified by comparison to test data from the Westinghouse PRHR test facility. Additional validation ofintegral effects is provide by comparison to the SPES-2 SGTR

( and MSLB tests. In all cases, the results show an excellent capability of the code to predict PRHR behavior under the range of conditions analyzed in the non-LOCA and SGTR events. Furthermore, l

the code provides flexibility through user-selected input and options to provide conservative responses suitable for bounding design-basis calculations.

LOFTRAN-AP code simulations of SPES-2 SGTR test data show that LOITRAN accurately predicts the operational behavior of the passive systems as well as the integrated plant operation under transient conditions. Comparison to the test data shows good agreement with all key parameters identified in the PIRT. In particular, the code did a good job of predicting CMT and PRHR behavior. Combined f

rn:\ap6000061-non\206I w-9.non:1b-070995 6-5 REVISION: 0

. _. - . . . . _ _ _ . _ _~ _.

. '[

e' ?

t t

with analysis assumptions, which maximize break flow, it is apparent that LOFITR2-AP pmvides a [

I good model for conserva'sve, design-basis SGTR calculations.

i Comparison of LOFIRAN-AP simulations with the SPES-2 MSLB test data demonstrated that- [

LOFTRAN AP accurately predicted the overall transient trends and provides conservative results - !

suitable for design-basis safety analyses. LOFTRAN provided good predictions of CMT and PRHR behavior for the SPES-2 MSLB test.

Based on comparison of LOFIRAN simulations to SPES-2 SGTR and MSLB tests presented in this !

. report LOFTRAN- AP adequately predicts the key parameters identified in the PIRT during the applicable time frame analyzed for the SSAR. Comparisons of LOFIRAN simulations to SPES-2 data I show that the LOFTRAN CMT and PRHR modules continue to provide an accurate representation, when coupled with other LOFIRAN modules. The SPES-2 test data showed that Individual CMT behavior is relatively symmetric; thus, it is amenable to simulations with the single CMT LOFTRAN ;

module. !

l This report concludes that the LOFTRAN code provides an accurate model of the AP600 plant over !

the range of conditions required for the analysis of design-basis non-LOCA and SGTR events. In !

conjunction with conservative input parameters based on established safety analysis methodologies, 'f LOFTRAN provides an excellent tool for design-basis safety analyses. j i

e k

i s

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7.0 REFERENCES

1. M. Lambert, LOFTRAN CMT Preliminary Validation Report, MTD1-GSR-002, (Proprietary),

j November 1994.

2. E. Carlin, M. Lambert, W. Scherder, LOFTRAN Preliminary Validation Reportfor SPES-2 Tests, PXS-GSR-001,(Proprietary), April 1995.

3. Burnett, T. W. T., LOFTRAN Code Description, WCAP-7907-P-A (Proprietary) and WCAP-7907-A (Non-proprietary), April 1984.

4. '1homas, C. O. (NRC), Leuer to P. Rahe (Westinghouse) Regarding Staff Acceptance for l Referencing of WCAP-7909 (Proprietary) and WCAP-7907 (Non-proprietary), "LOFTRAN Code Description," July 29,1983.

5. Lewis, R. N., et al., SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill, WCAP-10698-P-A (Proprietary) and WCAP-10750-A (Non-proprietary), August 1985.

6. Lewis, R. N., et al., Evaluation of Ofsite Radiation Dosesfor a Steam Generator Tube Rupture i Accident, Supplement I to WCAP-10698-P-A (Proprietary) and Supplement I to WCAP-10750-A (Non-proprietary), March 1986.

7. Lewis, R. N., et al., Evaluation of Steam Generator Overfill Due to a Steam Generator Tube Rupture Accident, WCAP-11002 (Proprietary) and WCAP-11003 (Non-proprietary), i February 1986.

8. Carlin, E. L., LOFTRAN & LOF77R2 AP600 Code Applicability Document, WCAP-14234 (Proprietary), November 1994.

9. Botti, S., et al., " Experimental Data Report - SPES test SP-NC Single Phase Natural Circulation (15% of Nominal Power)," SIET 0001 rd 89, December 1989.

10. Delose, F., Facility Description Reportfor AP600 Core Makeup Tank (CMT) Test Program, WCAP-14132 (Proprietary), July 1994.

I

11. AP600 Core Makeup Tank (CMT) Test Spec #ication, WCAP-13345, Rev. 4 (Proprietary), 3 August 1994.

I2. Aumiller, D. and L.E Hochreiter, Scaling Logic for the Core Makeup Tank Test, WCAP-13963 (Proprietary), February 1994.

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es i

l

13. SPES-2 Facility Description, WCAP-14073 (Proprietary), SIET Document #00183R192, May 1994.

I

14. Leonelli, K., Core Makeup Tank Test Data Report, WCAP-14217 (Proprietary), November 1994.

1

15. Crane Co. Technical Paper #410, Flow of Fluids Through Valves, Fittings, and Pipe,1978.

16. Wilson, J. F., et al, Steam Volume Fraction in a Bubbling Two-Phase Mixture, Trans. Am. Nucl.

Soc., 4, 356-357 (1961).

17, Za1oudek, F. R., Steam-Water Critical Flow From High Pressure Systems interim Report,"

HW-80535, January 1964.

18. Not used.

19. Conway L. et al., AP600 Design Certification Program SPES-2 Tests Final Data report WCAP-14309, March 1995.

20. Quick Look Reportfor Full Pressure Full Height Test 501211 in SPES-2 (Blind Test),

L'ICT-T2R-031, October 1994.

21. Not used.

22. NS-EPR-2648, Letter from E. P. Rahe, Jr., Westinghouse to C. O. Thomas (NRC), Summary of NRC/ORNUWestinghouse Technical Review Meeting of July 13-4,1982, dated August 27,1982.

23. Docket 50-244, Letter from J. E. Maier (RGR) to D. M. Crutchfiel (NRC), " Response to Safety Evaluation Report - NUREG-0916, Steam Generator Tube Rupture Incident, R. E. Ginna Nuclear Power Plant," dated November 22,1982,

24. AP600 FHFP Integral Systems Test Specification, WCAP-14053, Rev. 2 (Proprietary), April 1995. I l

25. NUREG-0800, Revision 2, "USNRC Standard Review Plan," July 1981.

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L APPENDIX A CMT COMPONENT TESTS l

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1 A.1 CMT Component Test Facility Description The AP600 Core Makeup Tank Component Test Facility was located at the Westinghouse Waltz Mill Site in Madison, PA within two steel frame, steel panel clad buildings. l A schematic representation of the AP600 Core Makeup Tank Test Facility is provided on Figure 3-2. l The facility consisted of the following major components and systems: ;

1. Core Makeup Tank - a single core makeup tank is modeled using a 10 ft. high,24-inch outside diameter carbon steel pressure vessel. A prototypic level instrument is also installed in the test CMT for evaluation.

t

2. Steam / Water Reservoir - a 10 ft. high, 36-inch inside diameter pressure vessel which represents the reactor coolant system.

3. Steam Supply System - a steam generator,1164 ft.8 dry steam accumulator and pressure ;

control valve designed to provide saturated steam over a range of pressures sufficient to capture all CMT modes of operation.

l

4. Test Facility Piping - simulates the primary features of the RCS to CMT balance line and CMT discharge line piping.

5. Test Instrumentation .

6. Data Acquisition System (DAS)

The test Core Makeup Tank was an instmmented test vessel (1/2 scale in height and in.77 scale in l diameter) designed to model the key thermal-hydraulle phenomena related to all modes of CMT ;

operation. The Steam / Water Reservoir (S/WR) simulated the remainder of the AP600 RCS in that it ;

provided a source of steam to the CMT, and stored water drained from the CMT. The test facility piping modeled the cold leg balance line piping from the RCS cold leg to the CMT and the CMT '

. drain line piping from the bottom of the CMT to the reactor vessel. Saturated steam was supplied to the S/WR using a high pressure steam generator, accumulator and pressure control valve to '

accommodate the entire range of test pressures. A data acquisition system (DAS) was provided to record signals from the various test instruments which include thermocouples, pressure sensors, and flow meters.

l The Core Makeup Tank test article, related piping, steam supply and ancillary support systems were 4 situated in a 32-foot wide by 46-foot long by 65-foot high building which was selected because of the headroom available o model actual plant elevations.

i m:\ap60rA2061 -non\2061 w- 1.non: lb-070995 A-1 REVISION: 0

The facility control room, which housed the test loop controls and data acquisition system, was located in an adjacent building. All test operations were conducted from the control room area.

A full description of the test facility is provided in Reference 10.

A.2 CMT Component Tests Used for LOFTRAN Code Validation Simulations of the CMT component tests provide primary validation of the LOFTRAN-AP CMT model. During the design-basis non-LOCA and SGTR transients analyzed with LOFTRAN-AP, subcooling exists in the RCS cold legs. During these events, the CMT exhibits recirculation rather than the draindown mode of injection, which is expected for a LOCA. Therefore, only the recirculation phase of the 500-series tests is used for validation of the LOFTRAN-AP CMT model.

The 500-series tests are documented in Reference 14.

Matrix tests 501 to 509 simulated heating of the CMT water by natural circulation and with subsequent draindown and depressurization. Only the natural circulation phase of the tests is simulated because this is the mode relevant to LOFTRAN-AP non-LOCA and SGTR transient analysis, and the draindown phase is outside the scope of the LOFTRAN-AP code. During the natural circulation phase of the tests, the water reservoir contained water (close to saturation) and steam; the level was above the line 2 inlet (Figure 4-1). Line I was closed, and natural circulation was initiated by fully opening the injection line (valve V3). This valve remained fully open throughout the natural circulation phase. During the transient, a pressure control valve operated to keep the reservoir pressure constant by admitting steam from the steam accumulator.

Bree parameters differentiate tests 501 to 509:

- The pressure target of the loop (1,085 or 1,835 psi)

The duration of the natural circulation phase (until one-fifth, one-half, or the entire CMT is heated)

. The setting of the injection line valve for the draindown phase (the draindown phase is not simulated in this report)

Only tests with the entire CMT heated are selected for simulation for LOFTRAN AP validation because these tests essentially duplicate the tests with a partially heated CMT. These tests are C064506 (Matrix Test 506) and C072509 (Matrix Test 509).

The parameters of test C064506 and C072509 are summarized in Table A-1. Full test results may be found in Reference 14.

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TABLE A 1 i CMT Soo-SERIES TESTS USED FOR LOITRAN-AP VALIDATION Limited f Test Run Test Pressure Drain Rate Number Date (psig) (spm)

C064506 8/29/94 1085 16 C072509 9/14/94 1835 16 l

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1 APPENDIX B SPES-2 TESTS i

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l 11.1 SPES-2 Test Facility Description 11.1.1 Introduction The SPES facility is an experimental plant located at SIET (Societa Informazioni Esperienze Termoidrauliche) laboratories in Piacenza, Italy. He SPES facility was substantially modified to simulate the AP600 plant while maintaining full-scale elevation, full-pressure, and full-power at a volume and power scaling factor of 1/395. The resulting SPES-2 facility retained some major components of the previous facility (rod bundle, pressurizer, steam generators), but had significant changes in the power channel and the primary pumps. In addition, all of the main coolant loop piping and the passive safety systems were expressly designed and constructed for SPES-2 in order to model the AP600. A complete description of the SPES-2 facility ic provided in the SPES-2 Facility Description Report (Reference 13).

B.I.2 Facility Scaling Summary ne SPES-2 facility was designed to simulate the following:

Re primary system ne secondary system up to the main steam line isolation valves (MSLIVs)

The passive safety systems: accumulators, core makeup tanks (CMTs), in-containment refueling water storage tank (IRWST), passive residual heat removal (PRHR), and automatic depressurization system (ADS)

Re nonsafety systems: normal residual heat removal system (RNS) and chemical and volume control system (CVS)

He overall scaling factor was specified to'be 1/395 and the main characteristics were specified to be:

Process fluid water

Loop number 2

Pump number 2

Primary design pressure 2900 psia

Secondary design pressure 2900 psia e Primary design temperature 690*F

Secondary design temperature 590*F L

Maximum power 9MW

= Elevation scaling 1:1 I

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l The SPES-2 scaling was intended to preserve the following parameters:

Fluid thermodynamic conditions

. Venical elevations

Power-to-volume ratio I

= Power-to-flow rate ratio

. Transit time of fluid

. Heat flux For example, the surge and passive safety system lines, where friction pressure drops are relevant to scaling of the AP600 plant, were designed to maintain the dimensions of the reference plant, whereas the facility piping was designed with the Froude number conservatism in order to preserve the slug to stratified flow pattern transition in the horizontal piping.

The passive safety systems were designed according to the criteria described above in order to reproduce, as accurately as possible, the thermal-hydraulic phenomena in AP600 during a transient.

Also, when deemed necessary, the layout of the connection lines was designed to preserve the similarity of the full-scale AP600 layout.

H.I.3 Facility Description As discussed above, the SPES-2 facility is a full simulation of the AP600 primary and passive core cooling systems. The stainless steel test facility uses a 97-rod heated rod bundle that has a uniform axial power shape and uses skin heating of the heater rods. Here are 59 heater rod thermocouples distributed over 10 elevations with most located at the top of the bundle to detect the possibility of bundle uncovery. The heater rods are single ended and are connected to a ground bus at the top of the bundle at the upper core plate elevation. All but two rods are designed to have the same power. Two heater rods are hot rods that have 19 percent higher power.

The primary system, as shown in Figure 3-2, includes two loops-cach with two cold legs, one hot leg, a steam generator and a single reactor coolant pump (RCP). The cold leg flow splits downstream of the simulated RCP into two separate cold legs, which then flow into an annular downcomer. The pumps can deliver the scaled primary flow, and the heater rod bundle can produce the scaled full-power level such that the AP600 steady-state temperature distribution can be simulated. The steam generators have a secondary-side cooling system that removes heat from the primary loop during simulated full-power operation. Startup feedwater and power-operated relief valve (PORV) heat removal is provided following a simulated plant trip.

The upper portion of the simulated reactor vessel includes an annular downcomer region, where the hot and cold legs, as well as the safety injection lines, are connected. We annular downcomer is connected to a pipe downcomer below the direct vessel injection (DVI) lines; the pipe downcomer then connects to the vessel lower plenum. In this fashion, the four cold leg, two hot leg characteristics mup60m2061-nonco61. 2.non:1b-070995 B-2 REVIsloN: 0

of AP600 can be preserved along with the downcomer injection. Here are turning devices to direct the safety injection flow downwards in the annular downcomer, as in the AP600.

A full-height, PRHR heat exchanger, constructed in a C-tube design, is located in a simulated IRWST that is maintained at atmospheric pressure. De line pressure drop and elevations are preserved and the heat-transfer area is scaled such that the natural circulation behavior of the AP600 PRHR heat exchanger is simulated.

l l De design of the CMTs is unique and has been developed by the SIET engineers so that the CMT l metal mass is scaled to the AP600 CMT. De SIET CMT design uses a thin-walled vessel inside a thicker pressure vessel, with the space between the two vessels pressurized to approximately 1015 psi.

In this marmer, the amount of steam that condenses on the CMT walls during draindown is preserved.

Since the CMTs are full height and operate at full pressure, the metal mass-volume ratio of a single pressure vessel would have been excessive, resulting in very large wall steam condensation effects.

A SPES-2 ADS combines the two sets of AP600 ADS piping off the pressurizer into a single set with the first , second- and third-stage valves. An orifice in series with each ADS isolation valve is used to achieve the proper scaled flow area. The three ADS valves share a common discharge line to a condenser and a collection tank that has load cells to measure the mass accumulation. A similar measuring arrangement is also used for the two ADS fourth-stage lines, which are located on the hot legs of the primary system.

Pipe breaks are simulated using spool pieces that contain a break orifice and quick-opening valve.

De break discharge is condensed and measured by collecting the flow into a catch tank.

The specifics of the key systems / components are discussed in References 13 and 19.

11.1,4 Instrumentation Data Acquisition System The SPES-2 facility instmmentation has been developed to provide transient mass and energy balances on the test facility. There are approximately 500 channels of instmmentation that monitor the facility and component pressure, temperature, and mass inventory.

A variety of different methods and components were used to measure the significant thermodynamic quantities that are direct (absolute and differential pressure, temperature, voltage, current, etc.) and derived quantities (mass velocity, flowrate, etc.) as shown below:

Ouantity Method or Component Pressure -

Pressure transmitters Differential pressure -

Differential pressure transmitters Temperature -

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Collapsed level -

. Differential pressure transmitters Density -

ydensitometers Velocity - Turbines / orifices, venturi tubes

Flow rate . -

Turbines / orifices, venturi tubes Integral flowrate -

Catch tanks ,

Electrical power -

Voltage drops and shunts i

i he flows into the simulated reactor system, such as the CMT discharge flow, the accumulator flow I and the IRWST flow, are measured using venturi flow meters. Flows out of the test facility, such as ,

break flow and ADS flow, are measured with a turbine meter and condenser / collectiont ' ank. De use i of condensers allows accurate mass flow versus time measurements of the two-phase ADS and break' {

flow streams. De use of collection tanks following the condensers provides redundancy for the ' l critical measurements of the mass leaving the test system. Differential pressure measurements are l

arranged as level measurements on all vertical components to measure the rate of mass change in the -

component. here are also differential pressure measurements between components to measure the j frictional pressure drop, both for single- and two-phase flow. De CMTs are instrumented with wall -

and fluid thermocouples to measure the CMT condensation and heatup during their operation. De i PRHR HX is also instrumented with wall and fluid thermocouples so that the tube wall flux can be ;

calculated from the data. Dere are thermocouples in the simulated IRWST to measure the fluid !

temperature distribution and to assess the amount of mixing that occurs. The rod bundle power is ;

measured accurately to obtain the rod heat flux and the total power input to the test facility. I De data acquisition and elaboration system collects and handles measured signals from the plant. He i large amount of operations necessary to the user are implemented in appropriate software procedures ;

in order to avoid errors and loss of information. :

B.1.5 Control Loops l During the test, control loops managed and controlled the key plant parameters. Most of the control i loops are electronic and are located on the control room main board. Dese main control loops i regulate:

Primary pressure and level f'

. Steam generator pressure and level

CMT external containment air pressure I

SFW, NRHR, CVCS flow rates ;

Bundle power !

l

-t De basis for the SPES-2 bundle power decay is to simulate the heat flux versus time from the AP600 :

fuel rods, including stored energy and fission product decay heat. His power decay versus time has been determined based on AP600 LOFTRAN analyses. De fission product decay heat versus time is ;

based on the ANS 1979 decay heat standard plus two sigma uncertainty. De SPES-2 heat loss [

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compensation value (150 kW) is based on pre-operational testing and is added to the SPES-2 bundle power decay. The core power versus time simulation is discussed in detail in Reference 19.

He day before a test, the facility is checked out. On the day of the test, several key steps are performed to bring the plant up to initial conditions. When nominal conditions are reached, they are j maintained for about 500 seconds before starting the transient. To start the transient, a specific break

} valve (or valves if required) is opened to begin break flow. At this point, the transient follows a j course of events that is specific to the test procedure for that particular matrix test.

Ilowever, some events are common to most of the tests. Once a setpoint is reached initiating the reactor trip R signal, the main feedwater (MFW) isolation valves are closed and the power decay simuladon is begun. Upon S signal initiation, the CMT isolation valves and the PRHR isolation valves are opened, and the main steam line isolation valves are closed, all with a 2-second delay.

16.2 seconds after S signal, the RCP coastdown is initiated. ADS-1 is actuated on CMT volume of 67 percent with the other ADS stages following the delay time specified in the test procedure. Heat i loss compensation is terminated with ADS stage 1 actuation. He accumulators begin injecting when the primary system pressure falls to ~700 psia. De IRWST begins injecting water when the primary system pressure is 26 psia. De test is terminated when final conditions are achieved as specified in the test procedure. The specific facility operation and configuration for each test used in LOFTRAN-AP validation are discussed in subsection B.3.

B.2 SPES-2 Tests Used for LOFTRAN Code Validation Simulations of the SPFS-2 SGTR and MSLB tests validate the integral AP600 plant response with the passive safety systems. Rese tests also provide additional validation of the LOFTRAN-AP CMT and PRIIR models.

For the LOFTRAN-AP code validation, data from the following SPES tests is used:

Matrix Test No. 9 -

Steam generator tube rupture with nonsafety systems operational and operator action for mitigation (SPES-2 test S01309)

Matrix Test No.10 -

Steam generator tube mpture without nonsafety pumped injection / heat removal systems and without operator action (SPES-2 test 501110)

Matrix Test No. I1 -

Steam generator tube rupture without nonsafety pumped injection / heat removal systems and with no operator action other than manual ADS actuation (SPES-2 test S01211, blind test)

Matrix Test No.12 -

Large single-ended steam line break at hot standby conditions without nonsafety systems and with no operator action. (SPES-2 test S01512, blind test) m Anp600C061 -oonu061 w-2.non: 1 b-070995 REVislON: 0 B-5

BJ Test Descriptions His section summarizes the SPES-2 SGTR and MSLB tests. Complete descriptions of the tests are available in the final test report (Reference 19).

H3.1 Design-Basis Steam Generator Tube Rupture with Nonsafety Systems i Operational and Operator Action for Mitigation (SPES-2 Matrix Test S01309)

Dis section summarizes the details of the SPES-2 matrix test S01309 that are relevant to the l LOITRAN-AP simulation. Additional details and data for matrix test S01309 can be found in l Reference 19.

I his matrix test simulated an SGTR with both the passive and nonsafety systems operational and with I operator action for mitigation. Here were two operator actions simulated for this test: cooldown of l the primary system by dumping steam through the intact SG power-operated relief valve (PORV) to l

obtain hot leg subcooling while limiting the overall cooldown rate of the primary system; and subsequent controlled depressurization of the primary system to terminate primary-to-secondary j l

l leakrge using an ADS-1 valve. Bere was no CMT draindown during this transient. Berefore, there was no action of the ADS primary system depressurization function, no IRWST injection, and no significant accumulator injection. De chemical and volume control system (CVCS) and startup i feedwater system (SFWS) were automatically initiated and controlled throughout this test.

1 Plant personnel manually performed the following actions: -)

Five minutes after the S signal, the faulted steam generator was isolated by closing the SFW isolation valve.

From 2381 to 2669 seconds and reoccurring every 40 seconds, the ADS-3 valve was mistakenly opened for 10 seconds instead of the SG-A PORV.

Beginning at approximately 3000 seconds and reoccurring every 40 seconds, the SG-A PORV was opened for 10 seconds until it was left fully open at 4074 seconds.

Beginning at 5276 seconds and reoccurring every 30 seconds until the test was terminated, the ADS-1 valve was opened for 10 seconds.

De single SGTR was simulated via a line connected from the primary side (coolant pump B suction piping) to the secondary side of SG-B (approximately 3.9 ft. above the tube sheet), with a break orifice diameter scaled to simulate 1.2 times the area of a single AP600 SG tube diameter. Table B-1 shows the initial test conditions. De sequence of events for S01309 is listed in Table B-2.

I l

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l.

i l

De event phases identified to evaluate the thermal-hydraulic phenomena that occurred within the primary and safety systems for the SGTR event are as follows:

j

Initial depressurization phase

= Pressure decay phase Initial Depressurization Phase (0 to 910 seconds)

De initial depressurization phase began with the break at time zero and continued for 900 seconds l until the primary system depressurization was slowed (when primary system pressure was supported by

! the saturation pressure of the upper plenum and hot legs). His period included the following major events: initiation of the break at zero seconds; initiation of CVCS makeup flow to the primary system when pressurizer level decreased to the low-level setpoint of 10.2 ft. at [ ]'6#; shut-off of the one operating pressurizer internal heater when the pressurizer level decreased to [

l ]'6#; and simultaneous reactor trip (R) signal and safeguards (S) signal initiation when the pressurizer level decreased to the low-low level setpoint of 2.2 ft. at [ ]'6' 1

I l

Be R and S signals initiated the following actions: the main steam line isolation valves (MSLIVs) and main feedwater isolation valves (MFWIVs) closed; the CMT injection line valves and the PRHR 11X return line isolation valve were opened - all with a 2-second delay; power decay was initiated after a 5.7-second delay; RCP coastdown was initiated after a 16.2-second delay; and the CVCS setpoints were reset to ON when pressurizer level was less than 10 percent (2.2 ft.) and to OFF when pressurizer level was greater than 20 percent (4.4 ft). Also during the initial depressurization phase, j operator action to isolate startup feedwater flow from the faulted SG-B was simulated at 767 seconds I (S signal + 5 minutes), and operator action to ini-iate additional primary system cooling by opening the steam generator A PORV was simulated beginning at S signal + 7 minutes. The ADS-1 valve was inadvertently opened with the SG-A PORV at approximately 900 seconds.

Pressure Decay Phase (910 seconds to end-of-event) l De pressure decay phase began at approximately 900 seconds when the system pressure was initially i supported at the saturation pressure determined by the primary system core outlet temperature. The ADS-1 valve was opened by the facility control computer at about 900 sec., causing a drop in pressure from approximately [ ]'6#.

His phase of the event was characterized by a slow decrease in the overall system pressure and temperature. Operator actions to obtain hot leg subcooling margin and to reduce the primary system temperature and pressure were not effectively initiated until approximately 4100 sec. and $300 sec., v respectively. As a result, throughout most of the pressure decay phase, the hot leg side of the primary system was close to or below saturation, and primary pressure was mainly dictated by the pressurizer steam bubble pressure.

m Aa@0m2061 -non\2061 w-2.aon: lt>-070995 B-7 RmsloN: 0

l The test was terminated when the primary- and secondary-system pressures equalized.

11.3.2 Design-Hasis Steam Generator Tube Rupture without Nonsafety Systems and No Operator Action to Isolate the Steam Generator (SPES 2 Matrix Test S01110)

This section summarizes the details of test 10 that are relevant to the LOFTRAN-AP simulation.

Additional details and data for test S01110 can be found in Reference 19.

This matrix test simulated an SGTR without any nonsafety systems operating or operator actions, and with only the automatic passive safety systems used for accident mitigation. He pressurizer internal heaters were supposed to shut-off at break initiation and the CVCS, NRHR function, and SFWS were shut-off for this test. During the test, heaters were mistakenly left on for 250 sec. at a power of approximately 11 kW. During mitigation of the SGTR, there was no CMT, accumulator, or IRWST injection throughout the transient. The single SGTR was simulated via a line connected from the primary side (RCP B suction piping) to the secondary side of SG-A (approximately 3.9 ft. above the tube sheet) with a break orifice diameter scaled to a single AP600 SG tube diameter.

Table B-3 provides the initial test conditions. He sequence of events for S01110 is listed in Table B-4.

Since this SGTR event did not result in ADS actuation, only the first two event phases observed in loss-of-coolant (LOCA) recovery occurred. The event phases are as follows:

Initial depressurization phase

. Pressure decay phase Initial Depressurization Phase (0 to 1050 seconds)

The initial depressurization phase began with the initiation of the break at time zero and lasted until the primary system pressure was supported by the saturation pressure for the upper plenum and the hot legs. This period included the following events: initiation of the break, the activation of the R and S signals, closure of the MSLIVs, opening of the CMT injection line valve, opening of the PRHR HX return line isolation valve, and closing of the MFWIVs, all within a 2-second delay. Rod bundle power was reduced to 20 percent with a 5.7-second delay. Decay power simulation was initiated with a 14.5-second delay, and RCP coastdown was initiated after a 16.2-second delay.

Pressure Decay Phase (1050 Seconds to End-of-Event)

The pressure decay phase began when the system pressure was supported by the saturation pressure in the hot leg fluid of the primary side and continued until the end of the test. This phase was characterized by a slow decrease in the overall system pressure and temperature. The core power (decay heat loss compensation) was reduced from 275 kW to 220 kW.

mAaptaA2061 -non\2061 w-2.non: n>-070995 B.8 REVIsicN: 0 m . . . .. . __ __ _

e At 2000 seconds with core power at 245 kW, the PRHR was removing approximately 83 kW; the CMTs provided approximately 84 kW of effective heat removal when the cold CMT water replaced the hot water entering the CMTs through the balance lines; the break flow removed about 7 kW; and facility heat losses were approximately 125 kW. The total heat removal exceeded the heat input (299 kW versus 245 kW); there was no boiling in the power channel, and system pressure decreased slowly.

From 1500 to approximately 3000 seconds, the top of the upper plenum and upper head drained and I partially refilled the pressurizer.

i l

At 3000 seconds into the event, the primary- and secondary-side pressure equalized, the SG U-tubes begin to drain, and the break flow and the heat transfer to the secondary side decreased, j l

l l Periodic boiling in the core and related oscillations in temperature, void fraction, flow through the I core, and system pressure oscillations began in the pressure decay phase at about 3000 seconds into the event and continued throughout the rest of the event.

De U-tubes of SG-B were drained and the pump suction B was partly drained at about 6000 seconds into the event. For SG-A, the flow continued until the end of the test; however, there was an oscillating void fraction at the top of the SG-A U-tubes. The test was terminated at about 7500 seconds after the primary- and secondary-side pressures equalized.

B.3.3 SGTR with Inndvertent ADS Actuation (S01211)

This matrix test simulated a double-ended rupture of a SG tube followed by inadvertent ADS actuation. This test was performed without any nonsafety systems operating. The CVCS, NRHR, and SFWS were turned-off for this test. He SG FR was simulated via a line connecting the primary side (coolant pump B suction piping) to the secondary side of SG-B (3.9 ft. above the tube sheet), with a break orifice diameter scaled to simulate [ ]** times the area of an AP600 SG tube to obtain the same flow as a double-ended break of an AP600 SG tube. The ADS-1 flow path was opened 2.5 minutes after the reactor trip (R) and the safety systems actuation (S) signals were generated.

Tables B-5 and B-6 show the initial conditions and sequence of events for this test.

Only the initial depressurization phase of this test is relevant to LOFTRAN-AP validation. After ADS-1 actuation, the event became a LOCA. He phase began by opening the SGTR break valve at time zero and ended when ADS-1 was actuated at 626 seconds. When the break valve was opened, primary system fluid flow to the secondary-side SG-B resulted in a decrease in pressurizer level and pressure. The pressurizer level decreased to [ ]** at [ ]**, simultaneously actuating both the R and S signals. He SG main steam line isolation valves (MSLIVs) closed. The CMT injection line isolation and PRHR HX return line isolation valves opened after a [ ]** delay. The heater rod power step changed from 100- to 20- percent power after a 5.7-second delay. The RCPs were shut off after a 16.2-second delay. Flow through the PRHR HX and CMTs began immediately m Aap600\2061 - non\2061 w- 2.noo: l b-070995 B-9 REVIsloN: 0

1 I

when the isolation valves opened. Rod bundle power remained at 20 percent through 14.5 seconds after the R signal. At this time, the SPES-2 integrated heater rod power into the primary system l matched the scaled AP600 core power decay. The SPES-2 heater rod power then decreased, !

simulating the scaled AP600 core decay heat but maintained an additional 150 kW, compensating for the SPES-2 facility heat losses.

I Pressurizer level rapidly decreased to [ ]"# when the R and S signals occurred due to rapid l cooldown and shrinkage of the power channel hot leg side water prior to RCP shut off. Pressurizer pressure decreased rapidly from about [ ]' 6

to [ ]*# as a result of this cooldown.

Pressure increased slightly after the RCPs were shut off as the hot leg side fluid temperature increased, then rapidly decreased toward the power channel hot leg side fluid saturation pressure.

1 l

Bere was little or no boiling in the power channel throughout the initial depressurization phase. After {

the RCPs coasted down, some heat transfer from the primary to secondary side occurred since primary system pressure was higher than the secondary-side pressure until about [ ]"".

i B.3.4 Large Steam Line Break at Hot Standby Conditions with Passive Safety Systems (S01S12) j l

Matrix test S01512 simulated a large steam line break with the facility at zero power and in hot standby conditions. Only passive safety systems were operating to mitigate the accident. The purpose of this test was to demonstrate that the CMTs would not drain and initiate the ADS; therefore, cooldown of the primary system was maximized by having no decay heat simulated, no heat loss compensation, and using three PRHR HX tubes. The CVCS and the NRHR did not provide pumped injection. The stanup feedwater system (SFWS) was not operated for this test since it would have been isolated by low-low T-cold in the AP600 plant.

I ne break was simulated by opening the SG-A PORV. The check valves in the main steam lines were removed to permit flow from the intact SG to the faulted SG until the individual SG steam line isolation valves closed. The SG-A PORV line had an orifice installed with a diameter of

[ ]*#, which corresponds to a single-ended steam line break area of [ ]"# in the AP600 planL l

There was no CMT draindown throughout the transient; therefore, there was no ADS actuation or IRWST injection. The facility prer,sure remained sufficiently high to prevent any accumulator injection until [ ]"# into the transient. De initial conditions and sequence of events for S01512 are listed in Tables B-7 and B-8.

l l

Two event phases are relevant to LOFTRAN-AP simulation:

Initial depressurization phase

Pressure decay phase mAa;WOA2061.non\2061 w-2.non: I b-070995 B.10 REVIsloN: 0 ;

. . a

O Initial Depressurization Phase The initial depressurization phase staned with the opening of the break valve (SG-A PORV). All power to the heated rods and pressurizer heaters was stopped immediately at break opening The safety system actuation (S) signal was actuated one second after the break opening signal. When the S signal was activated, the CMT and PRilR isolation valves were opened with a two-second delay, the SG-A and -B steam line isolation valves were closed with a [ J'6# M4 W &Rm um shutdown with a [ ]'** delay.

The opening of the break valve resulted in a rapid depressurizatian and level decrease in both SGs until the individual SG steam isolation valves closed at about 10 seconds. Steam blowdown from SG-A continued until the water inventory in SG-A flashed and boiled away. The large amount of heat removed from the primary system by the SGs (primarily the faulted SG-A), combined with PRHR IIX and recirculating CMTs, resulted in a rapid initial cooldown, water shrinkage, and depressurization of the primary system.

The rate of primary system depressurization began to decrease as the SG-A secondary side inventory decreased. This depressurization rate decrease was apparently due to a decrease in the heat transfer (decrease in the tube surface area covered with water) to SG-A. The primary system depressurization rate then increased when the pressurizer was completely drained. After SG-A had boiled dry at

[ ]'6', the primary system continued to depressurize toward the pressure / temperature of the intact SG-D.

Primary system flow through the faulted SG-A stopped at [ ]'6* when the SG-A dried out.

Flow continued through SG-B only until approximately [ ], at which time the SG-B U-tubes began to drain. Also at this time, the rate of the primary-side pressure decrease slowed, since primary-side pressure was controlled and matched the SG-B saturation pressure corresponding to the SG-B secondary temperature of [ ]'*' These conditions ended the initial depressurization phase.

Pressure Decay Phase The pressure decay phase for this test began at [ ]'6' and ended at approximately

[ ]'*', when the test terminated. This steam line break event was characterized by a slow, continuous decrease in the primary system pressure in conjunction with the intact SG-B secondary-side pressure / temperature. The primary system pressure was maintained by continued voiding in the SG-B U-tubes, and by the expanding steam bubble in the power channel upper-head. As shown in data plots 21,23, and 31, the upper-head water level decreased slowly from approximately [ ]'** to

[ ]'**, and SG-B U-tubes were approximately [ l drained at approximately

[ ]"', at which time the test was terminated.

The power channel temperatures (with the exception of the upper-head) decreased continuously during this test. The primary system cooldown was due to heat removal by the PRHR HX, heat removal mAngWXM1-non00M w.2.non:lko70995 B.)) Revision: 0 1

O l

\

l resulting from energy stored in the CMTs during their recirculation mode of operation, and facility heat losses. Heat transfer to and from the SGs was limited because there was no secondary water in i SG-A, and SG-B U-tubes were voided. Here was, therefore, essentially no primary system flow through either SG. De primary system water inventory, with the exception of the upper head, remained subcooled throughout the test.

ne pressurizer, which emptied at approximately [ ]'"#, began to refill at approximately i

[ ]#. Derefore, at this time the volume addition due to CMT recirculation, and upper- ;

head and SG U-tube voiding was comparable to the primary water shrinkage (density increase) due to the temperature decrease. Pressurizer level had increased to approximately [ ]'6#. when the test was terminated. Part of this increase (approximately [ ]'6#) was due to water addition from the accumulators that began to inject beginning at [ ]'6#, Eroughout the test, the CMTs did not draindown, but maintained their recirculation mode of operation. Derefore, during the steam line break recovery, the primary loop cold leg piping remained water filled, so the CMTs remained water-filled and no ADS actuation was required.

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l m Aap60tuo61 -oon\2061 w-2.non: l t>070995 B.12 REVislON: 0 }

i

o TABLE B 1 COMPARISON OF SPECIFIED AND f ACTUAL TEST CONDITIONS FOR S01309 (Matrix Test 9) l Condition Specified Actual Comment Rod Power 4991.6 + 100 kW* ,3 , OK Pressurizer Pressure 2251 29 psia OK j Average HL Temperature 599.9 9F OK t

OK OK j f OK Reactor Vessel (Core) Inlet 529.5 19F OK Temperature Core Flow Rate 51.2 1 55 0 lbm/sec. Accepted Cold Leg Flow Rate 12.91 10.22 lbm/sec. OK OK Accepted OK DC-UH Bypass Flow Rate 0.39 0.11 lbm/sec. OK Pressurizer Level 12.4 i 1.25 ft. OK Accumulator Level 7.66 0.36 ft. OK OK Accumulator Water 68 1 9*F OK Temperatur OK Accumulator Pressure 711 14.5 psia OK l

OK IRWST Level 27.9 + 0.32 ft. OK Note:

[ jo m Aap60th2061 -noci 2061 w-2.non: l b-070995 B-13 REVISloN: 0

TABLE B 1 (Cont.)

COMPARISON OF SPECIFIED AND ACTUAL TEST CONDITIONS FOR S01309 (Matrix Test 9)

Condition Specified Actual Comment IRWST Water Temperature 68 9F OK u .c PRHR Supply Line > 212*F OK Temperature UH Average Temperature 529.5 1 9'F OK CL Balance Line Temperature > 329'F OK OK CMT level 20.5 ft.101 ft. OK OK CMT Temperature 68 9F OK OK SG Level 4.8610.49 ft. OK OK SG MFW Laperature 4391 12.6*F OK OK SG Pressure 711 129 psia OK OK m:\ar6000061 -oon\2061 w-2.non:1 b-070995 B-14 REVISION: 0

i TABLE B-2 SEQUENCE OF EVFETS FOR TEST S01339 (Matrix Test 9) i Event Specified Actual Time 0,ec.) f Break Opens 0 .

CVS On PZR Level = 10.0 ft.

PZR Internal Heater Off PZR Level = 6.6 fL Pressurizer Low Level PZR Low Level = 0.676m (R, S Signals)

MSLIV Closure PZR LL + 2 sec.

MFWIV Closure PZR LL + 2 sec.

CMT IV Opening PZR LL + 2 sec. .

t PRHR HX Actuation PZR LL + 2 sec. '

Scram PZR LI. + 5.7 sec. ,

Startup Feedwater L-010P = 0.676m f Reactor Coolant Pumps Tripped PZR LL + 16.2 sec.

SFW Isol to SG-B S + 5 min.

i ADS-1 CMT I2 vel 67% :,

+30 sec. )

SG-A PORV Open Permission '

S + 7 min.

Accumulators P.027P = 710 psia C

ADS-2 CMT level 67%

+125 sec. ;

ADS-3 CMT Level 67%

+245 sec. ;

ADS 4 CMT Level 20%

+60 sec.

IRWST Injection P-027P = 26 psia Note:

OS did not actuate due to CMT level, and accumulators and IRWST did not inject throughout this transient. j t

f mvu2061.nonco6 =-2.non:Ib-070995 B-15 REVISION: 0 j t

. - _- =_ _ ..

t

~ i TABLE B-3 COMPARISON OF SPECIFIED AND ALTUAL TEST CONDITIONS FOR S01110 (Matrix Test 10)

Condition Specified Actual Comment I Rod Power 4991.6 i 100 kW' ,_ OK Pressurizer Pressure 22512 29 psia OK j Average Hot Leg Temperature 599.9 2 5'F Accepted Accepted OK !

Accepted Reactor Vessel (Core) Inlet 529.5 3.6*F Accepted i Temperature l Core Flow Rate 51.2 0..',5 lbm/sec. OK Cold Leg Flow Rate 12.912 0.22 lbm/sec. OK Acapted (F-A02P)

OK ,

OK DC-UH Bypass Flow Rate 0.39 0.11 lbm/sec. OK l Pressurizer Level 12.4 21.25 ft. OK !

Accumulator Level 7.66 2 0.36 ft. OK l OK Accumulator Water 68 2 9'F Accepted" Temperature Accepted" Accumulator Pressure 711214.5 psia OK OK IRWST level 27.9 2 0.32 ft. - -

OK Note:

[ ]*** before time 0 Ambient air temperature in facility was unusually high due to plant heatup and hot summer weather.

l i

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r TABLE B-3 (Cont.) !

COMPARISON OF SPECIFIED AND ACTUAL TEST CONDITIONS FOR S01110 (Matrix Test 10) i Condition Specified Actual Comment IRWST Water Temperature 68 9 *F .s.c Accepted

PRHR Supply Line 347245'F Accepted. Sufficient to [

Temperature initiate natural circulation flow. I Ull Average Temperature 564.8 9F Accepted Cold Leg Balance Line 509 1 9*F Accepted j Temperature Accepted. Sufficient to initiate natural circulation flow.

CMT Level 20.5 ft. (full) OK l OK CMT Temperature 6819 F Accepted (T-A411E)*

Accepted

SG Level 4.86 1 049 ft. Accepted OK SG MFW Temperature 4391 12.6*F OK .

OK i i

SG Pressure 711 129 psia OK '

OK ;

Note: ;

Ambient air temperature in facility was unusually high due to plant heatup and hot summer weather.

i l

1 l

1 i

i m:\y60m2061-non\2061w 2.non:1b 070995 B.17 REVISION: 0 l

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$ l l

TABLE B-4 SEQUENCE OF EVENTS FOR TEST S01110 (Matrix Test 10)

Esent Specified Actual Time (sec.) l Break Opens 0 !

Pressurizer Low 1.evel R PZR Low level = 0.38m MSLIV Closure PZR LL + 2 sec. !

MFWIV Closure PZR LL + 2 sec.

CMT IV Opening PZR LL + 2 sec.

PRHR HX Actuation PZR LL + 2 sec.

Scram PZR LL + 5.7 sec.

RCPs Tripped PZR LL + 16.2 sec.

ADS 1 CMT Level 67%

+30 sec.

Accumulators P-027P = 710 psia ADS 2 CMT level 67%

+125 sec.

ADS 3 CMT level 67%

+245 sec.

ADS 4 CMT Ievel 20%

+60 sec.

IRWST Injection P-027P = 26 psia Note:

ADS, accumulators, and IRWST did not actuate or inject throughout this transient.

m:\np600 cool .nonco61 w-2.non: lti-070995 B.18 RmSION: 0

.- i t

t TABLEB5 l COMPARISON OF SPECIFIED AND ACTUAL TEST CONDITIONS FOR S01211 !

Statrix Test 11)

{

Condition Specified ___

Actual Comment i

Rod Power 4991.6 1 100 kW* 4.b.c OK l Pressunzer Pressure 2251 29 psia OK Average HL Temperature 599.919*F OK [

OK OK OK Reactor Vessel (Core) Inlet 529.5 i9 'F Accepted i Temperature Core Flow Rate 51.2 0.55 lbm/sec. OK f Cold Leg Flow Rate 12.91 0.22 lbm/sec. Accepted OK j OK l OK DC-UH Bypass Flow Rate 039 + 0.11 lbm/sec. OK f Pressurizer Level 12.4 1.25 ft. OK Accumulator Level 7.66 0.36 ft. OK OK Accumulator Water 68 9'F Accepted l Temperature Accepted Accumulator Pressure 711 1 4.5 1 psia OK OK IRWST Level 27.9 + 32 ft OK 1

IRWST Water Temperature 68 19 'F Accepted PRHR Supply Line >212*F Accepted l Temperature Note:

4893.7 kW before time 0 mAap6000061 -non\2061 w-2.non: lt>.070995 B.19 REvlsioN: 0

l TABLE B-5 (Cont.)

COMPARISON OF SPECIFIED AND ACTUAL TEST CONDrrIONS FOR S01211 (Matrix Test 11)

Condition Specified Actual Comment UH Average Temperature 529.5 9F a.bc Accepted l PR to CMT Balance Line 644 + 45 F N/A l Temperature N/A Cold Leg Balance Line >329 F OK Temperature OK CMT Izvel 20.5 ft. (full) OK OK CMT Temperature 68 9*F Accepted Accepted -

- l SG Level 4.86 1 049 ft. Accepted OK SG MFW Temperature 4392 12.6"F OK OK l SG Pressure 711 29 psia OK

- - OK l

l l

l m:\nph000061 -mon \2061 w-2. con: l two70995 B.70 REVISION: 0 l

l

TABLE B-6 SEQUENCE OF EVENTS FOR TEST S01211

=

Event Specified Instrument Channel Actual Tinea (sec.)

Break Opens 0 Z_002B0 ***

Pr.:ssurizer Low Level Pressurizer low L-010P level = 2.2 fL MSLIV Closure Pressurizer low Z_ANSO, F_ANS level + 2 sec.

Z_BMSO, F_B04S MFWlV Closure Pressunzer low Z_B02SO, F_B0IS level + 2 sec.

Z_A02SO, F_AOIS CMTIV Opening Pressurizer low Z_AM0EC, F-A40E level + 2 sec.

Z_BN0EC, F-B40E PRHR HX Actuation Pressunzer low Z_A81EC, F_A80EG level + 2 sec.

Scram Pressurizer low -

level + 5.7 sec.

RCPs Tnpped Pressurizer low I-AIP, S-AIP level + 16.2 sec.

1-BIP, S-BIP ADS-1 Pressurizer low level + 150 sec.

Z_00lPC ADS-2 Pressurizer low level + 245 sec.

Z_002PC Accumulators P-027P = 710 psia F_A20EG F_B20EG ADS-3 Pressurizer low level + 368 sec.

Z_003PC ADS-4 CMT Level = 3.9 ft. L_B40E

+60 sec. Z_0MPC, F-NOP IRWST Injection P-027P = 26 psia F_A60EG F_B60EG i

l mwwwooi-noeuo61.-2.non:lt-070995 B-21 REVISION: 0

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l TABLE B-7 I COMPARISON OF SPECIFIED AND ACTUAL TEST CONDITIONS FOR S01512 Condition (Instruments) Specined Actual Comment Rod Power (W-00P) 150 5kW OK Pressurizer Pressure (P-027P) 2251 29 psia OK Average Hot Leg Temperature 545.0 9F Accepted, avg hot leg is (T-A03PO/T-A03PL/ I ]"'

T-B03PO/T-B03PL)

Accepted OK Accepted, avg HL is

[ ]

Reactor Vessel (Core) Inlet 543.2 19F Accepted, avg core AT is l'F Temperature (T-003P) as expected Core Flowrate (F 003P) 51.2 0.55 lbm/sec. Accepted, cold leg total now is [ ]

Cold Leg Flowrate 12.91 2 22 0 lbm/sec. OK ,

(F_A0lP/F._A02P/F_B0lP/

F_B02P) OK Accepted, avg Loop B flow is 13.05 lbm/s OK DC-Ulf Bypass Flowrate 039 1 110 lbm/sec. OK (F_014P)

Pressurizer Level (L._010P) 6.56 ft. < PZR level OK s 8.2 ft.

t Accumulator Level 7.66 1 036 ft. OK l (L_A20E/L_B20E)

Accumulator Water 68 19F OK Temperature (T-A22E/ '

T-B22E) OK Accumulator Pressure 7111 4.5 1 psia OK (P-A20E/P-B20E)

OK IRWST Level (L_060E) 27.9 32fL __ _ OK i

e mAngue2061 mon \2061w 2.non:ltr070995 B-22 REVISION: 0

e l

jIf i TABLE B-7 (Cont.)

COMPARISON OF SPECIFIED AND ACTUAL TEST CONDITIONS FOR S01512 Condition (Instrunwnts) Specified Actual Comment IRWST Water Temperature 68 1 9*F OK

^'

(T-063E)

PRHR Supply Line > 212'F OK Temperature (T-A82E)

UH Average Temperature 543.219F Accepted, delayed flashmg of (T-016P) Upper Head maximizes the required CMT Makeup.

Cold Leg Balance Line > 329 F OK Temperature (T-A142PI/T-B142PL) OK CMT Level (L_A40E/ 20.5 ft. 0.1 ft. OK L_B40E)

OK CMT Temperature 68 1 9*F OK (T-A411E/T-B41IE)

OK SG Level (L_A20S/L_B205) 4.861 049 ft. OK OK SG MFW Temperature *

(T-A0IS/T-BOIS) ,

SG Pressure (P-AGtS/ 1001 129 psia OK i P-BG4S) {

OK -

l Note: l

Not applicable for this test.

l 4

m Ampwco61.nonucol w-2.non: I b-070995 B-23 REVISION: 0

o' TABLE B-8 SEQUENCE OF EVENTS FOR TEST S01S12 Event Specified Instrument Channel Actual Tirne (sec.)

Break Opens (PORV A) 0 Z-00280

S Sis:nal Break opening + 1 sec. N/A CMT-IV Opening S signal + 2 sec. Z_AN0EC, F-A40E f Z_B040EC, F-B40E PRHR HX Actuation S signal + 2 sec. Z_A81EC, F_A80EG MSLIV Closure S signal + 4 sec. Z_ANSO, F_ANS Z_BMSO, F_BMS RCPs Tripped S signal + 16.2 sec. DP-A00P j DP-BOOP SFW-A Flow Began N/A F-A20A SFW-A Flow Ended N/A F-A20A f

Accumulator / Injection P-027P = 710 psia F_A20EG F_B20EG - -

ADS-1 CMT level 67% L_B40E

i

+30 sec.

Z_00lPC ADS-2 CMT level 67% L_B40E *

+125 sec.

Z_002PC ADS.3 CMT level 67% L_B40E *

+245 sec.

Z_003PC ADS-4 CMT level 20% L_B40E *

+60 sec. Z_0(MPC, F-040P

IRWST Injection P-027P = 26 psia F_A60EG

F_B60EG Note:

ADS did not actuate due to CMT level, and the IRWST did not inject throughout this transient.

I m Aap600s2061 - nonuo61 w-2.non: 1 t, 070995 B.24 REVISION: 0 l

ML20086M916

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