ML20091P589
| ML20091P589 | |
| Person / Time | |
|---|---|
| Site: | 05200004 |
| Issue date: | 08/31/1995 |
| From: | Healzer J, Mcintyre T, Shiralkar B GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20091P581 | List: |
| References | |
| NEDO-32391, NEDO-32391-RC, NUDOCS 9509010267 | |
| Download: ML20091P589 (200) | |
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{{#Wiki_filter:_- w.,, GENuclearEnergy NEDO-32391 Revision C DRF A70-00002 Class 1 August 1995 Licensing Topical Report. l SBWR Test and Analysis Program Description SBWR
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$n$8W$? $$ C$ nnYA45165 San Jose. CA 95125-1014 400 9251005 (phone) 400 925-3991 (facsiman)
August 31,1995 MFN 169-95 Docket STN 52-004 s Document Control Desk U. S. Nuclear Regulatory Commission
- Washington DC 20555 Attention: Theodore E. Quay, Director ,
l Standardization Project Directorate l
Subject:
SBWR - Test and Analysis Program Descripdon, NEDO-32391, Revision C (Non-proprietary) j I
Reference:
- 1. MFN 166-95, J. E. Quinn (GE) to T. E. Quay (NRC), SBWR Test and- - !
J Analysis Program Description, NEDC-32391P, Revision C (Proprietary),' August 28,1995. This letter transmits the non-proprietary version of Revision C of the SBWR Test and Analysis Program Description (TAPD) report, NEDO-32391P, for your review. This report provides a comprehensive, integrated plan that addresses the testing and analysis elements needed for analysis of SBWR performance. Revision C replaces Revision B in total. The changes from Revision B to Revision C are identified by sidebars in the left-hand margins of the affected pages. Please dispose of all copies of Revision B or return them to GE Nuclear Energy, Attention: J. E. I.catherman, M/C 781. Sincerely, f\kuL O N _>
. James E. Quinn, Projects Manager I
Enclosure:
SBWR Test and Analysis Program Description (TAPD), NEDO-32391, Revision C cc: P. A. Boehnert (NRC/ACRS) - [7 paper copies w/ encl., plus E-Mail w/o encl.] I. Catton (ACRS) - [1 paper copy w/ encl., plus E-Mail w/o encl.] S. Q. Ninh (NRC) - [21 paper copies w/ encl., plus E-Mail w/o encl.] J. H. Wilson (NRC) - [1 paper copy w/ encl., plus E-Mail w/o encl.]
NEDO-32391 Revision C DRF A70-00002 Class 1 August,1995 SBWR Test and Analysis Program Description B. S. Shiralkar T. R. McIntyre ! J. M. Healzer W. Marquino H. A. Upton R. E. Gamble J. R. Fitch j Y.C.Chu j j P.F. Billig i J.C. Shaug 2 l l l Approved: j James E. Quinn, Projects Manager LMR and SBWR Programs l i I
NEDO-32391, Revision C IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY c-The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between the customer and GE, as identified in the purchase orderfor this report and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than the customer orfor any purpose other than thatfor which it is intended, is not authorized; and with respect to any unauthorized use, GE makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information containedin this document. IDENTIFICATION OF CHANGES The changesfrom the previous revision of this document revision are identified by sidebars in the left-hand margins of the affectedpages. I I i 1
NEDO-3239), Revision C TABLE OF CONTENTS 1.0 INTR OD U CTI ON . . . . . . . . . . . . . . .. . .. . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . .. . . . .. . . . . . . .. . . . . . . . . . . . .. . . . . 1-1 .... 1.1 Purpose.................................................................................................. 1-1 1.1.1 Scope..................................................................................................... 1-1 1.2 Background............................................................................................. 1-4 1.2.1 Use o f TRACG . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-5 ........ 1.2.2 Major S B WR Test Facilities ............. ...... .............. ............... .... ............. 1-8
-. 1.3 Strategy for Determination of Test and Analysis Needs........................... 1-17 1.4 Overall Test and Analysis Plan............................................................... 1-20 1.4.1 Relationship of TAPD Document to Overall TRACG Validation ............ 1-20 1.4.2 List of Reports to be Submitted to the NRC............................................. 1-21 ,
1 2.0 IDENTIFICATION OFIMPORTANTTHERMAL-HYDRAULIC . 1 PHENOMENA: TOP-DOWN PROCES S ................................ .... ....... . ..... .... 2-1 ! 2.1 In trod uc ti o n . . . . . . . . .. . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . .2-1 ............ l 2.2 Analysis of Events . . . .. .... .. ... .. ....... . ... . . ... . .. ... ... .. . . . . . ....... .. ........ . . . . .. . . . .. .. . .. . 2-2 l 2.2.1 Loss-of-Coolant Accident (LOCA).................. ....................................... 2-2 j 2.2.2 Antici pated Transients ... ... .... . . . . .... . ... ... .. ...... .. . .. ..... . .... .. .. ... ...... .. ... ... . . . .. .. 2-10 .i4 2.2.3 Anticipated Transient Without Scram (ATWS) .................. .................... 2-13 i 2.2.4 Stability................................................................................................... 2-14 2.3 Phenomena Identification and Rankmg Tables (PIRT) ............................ 2-26 2.3.1 Loss-of-Coolant Acciden t (LOCA). ..... .. ... . .... . ..... .. . .. . .. . . .. . . ..... .... . . .. . . . . . .. .. 2-27 2.3.2 Anticipated Transien ts . ... .. . .. . .. . .. . ... .. .. . . . . . . ... .. . . .. .. . . . .. . . .. . . .. .. .. .. . . ... . . ... . . . . . . .. 2-27 2.3.3 Anticipated Transient Without Scram...................................................... 2-28 j l 2.3.4 Stability............................................................................................ 2-29 l l I
,, 3.0 IDENTIFICATION OF SBWR-UNIQUE FEATURES AND PHENOMENA:
B OTTOM-UP PROCE S S .. .. . .. . .. . . ....... .. ......... ............ . . .. . . . ....... .... . ... .. . . ... . . . . . . . . . . . 3-1 3.1 In trod uc ti on . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3-1 .............. 3.2 Me th od olo g y . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.3 Results.................................................................................................. 3-3 3.3.1 RPV and Inte m ais (B 1 1 ) . ...... .. . ... .. ... .... . . . ..... .... .. .. . ..... . .. ..... .... ... .... . . .. . 3-3 3.3.2 Nuclear Boiler System (B 21).......................... ....... ................................ 3-3 3.3.3 Isolation Condenser System (B32)........................................ .................. 3-3 i 3.3.4 Standby Liquid Control System (C41).. ............................................. .... 3-3 3.3.5 Gravity-Driven Cooling System (E50) . ................ ......... ....................... 3-3 3.3.6 Fuel and Auxiliary Pools Cooling System (G21) ..................................... 3-3 11 L________-_____-_____________- -____-___ ___----___-- - -
. . _ _. _ . ._- . . = NEDO-3239'1, Revision C l l TABLE OF CONTENTS (Continued) ! 3.3.7 C ore (J - Se rie s) . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-3 3.3.8 Co n tainmen t (T 10) . . . . .. . . . . . . ... .. . . . . . . . . .. . .. . . . . .. . . . ... ... . . .. . . .. . . . . . . . .. . .. .. . .. . .. 3-3 . .. . . 3.3.9 Passive Containment Cooling System (Tl5) ........ .............. ................... 3-3 4.0 EVALUATION OF IDENTIFIED PHENOMENA AND INTERACTIONS..... 4-1 4.1 Composite List of Identified Phenomena and Interactions....................... 4-1 4.2 Analytical Evaluation of System Interactions...... ..... ............... ............. 4-1 - 4.2.1 Accident Scen ario Definition ... ... .... ....... .......... .................... ........... 4-2 4.2.2 Results from the Primary Systems Interactions Study . ........................... 4-2 4.2.3 Results from the Containment Systems Interactions Study................. .... 4-2 4.2.4 Summary of System Interaction Studies............... .. .................... .......... 4-3 4.3 S um m ary of Eval u ations . . ... . . . . . .. .. . .. . .... .. . . .. . .. . . . . ... . . . .. . . . . . .. . . . . . . . . .. . .. . . . . .. . . 4-3 4.3.1 LOCA............................................................................................... 4-3 4.3.2 Transients......................................................................................... 4-3 4.3.3 ATWS and S tability. . . . .. . . . . .. .. .. . . . . . . .. . . . . .. . . . . . . . . . ... .. . . .. . .. . . . . ... . .. . . . . . . . . . 4-3 5.0 MATRIX OF TESTS NEEDED FOR SBWR PERFORMANCE ANALYSIS... 5-1 5.1 Separate Effects Tests . . . . . . . . . . . ... . . . . . . . .. . . . . . . . . ... ... . .... . . . . ... . .. .. . .. . . . . . ... .. .. .. . . . 5-1 5.2 Component Perfonnance Tests................ ................ .. .. .... ........... ....... 5-1 5.3 Integral System Response Tests ....... .. ....... ........................................... 5-2 5.4 Plant Operatin g Data... . .. . . . . . ... . . . . . . . . . . .. .. ... . . .. . .. . . . ... . . . . . . . .. . . . .. .. . . . . . . . . . . . . . . 5-2 5.5 S ummary of Test Coverage.... . ......... ........... ......... .. ......... ......... .. .. .. .. 5-2 6.0 INTEGRATION OF TESTS AND ANALYSIS ................................................. 6-1 6.1 TRACG Qualification Plan .. .. .. . ....... .......................... . ....... ..... ............ .. 6-1 6.2 Use of Data for TRACG Model Improvement and Validation................. 6-1 '- 7.0
SUMMARY
AND CONCLUSIONS .... .......... .......... ......... ........... ......... . .. .. 7-1 s 8 .0 RE FERENCES . .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .8-1 i
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111
NEDO-3239), Revision C TABLE OF CONTENTS (Continued) APPENDICES A. TEST AND ANALYSIS PLAN (TAP) .......... .................... ................... A-1 A.1 In trod u c ti o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1 A.2 Test and Analysis Philosophy...... .. ..... . ... ...... .................................... A-1 A.3 Test and Analysis Plan .. . . . .. ........ . . ........ . . . . . .. . . ...... .. .. . . .. ... . ... .... . . . ... .. . .. . .. . A-4 Attachment Al -Test and Analysis Document Tables-of-Contents. ................. Al-1 B. SCALING APPLICAB ILITY .... .......... ..... ..... . . ..... .... . ....................... B-1 B.1 Scalin g S um m ary . . . . . . . . . .. . .. . . ... .. .... ... .. .. . . . .. .... . . ... . . . . .. .. . . . . . . .. . . . .. . . . . . . . B-1 C. TRACG INTERACTION STUDIES . ............... ..................................... C- 1 C.1 In tro d uc ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . C-1 C.2 Scenario Definition for Interaction Studies...... ...................... ..... .......... C-1 C.3 Primary System Interaction Studies .................... . ... ........ .................... C-2 C.4 Containment Interaction Studies................................. ............................ C-2 l C.5 Sum m ary of Interaction Studies ....................................................... ...... C-2 l l l 4' d'
NEDd-3239), Revision C i LIST OF TABLES F 1.2-1 Evolution of the General Electric BWR..................... ................ ...................... 1-11 1.2-2 SBWR Features and Related Experience ....................... ..... .............................. 1-12 ; 1.2-3 S B WR and ABWR Analysis Methods . . . . . . .. ... . . .. . ... . .. . . . . . . . . . . . . . . . . . . . . .. . . . ... . . . . . . .. . . . . . . 1-13 2.2-1 GDCS Line Break Sequence of Events ......................... ..................................... 2-16 2.2-2 LOCA Scenario with Diesel Generators Available-Additional S ystems Functional ............... ........... ......... ............ ........................ .... 2-18 2.2-3 LOCA Scenario with Offsite Power & Diesel Generators ~ Available........................................................................................................... 2-18 l APPENDICES , r A.2-1 Thermal-Hydraulic Test Data Groups and Description ....................................... A-44 A.2-2 SBWR Test Documentation Submittals .......................................... ........... ....... A-45 A.2-3 SBWR Analysis Documentation Submittals .. ................... ............................ .. A-47 A.3-1 Required Thermal Hydraulic Measurements-PCC Test........ . ............................ A-48 A.3-2a PANTHERS /PCC Steady State Performance Matrix - Steam Only Tests.......... A-49 A.3-2b PANTHERS /PCC Steady State Performance Matrix - Air-Steam Mixture Tests................................................................................................................. A-50 A.3-2c - PANTHERS /PCC Noncondensible - Buildup Test Matrix ................................. A-53 A.3-2d PANTHERS /PCC Pool Water level Effects - Test Matrix .......................... ..... A-53 A.3-3 PANTHERS /PCC TRACG Qualification Points......... ................... ................... A-54 A.3-4 Required Thermal Hydraulic Measurements - IC Test ........... .. ......................... A-55 : A.3-5a PANTHERS /IC Steady State Performance - Test Matrix................. ................. A-56 , A.3-5b PANTHERS /IC Startup and Operation - Test Matrix................ .. ..................... A-56 A.3-5c PANTHERS /IC Noncondensible Gas Effects - Test Matrix................ .......... .... A-57 A.3-5d PANTHERS /IC Water Level Effeets - Test Matrix ... .......... ........................... . A-57 A.3-6 PANTHERS /IC TRACG Analysis Cases .. .. . .. . ... . . . ... . . . .. . ... . .. . . . . . . . . . .. . . .. . . . . . .. . . . . .. . . A-58 . A.3-7 Panda Instrumentation Summary ........ .................................. ....... ................ ... A-59 A.3-8 Instrumentation Required for Test S 1 to S 13................ ......... ... ....... . ............ A-60 A.3-9a PANDA Steady State PCC Performance Test Matrix ......................................... A-61 A.3-9b PANDA Systems Test Matrix Summary................... ......................................... A-62 A.3-10a Initial Conditions for PAND A Test M3........................................ ..................... A-63 A.3- 10b Initial Conditions for PAND A Test M7.............................................................. A-63 A.3-11 SBWR Containment Conditions at 3600 see for Main , l- S teamline B reak LOCA . .. .. . ... .. . . . ..... . . .. .. ... ... . . ... . .. .. .. . . .. .. .... . . . . . . . . .. . . . . . . . . .. .. . . . .. . . .... A-64 A.3-12 Time Derivatives of Key PANDA Initial Conditions .. . ........... ......................... A-66 A.3-13 PANDA TRACG Analysis Cases . . . ....................................... ... ..... . .... ........ . ... A-67 Y
NEDO-32391, Revision C LIST OF TABLES (Continued) APPENDICES (Continued) A.3-14 GIST Test Matrix Initial Conditions (RPV at 100 psig) ...... .... ... .. ..... . ..... ..... A-68 A.3-15 GIST Runs With Existing TRACG Analysis............... .......... ........... . ............. A-70 A.3-16 GIRAFFE Test Matrix (Phase 1 Step- 1).................... ........ .............. ...... ......... A-71 A.3-17 GIRAFFE / Helium Integral Systems Test Matrix... ..... . ............................... .. A-72 A.3-18 GIRAFFE / Helium Base Case (H1) Initial Conditions . ....... .. .... ....... ...... .. . .. A-72
- A.3-19 GIRAFFE / Helium " Tie-Back" Initial Conditions....... .. .. .... ........ .... .... .. . .. A-73 A.3-20 GIRAFFE / Helium Test T2 Initial Conditions.. .... ............. ............. . . . . . . .. . A-74 A.3-21 G IRAFFE/S IT Test M atrix . . . . . . . . . . . .. . . . . . . .. . . . . . . . . .. . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . . . . . . ... . A-74 A.3-22 Test G S 1 Initial Conditions .. .. .... .. .. . .. .. ... ... . .. . .. . . . . .. . .. ....... .. ... . ... . ... . . . . .. . A-75 l
A.3-23 Basis For GIRAFFE / SIT Test Conditions .... . .................. ... . .. ...... .... .. . .... . A-75 A.3-24 PANTHERS PCC Structural Instrumentation ..... . ......... .. .. ... .. . . . . . . . . . . . . A-76 A.3-25 PANTHERS PCC Component Demonstration Test Matrix . . ......... .. .... ........ A-77 A.3-26 PAN'IIIERS/PCC LOCA Cycle Time History. ............. . ............. .. ... .. . . A-77 A.3-27 PANTHERS IC Structural Measurements .................. .. .... ...... .. ... . ..... ........ A-78 A.3-28 PANTHERS IC Component Demonstration Test Matrix..... ..... ......... ... ..... .... A-79 A.3-29 Comparison of Non-Dimensional Parameters Between SBWR and CRIEPI... .. . A-80 l 8 4 l _ _ _ _ _ _ - -
1 NEDO-32391, Revision C i 4 LIST OF FIGURES 1.I-1 TAPDFocus......................................................................................................1-3 1.2-1 Evoluti on of the B WR . . .. . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . .. . . .. . . .. . . . . .... . . . ... . . . .. . . . . . . . . .. . .. .. . . . . . . 1-14 J l 1.2-2 Evolution o f the B WR ... . . . . . . . . . .. . . . . . .. . . . . . . . . . . . . . . .. . . . . .. . . . . . . . .. .. . .. . . . . . . . ... . . . . . ... ... . . . . . 1-15 1.2-3 Comparison of BWR Containmen ts ........................ ........ ...... ........................... 1-16 1.3-1 Strategy for Determination of Test Needs...... .............. . .................................. 1-19 1.4-1 Technology B asis for SBWR Design ...... ............... ....... ... ............................. 1-22 1.4-2 Overall Test and Analysis Plan .... ...................... ............... ... ........ ..... .... ....... 1-23 , i 1.4-3 Road Map of SBWR TRACG Related Documentation . . . ........... .... ................ 1-24 2.2-1 S.BWR Passive S afety Systems .... ... .. ... .. . ...... ....... . ... ..... . . . .... . . ... . .......... . .... .... . 2-19 2.2-2 Phases of the LOCA Transient .... .. .. . .... . .. ........... ... ............. .... .. . . ... .... .. . .......... 2-20 2.2-3 GDCS Line Break-Chimney Two-Phase Level vs. Time ..... ... .. .. .................. 2-20 2.2-4 GDCS Line Break Containment Pressure and Temperature vs. Time... ........... .. 2-21 2.2-5 GDCS Line Break Decay Heat and PCC Power vs. Time .. . .............................. 2-22 2.2-6 MSIV Closure ATWS with Boron Injection ..... ... ............. ............................... 2-23 2.2-7 SBWR Stability Design Criteria and Performance.............. . ..... ...................... 2-25 2.2-8 SBWR Power / Flow Map Comparison With Calculated Stability Limit............... 2-25 3.2-1 S B WR Prod uc t S tructure . . . . . . . . . . . .. . . .. . . . . . . . . . . . . . .. . .. . .. .. . . . .. .. . .. . .. . . . . .. . . . . . . . . .. . . . . . . . 3-2 6.1-1 Technology B asis for SBWR Design ......................... ..... ............. .... ............... 6-2 APPENDICES A.3-1 Passive Containment Cooler Test Article.. ........................ ...... .. .................... . A-81 A 3-2 PANTHERS /PCC Test Facility Schematic .... ................................................... A-82 A.3-3 PCC Heat Exchanger Operational Modes ... .......... .... ................. ..................... A-83 A.3-4 Comparison of PANTHERS /PCC Steam-Air Test Range to S'BWR Conditions . A-84 . A.3-5 TRACG PANTHERS /PCC Qualification Points....... . .... ........... ... ......... ........ A-85 A.3-6 Isolation Condenser Test Article......................... ... ... ......... ........ . .... .. ..... ... . A-86 ,. A.3-7 PANTHERS /lC Test Facility Process Diagram .. . ...... .. .. .... ... ... .... ... ...... . A-87 A.3-8 PANDA Facility: IC/PCC Test Units...... . .... ...... .... .. . .... ............ .. . .. ........ A-88 A.3-9 PANDA Facility Schematic.......... ... ... ..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-89 ' A.3-10 PANDA Facility: Configuration of Vessels .. ........... .... ... . ... ..... ..... . ..... A-90 A.3-11 PANDA Facility: PCC 3 Steady State Supply Line ... .... ... ............. . . .. ....... A-91 l A.3-12 Deleted.............................................................................................. ...... A-92 A.3-13a PANDA Instrumentation: Condenser, Pool, and Vessel Temperatures..... .......... A-93 A.3-13b PANDA Instmmentation: Mass Flow Rates ............ ............. . .......... ........ .... .. A-94 A.3-13c PANDA Instrumentation: Absolute and Differential Pressures............. . .......... A-95 l A.3-13d PANDA Instrumentation: Noncondensible (C) and Phase Detectors.................. A-96 t i
NEDO-32391, Revision C LIST OF FIGURES (Continued) APPENDICES (Continued) A.3-14 PANDA Steady State Test Instrumentation / Configuration .. ... . . ..................... A-97 A.3-15 Comparisen of PANDA Steady State Test Range to SBWR Condition .... .. ...... A-98 A.3-16 GIST Facility S ehem atic .... .. ..... .. . ... . ...... . .. . . . . . . . .. ... . .... .. . .. . . . .. ... . ... .. . ... . . A-99 A.3-17 GIST Facility Piping Arrangement .... . ..... .......... ......... . . ... ............ . .. ...... A-100 A.3-18 GIRAFFE Test Facility Schematic (Phase 1) ........ ... . ..... ....... ..... ..... .. .... ..... A-101 A.3-19 GIRAFFE IC/ PCC Unit ......... ....... ............... ..... ....... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-102 A.3-20 GIRAFFE Test Facility Schematic (Post Phase 1)... . . ..... .. .. ........... ...... .. A-103 A.3-21 GIRAFFE PCC Unit (Shortened Tubes) .............. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A- 104 A.3-22 PCC Startup Initial Condition Map ...... ..... .. ..... ...... ....... .... ................ . ..... . A-105 A.3-23 IC Cyc le Types . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . .. . . . . . . . . . . . . . . .. A- 106 A.3-24 CRIEPI Test Loop Outline .. . .. . .. . ... . . .. .. ..... . . . . .. . .. .. . .. ... . ... . . . ..... .. . ... ... . ... .. .. . A- 107 I l A.3-25 CRIEPI Facility Stability Map Under Lower Pressure Startup Conditions and Representative Parameters for SBWR ... ...... . ...... . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-108 i 4 l l l 5 b Ykkk
NEDO-32391, Revision C ABBREVIATIONS AND ACRONYMS ABWR Advanced Boiling Water Reactor AC Alternating Current ADS Automatic Depressurization System APRM Average Power Range Monitor
.- ARI Alternate Rod Insertion ASME American Society of Mechanical Engineers ATLAS GE's 8.6 MW Heat Transfer Loop ~
ATWS Anticipated Transients Without Scram Bldn Blowdown BO Boiloff BWR Boiling Water Reactor CACS Containment Atmospheric Control System CCFL Counter Current Flow Limiting CISE Centro Informazioni Studi Esperienze COL Combined Operating License CPR Critical Power Ratio ( CRD Control Rod Drive CTP Core Thermal Power CRIEPI Central Research Institute of Electric Power Industry CSAU Code Scaling, Applicability and Uncertainty l CSHT Core Spray Heat Transfer DBA Design Basis Accident DC Downcomer DPV Depressurization Valve DW, D/W Drywell EBWR Experimental Boiling Water Reactor ECCS Emergency Core Cooling System EOPs Emergency Operating Procedures l FAPCS Fuel and Auxiliary Pool Cooling System
. FIST BWR FullIntegral Simulation Test FIX Swedish Test Loop Used for Testing Extemal Pump Circulation FMCRD Fine Motion Control Rod Drive FRIGG Research Heat Transfer Loop Operated for Danish Atomic Energy Commission FW Feedwater FWCS Feedwater Control System GDCS Gravity-Driven Cooling System GE General Electric Company ix
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NEDO-32391, Revision C ABBREVIATIONS AND ACRONYMS (Continued) GEXL General Electric Critical Quality Boiling Length Correlation GIRAFFE Gravity-Driven Integral Full-Height Test for Passive Heat Removal GIST GDCS Integral System Test HCU Hydraulic Control Unit .. HVAC Heating, Ventilating and Air Conditioning IC Isolation Condenser
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ICS Isolation Condenser System INEL Idaho National Engineering Laboratory LASL Los Alamos Scientific Laboratory LB Large Break LOCA Loss-Of-Coolant Accident LOOP Less Of Offsite Power LPCI Low Pressure Coolant Injection MCPR Minimum Critical Power Ratio MIT Massachusetts Institute of Technology MPL Master Parts List MSIV Main Steamline Isolation Valve MSL Main Steamline MW Megawatt NBS Nuclear Boiler System NRC Nuclear Regulatory Commission ORNL Oak Ridge National Laboratory P&ID Process and Information Diagram PANDA Passive Nachwarmeabfuehr-und Drueckabbau-Testanlage (Passive Decay Heat Removal and Depressurization Test Facility) PANTHERS Performance Analysis and Testing of Heat Removal Systems . PAR Passive Autocatalytic Recombiners I PCCS Passive Containment Cooling System , l PCT Peak Cladding Temperature PIRT Phenomena Identification and Ranking Tables i PSTF Pressure Suppression Test Facility j QDB Qualification Data base i RC&IS Rod Control and Information System I RPV Reactor Pressure Vessel RWCU Reactor Water Cleanup ! SB Small Break l l l x
NEDO-32391, Revision C i i l ABBREVIATIONS AND ACRONYMS (Continued) SBWR Simplified Boiling Water Reactor S/C Suppression Chamber (wetwell) SDC Shutdown Cooling SIET Societa Informazioni Esperienze Termoidrauliche o SLCS Standby Liquid Control System SPERT Special Power-Excursion Reactor Tests SRV Safety / Relief Valve
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SSAR Standard Safety Analysis Report , SSLC Safety System Logic Control i SSTF Steam Sector Test Facility TAPD Test and Analysis Program Description TCV Turbine Control Valve THTF Thermal-Hydraulic Test Facility TLTA Two-Loop Test Apparatus TPS Turbine Protection System TRAC Transient Reactor Analysis Code TRACG Transient Reactor Analysis Code, GE version IT Turbine Trip UCB University of Califomia, Berkeley ; VB Vacuum Breaker WW Wetwell l 1 ) I Xi
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- NEDO-32391, Revision C I
1.0 INTRODUCTION
1.1 Purpose he pu pose of the Simplified Boiling Water Reactor (SBWR) Test and Analysis Program j Description (TAPD) is to provide, in one document, a comprehensive, integrated plan that addresses the testing and analysis elements needed for analysis of SBWR steady state and' transient i . perfonnance. The program was developed by: i Study of the calculated SBWR transients and identi6 cation ofimportant phenomena. Identification of the unique SBWR design features and their effect on transient i performance. 1 2 Systematic definition of expenmental and analytical modeling needs. 1 t Evaluation of the current experimental and analytical model plan against these needs Definition of modifications as necessary. I his document describes the steps in this process leadmg to the final Test and Analysis Plan ! (Appendix A). The TRACG computer code is used for the analysis of SBWR transients, IAss-of- l 1 Coolant Accidents (LOCAs), Anticipated Transients Without Scram (ATWS) and stability. The Test l
- Plan has been cross-referenced against the identified phenomena to create the TRACG Qualification
- Matrix. Section 1.3 describes in more detail the strategy employed to arrive at these objectives. The use of specific tests in the development of TRACG models, for test predictions and for post-test validation, is addressed in this report. Descriptions of the SBWR-specfic test facilities and their l fidelity with respect to scaling the SBWR plant are provided in Appendices A, B and Reference [32].
He SBWR TAPD thus provides the technology basis for detennining the peiformance of the plant for transients and accidents. It ties together the ongoing diverse experimental and analytical efforts in support of SBWR certification. The ultimate output from this activity is a set of validated j analytical methods (primarily the TRACG computer code) for SBWR performance analysis. i 1.1.1 Scope ]- he SBWR Test and Analysis Program Description is directed at providing a sound technology , basis for the prediction of SBWR system perfonnance during normal operation, transients and LOCAs. He document scope includes (1) steady state operation and startup conditions, (2) transients 4 and ATWS, (3) stability, and (4) LOCA. LOCA response covers the vessel response [ levels in the clumney and dowrcomer and peak cladding temperature (PCT)] with operation of the Emergency . Core Cooling Systems (ECCS), as well as the containment pressure and temperature response to postulated breaks. Long-tenn core cooling by inventory makeup is also considered. he document does not address " severe accident" issues. The requirement to design the contamment to handle hydrogen generation assuming 100% metal-water reaction is, however, addressed as a Design Basis requirement. Issues related not to thermal-hydraulics but, for example, to material properties, crack resistance, water chemistry, etc., are not covered in this plan. 1-1
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NEDO-32391, Revision C
! he TAPD focus is illustrated in Figure 1.1-1. Transients and accidents, short of severe core damage, have been analyzed and the experimental and modeling needs incorporated into the plan. In the time domain, the focus of the studies has been on the first three days following a postulated accident or transient. Quasi-steady state conditions prevail well before this point in time. Interactions ;
with active systems such as the Fuel and Auxiliary Pool Cooling System (FAPCS) have been studied. ; j No new phenomena are introduced beyond this point. ; j He experimental and analytical modeling needs were analyzed in the context of the applicable ; criteria of 10CFR52.47(b)(2)(i)(A), which require in part that: , , ) - he performance of each safety feature of the design has been demonstrated through either analysis, appropriate test programs, expedence, or a combination thereof. , Interdependent effects among the safety features of the design have been found to be acceptable by analysis, appropriate test programs, experience, or a combination thereof. Sufficient data exist on the safety features of the design to assess the analytical tools used for safety analysis over a sufficient range of normal operating conditions, transient conditions, and specified accident sequences, including equilibrium core conditiors. 3 He term " safety feature" in the precedmg paragraph is understood to include safety-related . l passive systems as well as other active systems which may be available to operators during accidents
- or transients. De Bottom-Up process described in Section 3 specifically examines all SBWR-unique
features that are relevant to safety. Issues related to these features have been evaluated and the l
- supporting technology basis (analysis, experimental data, plant data) documented. Interdependent !
j effects among safety features have been specifically considered. Analyses have been performed l j (Appendix C) to screen interactions that deserve experimental validation. Finally, a test program has j i been established which provides a sufficient data base for the qualification of the 'IRACG Code for l SBWR safety analysis. ! i i I I 4 1 1 l ' l-2
1 1 NEDO-32391, Revision C l l J e SEVERE ACCIDENTS DBA D a E i TRANSIENTS I STARTUP j NORMAL OPERATIONS 3 TIME (days) , Figure 1.11 TAPD Focus 4 4 1-3
NEDO-32391, Revision C 1.2 Background l SBWR Design Evolution: he SBWR design is an evolutionary step in boiling water reactor (BWR) design which traces , its commercial demonstration and operating plant history back before 1960 (Figure 1.2-1). Since its , inception, the BWR has had plant simplification as a goal for each product improvement (Figure 1.2-2). De SBWR has major simplifying improvements drawn from predecessor designs, notably ! pressure-suppression contamment, natural circulation, isolation condenser handling of waste heat, and , gravity <lriven makeup water systems (Table 1.2-1). He incorporation of these features from i predecessor designs into the SBWR has emphasized employment of passive means of dealing with operational transients and hypothetical LOCAs. De result of this evolution of previously licensed , , plant features is simplified operator response to these events (most plant upset conditions are dealt with in the same manner, as typified by the hypothetical steamline break), and a lengthened operator response time for all hypothetical events (from minutes for previously licensed reactors to days for
' the SBWR). Most features of the SBWR have been taken directly from licensed commercial BWRs and reviewed and redesigned as appropriate for the SBWR (Table 1.2-2). For example, the Residual Heat Removal (RHR) heat exchangers in currently operational BWRs provide an extensive experience base for the design / fabrication / testing / operation / maintenance of condensing mode heat exchangers. The SBWR Passive Containment Cooling System (PCCS) heat exchangers take [
advantage of this experience, are designed specifically for the SBWR requirements, and are being thoroughly tested as described in Appendix A to this report. The SBWR draws together the best of previously licensed plant features to continue the simplification process. As an example, the evolution of the contamment is shown in Figure 1.2-3. , Analysis and Design Tools: As implied above, data available from operating plants and from the testing and licensing efforts done to license the predecessor designs (most recently, ABWR) is the pdncipal foundation of SBWR technology. As a measure of the SBWR's reliance on demonstrated technology, approximately 50% of the content of the SBWR SSAR is technically identical or technically similar (with minor differences) to the ABWR SSAR [31]. The 930 reactor-year data base [40] of feature performance in operating reactors, combined with the recent thorough licensing review of the ABWR , (Final Design Approval received July 1994), provides well-qualified foundation from which to make , r the modest extrapolations to the SBWR. - To make that extrapolation, GE has developed one computer code (TRACG) to use for design and for three out of the four most limiting licensing analyses. He TRACG Code, validated by , operating plant experience and appropriate testing, is used to analyze the challenges to the fuel (10CFR50.46 and Appendix K, SSAR Section 6.3), the challenges to the containment (SSAR Section 6.2), and many of the operational transients (MCPR, SSAR Chapter 15). De radiological responses to hypothetical accidents are also presented in SSAR Chapter 15, but do not use TRACG for analysis. Hus, 'IRACG draws from the very large data base of licensed BWRs which includes all features of ! the SBWR (albeit in various configurations) and appropriate testing, and allows direct application to j SBWR design and analysis. ! 1-4 .
NEDO-32391, Revision C 1.2.1 Use of TRACG i De TRACG Code and its application to the SBWR is documented in a series of GE Nuclear l Energy Topical Reports ([1], [2], and [7]).
'IRACG is a GE proprietary version of the Transient Reactor Analysis Code (TRAC). It is a best-estimate code for analysis of BWR transients ranging from simple operational transients to design basis LOCAs, stability, and A'IWS. j 1.2.1.1 Background . 'IRAC was originally developed for pressunzed water reactor (PWR) analysis by Los Alamos National Laboratory (LANL), the first PWR version of TRAC being TRAC-PI A. De development ;
of a BWR version of TRAC started in 1979 in a close collaboration between GE and Idaho National Engineering Laboratory. He objective of this cooperation was the development of a version of
'IRAC capable of simulating BWR LOCAs. He main tasks consisted of improving the basic models in TRAC for BWR applications and developing models for the specific BWR components. This ;
work cut^.ated in the mid-eighties with the development of 'IRACB04 at GE and 'IRAC- ! BDl/ MODI at INEL, which were the first major versions of TRAC having BWR LOCA capability. Due to the joint development effort, these versions were very similar, having vutually identical basic and component models. The GE contributions were jointly funded by GE, the Nuclear Regulatory , Commission (NRC) and Electric Power Research Institute (EPRI) under the REFILUREFLOOD and FIST programs. De development of the BWR version has continued at GE since 1985. The objective of this development was to upgrade the capabilities of the code in the areas of transient, stability and ATWS j applications. Major improvements included the implementation of a core kinetics model and addition i of an implicit integration scheme into TRAC. The contamment models were upgraded for SBWR j applications, and the simulation of the fuel bundle was also improved. 'IRACG was the end result of this development. 1.2.1.2 Scope and Capabilities } TRACG is based on a multi-dimensional two-fluid model for the reactor thennal-hydraulics ,- . and a three-dunensional neutron kinetics model. he two-fluid model used for the thennal-hydraulics solves the conservation equations for mass, momentum and energy for the gas and liquid phases. 'IRACG does not include any assumptions of thennal or mechanical equilibrium between phases. De gas phase may consist of a
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l mixture of steam and a noncondensible gas, and the liquid phase may contain dissolved boron. He thennal-hydraulic model is a multidimensional formulation for the vessel component and a one-dimensional formulation for all other components. De conservation equations for mass, momentum and energy are closed through an extensive set of basic models consisting of constitutive correlations for shear and heat transfer at the gas / liquid
- interface as well as at the wall. The constitutive conclations are flow regime dependent and are determined based on a single flow regime map, which is used consistently throughout the code.
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NEDO-32391, Revision C i 1 In addition to the basic thermal-hydraulic models, TRACG contains a set of component models for components, such as channels, steam separators and dryers. TRACG also contains a control system model capable of simulating the major control systems such as reactor pressure vessel (RPV) : pressure and downcomer sensed water level. [ The neutron kinetics model is consistent with the GE core simulator code PANACEA. It soh'es a modified one-group diffusion model with six delayed neutron precursor groups. Feedback is ! provided from the thermal-hydraulic model for moderator density, fuel temperature, boron !
- concentration and control rod position. -
i i The TRACG structure is based on a modular approach. The 'IRACG thermal-hydraulic model : I contains a set of basic components, such as pipe, valve, tee, channel, steam separator, heat exchanger , _ and vessel. System simulations are constructed using these components as building blocks. Any i number of these components may be combined. The number of components, their interaction, and , 1 the detail in each component are specified through code input. TRACG consequently has the capability to simulate a wide range of facilities, ranging from simple separate effects tests to complete j
. plants. !
4 'IRACG has been extensively qualified against separate effects tests, component performance data, integral system effects tests and full-scale plant data. A detailed documentation of the qualification is contained in the TRACG qualification report NEDE-32177P [2]. l i 1.2.1.3 Scope of Application of TRACG to SBWR [ The TRACG computer code has been qualified to Level 2 status at GE-NE. Thus, the code configuration is controlled, and the models and the resuks of validation testing have been reviewed
- and approved by an independent Design Review Team. In the development process, the separate
- effects and component data were used for model development and refinement.
The total effort and extent of qualification performed on TRACG, since its inception in 1979, now exceeds, both in extent and breadth, that for any other engineering computer program which GE j has submitted to the NRC for design application approval. The Level 2 application of 'IRACG ! includes LOCA analyses, transients, ATWS and Stability Analyses for the reactor and contamment. Table 1.2-3 compares the analytical methods used for ABWR and SBWR analysis. The table shows that GE has taken a major step forward in utihzmg one code (IRACG) for the bulk of the safety analysis. This results in greater consistency and simplification of the analysis process. The use of e TRACG to unify the LOCA analysis for the reactor vessel and contamment is particularly important for the SBWR because the two regions are closely coupled during the transient. While TRACG is used for all the analyses given in Table 1.2-3, the application of TRACG in the design process is different for ATWS and stability. For LOCA (ECCS and contamment) and transient analysis, GE performs SSAR calculations utihzmg a best-estimate analytical technique which reahstically describes system behavior and appropriately considers uncertainties in the analysis methods and inputs per the requirements of 10CFR50.46(a)(1)(i). 'Ihe ATWS calculations are performed as best-estimate calculations. For stability analysis, NRC approved methodology (FABLE) is used in the design process for determination of core and channel stability margins.
'IRACG is used for the evaluation of overall plant stability. 'IRACG has also been used to study the possibility of oscillations during the plant startup transient.
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NEDO-32391, Revision C 1.2.13.1 Transient Analysis
'IRACG is used to perform safety analyses of nearly all of the Anticipated Operational Occurrences (AOO) described in SSAR Chapter 15, and of the ASME reactor vessel overpressure protection events in SSAR Chapter 5. The Loss of Feedwater Heating and the Control Rod Withdrawal Error events presented in SSAR Chapter 15 are analyzed using the GE 3-D core simulator model. Other SSAR Chapter 15 exceptions are the control rod drop and the fuel-handling accidents, and radiological calculations for all postulated accidents.
The analysis determines the most limiting event for the AOOs in terms of Critical Power Ratio (CPR) and margin loss (ACPR) and establishes the operating limit minimum CPR (OLMCPR). The OLMCPR includes the statistical CPR adder which accounts for uncertainty in calculated results arising from uncertainties associated with the TRACG model, initial conditions, and input parameters. Sensitivity analysis of important parameters affecting the transient results is performed using TRACG. Concepts derived from the Code Scaling, Applicability, and Uncertainty (CSAU) methodology are utihzed for quantifying the uncertainty in calculated results. He analysis also determines the most limiting overpressure protection events in terms of peak vessel pressure. He results are used to demonstrate adequate pressure margin to the reactor vessel design limit with the SBWR design safety / relief valve capacity. He overpressure protection analysis is performed based on conservative initial conditions and input values. 1.2.13.2 ATWS Analysis TRACG is used for evaluation of the ATWS events in SSAR Chapter 15. The analysis determines the most limiting ATWS events in terms of reactor vessel pressure, heat flux, neutron flux, peak cladding temperature, suppression pool temperature, and contamment pressure. The results are used to demonstrate the capability of the SBWR mitigation design features to comply with the ATWS licensing criteria. 1.2.133 ECCS/LOCA Analysis TRACG is used for evaluation of the complete spectrum of postulated pipe break sizes and locations, together with possible single active failures, for Section 6.3 of the SBWR SSAR. His
. evaluation determines the worst case break and single failure combinations. The results are used to demonstrate the SBWR Emergency Core Cooling System (ECCS) capability to comply with the licensing acceptance criteria.
A sensitivity analysis of important parameters affecting LOCA results is performed using TRACG. For the SBWR, the LOCA analysis results are adjusted so that they provide 95% probability LOCA results for use as the licensing basis. The SBWR LOCA results have large margin with respect to the licensing acceptance criteria. 1.2.13.4 Containment Analysis TRACG is also used for evaluation of containment response during a LOCA. He analysis detennines the most limiting LOCA for contamment (or Design Basis Accident, DBA) in terms of 1-7
NEDO-32391, Revision C containment pressure and temperature responses. He DBA is determined from consideration of a , full spectrum of postulated LOCAs. The results are used to demonstrate compliance with the SBWR containment design limits. Sensitivity of the containment response to parameters identified as important is evaluated using TRACG to assess the effect of uncertainties of these parameters on the containment responses. The procedure derived fmm the CSAU methodology (Subsection 1.2.2) is used for mis purpose. 1.2.2 Major SBWR Test Facilities GE has used a procedure similar to the Code Scaling, Applicability and Uncertainty (CSAU)
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methodology developed by the NRC [4), [6] and submitted to the NRC by GE letter [41]. His procedure developed a list of phenomena important to the SBWR behavior in a large number of anticipated and hypothetical events and matched them against infonnation available from operating plant and'or test experience. De Phenomena Identification and Rankmg Table (PIRT) discussed in Section 2 of this report identifies specific governing phenomena, of which a significant fraction were
- concluded to be "important" in prediction of SBWR transient and LOCA performance. 'IRACG contains models capable of simulation of each of the important phenomena, and each has been qualified by the successful predictions of at least one, and in most cases, several test data sets. The PIRT defines more than 900 specific data sets, from 42 different tests and test facilities, that make up the TRACG qualification data base. Data from separate effects tests, component tests, systems and systems interaction tests, and operating plant experience have been predicted by TRACG in its validation.
Early in the SBWR program one piece ofinformation was identified as needed for the SBWR for which there was no information in the data base: that is, a heat transfer conclation for steam l condensation in tubes in the presence of noncondensible gases. A test program has since been conducted to secure this infonnation, reported to the NRC in Reference 19. The Single Tube Condensation Test Program was conducted to investigate steam condensation l inside tubes in the presence of noncondensibles. The work was independently conducted at the University of California at Berkeley (UCB) and at the Massachusetts Institute of Technology (MIT). De work was initiated in order to obtain a data base and a correlation for heat transfer in similar conditions as would occur in the SBWR PCCS tubes during a DBA LOCA. Bree researchers utilized three separate experimental configurations at UCB, while two researchers utihzed one
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configuration at MIT. He researchers ran tests with pure steam, steam / air, and steam / helium l mixtures with representative and boundmg flow rates and noncondensible mass fractions. The experimenters found the system to be well behaved for all tests, with either of the noncondensibles, .. for forced flow conditions similar to the SBWR design. The results of the tests at UCB have become the basis for the condensation heat transfer correlation used in the TRACG computer code. While all SBWR features are extrapolations from current and previous designs, two features (specifically, the Passive Containment Cooling System and the Gravity-Driven Cooling System) represent the two most challenging extrapolations. Thecefore, it was &Med for these two cases, to obtain additional test data, which could be used to demonstrate the capabilities of TRACG to successfully predict SBWR performance over a range oiconditions and scales. Blind (in some cases double blind) predictions of test facility response use only the internal correlations of TRACG. No 1-8
( l NEDO-32391, Revision C
" tuning" of the TRACG inputs is to be performed, and no modifications to the coding are anticipated as a result of these tests.
For the case of the PCCS, it is planned to predict steady state heat exchanger performance in full-vertical-scale 3-tube (GIRAFFE),20-tube (PANDA), and prototypical 496-tube (PANTHERS) l configurations, over the range of SBWR expected steam and noncondensible conditions (Appendix A). This process addresses scale and geometry diffemnces between the basic phenomena tests performed in single tubes, and larger scales including prototype conditions. Transient performance is
, similarly investigated at two different scales in both GIRAFFE and PANDA.
TRACG GDCS performance predictions were performed against the GIST test series. Pre-test predictions have also been perfonned for the PANTHERS and PANDA steady state tests. 1.2.2.1 Major SBWR-Unique Test Programs As noted pmviously, the majority of data supporting the SBWR design came from the design and operating experience of the previous BWR product lines. SBWR-unique certification and confhmation tests are briefly described below. They will be discussed in detailin Appendix A to this report. 1.2.2.1.1 GIST GIST is an experimental program conducted by GE to demonstmte the Grasity-Driven Cooling System (GDCS) concept and to collect GDCS flow rate data to be used to qualify the TRACG computer code for SBWR applications. Simulations were conducted of DBA LOCAs representing main steamline bmak, bottom drain line break, GDCS line break, and a non-LOCA loss of inventory. Test data have been used in the qualification of TRACG to SBWR and documented in Reference 42. Tests were completed in 1988 and documented by GE in 1989. GIST data has been used for validation of certain features of TRACG. 1.2.2.1.2 GIRAFFE GIRAFFE is an experimental program conducted by the Toshiba Corporation to investigate
., thermal-hydraulic aspects of the SBWR Passive Contamment Cooling System (PCCS). Fundamental steady state tests on condensation phenomena in the PCC tubes wem conducted. Simulations were l run of DBA LOCAs; specifically, the main steamline break. These tests have been completed. . GIRAFFE data will be used to substantiate PANDA and PANTHERS data at a different scale and to support validation of certain features of TRACG. Also, two additional series of tests will be conducted in the GIRAFFE facility: the first will demonstrate the operation of the PCCS in the presence of lighter-than-steam noncondensible gas; the second will proside additional information regarding potential system interaction effects in the late blowdown /early GDCS period.
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NEDO-32391, Revision C 1.2.2.1.3 PANDA PANDA is an experimental program to be nm by the Paul Scherrer Institut in Switzerland. PANDA is a full-vertical-scale 1/25 volume scale model of the SBWR system designed to model the thermal-hydraulic perfonnance and post-LOCA decay heat removal of the PCCS. Both steady state and transient perfonnance simulations are planned. Testing at the same thermal-hydraulic conditions as previously tested in GIRAFFE and PANTIERS will be performed, so that scale-specific effects may be quantified. Blind pre-test analyses using 'IRACG will be submitted to the NRC prior to start of the testing. PANDA data will be used directly for validation of cenain features of TRACG. . 1.2.2.1.4 PANTHERS . PANTHERS is an experimental program to be performed by SIET in Italy, with the dual purpose of providing data for TRACG qualification and demonstration testing of the prototype PCCS and IC heat exchangers. Steam and noncondensibles will be supplied to prototype heat exchangers over the complete range of SBWR conditions to demonstrate the ' capability of the equipment to handle post-LOCA heat removal. Testing at the same thermal-hydraulic conditions as performed in GIRAFFE and PANDA is planned. Blind pre-test analyses of selected test conditions using TRACG have been submitted to the NRC prior to the stan of testing [35]. PANTHERS data will be used directly for validation of certam features of TRACG. In addition to thermal-hydraulic testing, an objective of PANTHERS is to investigate the stmetural adequacy of the heat exchangers. This objective is beyond the scope of this report. 1.2.2.1.5 Scaling of Tests A discussion of scaling of the major SBWR tests is contained in Reference 32. That report contains a complete discussion of the features and behavior of the SBWR during challenging events. It includes the general (Top-Down approach) scaling considerations, the scaling of specific (Bottom-Up approach) phenomena, and the scahng approach for the specific tests discussed above. The detailed quantitative analyses of the major SBWR test facilities fonnerly contained in Appendix B has been incorporated in the Scaling Report [32]. l l l l l l-10
NEDO-32391, Revision C Table 1.2-1 Evolution of the General Electric BWR i Product Line Year of Number Introduction Characteristic Plants / Features B W R/l 1955 Dresden 1, Big Rock Point, Humboldt Bay, KRB, Dodewaard Natural circulation (HB, D) Internal steam separation Isolation Condenser Pressure suppression containment B W R/2 1963 Oyster Creek Large direct cycle BWR/3/4 1965/1966 Dresden 2/ Browns Ferry , Jet pump driven recirculation Improved ECCS: spray and flood 4 Reactor Core Isolation Cooling System (replaced Isolation Condenser)(BWR/4) B W R/5 1969 LaSalle
- Improved ECCS systems Valve recirculation flow control B W R/6 1972 Grand Gulf Improved jet pumps and steam separators Improved ECCS performance Gravity containment flooder l, ABWR Internal recirculation pumps Fine Motion Control Rod Drives SBWR Gravity flooder, passive containment cooling Return to Isolation Condenser Return to natural circulation 1-11 ,
NEDO-32391, Revision C Table 1.2-2 SBWR Features and Related Experience I SBWR Feature Plants Testing l IC Dodewaard, Dresden 1,2,3, Big Operating Plants Rock Pt., Tarapur 1,2, Nine Mile Pt. 1, Oyster Creek, Millstone 1, l Tsuruga, Nuctenor, Fukushima 1 i Natural Circulation Dodewaard, Humboldt Bay Operating Plants - i
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i Squib Valves BWR/1-6 and ABWR Operating Plants (SLCS, GDCS, DPVs) SLC Injection Valves IEEE 323 Qualification . Testing i I Gravity Flooder BWR/6 Upper Pool Dump System, Operating Plants Suppression PoolFlooder System PreoperationalTesting Internal Steam BWR/1-6 and ABWR Operating Plants Separators i Chimacy (Core to Steam Dodewaard, Humboldt Bay Operating Plants Separators) I FMCRDs ABWR ABWR Test / Development Program 1 (Demonstration at LaSalle Plant) 4 l l Safety Relief Valves All BWRs Operating Plants ) 1 (SRVs) Pressure BWR/1-6and ABWR Mk I, Mk II, Mk III and ABWR Suppression Tests Horizontal Vents BWR/6 and ABWR MkIII Testing l ABWR Testing 3 Quenchers BWR/2-6 and ABWR Mk I/II/III Testing ' l Operating Pl.nts ' PCC (DualFunction Heat BWR/6, RHR HX Steam Operating Plants, Exchangers) Condensing Mode PANDA, GIRAFFE, - PANTHERS l l J t ; 1 1-12 ; 4 1 4
I NEDO-32391, Revision C Table 1.2-3 SBWR and ABWR Analysis Methods Analysis Type Analysis Method ABWR SBWR Steady state ISCOR/RODAN ISCOR/TRACG Transients Pressurization 0,: VN/ FASC TRACG Loss of feedwater PANACEA PANACEA heating Other REDY/ FASC TRACG ATWS REDY/ FASC TRACG Stability FABLE /REDY FABLE /TRACG LOCA/ECCS SAFER TRACG LOCA/ containment Pressure / temperature M3CPT/SUPERHEX TRACG response Loads Approved Methodology Approved Methodology e i 1-13
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, KAHL - SANTA MARIA l FUKUSHIMA ' DE GARORA s i I 7 i ' SHOREHAM HATCH 2 l HATCH 1 l l LIMERICK 1
- I BROWNS BROWNS ,
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GRAND GULF lu$b Figure 1.2-1 Evolution of the BWR l l l e e *e % s.
i9n ' ! R R 3 , W wBA W C e B S y t i i c l p R W m B 7 2 n i s ht f e l% B e s o R K hg d s d r a i t n o Ww }c N e u y r w l o v D o E t 2-2 n 1 o e i r t u g u h i l F Af E o v
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y _ O 5
NEDC-32391P, Revision C GE ProprietaryInformation ISOLATION CONTAINMENT ISOLAllON CONDENSER CONDENSER FLOODER g t 1 CONTAINhENT FLOODERN \, o , DRY MARKI MARKIl MARK III ABWR SBWR PRESSURE SLTPRESSION NO YES YES YES YES YES NUMBER OF BARRIERS , CONTAINMENT 1 2 2 3 2 2 FISSION 2 4 4 4 4 4 2.5 0.4 0.5 1.6 0.5 0.3 VOLUME (nulhon ft3 ) 0.3 1.7 1.3 1.3 1.3 1,3 ITAT CAPACrrY (BTU x 10') 50 62 45 15 45 55 DESIGN PRESSURE (osis) 50 44 42 9 39 42 f.OCA PRESSURE (osis) Figure 1.2-3 Comparison of BWR Containments I l 1-16 1 l
NEDO-32391, Revision C 1.3 Strategy for Determination of Test and Analysis Needs The process of defining test and analysis needs for analysis of SBWR transient and accident performance is based on developing a thorough understanding of the key phenomena to be simulated and modeled. Once such a list of phenomena and intemctions between systems is compiled, the test and analysis plans can be checked against it to determine their sufficiency. In this study, a dual approach was used to arrive at a comprehensive list of controlling phenomena. Figure 1.3-1 shows the overall strategy. The Top-Down process starts with the calculated scenarios for the classes of transients and accidents to be studied. The scenario is divided into different phases based on the key events in the evolution of the transient. For example, the LOCNcontamment scenario can be divided into (1) the Blowdown phase, where the reactor vessel depressurizes, enabling the Grasity-Driven Cooling System (GDCS) to start injecting viater into the reactor vessel; (2) the GDCS phase during l which the GDCS tanks drain into the reactor pressure vessel; and (3) the Long-tenn cooling phase, after the GDCS tanks have drained and the Passive Contamment Cooling System (PCCS) removes decay heat and recycles condensed steam to the reactor vessel. For each phase of the transient, phenomena that might be important were listed and rankcd to produce Phenomena Identification and Rankmg Tables (PIRT). These tables were developed for each region of the reactor vessel and contamment. This Top-Down process and the results are described in Section 2. In the Bottom-Up process, unique SBWR design features were listed. Phenomena and issues related to these features that might influence SBWR operation and transient behavior were then compiled. This list was then reviewed and ranked by an independent team of experts. The resulting table of important phenomena and interactions is thus developed by an approach that is different from that used for the PIRT. Of course, both approaches require familiarity with SBWR transients and phenomena. This Bottom-Up process is described in Section 3. The information developed through both approaches was combined into a comprehensive tabulation of SBWR phenomena. Because the Bottom-Up approach focused on SBWR-unique features, the PIRT contains ' generic' SBWR phenomena (common to all BWRs) that were not picked up by the SBWR-unique issues. On the other hand, because the Bottom-Up approach starts with specific SBWR components and systems, it was more suitable to identify interactions between components and the various SBWR systems. The composite table can be found in Section 4.1. All the phenomena and interactions identified as important were evaluated. A Qualification Data base sheet was prepared for each phenomenon, issue or interaction, showing the expected range of SBWR parameters, the range of test data available and an analysis of the adequacy of the data base. This led to the identification of needs for additional test data or for TRACG qualification, which were
. factored into the test plan. The component and system interactions were also treated in the same manner. Numerous SBWR scenarios were analyzed to screen interactions that merited fmther study or experimental validation. This set was then compared with available integral system data that would capture these interactions. The test plan was amended to incorporate identified gaps in the data base. The results of the analytical studies are summarized in Section 4.2. Further details on the calculations are contained in Appendix C, The iterative evaluation process discussed above results in the TRACG Qualification Matrix (Section 5). The Qualification Matrix is a rearrangement of the Test Matrix showing how the identified phenomena are covered by specific tests. The Qualification Matrix has been divided into 1-17
NEDO-32391, Revision C four categories: Separate Effects Data, Component Data, Integral System Data, and BWR Operatmg Plant Data. The Test and Analysis Plan is discussed in Appendix A. It includes a brief description of each major SBWR test facility, and the test matrix, which contains the test conditions and the purpose and projected use for each category of tests. Planned analyses with TRACG for pre- and post-test calculations are identified. Detailed scahng studies were performed on the GIST, GIRAFFE, PANDA and PANTHERS facilities. The results show that the facilities are properly scaled to yield , data for cenification. Results of the scaling studies have been summarized in Reference 32. Section 6 shows how the data will be used for TRACG development and validation. Separate , effects and component data are used mainly for model development. Because interactions among components are pmsent during the overall system response of integral test facilities, these data validate the overall performance of the TRACG Code for prediction of complex system response characteristics. Integral system tests provide confinnation of the validity of the models. The feedback from these tests may also be used to improve nodalization in the TRACG representation of the test facility and, possibly, the SBWR. i 1-18
1 NEDO-32391, Revision C TOP-DOWN BOTTOM-UP
$8WR TRANSIENT SBWR-UNIQUE SCENARIOS FEATURES (SEC. 2.2) (SEC. 3.2)
L 1P 1I RELATED PIRT PHENOMENA (SEC. 2.3) (SEC. 3.2) PHENOMENA TO l : BE EVALUATED : l (SEC. 4.1) t Ir
- EVALUATION INTERACTION TEST PLAN STUDIES OF PHENOMENA :
(APPENDIX A) (SEC. 4,2)
- (SEC. 4.3) n o
SCALING QUAllFICATION TRACG MODEL (APP. B & REF. 32) MATRIX IMPROVEMENT (SEC. 5.5, 6.1) (SEC. 6.2) Hgure 1.51 Strategy for Determinadon of Test Needs 1-19
NEDO-32391, Revision C 1.4 Overall Test and Analyts Plan This section shows the relationships between the various testing, qualification, licensing and design activities. In this study, the overall TRACO qualification needs are detennined and additional SBWR related testing is defined as shown on Figure 1.4-1. As mentioned in the previous section, the primary output from the test and qualification activities is a fmal version of the TRACG computer program, which has been comprehensively validated for application to the SBWR. Figure 1.4-2 shows this process, which qualifies TRACG against large-scale component and integral system test data. A Licensing Topical Repon describing TRACG Qualification against SBWR related test data will be prepared and submitted to the NRC for review and approval. Upon completion of the technology-related activities, the SSAR calculations in Sections 6 and 15 will be re-perfonned with the fmal version of the TRACG Code. 1.4.1 Relationship of TAPD Document to Overall TRACG Validation TAPD describes the process for determining the necessary testing and analysis activities in support of SBWR technology. The output from this document is a list of the required tests and analysis tasks. This report is supplemented by numerous other reports on test results, TRACG models, qualification and application methodology. The purpose of this section is to describe the various documents that are being submitted to the NRC for review, their relationships to one another, and their roles in providing the information needed for the validation and application of TRACG. The CSAU road map, Figure 1.4-3 (from Reference 4), is a convenient means of describing how the necessary information is being provided. This road map identifies all the steps needed for validation and application of a computer code, staning from the selection of the application and the frozen code. The CSAU framework consists of three major elements comprising 14 steps. The first element relates to requirements and code capabilities. This is the process of derming the transient scenario to be analyzed (Step 1), selecting the nuclear power plant (Step 2), and development of the phenomena identification and rankmg table (PIRT) (Step 3). A frozen version of the code is selected (Step 4) and the documentation is provided on the mcdels in the code (Step 5). Comparison of the model capabilities with the phenomena to be modeled establishes the applicability of the code in Step
- 6. Element 2 is termed Assessment and Ranging of Parameters. 'Ihe major steps in this element are to establish the assessment matrix (Step 7), perform assessment of the code against separate effects tests (SETS) and integral effects tests (IETS) to determine the appropriate nodalization to be used (Step 8),
and to determine code biases and uncenainties (Step 9), as well as any bias and uncenainty due to the effect of scale (Step 10). The third element is comprised of sensitivity and uncenainty analyses. The ,. effects of reactor input parameters and operating state are evaluated in Step 11 to determine code biases and uncertainties. Calculations (Step 12) are then performed to determine the sensitisity of key parameters to the various biases and uncertainties identified in Steps 9-11. These biases and uncertainties are combined in Step 13 to determine the total uncertainty for the transient under consideration (Step 14). The TAPD addresses steps 1,2,3,4,6 and 7. PIRTs are developed for various transients, model capability is evaluated and the assessment matrix is established. 1-20
NEDO-32391, Revision C i The TRACG models are described in Reference 1, TRACG Model Description. This repon was submitted to the NRC in Febmary 1993 and is being revised to expand the description of the models and correlations. This repon addresses Step 5 in Figure 1.4-3. Reference 2, TRACG Qualification, describes the developmental acaument of TRACG, as well as comparisons with separate effects tests, integral effects tests and BWR plant data. SBWR-specific facilities such as GIST and GIRAFFE are included in this list of comparisons. Several major
. SBWR related tests are currently underway. A supplementary repon entitled "TRACG Computer i
Code Qualification for the SBWR", will be submitted after the tests are completed and analyzed. These two repons will address Steps 8,9 and 10. In addition comparing the results of TRACG
= analyses with data, the nodalization to be used for reactor and contamment analysis will be defined and model biases and uncenainties will be determined and included in the supplementary report.
Reference 7, Application of TRACG Model to SBWR Licensing Safety Analysis, is intended to address the remaining steps in the CSAU methodology (Steps 11 through 14). In the report previously submitted to the NRC, this process was completed for only operational transients. The report will be revised to incorporate the corresponding analysis for LOCA (ECCS and contamment) application. 1.4.2 List of Reports to be Submitted to the NRC The following is a list of Licensing Topical Repons planned to be submitted. (See Appendix A, Attachment 1. i TRACG Model Description, NEDO-32176 and NEDE-32176P , Revision 1. i TRACG Qualification for SBWR, new. (Supplement to TRACG Qualification, NEDE-32177P) Application of TRACG Model to SBWR Licensing Safety Analysis, NEDO-32178 and NEDE-32178P, Revision 1 In addition, there will be a Licensing Topical Repon covering the SBWR Test Program.. Additional information will be provided through a number of supplemental reports. These consist of data repons and prelimmary validation reports for each major test facility. A complete listing of these repons and their Tables of Contents are provided in Appendix A. 't 4 il , 1-21
OPERATING EX EXISTING BWR EXISTING BWR REACTOR I B TECHNOLOGY TECHNOLOGY EXPERIENCE TECHNOLOGY I COMPONENT PERFORMANCE SEPARATE QUALIFICATION INTEGRAL EFFECTS NEEDS GYSTEMS QUALIFICATION QUALIFICATION NEEDS NEEDS
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SEPARATE COMPONENT INTEGRAL SYSTEMS EFFE m QUAUFICATION PERFORMANCE QUAUFICATION AND > QUAllFICATION CONCEPT DEMONSTRATION n o L 62 Y 1> u -m u
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COMPONENT DEMONSTRATION TESTS Figure 1.4-1 Technology Basis for SBWR Design
4 NEDO-32391, Revision C MODEL LEVEL 2 MODEL FINAL DEVELOPMENT TRACG REFINEMENT VERSION i f I I i 1 i i i 1 1 I I I i 1 1 1 1 1 i i I
.. ir I i i 1 PLANNED NEW i i TESTING TESTS 1 1 1 1 t i 1 !
I e I I i i t 1 1 I I I I I I i 1 1 1r h PRE , POST TEST SAMPLE QUALIFICATION ANALYSIS REANALYSIS E 4 ir NRC NRC
- LICENSING LTR Y RNEWS REVIEWS u
4 s 1r DESIGN U AE Mgure 1.4-2 Overall Test and Analysis Plan 1-23
1 NEDO-32391, Revision C r flfENI.1 REO(AREMENTS F SPECIFY SCENARIO SELECT FROZEN s CODE AND CODE l ) c-S
, 1 NPP
_v_____ PROVIDE COMPLETE DOCUMENTATION C ' TAPD p l uSREDE IDENTIFY AND RANK l DE LO ME SE MENT PHENOWNA (PIRT) MODEL & CORRELATIONS DE I l l L__l___ _ .
! V l MCG MODEL l O6 U
"Ilo"DjNE APPUCABiUTY l REPORT, ,
NEDE.32176, Rev. I l ELEMENL2 ASSESSMENT l ASSES $$ENT AND RANGWG OF PARAMETERS _ _ _ _ _ _-_g_ uATRLx _) l DERNE ZATON CALCULATONS l I Y l V Y I TS U NP Y SEP E, ECTS INTEG EMS 4 3 g DOCUMENT DOCUMENT l 1 I a go YES I CHANGE NO l BLAS AND DETERMINE CODE l l UNCERTAINTY h AND EXPERIMENT ACCURACY TRACG QUALIFICATION l t REPORT l NEDE 32177 + SUPPLEMENT BIAS AND UNCERTAINTY w s DETERMINE EF CT l I I____ ______. .__________j _ _ _ - _ _ ___ M- ,
! Y UIETAl
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t i I NPneun% .- CALCULATIONS l fif E NIJ w 7 COMBINE BIASES A>O UNCERTAINTIES r - - - - IlNThD%N Uh4TATION IN TRACG l SENSmVITY AND DATA BASE, CODE ETC.l UNCERTAWTY l ANALYSAS Y _ _ _ _ _ ll MPUGTION TO SBhT, TOTAL UNCERTAINTY gNEDE 32178, Rev.1 TO CALCULATE I '#fSdiC?FICT? I i._________________ Figure 1.4-3 Road Map of SBWR TRACG Related Documentation 1-24
NEDO-32391, Revision C 2.0 IDENTIFICATION OF IMPORTANT THERMAL-IIYDRAULIC PHENOMENA: TOP-DOWN PROCESS 2.1 Introduction As explained in Section 1.3 and illustrated in Figure 1.3-1, the process of defining test and analysis needs for analysis of SBWR transient and accident performance is based on developing a thorough understanding of the key phenomena to be simulated and modeled. This is done in this report in two ways: (1) a Top-Down process based on analyses and sensitivity studies, and (2) a Bottom-Up process based on examination of individual design features. The Top-Down process identifies phenomena and their imponance based on how the overall system behaves; the Bottom Up process, by component and subsystem requirements. This section discusses the Top-Down approach, leading to Phenomena Identification and Rankmg Tables (PIRT). Chapter 3 discusses the Bottom-Up process. They are merged in Section 4. The PIRT is a summary of analytical modeling needs for a physical system (in this case, the SBWR). The principal feature of the PIRT is an assessment of the "imponance" of each modeling need by interdisciplinary teams of experts. The approach used in the SBWR follows the methodology of Boyack, et al. [6]. TRACG calculations established the scenarios of various events (LOCA, anticipated transients, ATWS and stability). These are described in Section 2.2. The descriptions stress the phenomenological evolution of the transients. A detailed description of the sequence of events can be found in the SSAR [3]. (It is noted that, due to modeling and design changes since SSAR submittal, the event sequences have been updated somewhat from the SS/a versions.) The analyses were then reviewed by interdisciplinary teams to identify each thermal- l hydraulic phenomenon that plays a role in the analysis, and to rank all of them in terms of :
"importance"; that is, degree of influence on some figure of merit (e.g., two-phase level inside l the shroud, containment pressure). NEDC-32391P Supplement I has a more detailed description
]
of the PIRT parameters and the rationale for their relative importance. Both Section 2.3 and i I
, NEDC-32391P Supplement I refer back to the scenarios developed in Section 2.2. The organization structure of these sections is summarized below:
Event Categories Analysis of Events PIRT Summary PIRT Detailed Description
~~
LOCA/ECCS Section 2.2.1.1 Section 2.3.1.1 Sup1.1 Sec. S1.3.1 LOCA/ Containment Section 2.2.1.2 Section 2.3..l.2 Supl.1 Sec. S1.3.2 Transients Section 2.2.2 Section 2.3.2 Supl.1 Sec. S1.3.3 j ATWS Section 2.2.3 Section 2.3.3 Supl.1 Sec. S1.3.4 l 4 Stability Section 2.2.4 Section 2.3.4 Supl.1 Sec. S1.3.5 l 2-1
~ NEDO-32391, Revision C 2.2 Analysis of Events 2.2.1 Loss Of-Coolant Accident (LOCA) Chapter 6 of the SSAR includes the entire matrix of calculations for postulated pipe rupture locations and single failures. For a complete PIRT evaluation, the entire spectrum of events must be covered, including analyses with less limiting conditions than the design-basis cue with ' no auxiliary power. The approach followed in this study is to focus initially on the design basis cases, in terms of the equipment and systems available. This leads to the most severe consequences and the greatest challenges to the analytical models in modeling the phenomena. , j The next step was to examine the possible interactions with other systems that might be i available, even though they are not classified as engineered safeguard features for the' event. To 4 facilitate understanding, a large break in the Gravity-Driven Cooling System (GDCS) line has 4 been chosen to illustrate the sequence of events during the LOCA. The sequence of events is
- similar for all the LOCA events, particularly after initiation of the GDCS flows, when the vessel and containment transients cre closely coupled. While there are some differences in the i assumptions made for analysis of the different breaks, these are not very important in determining the phenomenological progression of the LOCA or the importance of various parameters. The limiting LOCA from the perspective of margin to core uncovery is the GDCS line break; from the viewpoint of containment pressure, it is the large steamline break. A i
schematic of the SBWR's passive safety systems is shown in Figure 2.2-1. The overall LOCA sequence can be divided into three periods: blowdown period, GDCS period and the long-term cooling PCCS period. These periods are shown in Figure 2.2-2. The l Blowdown period is characterized by a rapid depressurization of the vessel through the break, i j safety relief valves (SRVs) and depressurization valves (DPVs). The steam blowdown from the j break and DPVs pressurizes the drywell, clearing the main containment vents and the PCCS 1 vents. First, noncondensible gas and then steam flows through the vents and into the suppression pool. The steam is condensed in the pool and the noncondensible gas collects in the wetwell air space above the pool. At about 500 seconds, the pressure difference between the vessel and the drywell is small enough to enable flow from the GDCS pools to enter the vessel. This marks the beginning of the GDCS period, during which the GDCS pools drain their inventory. Depending on the break, the pools are drained in between 2000 and 7000 seconds. The GDCS flow fills the l vessel to the elevation of the break, after which the excess GDCS flow spills over into the drywell. The GDCS period is characterized by condensation of steam in the vessel and drywell, depressurization of the vessel and drywell and possible openings of the vacuum breakers which ,
- retums noncondensible gas from the wetwell airspace to the drywell. The decay heat eventually overcomes the subcooling in the GDCS water added to the vessel and boiloff resumes. The drywell pressure rises until flow is reestablished through the PCCS. This marks the beginning of l the Long-term PCCS cooling period. During this period, the noncondensible gas that entered the drywell through the vacuum breakers is recycled back into the wetwell. Condensation of the
.boiloff steam in the PCCS is recycled back into the vessel through the GDCS pool. The most important pan of the LOCA transient for vessel response is the blowdown period and the early l part of the GDCS period when the vessel is reflooded and inventory restored. For some breaks, l 2-2
l
- NEDO-32391, Revision C l the equalization line from the suppression pool to the reactor vessel may open during the long-
! term cooling period to provide the vessel an additional source of makeup water. l J ;
1
! 2.2.1.1 Primary System Respon e for the GDCS Line Break i
! 1 The GDCS line break scenario is a double ended guillotine break of a GDCS drain line. l There are three GDCS pools in the SBWR containment, each with its own drain line from the ; j- pool to the vessel. Each drain divides into two branches before entering into the pressure vessel. ;
- Each branch has a check valve followed by a squib operated injection vabe and finally a nozzle
! in the vessel wall to control the blowdown flow in case of a break. The check valve prevents
!- backflow from the vessel to the pool. The GDCS break is assumed to occur in one branch, i between the squib operated valve and the nozzle entering the vessel. Additional assumptions for ! j the LOCA analysis include a simultaneous loss of anxiliary power and no credit for the on-site : diesel generators. He only AC power assumed available is that from battery powered inverters. ! i
- Blowdawn Period- At break initiation, the assumed simultaneous loss of power trips ,
i the generator, causing the turbine bypass valves to open and the reactor to scram. The l
- bypass valves close after 6 seconds. No credit is taken for this scram or the heat sink l
- provided by the bypass. The power loss also causes a feedwater coastdown. Dryweil ;
l cooling is lost and the control rod drive (CRD) pumps trip. The blowdown flow quickly ; increases the drywell pressure to the scram setpoint, although no credit is taken for this i safety function. i High drywell pressure isolates several other functions, including the Containment - ; ! Atmosphere Control System (CACS) purge and vent, Fuel and Auxiliary Pool Cooling j
- System (FAPCS), high and low conductivity sumps, fission product sampling, and l reactor building Heating, Ventilating and Air Conditioning (HVAC) exhaust.
i Loss of feedwater and flow out the break caus.: the measured water level in the ! j downcomer to drop past the Level 3 (L3) scrnin setpoint. The " measured" or " sensed" ; i downcomer level corresponds to the static head in the downcomer above the lower j l instrument tap used for the wide range level instrument. This setpoint is assumed to ! scram the reactor. The scram will temporarily increase the rate of measured downcomer , level drop and the Level 2 (L2) trip will quickly follow the L3 trip. This trip will isolate !
. the steamlines and open the isolation condenser (IC) drain valves, but no credit is taken j l in the safety analysis for heat removal by the IC. After L2, the rate of decrease in the i
!- downcomer sensed level will slow and, witheat extemal makeup, the Level 1 (L1) trip l will be reached, but not for several minutes. During this delay, the IC, if available, , would be removing energy and reducing pressure and break flow. After a 10-second ; delay to confirm the L1 condition, the Automatic Depressurization System (ADS) logic l l will start a timed sequential opening of depressurization and injection valves. Four SRVs , { (two on each steamline) open first. The remaining four SRVs open 10 seconds later to ! i l stagger SRV line clearing loads in the suppression pool and minimize downcomer level j swell. Similarly, opening of the depressurization valves (DPVs) is delayed 45 seconds. t Two DPVs on the main steamlines open first, followed in 45 seconds by two additional j DPVs. He remaining two DPVs open after an additional 45 seconds. Ten seconds after j l the last DPV opens, the six GDCS injection valves are opened. When the GDCS ! i ! l~ 2-3 l
. - - - - ..n, - .- . -- -- - - . -.- _ _ .- -_ _ __ __ _ _ _ _ _
\
NEDO-32391, Revision C injection valves first open, the hydrostatic head from the pool is not sufficient to open the check valves 'and GDCS flow does not begin immediately. When the GDCS check valves do_ open, the cold GDCS water funher depressurizes the vessel. Blowdown through the break and the SRVs'and DPVs causes a level swell in the downcomer and
- chimney, which collapses at the end of the blowdown period, with the GDCS injection.
- GDCS Period- The GDCS flow begins refilling the vessel and the downcomer two-phase level rises. When the two-phase level reaches the break, the GDCS flow spills back into the drywell. For the GDCS break, the flow of GDCS water is sufficient to raise the downcomer two-phase level above the break, until the pools empty, then the level drains back to the break elevation. Inside the core shroud, the two-phase level in ~
the chimney also decreases after depressurization, but is restored after the GDCS refills the vessel. Figure 2.2-3 shows the chimney two-phase level during the first 25 minutes of the transient. The two-phase level swell during the initial blowdown and opening of the SRVs and DPVs is not shown in the figure (note the level drop and then rise during the GDCS period as the vessel is refilled). 2.2.1.2 Containment Response for the GDCS Line Break Containment response calculations assume loss of all AC power except that available from battery powered inverters, reactor power at 102% of rated power and no credit for IC operation. The single failum used is the failure to open a check valve in one of the GDCS pool drain lines. Initial conditions are containment normal operating pressure and temperature, with the suppression pool at its maximum allowable operating temperature.
- Blowdows Period - The blowdown for the GDCS line break occurs from the vessel side of the broken line. Simultaneously, the pool side of the broken line drains the inventory of the one affected GDCS pool into the containment. The check valve keeps the vessel from blowing down through the unbroken branch of the GDCS line. As noted l earlier, the break flow is initially a liquid blowdown, and after the downcomer two-phase level falls below the GDCS line elevation, the break becomes a vapor blowdown. The l ADS, activated by the measured downcomer level, opens the SRVs and the DPVs. He flashing liquid (and later, steam) entering the drywell increases its pressure, opening the main contamment vents and sweeping most of the drywell noncondensible gas through ,
the main vents, the suppression pool and into the wetwell airspace. He steam flow through the vents is condensed in the suppression pool. During the blowdown phase of the transient, the majority of the blowdown energy is transferred into the suppression pool through the main vents. Within the pool, temperature stratification occurs, with the blowdown energy being absorbed primarily in the region above the open vents. The increase in drywell pressure establishes flow through the PCCS, which also absorbs part of the blowdown energy. For the GDCS break, this period of the accident lasts less than 10 minutes. The peak containment pressure in the shon term is primarily set by the compression of the noncondensibles initially in the drywell into the wetwell vapor space. The controlling parameters are the ratio of the drywell to wetwell vapor volumes, and the temperature at the top of the suppression pool, which sets the steam partial pressure. l- 2-4
Y l I
! NEDO-32391, Revision C l i !
j
- GDCS Period- Once the vessel pressure drops below the setpoint of the check valves i I
in the two unbroken GDCS lines, the GDCS pools begin to empty their inventory into the 1 vessel. The subcooled GDCS water quenches the core voids, stopping the steam flow !
; l from the vessel. He GDCS flow refills the vessel to the elevation of the break and then ;
i spills over into the drywell. Spillover from the break into the drywell begins at about 20 1 minutes into the accident and continues throughout the GDCS period of the accident. [ ]( Once the GDCS flow begins, the drywell pressure peaks and begins to decrease. The i j ,, decrease in drywell pressure stops the steam flow through the PCCS and main vents. l l The drop in drywell pressure is sufficient to open the vacuum breakers between the - i drywell and the wetwell airspace several times. - Once the GDCS flow begins to spill ;
- from the vessel into the drywell, the drywell pressure drops further and additional l vacuum breaker openings occur. Some of the noncondensible gas in the wetwell airspace j 7
j 'is returned to the drywell through the vacuum breakers. The GDCS period of the ;
- transient continues until the GDCS pools empty and the decay heat is able to overcome ;
i the subcooling of the GDCS inventory in the vessel. Then, the drywell pressure rises and t ! flow is re-established through the PCCS. The PCCS heat removal capacity, even while i recycling noncondensible Eas back to the wetwell, is sufficient to handle the steam l ] generated by decay heat, and the main vents are not reopened. Any uncondensed steam j j condenses and deposits its latent heat in the portion of the suppression pool above the j ! outlet of the PCCS vent. His period of the accident is expected to last approximately 3 j j hours for the GDCS line break, i l
- Long-Term PCCS Period- After the drywell pressure transient initiated by the GDCS l l flow is over, the drywell pressure settles out, slightly above the wetwell airspace l' j pressure. A drywell-to-wetwell pressure difference is established which is sufficient to i open the PCCS vent and drive the steam generated by decay heat through the PCCS. The
- drywell pressure and temperature during the first 12 hours of the GDCS line break j transient are shown in Figure 2.2-4. The drywell pressure rises rapidly during the l blowdown period, decrea.ces at GDCS initiation, drops as the GDCS spills into the j drywell and finally levels off as boiloff resumes. The temperature shown is for a node
, high in the drywell. At this location, the temperature rises during blowdown, then actually superheats during the GDCS period, but levels off as flow to the PCCS resumes. l In lower regions of the drywell, affected by GDCS spill, the temperature may drop l ,, during the GDCS period. Figure 2.2-5 shows the PCCS power during the first 12 hours j of the transient. Also shown is the decay heat. During the blowdown period, the PCCS picks up part of the energy released during the blowdown, most of which is deposited in j, the' suppression pool. During the GDCS period, steam flow to the PCCS stops and the PCCS power drops to zero. As soon as the decay heat can overcome the GDCS l subcooling, boiloff and steam flow to the PCCS resumes and by between 3 and 4 hours, i the PCCS power increases back to nearly equal to the decay heat power. l By way of comparison, the drywell pressure at the beginning of the long-term period for the GDCS line break is below the drywell pressure for the large steamline break. During the 72 hours which defines the long-term cooling period, the drywell pressure remains below the large steamline break pressure. ' As with other breaks, the drywell pressure established at the end of the GDCS period defines the contamment behavior during the long-term cooling period. b l 2-5 l
. ~. -
NEDO-32391, Revision C For this panicular break, depending on which GDCS line is broken, the downcomer level may slowly drop during the long-term cooling period because part of the inventory that is boiled off and condensed in the PCCS may be returned to the GDCS pool with the break. This part of the PCCS flow will drain into the lower drywell instead of returning to the vessel. To avoid uncovering the core, an equalization line between the vessel and suppression pool is designed to
. l open before the downcomer water level can drop below one meter above the top of the core.
This ensures sufficient liquid inventory to keep the core covered, even if the boiloff continues. l For some breaks, the water level in the lower drywell may rise enough to reach the spillover . holes in the main vents. Inventory added to the lower drywell past this point is returned to the suppression pool and back to the vessel through the equalization line. Analysis of the GDCS break indicates that for this break, the drywell water les el will not reach the spillover holes. - During this final period of the transient, drywell pressure will rise slowly. This results from a slow increase in the wetwell airspace pressure, due to the assumed leakage flow between the drywell and wetwell airspace and conduction acrou the wall separating the drywell and wetwell. This energy addition is partially offset by heat losses to the surroundings from the outside wetwell wall. Without the leakage, the containment pressure remains nearly constant during the long-term period of the transient. 2.2.1.3 GDCS Line Break Summary Although the discussion of the GDCS line break has been described in two parts, the primary system and containment response are not independent, particularly after the blowdown period. The sequence of events occurring in the GDCS line break transient is summarized in Table 2.2-1. He occurrences listed as " symptoms" in the first column result in " actions", which are the corresponding entries in the second column. The timing of the symptoms is also shown. For the GDCS break, the reactor core does not uncover, so there is no cladding heatup above saturation temperature of the coolant. In evaluating the "importance" of various phenomena in the PIRT process, the phenomena associated with cladding heatup (e.g., radiation heat transfer, metal-water reaction) are comparatively unimportant, while phenomena associated l with the two-phase level inside the core shroud (e.g., decay heat, energy release from heat slabs) are comparatively important. For the containment, after tiie blowdown and release of energy to the suppression pool, the effectiveness of the PCCS controls the containment response, with no , l pumped pressure and decay-heat removal temperature increase system slowly until theavailable. In theperiod, end of the 72-hour long-term at whichcooling time period credit for non-safety decay-heat removal systems is permitted. Thus, containment pressure and ,. temperature become the primary figures of merit for the containment and the phenomena affecting them are important. The LOCA scenario develops slowly for the SBWR. The accident detection system logic functions almost instantaneously, but thereafter, the time scales are measured in hours rather l than seconds. The chimney two-phase level (Figure 2.2-3) dips briefly about 10 minutes into the LOCA due to void collapse following GDCS injection. For the GDCS line break, the minimum chimney level occurs at about 7 hours after the break. At this point in time, the core void fraction is very small, and the chimney and downcomer levels are almost the same. This slow response, which is due to the large volume of water in the reactor vessel and GDCS pools, l 2-6 i -
= - . . - . - . -. ._ -- - . - _.
I NEDO-32391, Revision C makes the LOCA a very slow moving event from the reactor systems and operator response standpoint. Similarly, containment response (Figure 2.2-4) is gradual, not reaching the design pressure even 72 hours after the break. This slow response permits well-considered, deliberate
- operator actions.
I 2.2.1.4 Main Steamline Break i- In this subsection, the important features of the transient resulting from a large break in the main steamline are described. The emphasis is on those features that are different from the
< GDCS line break scenario.
- Blowdown Period - At break initiation, the blowdown flow quickly increases the drywell pressure to the scram setpoint, and a control rod scram occurs. The high velocities in the steamline initiate closure of the Main Steamline Isolation Valves (MSIVs) and the reactor isolates in 3 - 5 seconds. This trip also opens the Isolation Condenser (IC) drain valves, but no credit is taken in the safety analysis for heat removal j by the IC. High drywell pressure isolates several other systems, including the j Containment Atmosphere Control System (CACS) purge and vent, Fuel and Auxiliary Pool Cooling System (FAPCS), high and low conductivity sumps, fission product sampling, and reactor building Heating, Ventilating and Air Conditioning (HVAC) exhaust.
Loss of feedwater and flow from the break cause the vessel water level to drop. Without i external makeup, the Level 1 (L1) trip will be reached in about 6 minutes. During this period, the IC, if available, would be removing energy and reducing pressure and break t
- flow. After a 10-second delay to confirm the L1 condition, the Automatic Depressurization System (ADS) logic starts a timed sequential opening of depressurization and injection valves. Two SRVs on the unbroken steamline open first.
- The remaining two SRVs open 10 seconds later to stagger SRV line clearing loads in the suppression pool and to minimize vessel level swell. The sequence of opening of the DPVs and the GDCS injection valves is similar to that for the GDCS line break described earlier. However, because of the large steam break, the vessel depressurizes faster and GDCS injection begins earlier, at about 500 seconds versus 600 seconds for the GDCS line break. Blowdown through the break, the SRVs, and the DPVs causes a level swell l in the vessel. The two-phase level in the downcomer decreases at the end of the blowdown period, when GDCS injection begins.
>- In the containment, the steam entering the drywell increases its pressure, opening the , main containment vents and sweeping most c' the drywell noncondensible gas through the main vents, through the suppression pool, and into the wetwell airspace. During the blowdown phase of the transient, the majority of the blowdown energy is transferred into ] the suppression pool by condensation of the steam flowing through the main vents. The increase in drywell pressure causes flow through the PCCS, which also absorbs part of j j l the blowdown energy. The ADS, activated by the measured downcomer level, opens the SRVs and the DPVs and augments the steam flow to the suppression pool and drywell, respectively. This period of the accident lasts less than 10 minutes.
- l 2-7 l
1 4
NE00-32391, Revision C l
- GDCS Period - De GDCS flow begins refilling the vessel and the downcomer two-phase level rises. When the two-phase level reaches the elevation of the open DPVs, the GDCS flow spills back into the drywell. Inside the core shroud, the two-phase level in the chimney also decre:ses after depressurization, but is restored after the GDCS refills l the vessel. The minimum two-phase level in the chimney is of the order of 3-4 m above the top of the core; there is substantial margin to core heatup.
- Quenching of voids in the core by the GDCS flow reduces the steam outflow from the vessel to the drywell. Once the GDCS flow begins, the drywell pressure peaks and
~
begins to decrease. The decrease in drywell pressure stops the steam flow through the PCCS and main vents. This pressure decrease may be sufficient to open the vacuum ~ breakers between the drywell and the wetwell airspace. Once GDCS flow begins to spill from the vessel into the drywell, the drywell pressure drops further and additional vacuum breakers may open. If the vacuum breakers open, some of the noncondensible gas in the wetwell airspace will retum to the drywell through the vacuum breakers. The GDCS period of the transient continues until the water level in the GDCS pools equalizes with the collapsed level in the downcomer of the reactor pressure vessel and the decay heat is able to overcame the subcooling of the GDCS inventory in the vessel. Then, the drywell pressure rises and flow is re-established through the PCCS. The PCCS heat removal capacity, even while recycling noncondensible gas back to the wetwell, is l sufficient to transfer the steam generated by decay heat without reopening the main vents. This period of the accident is expected to last for less than one hour.
- Long-Term PCCS Period- After the drywell pressure transient initiated by the GDCS flow is over, the drywell pressure settles out, slightly above the wetwell airspace pressure. The Main Steamline break is the limiting break in terms of containment pressure and temperature. This part of the containment transient is similar to that for the GDCS line break. However, unlike the GDCS line break, the steam generated by the decay heat is condensed and all of it is returned to the vessel through the GDCS lines.
l Thus, there is no long term drop in the downcomer and chimney water level due to boiloff. A larger amount of water inventory is retained inside the vessel and a smaller amount in the lower drywell. 2.2.1.5 Small Breaku .- The thermal hydraulic phenomena which characterize the small breaks in the SBWR are very similar to those for the large steamline break. His is because once the downcomer level ,. drops below the Level 1 set point, the reactor is automatically depressurized through the SRVs and DPVs. For small breaks (depending on the size and location), it may take several minutes before the reactor is scrammed on low water level (Level 3), and still longer before the ADS is actuated. For a steamline break having an area equivalent to 2% of the main steamline cross-l sectional area, the measured downcomer water level will boil off to reach Izvel 1 in about one hour. During this period, the break flow exceeds the condensing capacity of the PCCS and results in clearing the top row of horizontal vents. This results in energy addition to the portion of the suppression pool above the top vents,'and increases the pool surface temperatures. The SBWR incorporates an ADS trip on high pool surface temperature to mitigate this effect. l 2-8
NEDO-32391, Revision C 2.2.1.6 Non-Design Basis LOCAs The discussion to this point hu focused on LOCA scenarios with design basis assumptions. With regards to system availability, the primary assumptions were to assume failure in an active system or component and loss of offsite power and diesel generators. The consequences of relaxing these assumptions towards a "best estimate scenario" are examined in this subsection. Single Failures: In the SBWR, the active component failures considered are the failure of a valve in the o GDCS line to open and the failure of a DPV to open. Scenarios without failures have been analyzed. With no failures, design margins are increased. No new thermal-hydraulic phenomena or intemctions are introduced because the differences relate simply to the number of GDCS lines available (quantity of GDCS flow) or the number of DPVs available for depressurization (amount of steam blowdown flow and rate of depressurization). While no new phenomena are introduced, these events do provide a wider range of parameters which is useful for code validation. Tests with both types of single failure and ones without any failure are included in the LOCA simulations performed in the GIST facility. Isolation Condenser Operation: < For LOCA analysis, the IC is not treated as an engineered safety feature and no credit is taken in the safety analysis for its operation. The valve in the condensate return line will l open in a realistic scenario. This increases the vessel liquid inventory before ADS and reduces the steam load on the containment. LOCA scenarios with the IC operational have been included in the consideration of important phenomena in Sections 3 and 4. These phenomena include the IC condensation efficiency, steam quenching in the reactor vessel downcomer, and interactions between the IC steam flow and the steam flow through the DPVs on the same nozzle. Diesel Generators Available: As shown in Table 2.2-2, additional systems become available when the diesel generators start up. Only the Control Rod Drive System in its high pressure injection mode is
, initiated automatically. This system injects water through the feedwater line into the downcomer. Scenarios with the CRD high pressure injection available are considered in Chapter 3 and Section 4.2. The Fuel and Auxiliary Pool Cooling System (FAPCS) will , also be available to the operator with the diesels operational. FAPCS isolates automatically on high drywell pressure. The operator can override the isolation manually. The FAPCS has several modes of operation. It can be aligned to function initially in the Low Pressure Coolant Injection (LPCI) mode. When core cooling is established, the FAPCS can serve as a Suppression Pool cooling system. It can also be used for drywell and wetwell spray.
- Interactions between the FAPCS and the passive safety systems (GDCS/PCCS) are considered in Chapter 3 and analyzed in detail in Section 4.2.
l 2-9
NEDO-32391, Revision C Offsite Power Available: Table 2.2-3 shows that the primary additional water makeup systems available with offsite power are the condensate and feedwater systems. Numerous auxiliary systems such as fuel pool cooling, drywell coolers, and drywell sump drain pumps would also be available. With feedwater and offsite power available, the accident becomes a relatively mild event. l After scram on high drywell pressure, the feedwater maintains normal downcomer water level for an extended period of time even for large breaks. 'Ihis allows the operator to initiate a controlled depressurization of the reactor. The water spilling out of the reactor collects in the lower drywell. For large breaks, the sump dram pumps will not be able to keep up with the break discharge. Eventually, water spills into the wetwell through the ' spillover holes in the pipes connected to the horizontal vents. The feedwater will be throttled back or turned off as the water level rises in the wetwell. 2.2.2 Anticipated Transients As with the LOCA, anticipated transients are discussed in the SSAR (Chapter 15) and results for specific events are not presented in this report. The PIRTs for anticipated transients were synthesized from consideration of the phenomena involved in various classes of events. 2.2.2.1 Fast Pressurization Events These .are the limiting pressurization events. Principal figures of merit on which "imponance" is defined are critical power (MCPR) and reactor pressure. Turbine Trips - initiated by trip of turbine stop valves from full open to full closed. Analyzed with bypass valves functional, and with bypass failure. Generator Load Rejection - initiated by fast closure of turbine control valves from partially open position to full-closed. This event is analyzed with bypass valves functioning, and with bypass failure. The turbine control valves may be initially at the ; same position (full are turbine admission) or at different positions (partial are turbine admission). ; Ioss of AC Power - Similar to load rejection; however, bypass valves are assumed to , i close after 6 seconds due to loss of power to condenser circulating water pumps. ' Main Steamline Isolation Valve (MSIV) Closure - In this case, the scram signal on valve position is further in advance of complete valve closure. This effectively mitigates : the shoner line length to the vessel available as a compression volume. l Ioss of Condenser Vacuum - This event is similar to the Loss of AC Power and a l Turbine Trip with Bypass. Because a turbine trip occurs at a higher vacuum setpoint i than the bypass valve isolation, the bypass valves are available to mitigt.te the initial ; pressure increase. i i l 2-10 i
NEDO-32391, Revision C 2.2.2.2 Slow Pressurization Events These are analyzed principally to ensure that they are bounded by the fast pressurization events. MCPR and reactor pressure determine "importance."
- Pressure Regulator Downscale Failure - Simultaneous closure of all turbine control valves in normal stroke mode. The triplicated fault tolerant control system prevents any single failure from causing this and makes its frequency below the anticipated abnormal
, occurrence category.
- Single Control Valve Closure - This event could be caused by a hydraulic failure in the valve or a failure of the valves rotor / actuator.
2.2.2.3 Decrease in Reactor Coolant Inventory Loss of feedwater flow is characteristic of this category of transient. The IC maintains l downcomer water level. Reactor water level in the downcomer is the principal figure of merit on which "importance" is defined. 2.2.2.4 Decrease in Moderator Temperature These events challenge MCPR and stability, which are the figures of merit on which "importance" is defined:
- Ioss of Feedwater Heating -initiated by isolation or bypass of a feedwater heater.
- Feedwater Controller Failure - hypothesizes an increase in feedwater flow to the maximum possible with all three feedpumps operating at maximum speed. Similar to turbine trip but with more severe power transient due to colder feedwater.
To determine the phenomena important in modeling anticipated transients, the sequence of events and system behavior for each class of events should be understood. To provide an example of this, the sequence of events for a fast pressurization transient is discussed below. For this class of transients,important phenomena are those affecting the MCPR and reactor pressure. 2.2.2.5 Generator Load Rejection Event Description A fast pressurization event will occur due to the fast closure of the turbine control valves (TCVs), which can be initiated whenever electrical grid disturbances occur which result in significant loss of electrical load on the generator. Closure of the turbine stop valves is initiated I by the turbine protection system. The valves are required to close rapidly to prevent excessive overspeed of the turbine-generator rotor. At the same time, the turbine stop or control valves are signaled to close, and the turbine bypass valves are signaled to open in the fast opening mode. The bypass valves are full open only slightly later than the turbine valves are closed, and can relieve more than one-third of rated steam flow to the condenser, greatly mitigating the transient. The bypass valves also use a triplicated digital controller. No single failure can cause all turbine bypass valves to fail to open l 2-11 l 2
NEDO-32391, Revision C on demand. De worst single failure can only cause one turbine bypass valve to fail to open on demand. l
. The closing time of the TCVs is short relative to the sonic transit time of the steamline, so j their closure sets up a pressure wave in the steamlines. When the pressure wave reaches the vessel steam dome, the flow rate leaving the vessel effectively undergoes a step change. The i area change entering the steam dome partially attenuates the pressure wave, propagating a !
weaker pressure disturbance down through the chimney and downcomer, increasing the vessel pressure, and mducing voids in the core. The void-reactivity feedback results in an increase in the neutron flux. A reflection of the pressure wave also travels back toward the turbine, , producing an oscillation in flow and pressure in the steamlines. , j Concurrent with closure of the turbine control valves, a scram condition is sensed by the i reactor protection system. A turbine stop valve position less than approximately full open triggers a scram, as does the low hydraulic fluid pressure in the turbine control valve solenoids which start their fast closure mode. The SBWR digital multiplexed Safety System Logic : Control (SSLC) will initiate a scram when any two turbine stop valves are sensed as closing, or , any two turbine control valves are sensed as fast closing. The; core reactivity is decreased by the control blade insertion and increased by the decrease in core voids and increase in inlet flow. The net effect may be either an immediate shutdown of ; the reactor and decrease in neutron flux (in cases where there are control blades partially inserted in high worth areas of the core) or a short period of increased reactivity and neutron flux followed by shutdown (in the safety analysis case where there are no control blades initially ; inserted, and a slower bounding CRD scram insertion time is assumed.) : In the case where the neutron flux undergoes a transient increase, the energy deposition in f the fuel pellet will increase clad heat flux. The minimum value of critical power ratio durmg : this transient is found to occur in the upper part of the bundle. Eventually, as the blades are fully inserted, the reactor is driven subcritical, power drops to ; decay heat levels, and clad temperature equilibrates near saturation temperature. He vessel pressure increase is terminated by the bypass valve opening. The downcomer l water level drops below the feedwater sparger and sprays subcooled water into the steam dome. This quenching of vapor also helps to terminate the pressure increase. If the bypass and feedwater systems are assumed to be unavailable, the duration of increased pressure would be .- < long enough to initiate the isolation condenser. < In the ASME overpressure protection analysis, the Isolation Condenser is not considered, , causing the pressure to slowly increase to the SRV opening pressure. The pressure increase is terminated immediately with SRV activation, and the maximum vessel pressure occurs at the vessel bottom. The overpressure protection case conservatively assumes the first scram signal to fail, and scram on neutron flux terminates the power increase in both turbine valve closure and the MSIV closure events. : The downcomer water level response in pressurization events is driven by the transfer of ! water from the downcomer to core and chimney caused by the collapse of voids in the core and l chimney regions. He sensed water level decreases rapidly below the L3 low water scram l setpoint. The feedwater system flow increases fast enough to prevent the L2 setpoint being i l 2-12
NEDO-32391, Revision C l reached in high frequency events (events where feedwater and bypass valves are available). The feedwater control system will demand maximum feedwater flow for approximately one minute, l until normal downcomer water level is restored. Without feedwater, the downcomer level drop will progress to L2, initiating the IC, isolating the MSIVs and transferring the CRD system to i high pressure injection mode. The IC can independently maintain the downcomer water level > near the L2 setpoint. CRD high pressure injection will cause the downcomer water level to
- slowly recover to above normal, and then automatically trip off.
4 1 2.2.3 Antidpated Transients Without Scram (ATWS) r , The most limiting ATWS event in terms of reactor vessel pressure, heat flux, neutron flux, ; ! peak cladding temperature, suppression pool temperature and containment pressure is the t inadvertent closure of all main steamline isolation valves with failure of rod insertion. This j event is described in Section 15.8 of the SSAR. It is the only ATWS event considered in j determining the phenomena needs for qualification of TRACG. l l j 4 2.2.3.1 MSIV Closure Transient f The incident is initiated by the inadvertent closure of all MSIVs which isolates the reactor ! vessel. If the control rod scram fails, the rapid increase of vessel pressure together with the l APRM not-downscale signal will generate an ATWS signal. This signal initiates the feedwater i runback and activates Alternative Rod Injection (ARI), FMCRD run-in and the boron injection j timer. If the alternate rod insertion fails, the squib valves in the standby liquid control system j (SLCS) will blow open after the boron injection timer runs out (180 seconds). The sodium j pentaborate solution is then released into the core bypass region. The highly enriched sodium ! pentaborate solution quickly mixes with the core coolant and achieves hot shutdown in less than ! 60 seconds. , ! The reactor pressure is limited to 10.2 MPa by the discharge of steam to the suppression pools through the safety valves as shown in Figure 2.2-6 A. The reactor thermal power (fuel rod j surface heat flux) peaks at about 7 seconds, and is reduced by the feedwater runback. Power reduction and safety valve discharge cause the vessel pressure to drop. At about 100 seconds, all the safety valves reclose. Four of the safety valves will open and close for another 9 cycles until i' the reactor achieves hot shutdown. The steam generated by the decay heat is then removed by
- the IC alone (capacity 4.5% of rated power), and no more steam is discharged into the suppression pool. This limits the suppression pool temperature to 329 K with an associated l ,, containment pressure of 0.121 MPa.
The downcomer water level decreases rapidly after the initiation of feedwater runback as shown in Figure 2.2-6 B. This in turn reduces the core flow rate and the reactor power as shown in Figure 2.2-6 C. The downcomer water level keeps decreasing since the water inventory is
, removed through steam discharge into the suppression pool. The only make-up water is from j the CRD flow which starts at about 80 seconds and accounts for only about 2% of the rated 1
feedwater flow. The water downcomer level finally starts to recover when the reactor reaches j hot shutdown. The steam will recycle through IC and the CRD flow provides the extra inventory 1 as shown in Figure 2.2-6 D. 1
! l 2-13
l NEDO-32391, Revision C The rapid pressurization of the reactor vessel collapses the core voids which results in a neutron flux surge to 330% of rated condition at 2 seconds. The increase of fuel temperature and core void limit the maximum value of the core power. The reactor is finally brought down by the reduced core flow and increased voids as shown in Figure 2.2-6 C. The thermal power of the reactor reaches its peak of 140% at about 19 seconds. The power then decreases with the reduction of the core flow and settles at about 25% of rated condition as the core flow is running at about 4% rated. This reduced core flow is the result of lower downcomer water level and reactor power which provides the buoyancy to drive the core flow. After attaining hot shutdown , from boron injection, the power follows the decay heat generation rate. The initial surge of power creates condidons for the boiling transition in some of the hot rods at the high powered channels. The peak cladding temperature reaches 408 C at 21 seconds - which is well below the design limits of 1200 C and fuel integrity is not compromised. 2.2.4 Stability Because the SBWR core flow is driven by natural circulation, the most limiting stability condition is at the rated power / flow condition. This is unlike operating forced-circulation BWRs, and it simplifies the stability analysis for the SBWR. For the SBWR, a stability criterion is used which is very conservative compared to l operating plants (Figure 2.2-7). The core decay ratio is maintained less than 0.4 and the channel decay ratio less than 0.3. The stability performance of the SBWR is evaluated at various conditions. 2.2.4.1 For Steady State Operation During steady state operation, the highest power / flow ratio occurs at 104.2% power and 100% flow condinons. The decay ratio is well within the conservative design criteria (Figure l 2.2-7). At reduced power level, the power / flow ratio is lower, so the decay ratios for both core j and hot channel are lower than at the rated condition. This conclusion is supported by Dodewaard test data as shown in the figure. The decay ratios during normal operation at , Dodewaard have been very low, with no indication of any incipient instability throughout its l long operating history. In Figure 2.2-8, the power / flow map of SBWR normal operation is , compared with the stability limit calculated in the Oak Ridge National Laboratory (ORNL) ] study. The results confirm that there is large margin for stability. This indicates that the SBWR l is very stable under normal operation conditions. .- l 2.2.4.2 For Anticipated Transients l Of the anticipated transients, the loss of 55.6 C (100 F) feedwater heating case gives the highest power / flow ratio. Loss of feedwater flow is another limiting event. However, the scram quickly mitigates the transient and the power conditions are reduced to hot shutdowr.. For both events, the decay ratios for core and hot channel meet the design criteria shown in Figure 2.2-7. In Figure 2.2-8, both of these transient events are seen to result in power / flow conditions that are well below the exclusion region. l 2-14
NEDO-32391, Revision C 2.2.4.3 For ATWS Conditions During ATWS conditions, the persistent high reactor power poses the most challenge to the stability criteria. However, feedwater mnback reduces the core power, and the SBWR's low i l power though thedensity also helps reduced downcomer waterto alleviate level the severity effectively decreases the coreof the flow ratechallenge and increases to the stab! the power / flow ratio to a higher value than those for the steady state and anticipated transient . conditions, the analysis of performance in the ATWS study indicates the reactor remains stable l and no power oscillation is predicted. Following the feedwater runback, both flow and power decrease, resulting in a more favorable power / flow ratio. The injection of boron will eventually shut down the reactor and terminate the transient. 2.2.4.4 For Startup During startup, there is a special concern that is not present at power. At very low flows, a periodic "geysering" flow oscillation can be postulated to occur caused by either of two t mechanisms. First, condensation of core exit vapor in the subcooled chimney region and the top of the core might cause a reduced pressure in the channels and a resultant flow reversal in the core. Oscillations of this kind are unlikely given the SBWR startup procedures, which are similar to those of the Dodewaard reactor (Dodewaard has experienced no "geysering" oscillation in its 22 refuel cycles of operation). ' Second, vapor production in the lower-hydrostatic-head chimney region could cause a reduction of hydrostatic head and a resultant core ; flow increase. This, in tum, could cause voids to collapse in the chimney, leading to a reduction in flow. Oscillations of this second kind have also never been seen at Dodewaard. Dhey were : to occur, they would be mild oscillations with little, if any, reactivity impact. i i : 1 ; i 1 l *, i i I i l 2-15
NEDO-32391, Revision C Table 2.2-1 GDCS Line Break Sequence of Events Symptom Action (s) Time (hr) Loss of offsite power Instantaneous GDCS line break. Generator trips, bypass 0. valves open and reactor scrams. Bypass valves close after 6 seconds. No credit for this scram or the bypass beat sink is taken in the SSAR Chapter 6 analysis Feedwater coastdown (diesel generators fail to start) Fuel pool cooling lost DW coolers lost CRD pumps trip High drywell pressure Scram Ino credit taken) 0.01 (Note 1) CACS (Cont. Atm. Control Sys) purge & vent isolates FAPCS (Fuel and Aux. Pool Cooling Sys.) isolation PCC condensation begins PCC pool boiloff begins, HX tubes remain covered
>72 hr Isolate high and low conductivity sumps, fission product sampling. reactor building HVAC exhaust Low water level L3 Scram 0.01 (Note 1)
Low waterlevel L2 1C drain valve opens (MSIV closure also initiates) 0.01 (Note 1) Isolate high and low conductivity sumps, fission product sampling, reactor building HVAC exhaust DW coolers isolate Low water level L1 ADS /GDCS initiation. Timed sequential opening of: 4 0.1 SRVs/4 SRVs/2 DPVs/2 DPV's/2 DPV's/6 GDC injection valves DW coolers isolate Same equipment which isolated on L2 receives redundant isolation signal. P < GDC pool head injection flow begins 0.2 Post LOCA radiolytic H2 and PARS (Passive Autocatalytic Recombiners) function. 0.2 (Note 2) O2 (PARS are not simulated in fuel peak temperature and minimum water level calculations) P dw < P ww - 0.5 psi Vacuum breakers open 0.3 GDCS pool empties DW pressure stabilized 2.4 DW-WW Ap initiates PCCS flow .- PCCS condensate returns to GDCS pool, drains to vessel and DW Reactor water level falls to Vessel to S/P equalization line opens, keeps core 6.6 .- one meter above top of core covered Liquid in DW reaches spillover Inventory added to DW now retums to S/P (then to 9.3 (Note 3) holes in main vents vessel) ___ Design-basis leakage and sen- Pressure rises slowly for 72 hours (defined as end of to 72 sible heat transfer from DW to design basis) WW causes gradualincrease of DW pressure l 2-16
4 NEDO-32391, Revision C Notes To Table 2.21: (1) Scram on high drywell pressure and level decrease to L2 occur within one minute of the line break. j (2) PARS will actuate as soon as they are exposed to radiolytic hydrogen, estimated to occur within a few ; minutes of the line break. ! (3) Increase of DW level to the spillover holes only occurs if it is assumed that inward flow through the l break cannot occur. Otherwise, the inventory spilled to the DW returns to the RPV through the break. O I I 9 l 2-17
NEDO-32391, Revision C Table 2.2-2 LOCA Scenario with Diesel Generators Available - Additional Systems Functional Symptom Action (s) Loss of normal AC Diesel Generator starts FMCRD mn-in backs up hydraulic scram Low water level L2 CRD initiates in high pressure injection mode Above actions arc antomatic, no operator action necessary. , Actions below require operatorintervention. Low water level L3 FAPCS LPCI mode, injection through FW system High pool temperature FAPCS Pool cooling mode,if adequate core cooling. Operator action required to over-ride system isolation. P cont > 14.2 psig FAPCS DW and WW spray T dw > ADS qualification FAPCS drywell spray temperature Low water level < L1 per Firewater EPG Containment pressure high DW Cooler or T dw > Tech Spec LCO 1 GDCS Poollevel < NWL - Trip CRD pumps 0.5m (2 of 3 pools) 1 2 days post LOCA Attach PCC vent fan Table 2.2-3 LOCA Scenario with Offsite Power & Diesel Generators Available ,. Symptom Action (s) ,. Low water level L3 FW and condensate injection j Pressure > nonnal setpoint Turbine bypass valves l 4 i l 2-18 l i
NEDO-32391, Revision C ISOLATION CONDENSER SSURNION FOR PASSIVE DECAY MAIN STEAM LINE HEAT REMOVAL I
.!'i
.-. .h l
.., L, . -
l l\ w N PCC CONDENFC4
![l )
<f .
FOR LONG TcRM
" CONTA:dMENT
_- ,f7 : COO'JNG
/
k {, 1 ['J -- N , N GRAVITY-DRIVEN CORE llll COOLING KEEPS CORE COVERED FOR ALL ACCIDENTS J
,=
. . . y EQUALIZER LINE d, :.
...' / .:..-
g
'g.
LARGE WATER POOL FOR POST-ACCIDENT FISSION PRODUCT i n , , RETENTION y 5
.. Figure 2.2-1 SBWR Passive Safety Systems os l 2-19
[ NEDO-32391, Revision C REACTOR r GDCS INITIATION DRM GDCS DRAINED REACTOR AND DRYWELL ,
.M WETWELL WETWELL c- GDCS -
LONG TEAM PCCS j PERIOD PERIOD t si ((
-10 minutes 2 hours // 3 days TIME Figure 2.2-2 Phases of the LOCA Transient 15
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NEDO-32391, Revision C i F
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NEDO-32391, Revision C 150. i 81 i i i i i i 6 8 I i i 6 6 i i i i i l i 6 8 1 NEUTRON FLUX - I 2 SIMULATED THERMAL POWER 100. 3 3 CORE FLOW _ 2 W _ 1 - m o - g - 1 *
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__ _ _ _ ~ . _ _ _ _ _ _ _ _ NEDO-32391, Revision C 2.3 Phenony.na Identification and Ranidag Tables (PIRT) The process of Top-Down analysis and qualification of the performance of the SBWR starts with the identificadon of the important physical phenomena. For this purpose, Phenomena l Identification and Rankmg Tables (PIRT) [6] were developed. This was done by assembling a team of experts knowledgeable about thermal-hydraulics and transient analysis, and obtaining consensus on the relative importance of various phenomena. Phenomena were given a rank between 0 and 9 based on their "imponance" as defined in Section 2.2. The ranking was done on
- a conservative basis, i.e. generally, phenomena were given a higher rank if there was any uncertainty as to its importance. This resulted in a large number of highly ranked phenomena. It is expected that a much smaller subset will actually prove to be "important" after the tests and ,
sensitivity studies are completed. Tables were developed for small break LOCAs, large break LOCAs, pressurization transients, depressurization transients and reactivity inserdon due to cold water injection. Plant startup was also treated as a category of operational transients because of the focus on the potential for geysering. Tables were also developed for ATWS (pressurization events) and for stability during normal operation and transients. In each case, the importance of the phenomena was evaluated for each reactor region: lower plenum, core, upper ; plenum / chimney, downcomer, etc., as well as for the containment. For the LOCA events, the tables were further subdivided into the blowdown, GDCS and long-term periods of the transients. It was apparent that for many transients and subregions, the phenomena of importance are the same as for operating BWRs. As an example, for pressurization transients, the most important parameters are the nuclear parameters (void, Doppler and scram reactivity), the l interfacial shear (void fraction), subcooled boiling and steamline dynamics. While all these phenomena appear in the PIRT, the phenomena that are unique to SBWR are given primary : emphasis in the following sections of this repon. These are primarily factors affecting the PCCS : performance, GDCS interactions and phenomena associated with natural circulation flow in the core. The PIRT tables are used for three purposes. First, the capabilities of the TRACG models are examined to see if all the relevant i phenomena can be treated. For this purpose, an evaluation of TRACG models is made with j refemnce to the PIRT parameters, to ensure that all relevant phenomena are modeled. This has ' been accomplished by verifying that a model with appropriate accuracy exists in TRACG for each phenomenon considered. Secondly, the qualification data base is examined for completeness against the important .- ! Examination of the l phenomena. phenomena rankedThe results
" Medium" of this evaluadon in importance areThe are also included. discussed in Chapter medium ranked phenomena 5.
l will be considered to augment the conservative ranking process adopted by the PIRT team. < l Finally, the PIRT is also used in the CSAU process for the determination of model bias and uncertainties. For this purpose, the phenomena ranked "High" in importance will be ranged and sensitivity studies performed to quantify the effect on an appropriate figure of merit. The results from this study will be documented in the Application of TRACG Model to SBWR Licensing of Safety Analysis (NEDE-32178P, Revision 1). I 2-26
NEDO-32391, Revision C s It is recognized that the PIRT is based on engineering judgment. If the planned tests reveal
] phenomena that were not considered in the development of the PIRT, they will be added to the tables, and their impact on the modeling evaluated. j 1
l 2.3.1 Loss-Of-Coolant Accident (LOCA) l j The overall transient consists of three periods: the blowdown period, the GDCS period and l l= the long-term cooling PCCS period. 1 For each of these periods, the important thermal-hydraulic phenomena were listed and i i, ranked. This was done by experts familiar with BWR and SBWR characteristics and with transient analysis. The group was interdisciplinary, drawn from several technical areas, such as SBWR design, methods development, and plant transient analysis. The phenomena were i classified by reactor and containment region (e.g., lower plenum, core, downcomer, chimney, drywell, wetwell, etc.). Phenomena are ranked separately for small and large breaks. Most of the phenomena and their rankings are similar for small and large breaks. While the front end of the accident progresses more slowly for the small breaks, the rapid depressurization by the ! Automatic Depressurization System on low sensed downcomer water level results in l characteristics similar to a large break. The liquid breaks like the GDCS line break and steamline i break are not shown separately, but the phenomena important to both have been grouped under "Large Breaks". 1 2.3.2 Anticipated Transients i Plant startup and three types of operating transients (pressurization, depressurization, and cold water transients) are evaluated. The importance rankings for various phenomena are tabulated by region. "Importance"is ranked by the influence these phenomena have on the , l Critical Power Ratio (CPR) and maximum pressure reached in the transient. For plant startup, the key criterion is the likelihood of large oscillations in the core flow and power. i The PIRT for transients has been revised. The discussion has been focused on specific transients. Clearer definition of the phenomena and re-evaluation of phenomena has resulted in l some changes to the relative rankings. A description of the phenomena and a rationale for the importance can be found in Supplement I to this report. In the PIRT for transients, the p pressurization transients considered are the turbine trip or load rejection without the opening of the bypass, and the inadvertent closure of the Main Steamline Isolation Valves (MSIVs). The cold water event is Loss of Feedwater Heating; inadvertent actuation of the RWCU/SDS and ICS are much milder in comparison. The Feedwater Controller Failure High is a cold water i event, followed by a turbine trip. It is bounded by the two above event categories. Depressurization transients are less limiting. The most severe event in this category is the Pressure Regulator Failure Downscale, in which all turbine control valves and bypass valves are assumed to fail open. Parameters ofimportance to all these transients can be grouped under the
- following categories
4 Parameters affecting initial operating state. This includes the initial natural circulation flow rate, the flow distribution among channels, power and void distribution, control rod distribution and separator carryunder. 2-27 i
NEDO-32391, Revision C Transient thermal hydraulic /neutronic response, characterized by pressure response, void fraction changes in the core, reactivity feedback from voids, Doppler broadening and scram, and the resulting transient core power and flow response. Margin to boiling transition, which is determined by the fuel rod thermal response and the relationship between the core thermal hydraulic conditions and critical heat flux. After the reactor has been scrammed, there is a long term phase of the transient that involves invemory control and bringing the reactor to a hot standby condition. This process does not involve challenging thermal hydraulic phenomena and is not considered here. This a phase of the transient does not affect the MCPR and peak pressure, which are the figures of merit for these transients. Issues related to the performance of the Isolation Condenser (IC) are considered in Section
- 3. Only the effects of the cold water injection in the downcomer have been included in the PIRT discussions in Section 2. Because of the neutronic coupling, the core region is by far the most important for the operational transients. Steamline dynamics play an important role in determining the pressurization rate and void collapse following a turbine trip or load rejection.
The key parameters for a depressurization event are core void fraction, flashing, and void reactivity feedback. The chimney void fraction and the separator carryunder characteristics are important in determining the initial operating state. l The plant startup transient is not a MCPR limiting transient. The concern here is the l margin to large oscillations in flow at low power. The conditions most likely to produce : oscillations in flow are at the incipience of void generation at the top of the core and in the chimney. The flow in the core is single phase and there are no significant voids in the core. Thus, void reactivity feedback is not important unless the oscillations become very large. The ! plant startup transient is influenced by core inlet subcooling, rate of heatup, subcooled boiling, and the chimney subcooling and void formation. In NEDC-32391P Table 2.3-3, the PIRT parameters for transients are listed by the region of the reactor vessel. Historically, the LOCA tables were developed first. The list of parameters considered for LOCA formed the starting point for the transient PIRT. Thus, there are some parameters that are not relevant for transients, which have been retained in the complete list. The individual items in NEDC-32391P Table 2.3-3 are discussed in NEDC-32391P Supplement 1. A total of 82 phenomena were considered and 24 were evaluated as having "High" importance for at least one of the transients. Another 15 are in the " Medium" category and the rest were ranked " Low". , 2.3.3 Anticipated Transients Without Scram ,, The PIRT for a pressurization event (MSIV closure, turbine trip) with failure to scram is shown in NEDC-32391P Table 2.3-4. Because the event is initiated as a normal pressurization event, a large number of phenomena typical of the early phase of the transient are the same as ; those in the first column of NEDC-32391P Table 2.3-3 for operational transients. NEDC-32391P Supplement I has more discussion of the PIRT rankings. Anticipated transients without scram progress through three phases to shutdown. The first phase is the initial transient resulting from the initiating event, for example inadvertent closure of the MSIVs. 'Ihis phase is similar to the operational transients and the same phenomena are [ important as for the corresponding operational transient with scram. The response is more 2-28
NEDO-32391, Revision C 4 severe because of the lack of an early scram., and boiling transition will occur. In the second ! phase, the lack of scram is sensed and mitigation actions are taken. These include feedwater ! runback to reduce flow and power, and the initiation of the boron timer. Core power is reduced to about 25 % of rated and energy removed by the periodic opening of the SRVs, discharging
! steam into the suppression pool. The parameters of most interest in this phase are the thermal hydraulic and neutronic interactions which lead to power reduction as the natural circulation j flow drops, and the energy deposition in the suppression pool. In the final phase, boron is
- injected, 3 minutes after the stan of the boron timer. Boron is injected into the core bypass
!' region, mixes with the water in the bypass, moves into the core and results in hot shutdown. 1 After the fission process has been shut down, decay heat is removed by the IC and further SRV i openings are not expected. The important phenomena in this period relate to the processes j governing the delivery of the boron to the core. 2.3.4 Stabil!ty l Section 2.2.4 describes the conditions for which SBWR stability is evaluated: steady state i operation, anticipated transients, ATWS, and stanup. Of these, the startup transient has been ! considered as part of the PIRT for transients. The phenomena of importance for steady state and ! anticipated transients are the same. The differences lie in the reactor operating conditions at I which the evaluations are conducted. For example, the limiting conditions of the highest l power / flow ratio are obtained at the end of the loss of feedwater heating transient. Large l margins to instability are calculated for all operational transients. ATWS events lead to the most j severe conditions for stability (power / flow ratio) and also to situations where the critical power i may be exceeded. However, no power oscillations are expected as margins to instability are maintained, and the power / flow ratio becomes more favorable following feedwater runback. l Aspects of film boiling are treated in the ATWS PIRT. For stability, phenomena affecting decay l ratio and the likelihood of oscillations are considered in NEDC-32391P Table 2.3-5. Stability of j a plant is significantly affected by the plant operational state: power / flow ratio, control rod l distribution, and axial and radial power distributions. These have not been included in the PIRT ! list of phenomena. Core stability is also determined largely by the fuel design: void coefficient, two-phase / single phase pressure drop, and fuel rod time constant for heat transfer. These j parameters have been included in the PIRT. A separat: table of the operational and design i
- parameters that govern stability is provided in NEDC-32391P Table 2.3-6. It should be noted that TRACG will only be used for the evaluation of ATWS and the potential for instabilities i during an ATWS event. NRC approved methodology (FABLE code) is being applied for I stability evaluations under steady state conditions and for operating states resulting from !
, anticipated transients.
The BWR stability phenomenon is of the " density wave" type. Perturbations in the void j,.. fraction (density) propagate through the core and other two-phase regions at the vapor velocity, introducing phase lags in the pressure drops. These penurbations in void fraction are associated i l with corresponding changes in the neutron flux. Changes in the neutron flux are fed back to the fluid in the core region as changes in the heat flux. The heat flux perturbations are attenuated in
- magnitude by the thermal inertia of the fuel rods. They also lag the flux and void penurbations i by an amount dependent on the time constant of the fuel rods for heat transfer. Three potential
- instability modes are considered in NEDC-32391P Table 2.3-5. Channel stability refers to the j- hydrodynamic stability of the fuel channel with the highest decay ratio. Channel stability is
- analyzed with no neutronics feedback and with a constant heat flux. It is not possible to produce J
2 29
'NEDO-32391, Revision C l
channel instability in a BWR core because of the neutronic coupling with adjacent channels. Channel stability can be a factor in exciting regional oscillations. Core wide stability refers to the excitation of the core in its fundamental neutronic mode. This is the most common type of instability observed in tests and a handful of events at operadng BWRs. The perturbations in the flows and flux are in phase across the core. Core wide instability involves loop type perturbations, and the downcomer region participates in the process. In a regional instability, a higher order mode of the neutronics, with its associated suberiticality, is excited by the hydraulics. They are more likely to occur in a large core, which has a smaller subcriticality for ' the higher harmonics of the neutronics . The core pressure drop is essendally constant. The flows and fluxes are out of phase in different regions of the core. Regional instabilides have been observed in a few operating BWR cores. Both core wide and regional instabilities have , been calculated with TRACG [NEDE-32177P]. In addition to channel and core stability, the overall plant stability is also analyzed. This examines the response of the plant to changes in set points of the control systems. Overall plant stability (generally an exercise in tuning the control systems) is not treated in this section. A discussion of the individual items in NEDC-32391P Table 2.3-5 can be found in Supplement 1. e' 2-30
NEDO-32391, Revision C 3.0 IDENTIFICATION OF SBWR UNIQUE FEATURES AND PHENOMENA: BOTTOM-UP PROCESS 3.1 Introduction This section describes the Bottom-Up process, one of two methods used to develop the test and analysis needs for SBWR. It complements the Top-Down process described in Chapter 2,
, with which it will be merged in Section 4.
The Top-Down process relies on the Phenomena Identification and Ranking Tables (PIRT) to identify thermal-hydraulic phenomena that are important for TRACG to model accurately and for which code qualification is required. The PIRT tables are an elaboration of the code qualification requirements from a microscopic or phenomenological point of view. The Bottom-up approach compiles a list of SBWR-unique features, associated thermal-hydraulic phenomena and supporting TRACG qualification data from a macroscopic or system / component perspective. The purpose is to evaluate the adequacy of the data base used to qualify TRACG in the areas in portant to SBWR system thermal-hydraulic response, j 3.2 Methodology Each of the 127 SBWR systems was reviewed to determine if the system was unique, had unique features, or if a standard BWR component or system was subject to an application different from that found in the BWR operating fleet. Those systems that did not directly affect the thermal-hydraulic response of the SBWR were not considered. System-unique features, the safety classification of the system, and the Master Parts List (MPL) number were documented. The principal design engineers were consulted with respect to the current reference system design and unique features, as well as References 3,31, 32, to detennine any new issues associated with that unique feature. For each of the issues, associated important thermal-hydraulic phenomena were identified. The SBWR Product Structure is Shown in Figure 3.2-1. S d }. 3-1
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l NEDO-32391, Revision C i l l 3.3 Results i This section provides a summary of results. l l 4 3.3.1 RPV and Internals (B11) Thirteen thennal-hydraulic phenomena were evaluated in detail.
- e
- 3.3.2 NuclearBoilerSystem(B21) I i
- Three thermal-hydraulic phenomena were evaluated in detail. ]
3.3.3 Isolation Condenser System (B32) ; Nine thermal-hydraulic phenomena were evaluated in detail. 3.3.4 Standby Liquid Control System (C41) ! Five thermal-hydraulic phenomena were evaluated in detail. j i 3.3.5 Gravity-Driven Cooling System (E50) j i Nine thermal-hydraulic phenomena were evaluated in detail. 3.3.6 Fuel and Auxiliary Pools Cooling System (G21) Two thermal-hydraulic phenomena were evaluated in detail. 3.3.7 Core (J-Series) In the area of the SBWR core, four issues / phenomena were identified as unique to the l SBWR. 1 3.3.8 Containment (TIO) l During the review of the SBWR design,34 important containment system thermal-hydraulic l phenomena were identified. l 1 3.3.9 Passive Containment Cooling System (T15) 1 The systematic review of the SBWR design identified 17 thermal-hydraulic phenomena related to the design of the PCCS. 3-3
NEDO-32391, Revision C 4.0 EVALUATION OF IDENTIFIED PHENOMENA AND INTERACTIONS The PIRT analysis in Section 2 identified High and Medium ranked phenomena for different types of transients and LOCAs. These were grouped by the period of the transient and listed separately for each region of the reactor vessel and containment. In Section 3, a Bottom-Up process was employed to identify SBWR-unique design features and associated phenomena and interactions. These were classified according to the SBWR system (e.g., FAPCS, Nuclear i Boiler, etc.) where the particular feature was found. Following the overall strategy described in l 2* Section 1.3, the highly ranked phenomena from these lists are now combined in this section to i yield a comprehensive, composite list of phenomena that need to be considered. The complete l list and the correspondence between the PIRT and Bottom-Up items can be found in NEDC-32391P Table SI-9 of the supplement to this report. The phenomena that were ranked Medium have also been tabulated and have been tracked separately. The list is composed of separate tables for phenomena and interactions for each type of transient (LOCA, operational transients, ATWS, stability). The list of interactions is screened in Section 4.2 and reduced to a fm~ al table I of phenomena for which data are needed for qualification of TRACG in Section 4.3. In NEDC-32391P Section 5, these tables are compared against the Test Plan to confirm that all elements of the tables are covered by tests. Specific test coverage for the medium ranked phenomena is also 4 addressed in NEDC-32391P Section 5. Where there is a lack of data, analysis will be performed. 4.1 Composite List ofIdentified Phenomena and Interactions 4.1.1 Loss-Of Coolant Accident (LOCA) This section discusses the phenomena and interactions important to LOCA. . 4.1.2 Anticipated Transients This section discusses the phenomena and interactions important to anticipated transients. 4.1.3 ATWS PIRT for ATWS was developed in Subsections 2.3.3. l
- 4.1.4 Stability PIRT for stability was developed in Subsection 2.3.4.
4 4.2 Analytical Evaluation of System Interactions The purpose of the system interaction study was: (1) to investigate the effects of both active l and passive systems which could be available to support Engineered Systems Feature (ESP) 4 systems during a LOCA; and, (2) to determine if interactions between the systems could degrade l the performance of the ESF systems from what it would be if they were acting alone. The study extends earlier work presented in Chapter 6 of the SSAR (Reference 3), which evaluated the l 4-1
NEDO-32391, Revision C effect of the break location and of various single failures. A part of this earlier study examined the possible adverse effect of reverse flow through the Isolation Condenser during an inadvertent 1 opening of a DPV. Additional analysis in Chapter 19 of the SSAR (Reference 3) examined use of non-safety grade engineered systems to prevent core damage. The present study examines both system interactions which could affect the SBWR primary system response, as measured by the fuel temperature and vessel water level, and system . interactions which could affect the containment response, as measured by the containment . temperature and pressure. The study was performed using the TRACG code with two different input models. System interactions affecting the primary system were studied with the TRACG input model used for LOCA analysis of the SBWR, which provides a detailed representation of
~
! the reactor core, vessel internals and associated systems, but a less detailed representation of the containment. For system interactions affecting the contamment, the TRACG input model for containment analysis was used. This input model provides a more detailed representation cf the containment and its systems but a less detailed reactor pressure vessel model. Both input models have been compared to assure that they predict similar global response behavior of the reactor pressure vessel and containment. The use of analysis methods is a practical and effective way to evaluate system interactions. The TRACG code and the input models for the primary system and containment which were discussed above include detailed modeling of the important passive and active systems available in the SBWR and can simulate the interactions between these systems during various accident scenarios. This makes it possible to semen a large number of possible system combinations and accident paths to identify those system combinations and accidents most likely to produce adverse interactions. Based on this type of study, final confirmation of interaction effects can then be obtained from integral tests. 4.2.1 Accident Scenario Definition The systems selected for the study were those that would likely be available during a LOCA and which could produce adverse interactions with the safety grade engineered systems for core and containment cooling. 4.2.2 Results from the Primary Systems Interactions Study ' Several different break locations were considered for the primary system interactions study. ,. 4.2.3 Results from the Containment Systems Interactions Study The containment system interactions study investigated interactions between available safety grade engineered systems as well as interactions of these systems with other systems which could be available for contamment cooling without a loss of power. l 4-2
NEDO 32391, Revision C 4.2.4 Summary of System Interaction Studies The system interactions considered in this study included those considered most likely to occur when some form of external electrical power was available and which were not clearly , beneficial to the operadon of the safety grade engineered safety systems. 4.3 Summary of Evaluations
'Ihis section summarizes the results of screening the phenomena listed in the tables of Section 4.1, primarily in the area of interactions, as a result of the studies of Section 4.2. This
, constitutes the final step in determining the needs for test data for TRACG qualification. These needs are detailed in Subsections 4.3.1 and 4.3.2 for LOCA and transients, respectively. Subsection 4.3.3 covers ATWS and stability. Section 5 then presents the results of comparing these needs against the test plan. 4.3.1 LOCA Table 4.1-la summarized the highly ranked phenomena for LOCA/ECCS. This table will be used in its entirety as a list of needs for qualification tests. Items listed as being moderately important in Table 4.1-lb will also be evaluated. 4.3.2 Transients All issues but one will be carried forward to Section 5 as needs for TRACG qualificadon. 4.3.3 ATWS and Stability For ATWS, the phenomena shown in Table 4.1-4a will be carried through to Section 5 for evaluation. Phenomena of moderate importance from Table 4.1-4b will also be evaluated. The majority of the phenomena are captured either by the Transient PIRT (neutronic and thermal hydraulic issues, Isolation Condenser, etc.) for the reactor parameters or by the Containment PIRT for SRV discharge to the suppression pool (critical flow, pool stratification and heatup, 3 etc.). Natural circulation at low downcomer water levels leading to reverse bypass flow providing an internal natural circulation path inside the core shroud is covered under LOCA. ~. l 4-3
l NEDO-32391, Revision C , l 5.0 MATRIX OF TESTS NEEDED FOR S8WR PERFORMANCE ANALYSIS The tables of highly ranked and medium ranked phenomena and interactions from Section 4 were compared with the original Test Plan as it existed when this study began. It was found that most of the identified effects were covered by the existing tests which could be used to , qualify TRACG. The list of test data has been restricted to include only tests that have been , analyzed, or will be analyzed, with the configuration-controlled version of TRACG. His results l in elimination of data from a number of facilities (simulations of earlier BWR types) that have , not been reanalyzed with the current version of the code. The early GIRAFFE data (Phase 1) are now being used for confirmatory purposes and have been marked as such. In a few cases,. additional testing or qualification was proposed and incorporated in the Test Plan. He resulting , matrix of tests needed for TRACG qualification is presented in this section. The tests have been divided into (1) Separate Effects Tests, (2) Component Performance Tests, (3) Integral System Tests, and (4) Operating Plant Data. He first two types of tests are suitable for model development, the latter two for checking the overall performance of the code. l Separate tables have been included for the High and Medium ranked phenomena. The Medium ranked phenomena are less important. Where test data are not available, analyses will-be performed to evaluate their significance. A separate PIRT was developed for stability in Section 2 and the High and Medium parameters identified in NEDC-32391P Section 4. However, in evaluating test coverage, these have been consolidated into three entries: STl for hydrodynamic or channel stability; ST2 for core wide stability; and ST3 for regional stability. In the interest of conciseness, the tables in NEDC-32391P Section 5 do not include identification of the phases of the LOCA or specific transient for which data are needed. This information is contained in the previous sections. Also, the tables in NEDC-32391P Section 5 do not contain sufficient detail to verify that each referenced test provides data over the applicable range of SBWR conditions. For the major SBWR facilities such as GIST, PANDA, GIRAFFE, and PANTHERS, relevant information is in Appendix A. For other test facilities, details on range of test data will be provided in the TRACG Qualification Report. 5.1 Separate Effects Tests The facilities are listed in Appendix A, where the type of test, test purpose and data available from each are also briefly described. 5.2 Component Performance Tests ne distinction between component tests and separate effects tests is that the component tests focus on overall component perfonnance. He level of instrumentation may not be as extensive as in a separate effects test. The distinction is often blurred. Both types of data have controlled boundary conditions and are suitable for model development as well as validation. l 5-1
NEDO-32391, Revision C A large number of phenomena related to the blowdown and refill processes in the Icwer plenum, bypass and core are covered by the component tests. Parallel channel effects and separator characteristics are also part of this data- base. Of special note are the DPV tests, where the blowdown capacities of full-scale DPVs were tested. Fdl-scale tests of an IC module are planned at the PANTHERS facility. A scaled model will be tested at PANDA. On the containment side, two full-scale PCCS modules will be tested at PANTHERS. A ,, 1/25 scale module will be tested in PANDA. A smaller module with three tubes has been tested l in the GIRAFFE test facility by TOSHIBA (supponing data). A large amount of data exists on the early blowdown response of pressure suppression type containments. . 5.3 Integral System Response Tests Integral system response tests model overall behavior of a facility subjected to transients simulating specific accidents or transient events. Tests are performed on a scaled simulation of the reactor system. This section discusses the integral systems testing of the SBWR. 5.4 Plant Operating Data The transient response of the SBWR is similar to that of other BWRs for operational transients in many respects. Plant data are very valuable in validating code performance for comple~. systems involving an interplay between thermal hydraulics, neutron kinetics and control spl:m response. 5.5 Summary of Test Coverage NEDC-32391P Sections 5.1 through 5.4 identified the test facilities and BWR plants from which data have been used (or will be used) for TRACG qualification. This information was tabulated for each of the identified important phenomena, by category of tests (separate effects, component performance, etc.). l 5-2
XLOO-32391, Revision C 6.0 INTEGRATION OF TESTS AND ANALYSIS This section examines the tasks necessary to complete the qualification of TRACG. Figure 6.1-1 shows the " Road-Map" of how the new and existing test data support SBWR certificadon. 6.1 TRACG Qualification Plan
. This section discusses work required to qualify TRACG for SBWR application.
The Analysis Plan in Appendix A identifies the specific tests for which blind predictions and post-test analysis will be performed. 6.2 Use of Data for TRACG Model Improvement and Validation The TRACG computer code is qualified to Level 2 (verified, production) status at GE-NE. Thus, the code configuration is controlled, and the models and the results of validation testing have been reviewed and approved by an independent Design Review Team. In the development process, the separate effects and component data were used for model development and refinement. These data also provided guidelines for the nodalization which was used for all the SBWR calculations. The new data and the results of the post-test analyses will be used in the same way. If changes are necessary to the TRACG models, a new version of the code will be created and brought to a controlled Level 2 status under the GE-NE quality assurance procedures. If changes in the nodalization are indicated, calculations affected by the changes will be redone and ' reverified. 1 09 4 l 6-1
OPERATING EXISTING BWR EXISTING BWR REACTOR I EXISTING SBWR TECHNOLOGY TECHNOLOGY EXPERIENCE l TECHNOLOGY l GIST PANTHERS /PCC OPERATING GIST PANTHERS /IC REACTOR PANTHERS /PCC PANDA PANDA EXPERIE!.CE PANTHERS /IC GIRAFFE CRIEP! GEYSERING PANTHERS IC 4T MKit DODEWAARD STARTUP PSTF MKill PSTF MKill 4T MKit SEPARATE COMPONENT INTERGRAL SYSTEMS EFFECTS QUALIFICATION PERFORMANCE QUALIFICATION AND Q QUALIFICATION CONCEPT DEMONSTRATION 4 n N
?
tO .*
, _x' _ ^
( 3
,'g g , & TRACG ;
mANAtysia s i o
'x s
~ ,;
. s Tttjj$fD$$1Ghi.4)%;i6 gman- + s
' sc ,
's I:iifhkN ' ^ s; l COMPONENT DEMONSTRATION PANTHERS /IC TESTS PANTHERS /PCC STRUCTURAL DEMO > 4
( STRUCTURAL DEMO ! VACUUM KER DEPRESSURIZATION 88fgg g VALVE DEMO Figure 6.1-1 Technology Basis for SBWR Design i
NEDO-32391, Revision C . i ! i r I ! 7.0
SUMMARY
AND CONCLUSIONS I The Test and Analysis Program Description (TAPD) systematically defined test and l i analysis needs using Top-Down and Bottom-Up approaches to identify key phenomena, issues l and interactions between phenomena and systems (Sections 2,3, and 4 and Appendix C). These : i needs were compamd to the existing test plan and the existing TRACG qualification plan, and
- modifications were made where necessary to fill in gaps in the database and the TRACG ,
! qualification base (Sections 5 and 6). The Test and Analysis Plan defined the remaining i# activities for closure (Appendix A). Test facility scaling was addressed quantitatively in . ; l Reference [32]. This document supersedes previous GE-NE submittals with regard to test j, objectives, test conditions, data use, and anticipated test analysis.
! Several changes in the test and analysis programs resulted from the study documented here. l 1 - A number of tests were added. In severalinstances, tasks to be performed have been defined in , more detail, and the focus and data usage from some facilities was modified. The following summarizes the key changes:
- Test Plan j I
- GIST: No changes in testing. Data usage focused on TRACG qualification of GDCS j injection into RPV: GDCS flow, GDCS initiation time, and RPV levels. ,
1
- GIRAFFE: Phase 1 and Phase 2 data usage changed from primary qualification of l TRACG to support use. Helium and systems interaction testing (SIT) added.
. i
- PANTHERS /PCC: No changes in testing or data usage.
- PANTHERS /IC: Test matrix revised to measure performance at lower pressures.
l
- PANDA: Program added to list of tests required for certification. Test matrix expanded l from two to nine transient tests. Program becomes the primary containment and systems mteraction data base.
l Analysis Plan
- GIST: Analysis completed.
l
- GIRAFFE: Helium test and systems interaction test analyses added for TRACG analysis.
I,,
- PANTHERS /PCC: Sixteen specific runs identified for TRACG analysis.
. t PANTHERS /IC: Six specific runs identified for TRACG analysis. I f ..
- PANDA: All six steady state tests and nine LOCA tests identified for TRACG analysis.
l OTHER TESTS: TRACG analysis of five other tests (1/6 scale Boron mixing, CRIEPI ! Geysering, PSTF/Mk III,4T/Mk II, and PSTF Stratification) and one operating plant i experience (Dodewaard startup) to address specific identified qualification needs. 3
- The TAPD specifically addresses the requirements of 10CFR52.47 by establishing that a technology basis (a combination of test data, analysis and plant data) exists for the SBWR safety features, for interdependent effects between safety features, and for qualification of the TRACG code used for SBWR safety analysis. Specifically:
l 7-1
NEDO-32391, Revision C
- 10CFR52.47 requires that 'The performance of each safety feature of the design has been demonstrated through either analysis, appropriate test programs, experience, or a ;
combination thereof." The studies summarized in Sections 2, 3 and 4 defined the , phenomena important to SBWR safety in two independent ways. These are merged in Section 5 where the testing and experience bases applicable to each are shown. Each important phenomenon is covered by at least one separate effects test, component test, integral systems test, or operating reactor datum. 10 CFR52.47 requires that " Interdependent effects among the safety features of the - design have been found to be acceptable by analysis, appropriate test programs, l- experience, or a combination thereof." The studies summarized in NEDC-32391P Section 4 and Appendix C identified the important interactions. For most of these, analyses or tests already planned suffice to show the effects are negligible or bounded. For a few, additional tests were judged to be necessary. These have been added to the SBWR program.
- 10CFR52.47 requires that " Sufficient data exist on the safety features of the design to assess the analytical tools used for safety analysis over a sufficient range of normal operating conditions, transient conditions, and specified accident sequences, including equilibrium core conditions." The matrix of tests and operating plant data shown in
-l NEDC-32391P Section 5 identifies elements which have been used to date (X entries), elements in which existing test data will be used (Q entries), and elements in which forthcoming test data will be used (T,Q entries) to qualify the SBWR analydcal model, TRACG. These are collected in Section 6 to show the composite TRACG qualification plan. GE-NE believes that if the overall TRACG qualification plan described in Section 6, and the SBWR-specific test programs (and associated TRACG analyses) described in Appendix A, are completed with no major surprises, it will be possible to conclude that the provisions of 10CFR52.47(b)(2)(i)(A)(1), (2), and (3) have been satisfied. Oe l- 7-2
NEDO-32391, Revision C i
8.0 REFERENCES
a [1] TRACG Model Description, J.G.M. Andersen, Md. Alamgir, Y.K. Cheung, L.A. Klebanov, J.C. Shaug. NEDE-32176P, Licensing Topical Report, January 1993. l[2] TRACG Qualification, J.G.M. Andersen, M.D. Alamgir, J.S. Bowman, Y.K. Cheung, j L.A. Klebanov, W. Marquino, M. Robergeau, D.A. Salmon, J.C. Shaug, B.S. Shiralkar, F.D. Shum, K.M. Vierow, NEDE-32177P, Licensing Topical Report, January 1993. (3) SBWR Standard Safety Analysis Report,25A5113, Rev. A. [4] Quantifying Reactor Safety Margins, NUREG/CR/5249, EGG-2552, Rev. 4. (5} 1mplementation of TRACG for Licensing Analysis, J.L. Rash to R.C. Jones Jr. (NRC, NRR Chief Reactor Systems Branch), MFN 042-92 and JSC-92-010, March 9,1992. [6] Quantifying Reactor Safety Margins, B.E. Boyack et. al, Nuclear Engineering and Design (Parts 1-4),119, Elsevier Science Publishers B.V. (North Holland),1990. l(7) Application of TRACG Model to SBWR Licensing Safety Analysis, NEDE-32178P, H.T. Kim, February 1993. (8} AEOD Concerns Regarding The Power Oscillation Event at LaSalle 2 (BWR-5), USNRC, AEOD Special Report S803,1988. [9] Leibstadt Stability Tests During Startup Testing (contained in Reference 5).
- [10] Stability Investigations ofForsmark-1 BWR, B. Anderson, et. al., International Workshop i on BWR Stability, October 1990.
[11) Vermont Yankee Cycle 8 Stability and Recirculation Pump Trip Test Report, NEDE-I. 25445, August 1982. [12] ODYSYO1/ODYSYO2 Qualification Report, NEDE-30227, (includes Peach Bottom-2 i
- Stability Test Data), July 1983.
i (l31 The Startup of the Dodewaard Natural Circulation BWR - Experiences, W.H.M. Ntssen et. al., N.V. GKN, Netherlands, Waalbandijk 112a, 6669 MG Dodewaard (Ah? '92), i , Tokyo, Japan (Detail measurements are contained in GKN-report 92-017/FY/R), October l 25-29, 1992.
. [14} Stability Monitoring of a Natural-Circulation-Cooled Boiling Water Reactor, T.H.J.J.
van der Hagen, Thesis, Delft Univ. of Technology, The Netherlands,1989. [15} Measurements at Various Pressures at the Dodewaard Natural Circulation Boiling Water Reactor in Cycle 23, T.H.J.J. van der Hagen, GKN-Report 93-023/FY/R,1993. [16] Hatch Unit 2 Two-pump Trip Test, November 18,1978 (data contained in Reference 5). [17] Hydrodynamic and Heat Transfer Measurements on a Full-Scale Simulated 36 Rod BWR Fuel Element with Non Umform Axial and Radial Heat Flux Distribution, O. Nylund, et. al., FRIGG-4, ASEA-ATOM, December 1970. (18} ABWR Horizontal Vent Containment Tests, NEDC-31393 - SS Test Series. (19] MIT and UCB Separate Effects Tests for PCCS Tube Geometry, " Single Tube Condensation Test Program", NEDC-32301. 3 8-1
. NEDO-32391, Revision C } (20] Mark III Confirmatory Test Programm, Phase 1 Large Scale Demonstration Station j Tests, NEDM-13377, October 1974.
- [21] MSC/NASTRAN Manual Version 67 Macneal-Schwendler Corporation, Los Angeles, CA.
i (22] ABWR Horizontal Vent Containment Tests, NEDC-31027 - FS series.
- (23] Steam Purge Clearing Tests, NEDE-28853.
(24} 1/3 Area Scaled- Single Vent Row Tests, NEDE-21596P. - (25} 1/9 Area Scaled-3 Vent Row Tests, NEDE-24720P. , [26] JSBWR Phase 2 Report, EBWR and VK50 Tests. , , l l(27] Thermo-Hydraulic instability of natural circulation BWRs (Explanation on instability mechanisms at startup by homogeneous and thermodynamic equilibrium model , considering flashing effect.) by Fumio inada and Tomio Ohkawa, Int. Conf. New Trends ' in Nuclear System Thermohydraulics, pp. 187-193,1994. e i28] BWR 5 - 1/6 scale Boron Mixing Tests, NEDE-22267. [29] ABWR - 1/6 scale Boron Mixing Tests, NEDC-30326.
- [30] Document No. BN-TOP-3, Bechtel Power Corp, San Francisco, CA, August 1975. -
l (31] Design and Analysis Similarities Between the ABWR and SBWR, NEDC-32231, December 1993. [32] Scaling of the SBWR Related Tests, NEDC-32288 Rev.1, R.E. Gamble, A. Hunsbedt, j F.J. Moody, M.E. Parker, G. Yadigaroglu. l (33} Sixteen-Rod Heat Flux investigation, Steam-water at 600 to 1250 psia, E. Janssen, GE. , I (34] The Effects of Lengths and Pressure on the Critical Heat Fluxfor a Closely Spaced 19- l l Rod Bundle in Forced Convective Boiling, Dept. of Chem Eng., Engineering Research } ! Lab Heat Transfer Research Facility, Columbia University, NY. t
\35} NRC Requestsfor AdditionalInformation (RAls) on the Simplified Boiling Water Reactor (
l
- (SBWR) Design, Letter MFN No. 078-94 from P.W. Marriott (GE) to Richard W.
j Borchardt (USNRC), March 31,1994. , (36] Quantifying Reactor Safety Margins - Application of Code Scaling Applicability and
- f. Uncertainty Evaluation Methodology for a Isrge-Break lass-of-Coolant Accident, B. ,.
Boyack et al., NUREG/CR-5249,1989. i [37] (See Reference 7). Mark 11 Pressure Suppression Test Program Phase 11 and Ill Tests, NEDE-13442P-01, - ) l [38] May 1976. Mark 11 Pressure Suppression Test Program Phase 11 and 111 Tests, NEDO- ' 4 13465, October 1976.
- . \39] Hierarchical, Two-Tiered Scaling Analysis, Appendix D to An Integrated Structure and l'
Scaling Methodology for Severe Accident Technical Issue Resolution, Nuclear l Regulatory Commission Report, NUREG/CR-5809, EGG-2659, November 1991. [40] GE COMPASS data base, run dated July 1994. 3 [41]' GE Master File Number MFN 042-92, Letter J.L. Rath (GE) to R.C. Jones (NRC), a implementation of TRACGfor Licensing Analysis, March 9,1992. 8-2 l
NEDO-32391, Revision C (42} GIST Final Test Report, SBWR Program Gravity-Driven Cooling System Integrated Systems Test, GEFR-00850, October 1989. [43} GIRAFFE Passive Heat Removal Testing Program, by K.M. Vierow, NEDC-32215P, June 1993. (44) Test Reportfor SBWR Depressurization Valve Operational and Flow Rate Test, WYLE Report #41152-0, September 1990.
,, (45) Startup of the Dodewaard Natural Circulation Boiling Water Reactor, Hagen, T.H.J.J.
van der, Karuza, J., Nissen, W.H.M., Stekelenburg, A.J.C., Wouters, J.A.A., GKN Report 92-017/FY/R,1992.
. [46) Full-Scale Mark Hi Tests, Test Series 5707, NEDE-21853-P, August 1978.
(47] Mark Hi Conprmatory Test Program,1B Scale Condensation and Stratipcation Phenomena, Test Series 5807, NEDE-21596P, March 1977. (48) A Comparison of the RELAP Simulation of Mark HI Suppression Pool 7hermal Stratipcation with Data from the Pressure Suppression Test Facility, NEDE-21957P, September 1978. (49] Transient Behavior of Natural Circulation for Boiling Two-Phase Flow (2nd Report:
- Mechanism of Geysering), M. Aritomi, et al., Transactions of First JSME/ASME Joint I
International Conference on Nuclear Engineering (ICONE-1), Tokyo, Japan, pp. 87-94, November 4-7,1991. [50] PANTHERS Test Plan & Procedure, SIET document No. 0098 PP 91, Revision 1. (51} Technical Specipcation For LC & PCC Instruments Installation, SIET document No. 00157 ST 92, Revision 1, January 1,1994.
\52} PANTHERS-PCC Test Facility Instrumentation, Data Acquisition, & Processing Specipcation, SIET document No. 00095 RS 91, Revision 1, June 8,1994. '
(53] Isolation Condenser and Passive Containment Condenser Test Requirements, GE Nuclear Energy,22A6999. [54] PANDA Test Specifcation, GE Nuclear Energy,22A5587. PANDA Steady State PCC Performance Tests, Test Plan and Procedures, Paul Scherrer l(55} Institut, document No. ALPHA-410.
. [56] PANDA Pre-Test Analysis, NUCON Report 40315-NUC-94-7034, GE Nuclear Energy Letter MFN 119-94.
[57] GIRAFFE Test Specipcation, GE Nuclear Energy, 25A5677.
~
(58] Guidefor Quality Assurance ofNuclear Power Plants, Electrotechnical Standard Survey Committee, Japan Electric Association, JEAG 4101-1990. [59] PANTHERS Pre-Test Calculation (contained in Reference 35). (60) Thermally Induced Flow Instabilities in Two-Phase Mixtures, Ishii, M., and Zuber, N., 4th Intemational Heat Transfer Conference, Paris, Paper No. B5.11,1970. (61] Drift Flux Modelfor Large Diameter Pipe and New Correlation for Pool Void Fraction, { Kataoka, I. and Ishii, M., Intl. J. Heat Mass Transfer, vol. 30,1927-1939,1987. J 8-3
NEDO-32391, Revision C (62) Scaling and Analysis of Mixing in Large Stratified Volumes, Peterson, P.F., Intl. J. Heat Mass Transfer, vol. 37,97-106,1994. l(63} The Thermal Hydraulics of a Boiling Water Reactor, Lahey, Jr., R.T., Moody, F.J., American Nuclear Society,1977. (64) Thermo-hydraulic Instability of Natural Circulation BWRs at Low Pressure Start-up: Experimental Estimation of Instability Region with Test Facility Considering Scaling 4 low, Fumio INADA, Masahiro FURUYA and Akira YASUO, Central Research Institute
~
of Electric Power Industry, CRIEPI, Hiroaki TAB ATA and Yuzuru YOSHIOKA, Japan Atomic Power Company, JAPC, H. T. Kim, GE Nuclear Energy. Paper presented at ~ ICONE3, April 1995. , (65} TRACG Analyses of Flashing Instability During Start-up, Andersen, J.G.M. and l Klebanov, L.A., GE Nuclear Energy. Paper presented at ICONE3, April 1995. [66) Thermal-Hydraulic Oscillations in a Low Pressure Two-Phase Natural Circulation Loop at Low Poweis and High Inlet Subcooling, S.B. Wang, J.Y. Wu, Chin Pan and W.K. Lin, i National Tsing Hua University,4th International Topical Meeting on Nuclear Thermal . Hydraulics, Operations and Safety, April 6-8,1994. ^ (67] GIRAFFE Heat Removal Performance Tests Test Plan and Procedures, Toshiba l document no. TOGE110-TO7, revision 2, May 31,1995. [68] BWR Void Fraction Correlation, J.A. Findlay, G.E. Dix, NEDE-21933, August,1978. l (69] Critical Power and Press::re Drop Test-Step 11 Fuel Design Standardization Program for i BWR/2-S Reload Fuel, B. Matzner and D.A. Wilhelmson, NEDC-31499P, San Jose, CA, 1987. [10} Separate Effect Tests at UCBfor PCCS Tube Geometry, NEDC-32310. [71) Backpow Leakage from the Bypass Region for ECCS Calculations, General Electric
- Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K, NEDE-21156, January,1976.
(72) Supplemental Information for Plant Modspcations to Eliminate Signipcant In-Core Vibrations, NEDE-21156, January,1976. i (73] Vacuum Breaker Design Basis Accident Simulation Test Procedure, PCNVBR00001 Is.: 1 Rev.1 (BPD DIFESA E SPAZIO Report).
~
[74) Monticello inplant SRVDischarge High lxad Test, NEDC 20997, Rev. A. [75) Caorso SRVDischarge Tests, Phase 2, NEDO-25118. Caorso SRVDischarge Tests, Phase 1, NEDE-25100P. (76) (77] PANTHERS- PCC Data Analysis Report, SIET Document No. 0039 RA95, Rev. O, June 20,1995. (18] Development of a System for Catalytic Hydrogen Reduction in Severe Accidents in i Nuclear Power Plants (Title translated from the German), M. Eigenbauer, and M. Seidler, NIS Ingenieurgesellschaft Technical Report No.1141/1866/0, Sep.17,1991. (79] Letter from J.E. Quinn (GE) to T.E. Quay (NRC), SBWR - Pre-Test Analysisfor PANDA Test M3, MFN 161-95, August 21,1995. 8-4
. .t NEDO-32391, Revision C i .
,. [80] PANTHERSBC Test Plan, SIET document no. 00396RI95.
, [81] PANTHERSBC Test Procedures, SIET document no. 00395PP05. . t 8 4 d . I ge i a a s 7 J I 4 i 4 4 i
\
j ; i . l i-i k 1' E 1* 1 1 M 4 i- . 4 4 4 4 k 8-5
~
d 4 l NEDO-32391, Revision C
~
- - APPENDIX A - TEST AND ANALYSIS PLAN (TAP) i i
A.1 Introduction i This appendix identifies the specific tests and analyses that will be performed to meet the , I identified supplemental needs.
- The goal of the SBWR Test Program is to provide a sufficient database to support i certification of the SBWR as a standard design. Consequently, the scope of the test program goes !
). l beyond establishment of the TRACG qualification data base, in that demonstration testing of
] concepts unique to the SBWR, or equipment having design requirements not previously analyzed ;
~
or tested, is also included. This testing is also described in this appendix. In many cases, the same test data are used for both applications. t Section A.2 provides an overview of the philosophy used in determination of specific tests 4 and analyses, definition of test types, and an overview of the test effort. Section A.3 presents the ; Test and Analysis Plan. The following information is provided for each identified test:
- . Test Plan l
- A test description including overviews of test facilities, instrumentation, and [
procedures.
- Objectives for each test program (and specific tests, as applicable).
- Test matrices, cross referenced to the test objectives, anddescriptions of how the data
} will be used to meet the test objectives. ! l
- Justification of the test conditions. l l Analysis Plan !
- Test runs identified for TRACG analysis l
- Description of how the identified comparisons between test and analysis meet the
- qualification needs This document supersedes previous submittals with regard to test objectives, test conditions, 1 3
data use, and anticipated test analysis. l A.2 Test and Analysis Philosophy A.2.1 Test Types
' The overall goals of the SBWR Test and Analysis Program are to be met by several types of testing, in several different facilities, world wide. Testing is divided into:
- Thermal-Hydraulic Testing - provides data necessary for qualification of TRACG and for demonstration of the concepts ofpassive safety systems design. 'Ihermal-l hydraulic testing is further subdivided into (1) steady state and separate effects tests, (2) component performance tests, (3) integral systems tests, and (4) concept demonstration tests.
!
- Component Demonstration Testing - provides data on the capability of specific equipment to meet its design objectives.
4
l NEDO-32391, Revision C A.2.2 Test Overview SBWR thermal-hydraulic testing is summarized in Table A.2-1. The test program consists l of 124 steady state test conditions,15 transient perfonnance demonstrations, and 45 integral systems tests. Subsection A.3.1 describes each of the four facilities (PANTHERS, PANDA, GIST, and GIRAFFE) in which these tests will be or have been performed, and includes specific test objectives, test matrices and descriptions of how each of the test groups addresses the test ' objectives. Subsection A.3.1.8 also gives an overview of other data that will be used for TRACG qualification beyond the qualification described in Reference 2. - SBWR component performance tests are described in Subsection A.3.2, including testing of
. the PCC and IC heat exchanger components, depressurization valves (DPVs), and vacuum breaker valves (VB).
A.2.3 Test Approach
'Ihe philosophy of testing is to focus on those features and components that are SBWR-unique or performance-critical, and to test over a range that spans and bounds the SBWR parameters ofimponance. In general, TRACG is used to predict the SBWR parameter range for the spectrum of accidents and transients, and then that range is bounded in the test matrix. Some SBWR tests are performed in a scaled configuration. For these tests, the values of theimponant parameters are scaled to be consistentwith this philosophy. This approach is discussed in Reference 32 and Appendix B.
Additionally,it is the program philosophy to test in multiple scales wherever possible. In these cases, initial conditions for the various tests have been made as similar as possible. Multiple scale testing is useful, since it validates the scaling approach and allows a better understanding of the thennal-hydraulic phenomena involved. A.2.4 Analytical Approach The analytical approach to be used is consistent with that previously documented in the .- , TRACG Qualification Licensing Topical Report, (Reference 2). Briefly, the approach is to choose j a representative sampling of test data which comprise separate effects, component performance, j and integral systems effects, and to perform either pre-test or post-test analysis using TRACG. ~ ! Tests are chosen for analytical prediction Sased on their adequacy to demonstrate modelprediction i capability over the range of predictedSBWR conditions. Sufficient tests are chosen from ! certification data to establish model adequacy, Additional tests have been chosen from supporting [
' data to confum the certification predictions, over a wider range of test conditions, or at intermediate ;
points. It is planned to produce a number of " double blind" pre-test analyses for those cernfication ! data experiments not yet performed. Double blind indicates that the analyst has no information on ! either the results or the exact initial conditions of the experiments. These predictions are based on ] the as-designed facility configurations, and will be verified. 1 l' A-2 l 1
-. - _ _ - , . ~ - .- - _ _ . . - . - . - __ _ - - - - _ . - . _ _ .
l
. i j :
l NEDO-32391, Revision C l l 1 4 Following completion of individual tests, additional test runs will be analyzed with TRACG .
- and compared with the test results. These post-test analyses will be performed with the analyst !
having knowledge of the test results, but will utilize the same nodalization and modeling as the
" double blind" predictions, corncted, if necessary, to reflect facility as-built geometry and the actual initial conditions. The objective is to establish the adequacy of the TRACG model in this application. Allinput decks will be verified. j TRACG modeling or nodalization changes are not expected, but will be made if deemed i
- 1. necessary following an assessment of TRACG predictive capability.
-. A.2.5 Documentation of Tests and Analysis
- A.2.5.1 Test Documentation l 1
1 Testing is documented by submittal of a series of repons and other documentation that define i the configuration of each SBWR test facility, the evaluations performed in conducting the tests, ! i and the results of the testing. Table A.2-2 provides a listing of these submittals. In those cases : ) where the documentation has already been submitted, Table A.2-2 also includes the submittaldate,
. and a reference document identification.
Tables-of-Contents for those report types that apply to all four major test programs j [ Apparent Test Results (ATRs), Data Transmittal Reports (DTRS) and Data Analysis Repons ; i (DARs)] are included as Attachment A1. In addition, an SBWR Test Program Licensing Topical i Report will be submitted summarizing the results from all testing, and integrating the findings l l from all of the test programs. ; i l A.2.5.2 Analysis Documentation i 1 , ! The results of the TRACG analysis will be documented in the form of pre-test predictions i i for selected tests, preliminary validation results for each set of tests, and afinal TRACG Qualification Licensing Topical Report. The Licensing Topical Repons are listed in Section 1.4.2. ', The other analysis reports that are planned for submittal to the NRC are listed in Table A.2-3. i Each of the preliminary validation repons will include the objective of the qualification task, l ,, the use of the data, a description of theTRACG model and a discussion of the results. He i proposed Table of Contents for the Preliminary Validation Results documents is shown in l Attachment A1. he results of these post-test calculations as well as other supporting qualification j ., studies will be integrated into the final Licensing Topical Report (LTR), entitled "TRACG
- Qualification for SBWR". This LTR will supplement the previous LTR on TRACG Qualification,
- .and will include comparisons with data from all the SBWR-specific facilities. His LTR will I
discuss the overall strategy, nodalization of the reactor vessel as well as the containment, and the j evaluation of model uncertainties and bias for SBWR application. The detailed Table of Contents is l pmvidedin Attachment A1. P 1 1 i l A-3
- . .. - . . ~- _- - - - - l NEDO-32391, Revision C A.3 Test and Analysis Plan A.3.1 Thermal-Hydraulic Tests l A.3.1.1 PANTHERS /PCC A.3.1.1.1 Test Description ' Overview l PANTHERS /PCC (Passive Containment Condenser) testing is performed as a joint effort . j by GE, Ansaldo, ENEA, and ENEL at Societa Informazioni Esperienze Termoidrauliche (SIET) i in Piacenza, Italy. 'Ihe test facility consists of a prototype PCC unit, steam supply, air supply, and vent and condensate volumes sufficient to establish PCC thermal-hydraulic performance. Both - thermal-hydraulic and component structural demonstration tests are performed in this facility. This section covers the thermal-hydraulic portion of the testing; component structural performance tests are covered in Subsection A.3.2.1. ' The PCC condenser is afull-scale, two-module venical tube heat exchanger designed and i built by Ansaldo. Figure A.3-1 is an outline drawing of the heat exchanger assembly. It should be noted that the heat exchangeris a prototype unit, built to prototype procedures and using prototype materials. Three heat exchanger units (6 modules) would be found in an SBWR. The PCC is installed in a water pool having the appropriate volume for one SBWR PCC assembly. Instrumentation > Figure A.3-2 is a schematic of the PANTHERS /PCC facility. The primary instrumentation l specified is sufficient to ascertain heat exchanger thermal-hydraulic performance by performing ; mass and energy balances on the facility. Additionally, four heat exchanger tubes are instrumented in such a way that local heat flux information may be obtained. All test instrumentation is calibrated against standards equivalent to the U.S. National , Institute of Standards and Technology. Table A.3-1 defines the thermal-hydraulicmeasurements l taken during the PCC tests. Additional information may be found in the PANTHERS /PCC Test l Plan and Procedure (Reference 50), the Technical Specification of ICand PCC Instrument ; Installation (Reference 51), the PANTHERS PCC Test Facility Instrumentation, Data ; Acquisitions, and Processing Specification (Reference 52), and the IsolationCondenser and ! Passive Containment Condenser Test Requirements (Reference 53). l Test Method The majority of the PANTHERS /PCC testing is steady state performance testing. For these f
~
tests, the facility is placed in a condition where steam or air / steam mixtures are supplied to the ; PCC, and the condensed vapor and vented gases are collected. Allinlet and outlet flows are measured. The condensate is returned to the steam supply, and the vented gas is released to the l atmosphere. Once steady state conditions are established, data are collected for a period of approximately 15 minutes. The time-averaged data are reported and analyzed. Steady state tests using a steam / air mixture are perfonned as follows. The test loop and PCC condenser are first purged with steam to remove any residual air from the system and to heat the PCC pool to saturation. When the pool is boiling, the required steam flow rate is established, , followed by establishment of the required air flow rate to the PCC. The desired PCC inlet pressure ! l A-4 f
l NEDO-32391, Revision C is then established by adjusting the position of the vent tankflow control valve. When steady conditions have been established, data is taken for a period of approximately 15 minutes. A slightly different procedure is used for the steam-only tests. In this case, the vent tank is isolated by installation of a blind flange on the vent line. Following purging of the system, the desired steam flow rate is established. The inlet pressure is not controlled, but allowed to stabilize while maintaining full condensation at the desired steam flow rate. Again, data is then acquired for a period of approximately 15 minutes. PANTHERS /PCC transient condenser performance tests are used to establish noncondensible buildup effects and PCC pool water level effects. They are not intended to be integral systems tests. The noncondensible build-up tests are performed as follows. The test conditions are initialized, using the steam-only procedure describedin the steady state test section. When steady state conditions are established, the data acquisition system is staned, and air, helium, or an air / helium mixture is injected at the rate specified. The inlet pressure is allowed to increase as the noncondensibles collect in the vent tank, and the condensation process is degraded by the presence of noncondensibles in the PCC heat exchanger. The test is terminated when the PCC heat exchanger reaches its design pressure. For the pool water level tests, the procedure is to establish the initial conditions as described in the steady state air / steam mixture tests, then to initiate data acquisition. With the position of the vent flow control valve fixed, the PCC pool water level is allowed to decrease by either boil-off, draining, or a combination of the two. Inlet pressure to the PCC is allowed to rise, consistent with l the condensation process. The test is concluded when the desired pool water level range has been investigated. A.3.1.1.2 Test Objectives The test objectives of the PANTHERS /PCC Test Program are:
- 1. Demonstrate that the prototype PCC heat exchanger is capable of meeting its design requirements for heat rejection. (Component Performance) l 2. Provide a sufficient data base to confirm the adequacy of TRACG to predict the quasi-steady-heat rejection performance of a prototype PCC heat exchanger, over a range of air flow rates, steam flow rates, operating pressures, and superheat conditions, that span and bound the SBWR range. (Steady State Separate Effects)
- 3. Determine and quantify any differences in the effects of noncondensible buildup in the PCC heat exchanger tubes between lighter-than-steam and heavier-than-steam gases.
(Concept Demonstration) A.3.1.1.3 Test Matrix and Data Analysis Steady State Performance Tests
- Table A.3-2a shows the PANTHERS /PCC Steady State Performance Matrix for Steam-Only Tests. Thineen test conditions are included.
l A-5
-- - -~ , . - .. - . ___ - - - _ - - .- _- -_
d l- NEDO-32391, Revision C
- Test Conditions 37 through 43 (Test Group P1) are used to determine the baseline heat exchanger performance over a range of saturated steam flow rates without the presence ,
of noncondensible gases. Test Group P1 data are compared with design requirements i to meet Test Objective 1. Test Conditions 44 thmugh 49 (Test Group P2) address the effect of superheat conditions in the inlet steam. Test Conditions 38, 44, 45, and 46 , may be used to establish the effects of superheat at a relativelylow steam flow , condition, while Test Conditions 41,47,48, and 49 will give the same information at a L steam flow rate near rated conditions. 1
- s ; Table A.3-2b shows the PANTHERS /PCC Steady State Performar.ce Matrix for Air / Steam l Mixture Tests. As noted previously, the independent variables are steam mass flow rate, air mass ;
flow rate, steam superheat conditions, and absolute operating pressum. Figure A.3-4 shows the - i
- relationship between the steam and air flow rates specified for PANTHERS /PCC testing and the l SBWR expected range.
- Test Conditions 9,15,18, and 23 (TestGroup P3) will be used to compare heat rejection rates over a range of air flow rates to the saturated, steam-only condition determined from Test Condition 41 in the pure steam series. Holding steam flow constant at near rated conditions, these tests yield the effect of air on the condensation i
process. !
- Test Conditions 2,13,16,17,19,22, and 25 (Test Group P4) supplement Test Group :
P3, in that they define condensation performance at the extremes of the SBWR air / steam mixture ranges, and at several intermediate points. These tests will be used to quantify noncondensible effects at off rated conditions. They will be compared to the appropriate Test Conditions in the P1 group. i
- Test Conditions 35 and 36 (Test Group PS) funher supplement Test Group P4 by :'
extending the effect of noncondensible gases over the superheated steam range. These tests can be compared to Test Conditions 48 and 49 to establish the effect of air content at the same superheat condition, and to Test Condition 23 at the same air flow, but with saturated steam.
- Test Conditions 1,3,4,5,6,7,8,10,11,12,14, 20, 21, and 24. (Test Group P6) am :
lower priority tests. They are run at only a singleinlet pressure to supplement the previously identified tests by increasing the data density within the already established air / steam flow map. l l l Transient Test Conditions l Table A.3-2c shows the PANTHERS /PCC Noncondensible Buildup Test Matrix. Eighttest i conditions are specified as Test Gmup P7. In these tests, steam is supplied at a constant rate, and , l steady state conditions are established in a manner similar to that of the steady state performance . tests. Air, helium, or air / helium mixtures are then injected into the steam supply, with the vent line closed. The transient degradation in heat transfer performance will be measured, as a function of , the total noncondensible mass' injected. !
- Tests Conditions 50 and-51 provide a baseline condition with air as the only noncondensible. Air is similar to nitrogen in molecular weight, and is heavier than i A-6 I l
_ ~ _
l NEDO-32391, Revision C steam. Test Conditions 52 and 53 are similar to Test Conditions 50 and 51, but with the , steam supply superheated. Test Conditions 75 and 76 mpeat Test Conditions 50 and l 51, but utilize helium as the noncondensible gas instead of air. Helium is lighter than l steam, and will mix in a manner similar to hydrogen. The results of Tests 50 through l 53 plus 75 and 76 can be compared to establish performance diffemnces between j lighter-than-steam and heavier-than-steam gases as they build up in the heat exchanger j tubes. Test Conditions 77 and 78 can be used to evaluate the effect of an air and helium ) mixtum concurrently flowing into the heat exchanger. ;
- Test Group P7 data will be evaluated to meet the requirements of Test Objective 3.
Table A.3-2d shows the PANTHERS /PCC Pool Water level Effect Test Matrix. Eme test conditions are specified as Test Group P8. In these tests, steam and air / steam mixtures are supplied to the PCC heat exchanger, and steady state conditions established in a manner similar to the steady state performance tests. In these tests, however, the waterlevel in the PCC pool is allowed to drop and the PCC tubes to uncover. Both the PCC poollevel and the PCC heat rejection rate are monitored as a function of time.
- Test Conditions 54,55, and 56 establish the effect of waterlevel in the PCC pool for a range of steam and air / steam supply rates to the PCC. Data from Test Conditions 54, 55, and 56 can be compamd to Test Conditions 41,15, and 25, respectively, to obtain the effect oflowered water level on condensation performance. Test Conditions 54 and 55 can be compared to establish the effect of air content on the rate of pool boiloff.
- Test Groups P1 through P5, P7 and P8 provide a data base for TRACG qualification and meet Test Objective 2.
A.3.1.1.4 Justification of Test Conditions PCC Operation In the SBWR, the post-LOCA function of the PCC heat exchanger is to remove decay heat from the drywell and reject this energy to the atmosphere. This is the major difference between the
- SBWR and earlier pressure suppression containment designs. In earlier designs, the decay heat is transfered from the drywell to the wetwell via themain vent flow, where the energy is
% subsequently transferred to the ultimate heat sink by the Residual Hest Removal (RHR) system.
As in previous pressure suppression containment designs, the maximum drywell pressure is limited to the wetwell pressure plus the vent submergence head and any vent system flow losses. During a LOCA in the SBWR, the PCCS and the GDCS form a loop to keep the core covered with water and remove the decay heat. Steam coming off the core, leaves the RPV through the DPVs, enters the drywell, and flows to the PCCS. Condensate flow from the PCCS heat exchangers goes to the GDCS poolin the drywell. The GDCS delivers the water to the RPV where the decay heat of the core converts it to steam and starts the loop again. PCC Operational Modes he operational modes of the PCC heat exchanger can best be described in terms of the pressure difference across the unit. l A-7
+ - - - _ _ _ . - .
I l NEDO-32391 Revision C Figure A.3-3 illustrates several of a family of possible pressures along the flow path from the drywell to the suppression pool via the PCC heat exchanger. Note that on the drywell side, the pressum difference can vary only between that required to open the vacuum breaker and that required to open the main vent. Reference LOCA Condition - Curve 1 illustrates the SBWR post-LOCA condition with the PCC canying the decay . heat load. In this case, the drywellpressure is slightly greater than the PCC vent submergence pressure, but less than the LOCA vent submergence pressure. Thus water is forced out of the PCC vent line, clearing a gas venting path to the suppression pool. The - flow is forced through the PCC heat exchanger by the drywell to wetwell pressu e difference, and noncondensibles are vented into the suppression pool. PCC Capacity Greater Than The Decay Heat - Curves 2 and 3 of Figure A.3-3 illustrate a situation where most of the noncondensibles have been vented to the wetwell. These two curvesillustate two cases where the drywellis supplying nearly pure steam to the heat exchanger: Curve 3 has less noncondensibles than Curve 2. As the effects of noncondensibles degrading the heat transfer pmcess are reduced, the heat exchanger can reject more energy than is supplied to the drywell by decay heat, and the drywell pressure is reduced. The reduced pressure is no longer capable of keeping the PCC vent open, so suppression pool water panially refills the PCC vent pipe. The flow into the PCC heat exchanger is no longer driven by the drywell-to-werwell pressure difference, but by the lowered prnsare in the heat exchanger tubes due to the condensation process. The limit of this type of operation is shown on Curve 4, where the drywell pressure has fallen to below the wetwell pressure by an amountequal to the vacuum breaker opening pressure. Here, the vacuum breaker opens, returning noncondensibles to the drywell to re-enter the PCCS. 'Ihe capacity of thePCC to remove energy is temporarily degraded. PCC Capacity Less Than The Decay Heat - ,. Finally, Curve 5 of Figure A.3-3 illustrates the other extreme of PCC operation. In this case, the PCC cannot remove sufficient heat to reject the decay heat, and the drywell pressure rises. Again, flow is forced through the PCC by the drywell-to-wetwell pressure difference. However, the magnitude of the PCC driving pressure difference is limited by the presence of the main LOCA vents. If the main LOCA vents clear, then mass and energy will flow to the suppression pool via the main vent system and limit the drywell pressure. This pressure difference also determines flow through the PCC heat exchanger. In summary, there are two possible operating modes for the PCC heat exchanger: (1) a pressure drop driven mode, when the PCC vent is cleared of water, and flow is typically a mixture of steam and noncondensibles; and (2) a condensation pressure driven mode, when the PCC vent is partially filled with water, and the flow is nearly free of noncondensibles. These PCC operational modes are summarized below: l A-8
l NEDO-32391, Revision C ; i
. 1. - Pressure Drop Driven Mode PCC capacity 5 core decay heat I -
PCC flow is forced by the DW to WW pressure difference PCC flow is a rich mixture of both steam and noncondensible gas j 2. Condensation Pressure Driven Mode u PCC capacity 2 core decay heat i - PCC flow is induced by DW to PCC-Hx outlet pressure difference due to ] condensation 4 4 - PCC flow is rich in steam, but is lean in noncondensible gas
- PCC Purge and Vent Process
! A PCCS purge event can occur as a result of the system being called upon to remove decay ] heat after an extended period of inactivity or by an increase in the mass fraction of l noncondensible gas in the region of the drywell from which the system draws its inlet mixture. If the system is staning up after a period of inactivity, the condensers will contain a mixture of i ] steam and noncondensible gas in near thermal equilibrium with the sunounding po 1. The partial , j pressure of the steam will be approximately saturation pressure at the pool temperature and the ) 4 remainder of the mixture will be noncondensible gas. This mixture must be expelled from the j condensers before heat removal can begin. As steam is added to the drywell by the RPV, the drywell pressure will rise until the PCCS vents are cleared andthe initial steam /noncondensible inventory of the condensers is vented. The movement of the initial inventory out of the condensers will be accompanied by ingestion of a fresh steam /noncondensible mixtare at the 2 existing drywell conditions in the neighborhood of the PCCS inlets. Depending upon the fraction of noncondensible in the inlet mixture, and the decay power, the system may or may not be able to condense steam at the rate it is being added to the drywell by the RPV. Consider, first, the case where the PCCS heat removal rate at the existing inlet conditions is
- less than decay power. The situation is the same whether the PCCS is staning up from a period of inactivity or, while operating, is confronted with an increased noncondensible fraction in the
% inlet mixture. The drywell pressure will rise, thereby increasing the flow rate through the condensers from the drywell to the wetwell. The rise in drywell pressure also slightly increases } the condensation rate. Additionally, as steam is continuously added to the drywell by the RPV, 5 and a steam / gas mixture is transported through the condensers, the mass fraction of ! noncondensible in the inlet mixture will stan to decrease. At some point, the combination of j l increasing drywell pressure and decreasing noncondensible inlet mass fraction enables the PCCS i
- ' heat removal rate to match decay power and the drywell pressure stops rising. j Next, consider the case where PCCS heat removal rate at the existing inlet conditions is greater than decay power. Again, the situation is the same whether the PCCS is staning up from l a period ofinactivity or, while operating,is confronted with a decreased noncondensible fraction in theinlet mixture. The drywell pressure will start to drop, allowing water to reenter and close
- ' the vents. Unless the inlet conditions are pure steam, the PCCS will then start to accumulate noncondensible gas. Gas can accumulate in the vent pipes above the waterlevel, in the headers, l A-9 4
l NEDD-32391, Revision C and in the condenser tubes. The combination of accumulating noncondensible and, to a lesser extent, decreasing drywell pressure results in a decreasing condensation rate. Eventually, the condensation rate will drop below decay power and the drywell pressure will start to rise, initiating a new purge cycle. The presence of vacuum breakers in the SBWR leads to a potential interaction between PCCS purging and vacuum breaker operation. As discussed above, if the instantaneous PCCS heat removal exceeds decay power, the drywell pressure will decrease. When the difference between DW and WW pressure drops below the submergence head of the PCCS vents, water will - enter the vents and noncondensible will start to accumulate in the PCCS. The drywell pressure will continue to decrease until the combination of the lower pressure and the noncondensible accumulation drops the PCCS heat removal rate below decay power. If the PCCS noncondensible - inventory when the vents close is relatively small, and the mass fraction of noncondensible in the inlet mixture is also small, the drywell pressure can continue to fall until it drops below the , wetwell pressure by a suflicient amount to allow the vacuum breakers to open. The noncondensible which flows back to the drywell via the vacuum breakers increases the mass fraction of noncondensible in the inlet mixture, degrades condenser performance and leads to a new purge cycle. Thus, it can be seen that, depending on the attendant circumstances, a PCCS purge event may or may not lead to a vacuum breaker cpening. In discussing these two possibilities, GE has introduced the nomenclature " strong purge" to identify a purge event which leads to opening of a vacuum breaker and " weak purge" to identify one which does not. PANTHERS /PCC Operation PANTHERS /PCC matches the behavior of the SBWR unit. Steam for the tests comes from the neighboring power plant. Instead of a GDCS, the facility has a condensate tank which collects the condensate from the PCC and returns it to the power plant. The boundary conditions on the condensate line match that for SBWR. The water level in the condensate tank is held to the same water level as found in the GDCS pool of SBWR. The pressure in the tank is equal to the steam inlet pressure which is the same as the SBWR where the pressure above the GDCS pool is the drywell pressure. The PCC vent configuration differs among the types of tests and is discussed below. Steady State Tests The independent variables for the PANTHERS /PCC steady state tests are steam flow rate, air flow rate, and PCC inlet pressure. The design basis of the Passive Containment Cooling System (3 heat exchangers) provides the ability to reject allSBWR decay heat at approximately one hour post-LOCA. - Figure A.3-4 compares the range of test conditions for PANTHERS /PCC with the air and steam flow conditions for the SBWR main steamline and GDCS line break scenarios after one hour into a LOCA. The triangles representing the two breaks are constructed as the intersection of a vertical line, bounding the maximum steam flow, and a line drawn from the origin with a slope sufficient to envelope the calculated steam and noncondensible flow rates. The triangles are not one-to-one maps or time histories but, rather, bounds of the steam /noncondensible gas inlet conditions throughout the calculated SBWR LOCA scenario.
-l A-10
,~ -- . - - - ---- . _ _ - - - . . - - ~ - - . . - _ - 4 l
- l NEDO-32391, Revision C i
The triangles can be used to explain the progression of inlet conditions as the transient
- proceeds. This progression starts at the origin. In the periodimmediately following one hour, j i
subcooled GDCS water is absorbing the decay heat power and there is no flow to the PCCS. This i period is represented by the region near the origin. When the RPV water again reaches saturation, ! flow to the PCCS resumes and, at first, follows an approximately linear steam /noncondensible gas i flow trajectory corresponding to the noncondensible gas mass fraction in the region of the drywell ] which feeds the PCCS. This initiates the purging process which transports the drywell l l, noncondensible gas to the wetwell via the PCC units. At some point during the purge, the concentration of noncondensible gas in the drywell starts to drop and the steam /noncondensible 4 gas trajectory turns over. The steam flow continues to increase as the noncondensible gas l concentration in the inlet mixture decreases. The end of the purging process is represented by the extreme lower right corner of the triangle. The steam flow has now increased to its maximum l
- value (matching the decay power) and the noncondensible gas flow has dropped toessentially j j zero. From this point, the steam flow " walks" backward along its axis as the decay power slowly !
drops. l l l The difference in the steam /noncondensible gas envelopes for the GDCS line and main 1
- steamline break accident scenarios results from the behavior during the GDCS injection phase of the transient. For the main steamline break, a large fraction of the subcooled GDCS water is
- retained in the pools as the RPV two-phase water level rapidly recovers to the main steamline l elevation and equilibrates with the water in theGDCS pools. There is no vacuum breaker
- activation and, accordingly, there is a small noncondensible gas fraction in the drywell at the initiation of PCCS flow. PCCS flow initiates about one hour from the instant of LOCA at a decay
{ ! power close to its rated heat removal capacity and, as a result of the low drywell noncondensible ! gas inventory, it rises to match decay power relatively rapidly. For the GDCS break, the pools 1 } drain completely and RPV steaming does not resume until about 2.5 hours from the LOCA. During the GDCS injection period, there are multiple vacuum breaker activations leading to a
- l. relatively large noncondensible gas fraction in the drywell when PCCS flow initiates. The larger l PCCS steam flow does not match decay power until about 3.5 hours from the LOCA. PCCS
{ steam flow is significantly reduced from the main steamline break case. A bottom drain line break j would behave similarly to the GDCS line break. ! Them is no precise relationshio between Figures A.3-4 and A.3-3. Figure A.3-3 illustrates various flow conditions which can exist in the PCCS circuit between the drywell and the we well. is The purpose of Figure A.3-4 is to show that SBWR conditions are covered by the PANTHERS i test matrix. Curves 1 through 4 in Figure A.3-3 describe a sequence of conditions which can follow the initialascension of the PCCS heat removal to match decay power. At the end of the
;' purge, drywell pressure exceeds wetwell pressure by slightly more than the head required to clear
+ the PCCS vents. If the purging process has left the condensers in a relatively noncondensible gas-1 free condition (a " strong" purge), the drywell pressure will then start to fall as the PCCS removes slightly more than the decay power. Eventually, as it moves through the sequence from curves 1 to 4, the drywell pressuit will decrease to the point where the vacuum breakers open. This will reintroduce noncondensible gas to the drywell and drive the PCCS inlet conditions to the left and upwanis in the triangular regions of Figure A.3-4. Calculations performed to date have given no indication that this would lead to a " penetration" of the triangle boundary defined by the initial
. purge.
1 l A-l l
t i l NEDO-32391, Revision C 4 From Figure A.3-4, the test conditions clearly bound the possible SBWR range of air / steam and noncondensible flows. The third independent variable, PCC inlet pressure, is notindicated on this figure, but is shown for the various tests in Table A.3-2b. For this same time frame, the SBWR would be expected to have a PCC operating pressure near 300 kPa. Test Groups P3, P4, j and P5 typically have data taken at five pressures, ranging from 200 to 500 kPa, with one pressure j near the 300 kPa nominal value. All Test Group P6 data points are taken at a nominal PCC inlet pressure of 300 kPa, consistent with the P6 goal of increasing the data density near the post-LOCA SBWR operating conditions. l , i l As noted in the previous discussion of operating modes, the pressure drop from the drywell ! l (through the PCC heat exchanger to the PCC vent exit) cannot exceed a value equivalent to the ! a dificience between the main IDCA vent submergence and the PCCS vent submergence. The - j PCC pressure drop is one of the dependent variables measured during the testing. On the basis of ; this data, it is possible to establish the maximum flow rate through the PCC, independent of the !
- timeinto a postulated IDCA scenario. This is the basis for using the PANTHERS /PCC data to j 1
qualify TRACG for application at times earlier than one hour post-LOCA. l Transient Tests
- Transient tests are performed to assess two phenomena: the buildup of noncondensibles in i the heat exchanger, and the reduction of PCC pool waterlevel as the inventory is boiled away. In the noncondensible case, air and helium, representing heavier-than-steam and lighter-than-steam gases, are introduced atlow ,olume flowrates; the flow rate is low enough such that the performance may be considered quasi-steady. The tests begin with pure steam condensation and noncondensibles are added until condensation is essentiallystopped. Thus, the tests cover the 3 entire potential range of PCC operation from the standpoint of noncondensible inventory in the condenser. In the water level tests, through a combination of normal boil-off and draining of the 1
pool, the PCC poollevelis lowezed through a range that exceeds the SBWR inventory loss over a 72 hour period. Hence, both transient test types cover the entire applicable SBWR range. l l Pressure Drop Driven and Condensation Pressure Driven Modes As noted in the operational modes discussion, the PCC can perform in two modes: pressure drop driven and condensation pressure driven. Both of these conditions are simulated in the
- PANTHERS /PCC steady state tests.
The pure steam tests, Test Conditions 37 through 49 (Test Groups P1 and P2)are all .- performed with the PCC vent closed. Since there is no vent flow through the heat exchanger, all the steam is condensed within the PCC and steam is drawninto the heat exchanger by the , condensation process. These tests simulate the condensation pressure driven mode. . i In the remaining air / steam mixture tests, Test Groups P3 through P6, the PCC vent is open, and both the inlet flow rate and vent tank pressure are controlled. These tests duplicate the pressure drop driven mode. In this case there is flow through the heat exchanger,with the flow rate determined by the difference in pressure between the inlet supply and the vent tank. 1 t l A-12 1
}
l NEDO-32391, Revision C l A.3.1.1.5 TRACG Analysis Plan Table A.3-3 lists those PANTHERS /PCC tests that will be analyzed with TR ACG. Fifteen TRACG runs are included in this group, which is intended to demonstrate the capability of TRACG to predict the heat rejection rate of the PCC heat exchanger over a wide range j of conditions. The focus will be on rated conditions, with the qualification points also established j near the extremes of the SBWR range. Twelve of the qualification data points come from the ) l steady state performance test matrix (Test Groups P1 through P5), and the remaining three from 1 the transient group (two from P7 and one from P8). l l Figure A.3-5 illustrates the locations of the ten saturated condition steady state TRACG 1
. qualification points within the overall PANTHERS /PCC steady state test performance test matrix.
The remaining two conditions are superheated, and cannot be shown on this figure. ; i Analysis results will be compared with test data as defined in Table A.3-3. For the steady 1 state saturated and superheated steam conditions, the assessment of adequacy will be made on the l basis of total heat rejection rate and PCC pressure drop. For air / steam and helium / steam 1 mixtures, the degradation factor, defined as the ratio of the heat rejection rate in the noncondensible case to that in the pure steam case, will be the figure of merit. The air / steam mixture data are taken at five different pressures. The degradation factor will be based on the air / steam mixture case having the absolute pressure nearest to the pure steam case: 1 Pure Steam Condensation - Analysis of Test Conditions 41 and 43 demonstrates TRACG capability to predict pure saturated steam condensation rates at and above rated conditions. Test Condition 49 addresses superheat in this state. Air / Steam Mixtures- Analysis of Test Conditions 9,15,18, and 23 addresses the effects of noncondensible mass fraction at rated steam flow conditions, over the complete range of potentialair fractions. Test Conditions 2 and 22 address the effects of air in thelow steam flow range, but at the limits of air flows. Test Conditions 17 and 19 are in the intermediate range. Test Condition 35 addresses superheat effects. l Noncondensible Density - Analysis of Test Conditions 51, 76, and 78 addresses the buildup of noncondensibles in the PCC tubes, and will be predicted on a transient basis. Test Condition 51 uses air, Test Condition 76 uses helium, and Test Condition 78 uses both i helium and air. PCC Pool Level- Transient analysis of Test Condition 55 addresses the capability of TRACG to predict the effects of PCC pool water level. A.3.1.2 PANTHERS /IC A.3.1.2.1 Test Description Overview PANTHERS /IC (Isolation Condenser) testing is performed at Societa Informazioni
' Esperienze Termoidrauliche (SIET)in Piacenza, Italy. The tests areperformed in the same facility used for the PANTHERS /PCC program, but using several pieces of different equipment, in order l to better simulate the performance environment of the IC. For the IC testing, the facility consists
- l A-13
l l NEDO-32391, Revision C of a prototype IC module, a steam supply vessel which simulates the SBWR reactor vessel, a vent volume, and associated piping sufficient to establish IC thermal-hydraulic performance. Both j thermal-hydraulic and component demonstration tests are performed during these tests. This , l section covers the thermal-hydraulic portion of the testing; component structural performance tests l l are covered in Subsection A.3.2.2. The IC being tested is one module of a full-scale, two-module vertical tube heat exchanger designed and built by Ansaldo. Only one module unit is being tested because of the much higher energy rejection rate of the IC relative to the PCC unit, and inherent limitations of facility and . steam supply size. Figure A.3-6 gives an outline drawing of the heat exchangerassembly. Like the PCC unit, the IC is a prototype unit, built to prototype procedures and using prototype materials. Six modules (three heat exchanger units) of the type being tested are used in the - SBWR. The IC is installed in a water pool having one half the appmpriate volume for one SBWR IC assembly. Instrumentation Figure A.3-7 is a schematic of the PANTHERS /IC facility. The primary instrumentation specified is sufficient to ascertain heat exchanger thermal-hydraulic performance by performing mass and energy balances on the facility. Table A.3-4 defines the thermal-hydraulic measurements taken during the IC tests. Like the PCC testing, all test instrumentation is calibrated against standards equivalem to the U.S. National Institute of Standards and Technology. References 51 and 53 contain information on the IC instmmentation, as well as the PCC instrumentation. Additional information may be found in the PANTHER /IC Test Plan (Reference 80) and Test Procedures (Reference 81). Test Method PANTHERS /IC testing procedures are specific to the type of test being performed. In general, however, the procedure for the steady state tests will be as follows: The steam vessel and IC heat exchanger will be purged of initial air in a manner similar to that done with the PCC heat exchanger. The IC pressure will be at the design pressure or a lower value, depending on whether the test is also being used as a structural demonstration cycle. Subsection A.3.2.2 describes the PANTHERS /IC structural demonstration cycles. The IC is placed in operation by opening the IC drain valve. Steam supply to the steam vessel is then regulated such that the vessel pressure stabilizes at the desired value. Data will be acquired for a period of approximately 15 minutes. At this point, the steam supply can be increased or decreased to gather data at a different operating pressure, or testing may be terminated. In all cases, flowinto the IC will be natural circulation driven, as is the case for the SBWR. Noncondensible gas effects tests begin similarly until the point where pressure is stabilized at the desired value. For this case, a mixture of nitrogen and helium is injected into the IC supply line at a very low flow rate. The ratio of nitrogen to helium in the injectedflow will be 3.5:1, simulating the composition of radiolytic gases. Gas injection will continue until the IC inlet pressure increases to 7.653 MPag (1110 psig). The noncondensible flow rate is approximately 3 to 5 g/s. The lower IC vent is then opened, and the IC vented until the pressure returns to the initial operating pressure, or stabilizes at an intermediate value. If the pressure returns to itsinitial value, the test is terminated. If the inlet pressure stabilizes, the IC top vent will be opened, and the l A-14
l NEDO-32391, Revision C i l performance monitored until venting is complete, and the inlet pressure returns to the initial value. l The test is then terminated. Water level tests also begin with the IC in stable operation at the desired initial inlet pressure. The IC pool water level is then reduced and the IC perfonnance monitored. Water level will be l reduced until the IC inlet pressure reaches 8.618 MPag (1250 psig). The pool water level willthen be increased to normal and IC performance allowed to return to normal. 'Ihe test is then tenninated. A.3.1.2.2 Test Objectives The objectives of the PANTHERSAC Test Program are: j
- 1. Demonstrate that the prototype IC heat exchanger is capable of meeting its design requirements for heat rejection. (Component Performance)
- 2. Provide a sufficient data base to confirm the adequacy of TRACG to predict the quasi-steady heat rejection performance of a prototype IC heat exchanger, over a range of l operating pressures that span and bound the SBWR range. (Steady State Separate Effects)
- 3. Demonstrate the stanup of the IC unit under accident conditions. (Concept Demonstration)
- 4. Demonstrate the capability of the ICC design to vent noncondensibles and to resume condensation following venting. (ConceptDemonstration)
A.3.1.2.3 Test Matrix and Data Analysis Steady State Performance Tests As for the PANTHERS /PCC tests, the majority of the IC tests are steady state performance tests. Table A.3-5a provides the PANTHERSSC Steady State Performance Test Matrix. A total of ten test conditions are specified. Test Conditions 2 through 11 are identified as TestGroup II. These data will establish the IC heat rejection rate as a function ofinlet pressure. s Transient Test Conditions PANTHERSAC transient tests will demonstrate startup of the IC heat exchanger for full-scale thermodynamic conditions. These tests are designed to demonstrate heat exchanger performance; they are not intended to be integral systems tests. l Tables A.3-5b through A.3-5d give the PANTHERSAC Transient Demonstration Test Matrix. Five Test Conditions are specified. Test Condition 1 (TestGroup 12) is a set of two duplicate tests designed to demonstrate the stanup and operation of the IC in a situation comparable to a reactorisolation and trip (Table A.3-5b). This is a type 2 test as shown in Figure A.3-23. Test Conditions 12 and 13 (Test Group 13) will have an air / helium mixture injected slowly after the steam vessel pressure has been reduced to the value specified as " inlet pressure" in l Table A.3-5c. The IC will be vented when the inlet pressure reaches 7.653 MPag (1110 psig) or when the pressure peaks,if at a lower value. Re-establishment of condensation following venting l A-15
l NEDO-32391, Revision C will be recorded. Test Conditions 14 and 15 (Test Group I4) are repeats of Test Conditions 12 and l 13, but with the waterlevelin the IC pool allowed to drop, exposing the IC tubes (Table A.3-5d). Both the IC pool level rate and the IC heat rejection rate will be monitored as a function of time.
- Test Group 12 will demonstrate stanup of the IC under near prototype conditions, pmvide heat rejection data at a higher pressure than the data from Test Group II, and demonstrate test repeatability. Test Conditions 12 and 13 will demonstrate restart of condensation in the IC following venting noncondensible. Test Conditions 14 and 15 will establish the degradation of hnt rejection ability of the IC as the IC pool water -
level decreases.
- Test Groups Il and 12 will becompared with design requirements to meet Test ,
Objective 1.
- Test Groups II,12, and I4 provide a data base for TRACG qualification and meetsTest Objective 2.
- Test Group 12 demonstrates restart of the IC and meets Test Objective 3.
A.3.1.2.4 Justification of Test Conditions Steady State Tests The independent variable for the PANTHERS /IC steady state tests is the isolation condenser inlet pressure, which is equal to the steam vessel pressure. The isolation condenser is a natural circulation unit. He IC inlet pressures to be tested shown in Table A.3-5a span the entire operating range of l the SBWR The SBWR range is bounded by the SRV setpoints at 7.920 MPag (1150 psig) and the vessel depressurized state. This is consistent with the test pressures. Transient Tests The transient test independent variables are IC inlet pressure, total noncondensible gas a.ided, l and IC pool waterlevel. IC inlet pressures chosen are 0.48 MPag (70 psig) and 2.07 MPag (300 psig). Rese conditions were chosen because they represent typical non-LOCA operating conditions where an operator might have the IC in service. The ratio of air to helium in the injected gas was chosen to be representative of the oxygen to hydrogen ratio due to radiolytic - decomposition of water in the SBWR core. While the injection rate has not been determined at this time,it will be chosen such that quasi-steady operation of the heat exchanger occurs. For the pool water level tests, water levels at least as low as mid-height of the condenser tubes is specified provided the design pressure isn't exceeded. Bounding calculations based on decay heat rejection indicate that no more than one-third of the tubes may be uncovered during the 72 hour post scram period. Consequently, the defined testingbounds the SBWR range of conditions. t The PANTHERS /IC tests are component, not system, tests. The purpose of the transient tests is to measure the change in performance of the IC with (a) a known quantity of non-condensible gas present or (b) a change in pool water level. Although the test facility is similar to ' the arrangement found in the SBWR with the steam supply and condensate retum line connected 'l A-16 r
I l NEDO-32391, Revision C to a large pressure vessel, the transient tests do not exactly match the system performance an IC would experience in the SBWR. For example, the operation of the heat exchanger in PANTHERS , differs from the conditions it will encounterin the plant; i.e., steam and non-condensible gases are
" metered"into the test facility, while in the plant the conditions at theinlet of the heat exchanger z depend on the conditions in the RPV, and are not independent variables. However, the venting of the test unit will closely match the performance of the plant unit, and demonstrate that the IC can j vent the gases and resume condensation. {
A.3.1.2.5 TRACG Analysis Plan i
. Table A.3-6 lists those PANTHERS /IC tests that will be analyzed with TRACG. Six j TRACG runs are included inthis group, which is intended to demonstrate the capability of ;
TRACG to predict the heat rejection rate of the IC heat exchanger over the range of reactor l pressures where it will be expected to perform. Three of the six points come from the steady state performance test matrix (Test Group I2), with the remaining three points coming from the transient data set. Analysis will be compared with test data as defined in Table A.3-6. In all cases, the primary comparison will be on the tota! heat rejection rate. Additionally, for the transient cases, IC inlet pressure will be compared as a function of time: Pure Steam Condensation - Analysis of Test Conditions 2, 6, and 11 demonstates TRACG capability to predict pure steam IC condensation rates over the expectedSBWR operating range (7.92 to 0.21 MPag) (1150 to 30 psig). Noncondensible Buildup and Venting - Analysis of Test Conditions 12 and 13 demonstrates TRACG capability to predict the effect of noncondensible buildup in degradation of the overall heat transfer capability of the IC, including re-establishment of steam-only condensation following venting. IC PoolInet Effects - Analysis of Test Conditions 15 demonstrates TRACG capability to predict the effect of pool level on the degradation of IC performance. A.3.1.3 PANDA A.3.1.3.1 Test Description Overview PANDA is a large-scale integrated SBWR containment experiment that will be performed by the Paul Scherrer Institut in Wuerenlingen, Switzerland. The test facility is an approximately 1/25 volumetric, full scale height simulation of the SBWR containment system. Pressure vessels representing the reactor pressure vessel, drywell, wetwell and wetwell air space, and GDCS pool are interconnected with appropriate piping in order to simulate a variety of containment transients. The facility is equipped with three scaled PCC heat exchangers and one isolation condenser unit, each with its own water pool. De PCC and IC units are both scaled by holding the heat transfer tubes at full size, but reduced in number from the prototype. The configuration of the IC and PCC j units is illustrated on Figure A.3-8. The reactor pressure vessel volume is equipped with electrical j l A-17
l NEDO-32391, Revision C heaters to simulate decay heat and thermalcapacitance of the vessel and internals. The facility is capable of simulating SBWR accident scenarios starting approximately one hour into the LOCA. Figures A.3-9 and A.3-10 show a schematic of the PANDA test facihty and the arrangement ! of the PANDA test vessels, respectively. Two interconnected vessels are used forthe drywell and , wetwell volumes in order to simulate potential asymmetric effects. In addition to its transient capabilities, PANDA also has temporary piping connections such that a PCC heat exchanger may be tested in a quasi-steady manner. In this case, a connection is , made from the IC piping supply line to the inlet of PCC3. Steam can then flow directly from the RPV to PCC3, bypassing the drywells. PCC3 will vent to the wetwell and condensate will return i to the GDCS tank, using the normal piping arrangement. The temporary supply piping , arrangement is shown in Figure A.3-11. i Instrumentation The PANDA data acquisition system is capable of recording up to 720 channels with each channel mcorded once every two seconds. For the PANDA tests, 598 channels have been , assigned. The instrumentation is summarized in Table A.3-7, with approximate locations given in i Figures A.3-13a thmugh A.313d. Test instrumentation is calibrated against standards equivalent to the U.S. National Institute of Standards and Technology. Additional information may be found in the PANDA Test Specification, Reference 54, and in the Test Plan and Procedure, Reference 55. For the steady state PCC perfonnance tests, only a subset of the PANDA instrumentation is
- required. This subset of the instrumentation is defined in Table A.3-8; locations are shown on ,
j Figure A.3-14. l Test Method { 4 Steady state Tests to demonstrate PCC performance will be the first tests l'erformed in the i PANDA facility. For these tests, the facility will be configured as described above and as shown schematically in Figure A.3-14. t The facility will be preconditioned for testing using the electrical heaters in the RPV for the heat source. The RPV will be filled with water to an appropriate level above the top of the heaters,
- and the heaters turned on. Once the water has been heated to saturation conditions, the RPV can be
! used to provide steam for heating of the other PANDA vessels. In addition, the hot water in the ,, RPV will be used to heat waterin the auxiliary water system. Then the steam, hot water, and/or air . from the auxiliary water and air systems will be used to separately bring the GDCS tank, werwell . vessels, PCC3, and PCC3 pool to the desired pressures and temperatures. ..
, Once the desired conditions are achieved in each vessel, the appropriate connecting lines will 1
be opened, and the steam and air flow will be directed to PCC3. The power to the RPV heaters and the flow fmm the auxiliary air supply will be adjusted to obtain the desired steam andair flow rates, respectively. For the tests with no air flow, the PCC3 vent line will be closed and the j l condenser pressure will be allowed to come to the steady state equilibrium value consistent with the specified steam flow rate. , ! After steady state conditions have been achieved, the test will be initiated and the data will be f , recorded for a period of at least 15 minutes. The test will then be tenninated. ) l A-18 [
l NEDO-32391, Revision C Test procedures for the transient matrix tests have not yet been pmpared. It is anticipated that facility pre-conditioning to establish the initial conditions for the transient tests will be similar to l that described for the steady state tests in the preceding paragraphs. Once the initial conditions for a given test have been established, all control (except for the decay of RPV power) will be terminated, and the PANDA containment will be allowed to function without operator intervention, mirroring the SSAR assumptions for the SBWR. Details will be submitted in the Test Plan and Procedure for these tests. l A.3.1.3.2 Test Objectives
- . The test objectives of the PANDA Test Program are:
4
- 1. Provide additional data to: (a) support the adequacy of TRACG to predict the quasi-steady heat rejection rate of a PCC heat exchanger, and (b) identify the effects of scale on PCC performance. (Steady State Separate Effects)
- 2. Provide a sufficient data base to confirm the capability ofTRACG to predict SBWR a containment system performance, including potential systems interaction effects.
, (IntegralSystems Tests)
- 3. Demonstrate startup and long-term operation of a passive containment cooling system.
(Concept Demonstration) A.3.1.3.3 Test Matrix and Data Analysis Steady State Performance Tests A series of steady state tests will be conducted using one of the PANDA PCC condensers. As noted in the test method section, the facility will be configumd to inject known flow rates of saturated steam and air directly to the PCCS heat exchanger. The condenser inlet pressure will be maintained at approximately 300 kPa for tests with air injection by controlling the wetwell pressure. For tests with pure steam flow the condenser pressure will be allowed to come to the steady state equilibrium value consistent with the specified steam flow rate. The steam and air flow to the heat exchanger will be controlled and measured. In addition, the condenser drain flow will be measured. Table A.3-9a shows the PANDA Steady State PCC Performance test matrix. In this series of tests, six test conditions (S1 through S6) am included.
~~
The independent parameters am the steam and air mass flow rates. Conditions were chosen so that a direct comparison can be made to PANTHERS and GIRAFFE test points. Table A.3-9a identifies the test conditions in PANDA and the corresponding PANTHERS and GIRAFFE Test Conditions.
- Five Test Conditions (Test Conditions S1 through S5) am planned with various air flows and a constant steam flow of 0.195 kg/sec. In addition, one test will be performed with a pure steam flow equivalent to that expected to match the steam condensing capacity of the condenser (Test Condition S6).
l A-19
I i l - NEDO-32391, Revision C i
* - PANDA Test Conditions S1 through S6 provide a data base for TRACG qualification ,
to meet the requirements of Test Objective 1(a).
- De results of PANDA Test Conditions S1 through S6 will be compared with the '
i PANTHERS and GIRAFFE steady state performance data as noted in Table A.3-9a to meet the requirements of Test Objective 1(b).
- Tests S7 through S9 have been deleted. )
- Tests S10, S11 and S12 are at thesame conditions as Tests S3, S5 and S6, .
respectively. Dese three tests are to evaluate the repeatability of these earlier tests. ; Test S13 is at the same conditions as Tests S6 and S12,except that the PCC poollevel is at the bottom of the upper header to evaluate the effect of reduced heat transfer from the upper header. For all other tests the PCC pool level is at the normal level. j Transient Integral Systems Tests A series of nine transient integral systems tests is planned for the PANDA facility to provide l an integral systems data base for PCC system performance with conditions representative of the long-term post-LOCA SBWR containment response. The philosophy used to determine the test : matrix is to derme a base case test representing SBWR performance under SSAR LOCA conditions, and then to perform penurbations around that base case to establish system effects and
- systems interaction effects. Two tests have been intentionally left undermed, so that the experience gamed in the first seven tests may be utilized in their definition. Table A.3-9b summarizes the key characteristics of each test, and data use. It is planned to perform the tests in three groups of three ;
tests each. The first group will consist of tests M3, M4, and M7, the second group tests M5, M6, : and M8, and the final group the remaining tests M2, M9, and M10. Test M3 will be the first matrix test performed and is identified as the Base Case Test. 'Ihe l initial conditions for Test M3 are summarized in Table A.3-10a. These conditions were derived from the SBWR main steam break LOCA analysis at one hour after LOCA initiation. Additional information on the basis for this choice may be found in Subsection A.3.1.3.4. ; The following provides the purpose and additional descriptive information on each PANDA transient test: Test M1 was deleted and replaced with Test M10. Test M2 is a penurbation to Test M3 with all of the break flow steam directed into drywell DW2. DW2 has two PCC condensers. This test maximizes the steam content , of DW2 and the air content of DW1. It is the most asymmetric condition that can be ,. t established in PANDA. Test M2 results will be compared with Test M3 results to : quantify asymmetric effects on PCCS containment performance. Test M3 is the base case test, as defined in the previous paragraph. j Test M4 is a repeat of Test M3 to demonstrate transient system response repeatability. . L Test M5 provides data for PCCS stanup conditions similar to what might be expected in the SBWR following operation of the drywell spray. Test M5 will be initiated at the same conditions as Test M3. After one hour from the start of the test (i.e.,two hours l from the instant of theLOCA in the SBWR), conditions simulating the activation of l A-20 . I
l NEDO-32391, Revision C i i the SBWR drywell spray will be established. The PANDA capability for Spray flow to the drywell is less than the scaled maximum flow rate for the drywell-spray mode of the SBWR FAPCS. Preliminary TRACG calculations for the PANDA facility have indicated that the available spray ficw will not, by itself, be sufficient to cause the desired vacuum breaker action. Current plans are to supplement energy removal by the spray with a reduction (possibly to zero) of the heaterpower. TRACG calculations have also shown that the combination of the spray and the reduced power will achieve the desired objective. The plan is to continue the spray / power reduction for one hour with the intended result being a substantial increase in the drywell noncondensible inventory. He spray will then be turned off and the heaterpower ramped back to the i decay power level. This will set the stage for PCCS testan with a relatively large noncondensible fraction in the inlet mixture. Test M6 is a perturbation to Test M3 with the IC operating in parallelwith the threc ; PCC condensers throughout the test period. This test will provide data showing the ; interaction between the PCC condensers andthe IC, as well as the effect of the additional heat removal by the IC on containment and reactor system performance. {
- Test M7 will examine PCCS startup under conditions where the RPV is producing steam and the drywell is blanketed with noncondensible gas. The initial conditions for l the PANDA vessels are given in Table A.3-10b. In the context of the SBWR, this set -
of conditions can be viewed as the limiting result of a combination of RPV and DW heat removal mechanisms (PCCS, ICS, sprays, etc.) which have caused all the ! noncondensible gas, transponed from the drywell to the wetwell during the blowdown, to be redistributed back to the drywell. i Test M8 is a penurbation toTest M3, but with drywell-to-wetwell bypass leakage. He ; bypass leakage area will be set at ten times the allowable SBWR value as scaled to PANDA. This test will provide the effect of bypass leakage on containment per'armance. The original objective of PANDA Test M9 was toobtain conditions simulating the { transition from the GDCS injection phase to the long-term cooling phase of the post- ' LOCA transient. This remains the objective of recorti, but it is currently being balanced i against the capabilities of the PANDA facility and the possible desirability of ! examining an altemative parameter variation which may be ofinterest following the : performance of earlier tests in the PANDA matrix. Consequently, there is no further specification of Test M9 available at this time. Test M10 will have test conditions defined later, utilizing tb experience gained from the previous tests. These test will focus specifically on systems interactions. l PANDA tests M2 through M10 provide a data base for TRACG qualification that
]
meets Test Objective 2. ; I PANDA tests M2 through M10 address long-term operation of the PCCS. Tests M5 l i through M7, M9 and M10 address systems interaction and PCCS restan issues. These ) , tests meet the requirements of Test Objective 3. l A-21
l NEDO-32391, Revision C A.3.1.3.4 Justification of Test Conditions Steady State Tests The conditions specified for PANDA Tests S1 through S6 were tabulated inTable A.3-9a. As noted in Subsection A.2.3, the SBWR test program philosophy is to test in multiple scales, wherever feasible. As noted in Table A.3-9a, every PANDA steady state test shares test conditions with a condition from PANTHERS /PCC (see Subsection A.3.1.1) and GIRAFFE (see Subsection A.3.1.5). The specific conditions chosen duplicate PANTHERS /PCC Test Group P3 i conditions at a steam flow near the mid-range for SBWR LOCA conditions and for a range of air flow fractions bounding the SBWR range. Additionally, one pure-steam test was chosen at near the maximum for the SBWR range. , PANDA Tests SI-S6 are shown on the SBWR flow map in Figure A.3-15 This figure may be compared with Figure A.3-4 to see the similarity to the PANTHEPS/PCC matrix.
- The choice of there test conditions was also chosen to facilitate comparison of TRACG predictions at different scales. Tests S1 through S6 all had TRACG pre-test analyses performed
! and submitted. r A.3.1.3.4.1 Transient Integral Systems Tests i Choice of the Base Case . The integral system response tests are specialized with the goal of investigating the highly l ranked phenomena identified as Qualification Needs in NEDC-32391P Table 6.1-1. Since the j number of tests is limited, the choice of conditions must be made to address the potential for ~ systems interactions as well as individual system operations. Additionally, specific phenomena (e.g., drywell depressurization due to spray initiation, and PCC restart following noncondensible re-entry to the drywell) need to be addressed. Consequently, most integral systems test programs l tend to be perfonned by definition of a Bare Case Test, around which perturbations are made to
- assess the effects of specific systems, systems interactions, and phenomena of interest. This testing philosophy was chosen for the PANDA program.
The choice ofBase Case Test is centralin this philosophy. For PANDA, the decision was l made to use the SBWR main steamline break conditions at one hour post-LOCA initiation as the base case. This choice has both historical and technical reasons. Historically, GE pressure - suppression containment and LOCA/ECCS testing has used conservative FSAR assumptions in definition of base cases. From a technical standpoint, this choice also is rational: SSAR conditions give conservative, yet realistic conditions from which to start an experiment. Since the process by which these conditions are predicated are mechanistic in nature, it is relatively straightforward to vary other conditions mechanistically to address the perturbations required. These arguments for using the SSAR conditions for the SBWR base case remain as valid today as they have been in the l past. Since PANDA is primarily a containment response experiment, and the main steamline break is the limiting scenario for the SBWR, this scenario using SSAR assumptions was chosen for PANDA Test M3, the base case. 1 l A-22
l l NEDO-32391, Revision C i 4
- Determination of Base Case Initial Conditions Initial conditions are chosen for the PANDA Test M3 base case on the basis of the predicted
. l state of the RPV and containment at a one hourpost LOCA for a main steamline (MSL) break. He predictions are made using the SBWR TRACG integrated system containment model. His modelincorporates a representation of the RPV and the associated systems (ADS, GDCS) which ! simulate a containment response starting from the beginning of the LOCA, i.e. the instant of the , pipe break. The conditions at LOCA plus one hour are tabulated in Table A.3-11. These conditions must then be synthesized to prescribe the initial thermodynamic state of the PANDA vessels representing the RPV, GDCS pools, drywell, wetwell, and PCCS pools. De process followed addresses the differences between the SBWR and test facility configurations, and L ' averages multi-cell TRACG results into the single conditions possible to specify for the PANDA ] vessels. Facility limitations, such as dynamic load capability, must also be factored intothe choice of ccuditions. l This process introduces three potential sources of discrepancy between the SBWR TRACG ! calculation and the test facility. De first results from averaging the conditions in the multi-cell SBWR model. For example, the SBWR model uses eighteen cells (Rings 3 and 4) to represent the i drywell region above the RPV skirt.The PANDA vessels can be initialized at asin/e nominal l drywell condition (total pressure, partial pressure of noncondensible, and temperature). The second i potential source of discrepancy arises from the need to establish test facility conditions in which the vapor region in each vessel is in thermodynamic equilibrium with the vessel liquid. This is a practical consequence of the length of time it takes to pre-condition the facility and the absence of j an independent means of heating the vapor regions of the vessels. The third potentialsource of ; } discrepancy is introduced by the translation of instantaneous transient conditions from the SBWR , j model into initial steady state conditions for the test facility. i The first two of these potential discrepancies may be resolved by comparing Tables A.3-10a and A.3-11. Typically, differences within the"SSAR" PANDA Test M3 conditions are small. ; j For example, the eighteen cells in Rings 3 and 4 vary less than 1% in total pressure. The same is l
- true for the RPV steam dome, and wetwell air space. Consequently, any departures from
! thermodynamic equilibrium are small. Likewise, since the variation in totalpressures are small, the volume averaging used to determine the PANDA vessel pressures does not introduce a large j error. In the drywell, the air partialpressures vary between 9 and 17 kPa, nearly a factor of 2,
- .. representing an expected variation in air distribution. Since it is impractical to produce a j distribution of noncondensibles in the PANDA drywells, a volume weighted average is used.
[' This leaves the question of rate of change of the test conditions, the so-called " start-on-the-fly" approach. To address this issue, several key outputs from the TRACG SBWR simulation j wereinvestigated. Table A'312 presents the results of this investigation. The drywellpressure, wetwell pressure, wetwell air partial pressure, and the mid and upper drywell air panial pressures were chosen as key parameters, their time derivatives were calculated from the TRACG output. Comparing the derivatives with the PANDA initial conditions, all are seen to be at least four orders of magnitude less than the absolute values. Based on these results, the effect of starting the tests "on-the-fly"isjudged to be negligible. g l A-23
l NEDO-32391, Revision C Other Tests Once the base case is specified system and phenomenological investigations may be performed by penurbations around this base case test. The specific Qualification Needs on a test-by-test basis are listed in Table A.3-13. A.3.1.3.5 TRACG Analysis Plan Each of the nine PANDA steady state and the nine PANDA integral systems tests will have a TRACG analyses performed: post-test, or both pre- and post-test.
- Pure Steam Condensation - Analysis of Tests S1 and S6 demonstrate TRACG's capability to predict pure saturated steam condensation rates at and above rated conditions.
- Air / Steam Mixtures - Analysis of Tests S2 through S5 addresses the effects of noncondensible mass fraction in the PANDA PCC configuration.
- Drywell WetwellNoncondensible Distribution- Analysis of Tests M2, M3, and M5 through M10 addresses the effects ofinitial gas and vapor distribution within the containment system, including vacuum breaker flow, and demonstrate TRACG's capability to model integral systems performance.
- Systems Interactions - Analytical studies of systems interactions have identified vacuum breaker and IC operation as the most likely canaMates for systems interaction effects. Analysis of Tests M5 and M6 address TRACG's capability to modcl systems interactions.
- Bypass Leakage- The TRACG analysis of Test M8 provides qualification of bypass leakage modeling.
A3.1.4 GIST A3.1.4.1 Facility Description The Gravity-Driven Integrated Systems Test (GIST) was performed by GE Nuclear Energy in San Jose, California, in 1988. Testing is complete, and results were reported in Reference 42. The GIST facility was a section-scaled simulation of the 1988 SBWR design configuration, with a , ' 1:1 vertical scale and a 1:508 horizontal area scale of the RPV and containment volumes. Because of the 1:1 vertical scaling, the tests provided real-time response of the expected SBWR pressures and temperatures. . An integrated systems test was performed in order toinclude th' effects of various plant conditions on GDCS initiation and perfomance. Figure A.3-16 provides afacility schematic, and Figure A.3-17 shows the major intemonnecdng linc s. The GIST facility consisted of four pressure vessels: the RPV, upper drywell, lower dryell and the wetwell. The RPV included intemal structures, an electrically heated core, and bypass and chimney regions. Keyinterconnecting lines, such as drywell vents and depressurization lines with quenchers, were alsoincluded. The suppression pool /wetwell includes the water supply tank, a recirculation l A-24 i
i
- l NEDO-32391, Revision C i
pump system used to heat and cool the pool water, and the air lines for pressurizing the wetwell air space. 1 l The GIST facility was a simulation of the SBWR design as it existed in 1988. Several . I differences exist between the GIST configuration and the final SBWR design. These differences l { are listed and reconciled in Reference 32. These differences notwithstanding,the facility simulates allimportant GDCS refilling phenomena. The data on GDCS initiation, flow rates, and chimney ' ! and downcomer level response can be used for TRACG qualification. (Additional data are being )
- . obtained from the GIRAFFE / SIT tests, Subsection A.3.1.7). i i
- Onc hundred twenty test instruments were mounted on the vessels and piping in the GIST
- facility. These instruments were used to measure ADS initiation, drywell and pool temperatures, j' break flow rates, GDCS initiation and flow rates, and RPV conditions such as temperature, pressure and waterlevel.
1 A3.1.4.2 Test Objectives The test objectives for the GIST Test Pmgram were: i 1. Demonstrate the technical feasibility of the GDCS concept. (Concept Demonstration) ! 2. Pmvide a sufficient data base to confirm the adequacy ofTRACG to predict GDCS ! flow initiation times, GDCS flow rates, and RPV water levels. (Integrated Systems ! Test) His test addresses the key interaction between XL3 and phenomena El, E2, E3, j E7 and F1 listed in NEDC-32391P Table 5.3-la i A3.1.4.3 Test Matrix The GIST Test Matrix is shown in Table A.3-14. Twenty-six test conditions were specified. ! These 26 individual tests were divided into four test types, three of them loss-of-coolant accidents: ;
- Bottom Drain Line Break (BDLB) l
- Main Steamline Break (MSLB)
- -
- GDCS Line Break (GDLB) 4 M e No-Break (NB) 1 l A broad spectrum of test parameters was varied within each one of these test types. In each
'., one of the four test categories, a base test was performed and then subsequent tests were run where j only one parameter at a time was varied from that used in the base case. The GIST facility l j modeled SBWR plant behavior during the finalstages of the RPV blowdown. The tests started ; ; with the vessel at 100 psig and continued until the GDCS flow initiated and flooded the RPV.
j - Series BDLB (Bottom Drain Line Break) consisted of parametric variations around the base test case of a relatively small break below the core. Seven tests were run in this i configuration. Series MSLB (Main Steamline Break) consisted of eight tests, six of which were i parametric variations and two of which were duplicates to establish the repeatability of results. i i l l A-25
- - + '
l NEDO-32391, Revision C Series GDLB (GDCS Line Break) consisted of four tests. Variations in ADS configuration were the parameter in this series. Series NB (No-Break) consisted of seven tests. This series typically utilized conditions well removed from the SBWR 1988 design envelope. They form a data set at or outside the limits of SBWR, and are the most challenging for TRACG analysis. For example, this series included several tests where the wetwellinitial pressure was atmospheric, and no air-purge occurred since there was no break. The major difference between the 1988 GIST and current SBWR configurations is the location of the GDCS - pool. From the standpoint of GDCS injection, the GIST configuration is conservative relative to the SBWR because the GDCS driving head is always slightly less in GIST than in the SBWR. In the case of zero wetwellpressure, the GDCS injection head is a much less than in the SBWR. This makes GDCS injection in GIST more challenging. Analysis of GIST data as reported in Reference 42 has proven the technical feasibility of the GDCS concept and accomplishes Test Objective 1. The overall GIST data base provides a sufficient basis for TRACG qualification and accomplishes Test Objective 2. A.3.1.4.4 TRACG Analysis Plan As part of the GIST program, five TRACG comparisons were previously performed. De objective of this effort was to confirm the capability of TRACG to accurately predict the GIST facility response to a variety of LOCA initiating events. The principal areas of interest were the effectiveness of the modeling of the GDCS and the modeling of the RPV and containment at low De qualificationconsisted of post-test caMulations with TRACG and l pressure conditions. comparison against GIST data. Comparisons were made for RP v pressure, RPV collapsed water level, core AP, GDCS flow rate, and GDCS initiation time. Good agreement was found between test and calculation; the results are reported in Reference 2. GIST tests for which TRACG analysis were completed are identified in Table A.3-15. These tests represent the full spectrum of break types, a wide range ofinitial pressure vessel liquid inventory, variations of containment initial conditions, and several degrees of GDCS availability. l A.3.1.5 GIRAFFE l A.3.1.5.1 Test Description .. Overview GIRAFFE Isolation Condenser / Passive Containment Cooling testing was performed at the Toshiba Nuclear Engineering Laboratory in Kawasaki City, Japan. The results are reported in l Reference 43. The test facilityconsisted of five major components which represent the SBWR l primary containment and suppression chamber pools (S/C), the isolation condenser / passive j cantainment cooling heat exchanger, and the connecting piping. Separate vessels represented the recctor pressure vessel, drywell, wet- dl, GDCS and the IC/PCC pool, which houses the IC/PCC condenser unit. A schematic of the, y is shown in Figure A.3-18. ;
'l A-26 l
i l NEDO-32391, Revision C l The IC/PCC condenser tested was a full-length, thme-tube heat exchanger. The single unit could be utilized as either an IC or a PCC. Figure A.3-19 gives an outline drawing of the heat , exchanger assembly. The IC/PCC was installed in a water pool composed of a makeup pool with ! a chimney and cavity arrangement in which the IC/PCC unit was set. l These GIRAFFE tests were performed as developmental tests. Except for comparison of the GIRAFFE steady state PCC performance data with that from PANTHERS and PANDA, no additional analysis of this data is planned. O A.3.1.5.2 Test Objectives The objectives of the GIRAFFE Test Program are:
- 1. Provide a data base to support primary data taken at other scales to confirm the capability of TRACG to predict the quasi-steady heat rejection rate of a PCC heat exchanger. (Steady State Separate Effects)
- 2. Provide a data base to rupport primary data taken at other scales to confirm the capability of TRACG to predict PCCS system performance. (IntegralSystems Tests)
A.3.1.5.3 Test Matrix and Data Analysis Steady State Tests The majority of the GIRAFFE data am steady state performance data for the IC/PCC unit under PCC conditions. For these tests, the facilitywas placed in a condition where steam or nitrogen-steam mixtures were supplied to the IC/PCC; the condensed vapor and vented nitrogen j were directed to volumes modeled to act as thereactor vessel and suppression chamber pool respectively. Condensate outlet flows from the IC/PCC were measured by measuring the RPV l collapsed levelincrease, which,in tum, was used to determine heat removal rate by multiplying it by the latent heat of vaporization. The condensate was returned to the RPV, and the vented l nitrogen was released to the S/C gas space. Once steady state conditions wem established, data were collected for a period of approximately 10 minutes. The time averaged data were reponed and analyzed.
. Table A.316 shows the GIRAFFE PCC Steady State Performance Matrix used to provide 1
data in support of the test objectives. Thirteen test conditions are included. Thesetests are l l identified in the test repon as the Phase 1, Step 1 Tests, and comprise Phase 1 Test group. These tests cover the SBWR range of steam and air mass flow rates, as has been previously discussed in the PANTHERS /PCC section. Data from Phase 1 Test group provide a support data base for TRACG qualification and meet the requirements of Test Objective 1. Data from Phase 1 Test group will be compared to that from conesponding PANDA and PANTHERS tests to corroborate those results at a third scale. l l A-27
. _ _ _ _ ._._ _ _ . - _ _ - _ _ _ . _ _m. _ _ ___ . . _ _
h i l NEDO-32391, Revision C l l I A3.1.5.4 TRACG Analysis l A significant number of GIRAFFE TRACG comparisons have been performed as pan of i the qualification effort. De objective was to confirm the capability ofTRACG to accumtely. l predict PCC steady state performance. Results are reponed in Reference 2. , A.3.1.6 GIRAFFE / Helium l A.3.1.6.1 Test Description i Overview The GIRAFFE / Helium tests are being performed by the ToshibaCorporation at their ! Nuclear Engineering Laboratory in Kawasaki City, Japan. He purpose of these tests is to ; demonstrate the operation of thepassive containment cooling system (PCCS) in post-accident i containment environments with the presenc: of a lighter-than-steam noncondensible gas as well as a heavier-than-steam noncondensible gas. These tests will demonstrate SBWR containment . thermal-hydraulic performance, heat remova; capabi'ity, and systems interactions. Also, they will provide additional data for the qualification of containment response predictions in the presence of _ lighter-than-steam noncondensible gases by the TRACG computer program. De facility configuration is very similar to that used in the earlierGIRAFFE tests described in the previous section. The facility configuration is shown schematically in Figures A.3-20 and A.3-21. The primary facility changes from the earlier configuration include shonening the PCC l tube length (to 1.8 meters) and modifying the piping orifices toyield flow resistances which more closely model the current SBWR values. Additionally, provision has been made for the j continuous addition of helium to the drywellduring a test. Details are provided in the GIRAFFE / Helium Test Specification (Reference 57). ; The GIRAFFE / Helium tests are performed in accordance with Japanese Quality Assurance Standard JEAG-4101,1990 (Reference 58). Review of this standard against the requirements of ANSI /ASME NQA-1 has shown that the essential elements of NQA-1 are met by this standard. i Therefore, results from the GIRAFFE / Helium test program are appropriate for use as design basis , data, i instrumentation f Instrumentation utilized in the GIRAFFE / Helium test program is similar to that used in earlier GIRAFFE tests. Test instrumentation consists of 81 thermocouple measurements, 5 pressure measurements,19 differential pressure measurements, and 4 flow rate measurements. .. Test instrumentation is calibrated against standards equivalent to the U.S. National Institute of Standards and Technology. Detail of the instrumentation, including instrument lists, types, and ranges are included in the Test Plan and Procedure (Reference 67). Direct measurement of noncondensibles during the GIRAFFE / Helium test program will be ; performed by periodically taking samples of the process fluid at two points in the drywell and one pointin the wetwell during all of the tests. Samples will be analyzed using gas chromatography. , It is necessary to limit the total number of samples taken, so as not to affect the test results. He , samples will be taken at the three locations specified, once per hour during the conduct of the test. l A-28 i
b l NEDO-32391, Revision C his data will be used to validate indirect measurements of noncondensible concentration inferred l , from temperature measurements. i ! Method i GIRAFFE /Hclium testing follows a methodology very similar to that used in PANDA. l 1 Once the initial conditions for a given test have been establishd all control (except for the decay of , RPV power and helium injection,if called for) will be terminaeJ, and the GIRAFFE containment will be allowed to function without operator intervention (except that the vacuum breaker is f operated manually to simulate automatic operation in SBWR), mirroring the SSAR assumptions for the SBWR. Details are included in the Test Plan and Procedure (Reference 67) for these tests. l i, A.3.1.6.2 Test Objectives i The test objectives of the GIRAFFE / Helium Test Program are: !
- 1. Demonstrate the operation of a passive containment cooling system with the presence l of a lighter-than-steam noncondensible gas, including demonstrating the process of purging noncondensibles from the PCC condenser. (Concept Demonstration)
. 2. Provide a data base for computer codes used to predict SBWR containment system ! performance in the presence of a lighter-than-steam noncondensiblegas, including . potential systems interaction effects. (IntegralSystems Tests) i
- 3. Provide a tie-back test, which includes the appropriate Quality Assurance
! documentation to repeat a previous GIRAFFE test, thereby reinforcing the validity of ; . the previous GIRAFFE testing.
! i t r A.3.1.6.3 Test Matrix and Data Analysis 1 Helium Test Series ! The series of helium tests (designated as Test Group H) is performed to demonstrate the
- operation of the PCC system with the presence of a lighter-than-steam noncondensible gas. Four j tests with lighter-than-steam, heavier-than-steam, and mixtures of heavier- and lighter-than-steam noncondensible gases are included. Table A.3-17 provides the test matrix which gives the initial ;
drywell conditions and helium injection rate for each test. Each test will run for at least 8 hours, ; and demonstrate at least one purge / vent cycle of the PCC condenser. L. He following provides the purpose and additional descriptive information for each GIRAFFE / Helium test: Test H1 is the base case with nominal initial conditions the same as in PANDA Test
- M3. Initial conditions are given in Table A.3-18 Test H2 is a repeat of Test H1, but with helium replacing the total volume of nitrogen ,
in the drywell and PCCS. Test H3 will have the same initial total drywell pressure as Tests H1 and H2, but with the initial noncondensible fraction consisting of a helium / nitrogen mixture. , I
?
l l A-29 1 h
l NEDO-32391, Revision C F Test H4 will start with the same initial drywell conditions as Test H1, and will have constant helium injection to the drywell. The helium addition rate will be such that the heliumis injected over a period of one hour. De helium injection will be terminated 3 when the total mass of helium added is equal to the initial drywell helium mass for Test { H3. i System response from the four tests will be compared to establish the effect of lighter-than- ! steam noncondensible, or a mixture . of lighter-than-steam and heavier-than-steam noncondensibles, on the effectiveness of heat rejection by the IC/PCC heat exchanger. . GIRAFFE Tests H1 through H4 will demonstrate the operation of thePCCS with the presence of a lighter-than-steam noncondensible gas. These tests meet the requirements of Test ' Objective 1. l GIRAFFE Tests H1 through H4 provide data for TRACG qualification to accomplish Test
- Objective 2.
."Tie-Back" Test Series The " Tie-back" series of tests (designated as Test Group T) is performed to reinforce the validity of previous GIRAFFE testing that did not include sufficient documentation to qualify as design basis information. This series of two tests will be run in accordance with JEAG-4101 Quality Assurance Guidelines; in fact, one of these tests will be a repeat of an earlier GIRAFFE test. It is anticipated that the test results will match those of the earlier test,thus demonstrating its
- l technical accuracy. Test T1, the test chosen for repeat, is a main steamline break test. Test initial 4 conditions are given in Table A.3-19.
- Test T2 test conditions are very similar to Test H1, but have initial drywell nitrogen content l intermediate to Tests H1 and T1. Initial conditions for Test T2 air given in Table A.3-20 l -
Comparison of the results of Test Tl with the previous GIRAFFE main steamline j break test results will meet the requirements of Test Objective 3. The combination of GIRAFFE / Helium Tests H1 through H4, T1, T2, and PANDA l Tests M3/4, and M7 form a comprehensive data base for investigation of the operation ! of the PCC heat exchanger in the presence of noncondensibles, and meet the ! requirements of Test Objective'l. A.3.1.6.4 Justification of Test Conditions Choice of the Base Case - Test H1, defined as the Base Case for Test Groups H and T, utilizes the same initial l conditions as PANDA Test M3 (see Table A.3-10a). De justification for the M3 conditions given
; in Subsection A.3.1.3.4 also apply to GIRAFFE / Helium test Hl.
The decision to use common initial conditions for the GIRAFFE / Helium and PANDA base cases is also advantageous from the test philosophy standpoint to test at different scales. Tests H1 and PANDA M3 may be compared directly to determine any effect of scale on the results. l A-30 c _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ . _ . _ _
l NEDO-32391, Revision C Other Tests
'Ihe other tests specified as part of the GIRAFFE / helium program were defined in such a l way as to investigate PCC operation for a range of both lighter-than-steam and heavier-than-steam noncondensible conditions. Figure A.3-22 shows the initial conditions on an air / helium partial pressure map. Initial conditions for PANDA Tests M3 and M7 are also included on the figure. 4 l The figure clearly shows that PCC operation will be demonstrated over a very wide range of air / nitrogen to steam ratios, from nearly pure steam to pure air. !
Test H2 helium specification is unrelated to any scenario; all the nitrogen used in Test H1 is replaced with helium to obtain a one-to-one comparison of PCC system performance in the presence oflighter-than-steam and heavier-than-steam noncondensibles. The purpose of the H3 and H4 tests is to demonstrate the effect of a high concentration of a lighter than steam gas on the performance of the PCC. At the same time, the pressure capability of the facility must be considered. The design pressure for the drywell and suppression chamber is 0.6 MPa. In order to assure that the design pressure is not exceeded during Tests H3 and H4, a helium mass less than the scaled amount determined for a 100% fuel-clad metal water reaction is used. Therefore, Tests i H3 and H4 will utilize a helium mass of 0.97 kilograms, which is equivalent to 20% by volume of the scaled amount of hydrogen gas that would be generated by a 100% fuel-clad metal water reaction. This results in an initial GIRAFFE drywell concentration of approximately 23% helium by volume. Since this quantity is equal to approximately 60 times the PCC volume it is a sufficiently high quantity of helium to capture the prototypical behavior oflighter-than-steam gases on the performance of the PCC. Test Tl initial conditions are the same as the initial conditions used for the previous i GIRAFFE main steamline break test. For Test T2, the total nitrogen mass for the drywell and 1 ) suppression chamber is equal to the total nitrogen mass for Test Hl. The initialdrywell nitrogen mass for Test T2 is approximately midway between that for Tests H1 and T1. Therefore, in Test T2 the effect of the nitrogen distribution between the drywell and suppression chamber will be investigated. A.3.1.6.5 TRACG Analysis Plans All tests in the GIRAFFE / Helium H-series will have TRACG analysis performed on a blind
; post test basis. Although the tests will be performed prior to TRACG analysis, the analyst will have no knowledge of the test results while the analysis is being performed. Tests T1 and T2 will have TRACG analysis performed on a post-test basis.
1 A.3.1.7 GIR AFFE/ SIT (Systems Interaction Test) A.3.1.7.1 Test Description l Overview i i The GIRAFFE / SIT (System Interaction Tests) will be performed by the Toshiba Corporation at their Nuclear Engineering Laboratory in Kawasaki City, Japan. Test data will be obtained for TRACG qualification during the late blowdown /early GDCS phase of liquid line l breaks. , i l A-31 l i
l l NEDO-32391, Revision C i l The facility configuration is discussed in Subsection A.3.1.6.1 and is shown schematically in ! Figure A.3-18, with the addition of a second heat exchanger so that both the PCC and IC can be in i l operation simultaneously. The configuration of the IC is similar to the IC/PCC unit shown in . Figure A.3-19. The GIRAFFE / SIT tests will be performed in accordance with Japanese Quality Assurance i Standard JEAG-4101,1990 (Reference 58). Review of this standard against the requirements of i ANSI /ASME NQA-1 has shown that the essential elements of NQA-1 are met by this standard. Therefore, results from the GIRAFFE / SIT test program are appropriate for use as design basis - data. ! Instrumentation , i Instrumentation utilized in the GIRAFFE / SIT test program is similar to that used in earlier i GIRAFFE tests (see Subsection A.3.1.6.1). Added instrumentation for the IC unit includes tube, j steam box and water box fluid temperatures; steam supply line flow; steam box pressure; and i collapsed water levels in the IC pool,IC tube and water box. The collapsed chimney water levels j are measured over 4 intervals using pressure taps at 0.5 to 7.0m above the top of the heater section. Both the bypass region collapsed level and thedowncomer collapsed level are measured. ! Instrumentation details, including instrument lists, types, and range will be included in the test plan and procedures for these tests. l Method i GIRAFFE / SIT testing follows a methodology very similar to that used in PANDA and l GIRAFFE / Helium. Once the initialconditions for a given test have been established, all control i (except for the decay of RPV power and possibly the microheater power) will be terminated. 'Ihe i GIRAFFE RPV and containment will be allowed to function without operator intervention, l (except that the vacuum breaker is operated manually, to simulate automatic operation in SBWR) j mirroring the SSAR assumptions for the SBWR. Details will be identified in the Test Plan and Procedure for these tests. 1 I A.3.1.7.2 Test Objectives l ! In the initial GE evaluation, no need for these tests was identified. However on page 16 of i the TAPD Draft Safety Evaluation Report (DSER) the NRC staff notes, "While GE considers } MSLBs to be the limiting accident in terms of containment performance, both GDCS line breaks and bottom drain line (BDL) breaks are more limiting in terms of reactor vessel response, especially minimum water level. The staff has, therefore concluded that additional integral .. + systems tests are required as part of the design certification test program for the SBWR. The tests should be performed in an appropriately scaled facility that (a) represents the current design of the SBWR; (b) has the capability of simulating a range of design basis events, including GDCS line breaks and BDL breaks; and (c) has sufficient power and pressure capability to represent these 3 events prior to the initiation of GDCS injection." The GIRAFFE facility meets these criteria. Based on the above, the test objective of the GIRAFFE / SIT Test Program is:
; Provide a data base to confirm the adequacy ofTRACG to predict the SBWR ECCS performance during the late blowdown /early GDCS phase of a LOCA, with specific focus j on potential systems intewetion effects. (InregralSystems Tesrs) !
i l A-32
l i l NEDO-32391, Revision C i A.3.1.7.3 Test Matrix and Data Analysis l A series of four transient systems tests is planned to provide an integral systems data base ! for potential systems interaction effects in the late blowdown /early GDCS period. All four tests l l are liquid breaks: three GDCS line breaks and one bottom drain line break. Tests will be performed with and without the IC and PCC in operation, and two different single failures are i considered.The test matrix defining the four tests is given in Table A.3-21. Initial conditions for tests are provided in the test specification. l The initial conditions for all tests approximate SBWR conditions approximately 10 minutes
, post-LOCA, based on the breaks and equipment operations listed in Table A.3-21. All tests will
- run for approximately two hours. Containment related parameters will based on the appropriate '.
1 SBWR TRACG LOCA case at the time RPV pressure is 1.034 MPa (150 psia). Heater power will be decayed from this time to simulate decay heat and stored energy transfer.
- j 'Ihe RPV collapsed water level at the stan of the test will be determined by using the 4 TRACG GIRAFFE model.. Since GIRAFFE is not an exact " scale model" of the SBWR, it will
- not be practical to have the water / steam distribution in GIRAFFE be the same as in SBWR. For example, the GIRAFFE RPV lower plenum is shorter than the SBWR lower plenum.
Additionally, the GIRAFFE RPV material is thinner, and begins the LOCA simulation at a lower j temperature than the SBWR. As a result, a smaller amount of energy is transferred to the RPV l lower plenum fluid in GIRAFFE. Methods to better simulate this energy addition are being !
, investigated, and may affect the final definition of the initial RPV collapsed water level. ,
{ Additional details on the initial conditions for the GIRAFFE / SIT tests will be included in the l
- Test Plan and Procedure. i The following provides the purpose and additional information on each GIRAFFE / SIT test
- -
{ Test GSI is the base case test, a GDCS line break, with DPV failure as the single failure and neither the PCCS, nor the IC, in operation. This test has initial conditions i similar to GIST Test C01 A, and may be compared with GIST C01 A to evaluate the effects of configuration distortions in GIST and potential GDCS containment system i performance interactions. Test GS2 is the same as Test G1, except that the PCCS and IC are operating. Test GS2 i- results will be compared to those of Test GS1 for identification of potential systems j interactions associated with the IC and PCCS. Test GS3 is a bottom drain line break with DPV failure. For this test both the PCCS s and IC will be functioning. Data from Test GS3 will be examined for identification of - I potential systems interactions associated with the IC and PCC for a bottom drain line ,
< break.
Test GS4 is a GDCS line break, with the single failure being a GDCS valve failure in
; one of the other GDCS injection lines. As in Test GS3, both the PCC and IC will be in operation. This condition is expected to provide the slowest rate of recovery for the j chimney swollen water level. Data from test GS4 can be compared to test GS1 to identify potential interactions with the IC and PCC even though the single failures are different. ,
4 l A-33 i e
l NEDO-32391, Revision C GIRAFFE / SIT Tests GS1 though GS4 provide a data base for TRACG qualification that meets the GIRAFFE / SIT test objective. A.3.1.7.4 Justification of Test Conditions Choice of the Base Case Test Test GS1 conditions resulted in the lowest predicted chimney swollen water level, considering the various break locations, sizes, and single failum combinations. Additionally, the commonality of conditions between this case and that of GIST Test C01A allows a comparison between the GIST and GIRAFFE simulations. The differences between the GIST andGIRAFFE test configurations allow an assessnient of the effect of containment on GDCS performance. Other Tests The other test cases were defined with the objective of identifyingsystems interactions, should they occur. Since the primary focus of this testing is GDCS performance, the RPV l chimney swollen water level is the figure of merit in these investigations. SBWR TRACG predictions for several break locations, single failures, and IC/PCC operation combinations were performed. The additional tests, which are presented in Table A.3-23, were chosen based on the TRACG results. The " objective" column in this table indicates the major featum of each case which will be captured by the test. The rate of chimney swollen water level recovery after the time of minimum level is expected to be substantially different for the GDL and BDL break cases noted for those " recovery" objectives. Some SBWR TRACG cases evidenced chimney swollen water level oscillations which is attributed to GDCS flow quenching voids in the RPV and to break flow depressurizing the drywell. Thus, a wide range of conditions is represented by the four test cases selected. Test Initiation The tests will be initiated from a steady state condition at the time RPV pressure reaches 1.034 MPa. This is unlike the SBWR situation, where the accident starts at much higher pressures. This difference is justified since the void formation, which results from depressurization, occurs very soon after the pressure is reduced to the fluid saturation temperature. SBWR calculations indicate that void fractions increase in about 20 seconds to much larger values, then remain relatively constant until GDCS flow is injected on the order of 80 seconds later ., (causing some void collapse) or until pressures become much lower than 1.034 MPa. Preliminary GIRAFFE TRACG cases indicate a very similar behavior. Based on this general agreement, the selected strategy is appropriate. The water level instrumentation in the test facility can be used to .- determine vadations in void fraction, and this will be used to provide test-related confirmation of the strategy. A.3,1.7.5 TRACG Analysis Plan All four transient tests in the GIRAFFE / SIT series will have TRACG analysis perfonned on a blind post test basis. Although the tests will be performed prior to the TRACG analysis, the analyst will have no knowledge of the test results while the analysis is being performed. Exceptions will be information needed to conduct the analysis such as actual initial conditions, l A-34
l NEDO-32391, Revision C heater power and microheater power during the test. The assessment of TRACG's adequacy will be based on the ability to predict chimney and downcomer swollen and collapsed water level. A.3.1.8 Other Analyses Planned The previous sections have discussed the major SBWR-unique test programs and defined the test conditions to be analyzed with TRACG. This section will give a brief overview of these tests and the anticipated corresponding TRACG analyses. A.3.1.8.1 1/6 Scale Boron Mixing Test GE-NE has performed a set of boron mixinginjection tests for BWR/5 and BWR/6 geometries. These tests wem reported in Reference 28. The tests were performed in a 1/6 scale thme-dimensional model of a 218 in, reactor pressure vessel, and used the High Pressure Core Spray (HPCS) spargers as the primary injection location of the simulated boron solution. Using scaled boron injection rates of either 400 or 86 gpm, with and without HPCS flow, the parametric effects on mixing were examined in the upper plenum and core bypass regions. Two altemate injection locations were also examined. Standby Liquid Control injection locations are diffemnt in the SBWR from previous product lines, due primarily to the natural circulation recirculation featum of theSBWR. The SBWR utilizes direct injection into the core regien through the shroud at 16 locations. A series of TRACG predictiont of the BWR/5-6 data is planned. Specific test cases to be analyzed have not yet been identified. Primary data comparisons will be made against data for the
- mixing coeficient, which is defined as the concentration of injectedsolution at the measured q location divided by the concentration that would be present if theinjected solution were uniformly l mixed with the entire vessel inventory. Comparisons will be made at several locations. l A.3.1.8.2 CRIEPI Natural Circulation Thermal Hydraulic Test Facility The CRIEPI 7 MPa test facility is a parallel channel test facility intended to study the
~
stability characteristics of a natural circulation loop during startup conditions. Figure A.3-24 shows the configuration of the test facility. The two pamllel channels are 1.79m high and am equipped with heaters with a maximum power input of 64 kW each. At the channel exit, there is I an adiabatic chimney which is 5.7m high. The loop has a separator, a condenser and a subcooler i which are used to return the condensed steam to the downcomer. A preheater with a capacity of 150 kW controls the inlet temperatum to the channels. Tests have been run at low pressure to simulate low pressure loop startup. Flow measmements (natural circulation conditions) for different system pressum, power and channelinlet temperatures were obtained at the CRIEPI experimental facility. Instability similar to geysering was observed at some conditions. Test Conditions and Scalability to SBWR The results of the CRIEPI test and comparisons with TRACG are reported in severalpapers (References 64,27, and 65). l l A-35
l NEDO-32391, Revision C The basic equation of the drift-flux model were non-dimensionalized to arrive at the important non-dimensional numbers for the hydrodynamic stability. 'Ihe characteristic numbers are reported in Table A.3-29 for the SBWR and CRIEPI facility. The tests were not run at thefull power conditions shown in the table. The full power conditions were selected to match N g and N of the SBWR at full power. To arrive at the low power conditions, the power of the facility was ratioed down by the same amount as the SBWR for the desired conditions. The subcooling was set to match the SBWR also. A complete discussion of the method is described in Reference 64. The comparison in the table is performed for a representative case of 0.1 MPa system pressure. As shown in the table, the test facility compares very well with the SBWR. The most notable difference is in the flashing parameter, rN. This difference is because the CRIEPI facility . is about 70% as tall as the SBWR. The good general agreement in the important parameters of the SBWR and CRIEPI facility indicate that the results are applicable to the SBWR. Tests were run at pressures of 0.1,0.2,0.35 and 0.5 MPa. The results are shown in Figure A.3-25. The figure shows the instability region in the heat flux-channel inlet subcooling plane as developed in Reference 64. Additionally, the expected SBWR conditions during start-up for these pressures are shown on the figure. The results indicate a significant amount of margin to unstable behavior in the SBWR. The margin increases as the pressure is increased. Some additional tests in another facility showed unstable behavior over the entire range of conditions tested there (Reference 66). However, these tests were run at a much higher heat flux and subcooling than is representative of the SBWR. Figure A.3-25 shows that instability at higher heat flux and subcooling is consistent with the unstable region for the CRIEPI results. A.3.1.8.3 Dodewaard Plant Startup
'Ihe Dodewaard reactor is'a natural circulation BWR with internal free surface steam separation. The reactor, with a maximum thermal power of 183 MWth, is connected to a turbogenerator capable of producing 60 MWe. Initial startup of the reactor was in 1969, and it has been operating continuously since that time. While relatively small in size, it is thermodynamically and neutronically similar to the SBWR. The SBWR startup procedures will be similar to those of Dodewaard.
On February 15 and 16,1992, the reactor was started-up for its 23rd fuel cycle. During that - stanup, data were recorded to characterize the startup for potential TRACG analysis Data were taken at discrete time intervals during the startup. Typically, the reactor was in a state of semi-equilibrium during the measurement. The results of the measurement show early establishment of recirculation flow during low power operation. No indication of any reactor instability, including geysering, was observed. Data are reported in References 15 and 45. TRACG analysis of this startup is being performed. l A-36 )
i l NEDO-32391, Revision C j A.3.1.8.4 Containment System Response - PSTF Mark III I ) In the early 1970s, GE-NE performed several series of tests at the Pressure Suppression i Test Facility (PSTF) to support the Mark III containment design. The SBWR and Mark III J containments share a similar horizontal vent system geometry. i The '.est series chosen for comparison is PSTF Series 5703, which was reponed in Reference i
- 20. Test Series 5703 utilized a full-scale, three horizontal vent system with geometry very similar to that used in the SBWR. Three comparisons will be performed to test data from Runs 5703-1, f -2, and -3, for which simulated steamline break size was the primary variable.
j' A.3.1.8.5 Containment System Response - Mark II 4T t
- In the mid-1970s, GE-NE conducted a series of containment tests supporting the Mark II containment design in the 4T (Temporary Tall Test Tank) facility in San Jose, California.
Test Series 5101 is reponed in Reference 38. These tests were afull-scale, single-vent
- simulation of Mark II (vertical vent pipe) performance. Normally, the drywellwas heated to 150 C prior to test initiation to minimize steam condensation. One test, Run 33, used aunheated l drywell. Very different response was seen due to steam condensation in the drywell. i l Additionally, Tests 34 and 35 were performed specifically to investigate the effect of a wetwell-to-drywell vacuum breaker. (In the Mark II containment, pressurization of the wetwell air space by pool swell causes a short term opening of the vacuum breaker.)
i These three tests will be analyzed with TRACG. l A.3.1.8.6 Suppression Fw I Stratification - PSTF In the late 1970s, Io - eries of experiments were performed in the PSTF specifically to investigate pool condensanon and thermal stratification in the Mark III containment system. These data were initially reponed in References 46 and 47, and extensively analyzed in Reference 48. More recently, these data were reviewed as one element of an effon to define an appropriate l nodalization for the TRACG SBWR suppression pool, but specific comparisons to the data have not yet been performed. !.. 'Ihe tests reponed in Reference 46 utilized a full scale single cell 9-degree segment of the Mark III vent system and suppression pool,while those reponed in Reference 47 used a vent
. system and pool having the same full-scale height, but with flow areas and pool surface areas reduced by a factor of 3. Suppression pool temperatures were monitoird by an array of
- thermocouples suspended throughout the pool. Initial pool temperatures and blowdown flow rates i
weit measured. TRACG will be used to analyze Test 5707 Run 1 and Test 5807 Run 29. 4 i i ! l A-37 4
l NEDO-32391, Revision C A.3.2 Component Demonstration Testing A.3.2.1 PANTHERS /PCC I A.3.2.1.1 Test Description j Component testing of the prototype PCC heat exchanger is performed using the same hardwam and test facility as described in Subsection A.3.1.1. The component demonstration tests am very similar in conduct to the thermal-hydraulic testing. The test anicle (PCC module "A") is instrumented with strain gages, accelerometers, and thermocouples. Structural instrumentation is . shown on Table A.3-24. Data are collected during the thermal-hydraulictests, as well as the stmetural performance tests described in this section. A.3.2.1.2 Test Objectives The test objective of the PANTHERS /PCC Component Demonstration Test is: Confirm that the mechanical design of the PCC heat exchanger is adequate to assure its structuralintegrity over a lifetime that exceeds that required for application of this equipment to the SBWR. A.3.2.1.3 Test Matrix and Data Analysis The approach taken to address the test objective is to subject the equipment to atotal number of pressure and temperature cycles well in excess of that expected over the anticipatedSBWR lifetime. The test matrix is shown in Table A.3-25. De number of cycles was conservatively chosen as 10 LOCA cycles and 300 pressure test cycles. This represents five times the design requirement number of hypothetical LOCAs (2) and nearly 17 times the number of expected l pneumatic PCC. Note, no test cycles credit in for is taken accordance with 10CFR50, the thermal cycles Appendix experiencedduring the PCC J over the 60-year thermal-hydraulic testing in determination of this Component Demonstration Test Matrix. Two types of tests are performed during the PANTHERS /PCC component demonstration test: simulated LOCA pressurizations and simulated pneumatic leak test pressurizations. Simulated LOCA Pressurizations Simulated LOCA cycles are performed by pressurizing the PCC units with steam to simulate both the temperature and pressure effects of a LOCA. The PCC pool is at ambient temperature at the beginning of a test, but is allowed to heat up to saturation as each cycle proceeds. Table A.3-26 - gives the time history of the LOCA pressurizations. Each LOCA cycle lasts approximately 30 minutes. Ten cycles are performed. Simulated Pneumatic Leak Test Pressurizaticns Simulated pneumatic tests are performed by pressurizing the PCC heat exchange with air to 758 kPag (110 psig). He PCC pooltemparature is at ambient conditions during these pressurizations. The test }nessure is held for 2 minutes for each cycle. A total of 300 cycles are l A-38
.l NEDO-32391, Revision C
, performed. 'Ihe test data will be analyzed by review of strains and acceleration data against component acceptance requirements, both in terms of magnitude and frequency content. A.3.2.2 PANTHERS /IC A.3.2.2.1 Test Description Overview i. Component testing of the prototype IC heat exchanger will be performee! using the same 4 hardware and test facility as described in Subsection A.3.2.1. The component demonstration tests will be very similar in conduct to the thermal-hydraulic testing. The test article (the IC condenser unit) is instrumented with strain gages, accelerometers, and thermocouples. Structural instrumentation is shown in Table A.3-27. Data _will be collected during the thermal-hydraulic tests as well as the structural performance tests described in this section. Test Method
- Figure A.3-23 illustrates the four cycle types for the PANTHERS /IC Component Demonstration tests. For each cycle, the IC unit is initially brought to high pressure by purging the i air with high pressure steam. The rate of pressure increase simulates the SBWR startup, and is
, limited to pmvent undue thermal stresses. The drain valve is closed, and the unit fills with water to , the level of the iC pool. The IC poolis at ambient conditions. This simulates the standby mode of the IC in a SBWR when the plant is in operation. The tests begin when the unit has stabilized at its
- required initini inlet pressure (P1).
In Types 2 and 5 tests, the IC drain valve is openet, allowing the water to flow from the unit. This exposes the tubes to steam and begins the IC operation. The operator controls the steam flow ! to the steam vessel to bring the unit to pressure P2 and holds it for two hours. Following this, the unit is cooled down at a controlled rate without condensing steam. Type 2 test is the same as the Transient Demonstration Test Group 12 (Subsection A.3.1.2.3), which demonstrates the startup 4 and operation of the IC unit. Type 5 tests are similar to Type 2 tests, but use a different initial pressure. The Type 6 cycle simulates the pressures and temperatures the IC would experience during , normal plant startup and shutdown. From the initial inlet pressure of P1, the unit is cooled down
- at a controlled rate without condensing steam.
4 The Type 7 cycle simulates an ATWS event. From pressure P1, the drain valve is opened and the IC begins operation. The operator briefly brings the unit to a high pmssure (P2), which is similar to what an IC unit would experience at the start of an ATWS event. The pressure is then brought back to initial pressure or as close as is possible (P3) and held for two hours. Following this, the unit is cooled down at a controlled rate. A.3.2.2.2 Test Objectives The test objective of the PANTHERS /IC Component Demonstration Test is: Confirm that the mechanical design of the IC heat exchanger is adequate by assuring that the loads used in design envelope the loads expected during the SBWR service conditions. l A-39
4 l NEDO-32391, Revision C l A.3.2.2.3 Test Matrix and Data Analysis l The approach taken to address the test objective is to include sufficient number of load cycles to reveal any thermal racheting where the elastically calculated stress levels exceed the ASME i Code shakedown limits, so the measured deformations can be used to envelope the ASME altemative shakedown analysis approach. Specifically,it is planned to subject the IC to 20 load cycles with a large fraction of the cycles to include thermal transients which will be sufficient to meet the above criteria, as well as uncover unexpected vibrations or unacceptable crack indications l at welds. Prototype non-destructive tests (NDT) will be performed before and after the cyclic .
- testing. The test matrix is given as Table A.3-28. Note, credit may be taken forthe thennal cycles i experienced during the IC thermal-hydraulic testing in determination of this Component l Demonstration Test Matrix, provided that the cyclic structural test conditions are met during the thennal hydraulic testing.
The test data will be analyzed by review of strain and acceleration data against component acceptance requirements, both in terms of magnitude and frequency content. Evidence of crack initiation or growth will be obtained from comparison of the pre-test and post-test NDT. l' He test measurements will then be used in conjunction with other loads in an analysis to calculate the resultant stresses and cumulative fatigue effects to ensuring conformance with the ; l ASME code. i 4 } A.3.2.3 Depressurization Valve (DPV) i A.3.2.3.1 Test Description a A Depressurization Valve (DPV) test program was performed to confirm the adequacy of a ; squib-actuated valve to provide a reliable means of rapidly depressurizing the reac*orvessel. ! Performance tests were performed on the primer and pmpellant materials after exposme to the l SBWR environmental conditions. Functional tests were performed on a full-scale prototype valve - at the vendor's shop. The DPV was subjected to steam flow tests to measure the steam flow i capacity and reaction loads. Finally, the DPV was subjected to accelerated environmental aging of I the nonmetallic components, and dynamic testing. Results are reported in Reference 44. : i A.3.2.3.2 Test Objectives - l The test objectives of the DPV Test Program were:
- 1. Confirm that the DPV is a zero leakage valve, and that it opens on-demand with a -
j momentary electrical signal, opens within the required response time, and remains ( open without an extemal power source. i
- 2. Obtain data from flow testing to determine stresses in the DPV and confirm that the DPV saturated steam flow rate meets the minimum expected blowdown flow rate.
- 3. Obtain additional information on primer and propellant performance to pmvide evidence for later qualification testing.
'l- A-40
l
\
l- NEDO-32391, Revision C i i A.3.2.3.3 Test Matrix and Data Analysis ; 4 Samples of the primer and propellant materials were subjected to irradiation, accelerated thermal aging, and LOCA steam aging. Firing tests were subsequently performed andthe results ; l confirmed that the pressure output versus response time met the performance requirements for the l DPV. 1 Two full-scale pmtotype squib actuated DPVs were manufactured, assembled and tested by i Pyronetics Devices, Inc., a subsidiary of OEA, Inc., of Denver Colorado. Firing tests were
- performed on a full-scale valve under both a high pressure (1500 psig) condition at the valve inlet ;
{ and a low pressure (1 psig) condition at the valve inlet. A momentary electrical signal was l 1 supplied and it was confirmed that the valve opened within the required response time and ; remained open without an external power source. A thermal exposure heat transfer test was ; i l performed on the valve to assess the effects of ambient temperature and steamline temperature. It 3 was confirmed that the booster surface temperature was acceptable when the valve was exposed to j the SBWR environmental temperature conditions. A leakage test was performed for each valve
- metal diaphragm seal. Each seal was pressurized to 1650 psig and it was confirmed that there was zem leakage.
Flow and reaction load tests were performed on a full-scale valve at Wyle Laboratories of l Huntsville, Alabama. The test facility was modified to incorporate a prototypical SBWR steamline i section. The DPV was connected to this prototypical section and instrumented with pressure, i temperature, and strain gages, accelerometers and displacement transducers. Four steam : I . blowdown tests were performed. The test data confirmed that the DPV mass flow rate would be on the order of 2.4 x 106lbm/hr at an operating pressure of 1100 psia. l Potential environmental qualification effects were investigated by addressing two elements. One element was the accelerated aging of those DPV components that contain non-metallic materials to ensure their reliability under adverse in-plant conditions. The second element was to i subject a full-size prototype DPV to dynamically induced loads to simulate in-plant vibration. De j booster assemblies with the non-metallic materials were subjected to accelerated aging conditions
- and then successfully fired, confirming that adequate pressure was delivered. The dynamic l
simulation was performed on a triaxial seismic table at Wyle Laboratorics. He DPV was
- assembled using the aged components and then instrumented. The dynamic aging tests included i resonance search, vibration exposure (slow sine wave sweep) and a series of triaxial multi-frequency random input motion tests. It was confirmed that when signaled to actuate, the DPV opened and remained open.
j A.3.2.4 Vacuum Breaker Valve A.3.2.4.1 Test Description he vacuum breaker valve test program was designed to confirm that the vacuum breaker valve would provide a reliable leak tight boundary between the drywell and wetwell and prevent , the pressure in the wetwell from exceeding that of the drywell by more than three pounds per square inch. leak tightness is achieved by use of a nonmetallic main seal and a backup hard seat. De double seal design pmvides assurance that maximum leakage requirements will not be . exceeded in the event that an obstruction should lodge on either seat. A full scale prototypevalve l' A .
.. - - , - --- - ~ . -
l NEDO-32391, Revision C was built and subjected to flow testing to verify lift pressure, flow capacity, and stability at low ; flow. The primary nonmetallic seal was radiation and thermally aged. Following thermal aging, j the valve was dynamically aged and subjected to design basis accident conditions to confirm its leak tightness to steam. Finally, the fully aged valve was subjected to reliability testing to confirm j that its intrinsic reliability was consistent with the assumptions of the SBWR PRA. l i AJ.2.4.2 Test Objectives j He objectives of the vacuum breaker test program were to demonstrate that: l
- The vacuum breaker flow capacity could be made equivalent to 1.04 square feet I
- The vacuum breaker lift pressure was less than 0.5 psi.
- He disk was dynamically stable under low flow conditions.
- The hard seat equivalent flow area was less than 0.2 square centimeters. !
- ne main seal was air bubble tight as installed and has an equivalent leakage flow area !
ofless than 0.02 square centimeters to steam in the fully degraded conditionunder design basis accident conditions.
- The dynamic loads which result in lift of the disk were acceptable.
- The opening and closing reliability are maintained after subjecting the fully aged valve ,
to gritingestion. ! A3.2.43 Test Matrix and Data Analysis
- j. De vacuum breaker was air leak tested with a new seal and it was confirmed that the seal l l was bubble tight. The valve was then placed in the flow test facility and evaluated for lift pressure j l and low flow stability. He lift pressure and flow stability met requirements. He flow test l
- demonstrated that the valve stroke was not sufficient to meet minimum flow requirements. Since j the natural stability of the valve eliminated the need for a disk damper, the stroke was increased to f take credit for damper deletion. It was demonstrated that increasing the valve stroke results in j achieving the required flow performance. A seal was then aged with radiation and placed in the l l valve for thermal aging. The valve leak test was then repeated and it was shown that the seal was !
sir bubble tight. l i ! He valve was then placed on a shake table for fragility testing to determine at what !
~
[ acceleration, lift occuned. The valve was then subjected to ten Safe Shutdown Earthquake l
- acceleration time histories. Upon disassembly of the valve it was discovered that the ballast ring i and the position sensor screws had come loose due to failure to engage existing lock washers. j Screws had been ingested by the valve and hammered by the disk. Leak rate testing confirmed the i main seal was undamaged and the hard seat still exceeded leak tightness requirements despite i marnng.' The valve ruggedness and resistance to seal damage was demonstrated by this event. i The Design Basis Accident test demonstrated that the fully aged valve meets leak .
requurments at steam pressures and temperatures characteristic of a loss-of-coolant accident ! followed by water spray. De leak tightness of the valve was demonstrated by measuring the ! l A-42 f
- - , ~ , , -~ - ,,-o, -, - - -~ , -- - ---- e .-
l NEDO-32391, Revision C condensate from the steam that passed through the valve seals. During pmssure peaks, water sprays and 80 hours of endurance testing, no measurable condensate leaked through the valve. The test demonstrated the inherent steam leak resistance of the valve. The final test was the reliability testing, which subjected the fully-aged valve to grit ingestion to simulate possible environmental conditions that could affect bearing surfaces and seals during normal service. The valve was cycled thme thousand times to demonscate reliability at its mquimd statistical failum rate of 3x104 per demand. G l i i i i 1 4 l A-43 _ . _ ~ _ 3 __ - _ ___ ___ __ _l
NEDO-32391, Revision C Table A.2-1 Thermal Hydraulic Test Data Groups and Description Data Test Facility Group Conditions Description l PANTHERS /PCC P1 7 PCC steady state performance; saturated steam l l PANTHERS /PCC P2 6 PCC steady state performance; superheated steam ) l PANTHERS /PCC P3 4" PCC steady state performance; air / steam mixtures l PANTHERS /PCC P4 7** PCC steady state performance; air / steam mixtures l PANTHERS /PCC P5 2** PCC steady state performance; air / steam mixtures l PANTHERS /PCC P6 14 PCC steady state performance; air / steam mixtures PANTHERS /PCC P7 6 PCC performance; noncondensible buildup PANTHERS /PCC P8 3 PCC performance; water level effects l PANTHERSAC 11 10 IC steady state performance; inlet pressure effects PANTHERSBC 12 1* IC stan-up demonstration PANTHERSMC I3 2 IC restart demonstration, noncondensible venting PANTHERSAC I4 2 IC performance; waterlevel effects l PANDA /PCC S 7 PCC steady state performance; steam and air / steam mixtures l PANDA M3,4,7 3 Containment performance l PANDA M5,6,8 3 Containment performance l PANDA M2,9,10 3 Containment performance l GIRAFFE Phase 1 13 PCC steady state performance - steam and air / steam mixtures l GIRAFFE / Helium H 4 Containment performance - noncondensible density effects l GIRAFFE / Helium T 2 Containment p.formance " Tie-back" test l GIRAFFE / SIT GS 4 GDCS performance - integral systems tests GIST BDLB 7 GDCS performance - integrated system effects - bottom drain GIST MSLB 8 GDCS performance - integrated system effects - main steam . Gift GDLB 4 GDCS performance - integrated system effects - G DCS breaks GIST NB 7 GDCS performance - integrated system effects - transients ,
- Test to be performed twice to demonstrate repeatability.
** Test to be performed five times at different absolute pressures.
A-44
NEDO-32391, Revision C Table A.2-2 SBWR Test Documentation Submittals Test Submittal Title Document No. Actual Submittal Date PANDA Test Specification 22A5587 Rev. I 15 Feb 95 l As-Built Drawing Package MFN 044-95 27 Mar 95
. l Instrumentation Drawing Package QA Implementation Procedures PPCP-QA-01 16 Feb 95 Pre-Test Analysis (SI-S6) 40315-NUC-94-7034 27 Sept 94 l
l Test Plan and Procedures (SI-S9) ALPHA-410-0 20 April 95 Apparent Test Results (S1-S6) ALPHA-509-0 20 July 95 l l l Data Transmittal Repon (SI-S6) Test Plan and Procedure (M3,4,7) Pre-test Analysis M3 MFN 161-95 21 Aug 95 l Apparent Test Results (M3,4,7) Data Transmittal Report (M3,4,7) Test Plan and Procedure (M5,6,8) Pre-Test Analysis M5 ) 1 ) Apparent Test Results (M5,6,8) Data Transmittal Report (M5,6,8) I Test Specification (update for M1&9) 22A5587 Rev. 2 Test Plan and Procedure (MI,2,9) Pre-Test Analysis M2, M9 5 Apparent Test Results (MI,2,9)
- Data Transmittal Report (MI,2,9)
PANDA Data Analysis Report I GIRAFFE / Helium
- l Test Specification 25A5677 Rev. I 17 July 95 As-Built Drawing Package TOGE110-T19 Rev. 0 24 July 95 (MFN 124-95) j instrumentation Drawing Package TOGE110-T07 Rev. 2 14 July 95 l QA Plan TOGE110-701 Rev. I 27 April 95 l Test Plan and Procedures (T1, H1-H4, T2) TOGEl10-T07 Rev. 2 14 July 95
. l Apparent Test Results (H1, H2) l Pre-test Analysis (H1-H4) MFN 159-95 18 Aug 95 l Data Transmittal Report (H1, H2) Apparent Test Results (H3, H4) l Apparent Test Results (T1,T2) l Data Transmittal Report (H3, H4, T1, T2) i i A-45
NEDO-32391, Revision C Table A.2 2 SBWR Test Documentation Submittals (Continued) Test Submittal Title Document No. Actual Submittal Date GIRAFFE / SIT l Test Specification MFN 144-95 8 Aug95 l Test Plan and Procedures (GS1-GS4) Apparent Test Results Data Transmittal Report , GIRAFFE Data Analysis Report PANLERS/PCC Test Specification 23A6999 Rev.3 15 Feb 95 As Built Drawing Package many 30 Jun 94 QA Plan 006-QQ-92 8 Sept 94 Instrument Installation Spec. 00157S192 Rev.1 30 Jun 94 Pre-Test Analyses RAI 900.35 31 May 94 Data Acquisition Spec. 0095RS91 Rey,1 30 Jun 94 Test Man and Procedure 0098PP91 Rev. I 16 Aug 94 Process & Instrument Drawing 00209DD93 Rev. 4 12 Dec 94 Data Transmittal Report 00393RP95, Rev. 0 14 Apr 95,5 May 95 l PANTHERS /PCC Data Analysis Report 00394RA95, Rev. 0 6 Jul95 l PANTHERS /IC Test Specification 23A6999 Rev. 4 28 Apr 95 l As Built Drawing Package Many 21 Jun 95 l Test Plan 00396R195, Rev. 0 21 Jun 95 l Test Procedures 00395PP05 Rev. I 15 May 95 l l Process & Instrument Drawing 00210DD93 Rev. 3 21 Jun 95 l Apparent Test Results(Phase 1) Pre-Test AnalysisPackage MFN 097-95 5 July 95 l Apparent Test Results(Phase 3) , Data Transmittal Report PANHIERS/IC Data Analysis Report A-46
NEDO-32391, Revision C Table A.2 3 SBWR Analysis I Documentation Submittals Pre-Test Predictions PANTHERS /PCC-Complete (59] e l PANDA Steady State (S Series) - Complete (56] PANTHERS /IC PANDA M2 l PANDA M3 -Complete (79] PANDA M5 ; PANDA M9 GIRAFFE / Helium (blind post-test) GIRAFFE / SIT (blind post-test) l Preliminarv Validation Results l PANDA Steady State Tests , PANDA Transient Tests r PANTHERS /PCC PANTHERS /IC GIRAFFE / Helium GIRAFFE / SIT t F b W A-47 !
NEDO-32391, Revision C Table A.3-1 Required Thermal Hydraulic Measurements-PCC Test Accuracy Frequency j (2 Std. (samples Measurement Units Expected Range Dev.) per see) Pressures: 1 NMsible gas inlet kPa gage 0 - 760 (0 - 110) 2%* 0.1 Steaminlet (psig) 0 - 760 (0 - 110) 2% 0.1 . PCCinlet 30- 690 (5-100) 2% 0.1 Condensate tank gas space 30 - 690 (5 - 100) 2% 0.1
- PCC upper plenum 30- 690 (5- 100) 2% 0.1 ,
j Vent tank gas space 30- 690 (5-100) 2% 0.1
- DifTerentialpressures:
- Condensate tank / vent tank kPa (psi) d - 30 (0 - 5) 2% 0.1 Upper plenum / lower plenum 0 30(0-5) 2% 1 Condensate tank / upper plenum 30(0- 5) 2% 1 Flow Rates
- Steam inlet kg/s(Ib/s) 0 - 12 (0 - 25) 2% 0.1 Noncondensible inlet kg/s (Ib/s) 0 - 3 (0 - 5) 2% 0.1
- Condensate kg/s (1b/s) 0 - 12 (0 - 25) 2% 0.1
! Ventline gas kg/s (Ib/s) 0 - 3 (0 - 5) 2% 0.1
- Pool makeup 1/s(gpm) 0 - 13 (0 - 200) 2% 0.1
- Temperatures
- Steam inlet *C (*F) 100 - 177 (212 - 350) 3 (5) 0.1 l Noncondensible gasinlet 100- 177 (212 - 350) 3 (5) 0.1 Upper plenum 100- 171 (212 - 340) 3 (5)
, 0.1 PCC inlet 100- 171 (212- 340) 3(5) 0.1
- Iower plenum 10 171 (50 - 340) 3 (5) i Drain line 0.1 10 171 (50 - 340) 3 (5) i Drain tank 10 171 (50 - 340) 3 (5) 0.1 4
Ventline 10 171 (50 340) 3 (5) 0.1 . Vent tank 10 171(50-340) 3 (5) 0.1 PCC pool (6 places) 10 100 (50 - 212) 3 (5) 0.1 , Tube wall (inside & outside) 82 171 (180 - 340) 3 (5) 0.1 Pool makeup water 10 - 100 (50 - 212) 3 (5) 0.1 i Water levels (collapsed): , PCC pool m (ft) 3.5 - 5.0 (11.5 - 16.4) 0.03 (0.1) 0.1
- Drain tank 0 - 6.5 (0 - 21.2) 0.03 (0.1) 0.1 Drain line 0 - 6.0 (0 - 19.7) 0.03(0.1) 0.1 Vent tank 0 - 6.5 (0 -21.3) 0.03 (0.1) 0.1 Lower plenum, 0 - 3.0 (0 - 9.8) 0.03(0.1) 0.1 Other(indirect): MWth Heat rejection rate 0 - 15 0.3 0.02 System heatlosses 0 - 0.5 0.05 0.02
* % means percent of full-scale A-48
NEDO-32391, Revision C f l Table A.3-2a PANTHER /PCC Steady State Performance Matrix - Steam Only Tests Test Test Group Condition Steam Flow
- Air Flow
- Superheat i Number Number [kg/s (Ib/s)] [kg/s (Ib/s)] [ C( F)]
P1 37 0.45(1.0) 0(0) <10(18) P1 38 1.4(3.0) 0(0) . <10(18) P1 39 2.5(5.5) 0(0) <10(18) P1 40 3.6(8.0) 0(0) <10(18) P1 41 5.0(11.0) 0(0) <10(18) P1 42 5.7(12.5) 0(0) <10(18) P1 43 6.6(14.5) 0(0) <10(18) P2 44 1.4(3.0) 0(0) 15(27)* P2 45 1.4(3.0) 0(0) 20(36)* P2 46 1.4(3.0) 0(0) 30(54)* P2 47 5.0(11.0) 0(0) 15(27)* P2 48 5.0(11.0) 0(0) 20(36)* P2 49 5.0(11.0) 0(0) 30(54)*
- Nominal Value t Superheat conditions are relative to the steam partial pressure.
6 A-49
NEDO-32391, Revision C l Table A.3-2b PANTHERS /PCC Steady State Performance Matrix- Air-Steam Mixture Tests Test Test Steam Air Inlet Group Condition Flow
- Flow
- Pressure
- Superheati Number Number [kg/s(ib/s)] [kg/s(Ib/s)] [kPa (psia)] [ C( F)]
P3 9-1 5.0 (11.0) 0.076 (0.17) 296(42.9) <10 (18) P3 "-2 5.0 (11.0) 0.076 (0.17) 330(47.9) <10 (18) 9-3
- P3 5.0 (11.0) 0.076 (0.17) 385 (55.8) <10 (18)
P3 9-4 5.0 (11.0) 0.076 (0.17) 549 (79.6) <10 (18) P3 9-5 5.0 (11.0) 0.076 (0.17) 703 (101.9) <10 (18) P3 96 5.0 (11.0) 0.076 (0.17) 782 (113.4) <10 (18) P3 15-1 5.0 (11.0) 0.16 (0.35) 300(43.5) <10 (18) P3 15-2 5.0 (11.0) 0.16 (0.35) 329 (47.7) <10 (18) P3 15-3 5.0 (11.0) 0.16 (0.35) 441 (63.9) <10 (18) P3 15-4 5.0 (11.0) 0.16 (0.35) 500 (72.5) <10 (18) P3 15-5 5.0 (11.0) 0.16 (0.35) 648 (94.0) <10 (18) P3 15-6 5.0 (11.0) 0.16 (0.35) 790 (114.6) <10 (18) P3 18-1 5.0 (11.0) 0.41 (0.90) 284 (41.2) <10 (18) P3 18 2 5.0 (11.0) 0.41 (0.90) 300(43.5) <10 (18) P3 18 3 5.0 (11.0) 0.41 (0.90) 328(47.6) <10 (18) P3 18-4 5.0 (11.0) 0.41 (0.90) 467 (67.7) <10 (18) , P3 18 5 5.0 (11.0) 0.41 (0.90) 599 (86.9) <10 (18) P3 18-6 5.0 (11.0) 0.41 (0.90) 641 (92.9) <10 (18) P3 23 1 5.0 (11.0) 0.86(1.9) 296 (42.9) <10 (18) P3 23-2 5.0 (11.0) 0.86 (1.9) 329(47.7) <10 (18) , P3 23-3 5.0 (11.0) 0.86 (1.9) 437 (63.4) <10 (18) P3 23-4 5.0 (11.0) 0.86(1.9) 505 (73.2) <10 (18) P3 23-5 5.0 (11.0) 0.86(1.9) 584 (84.7) <10 (18) P4 2-1 1.4 (3.0) 0.014 (0.030) 179 (26.0) <10 (18) P4 2-2 1.4 (3.0) 0.014 (0.030) 201 (29.1) <10 (18) P4 2-3 1.4 (3.0) 0.014 (0.030) 299 (43.4) <10 (18) P4 13-1 2.5 (5.5) 0.16 (0.35) 244 (35.4) <10 (18) P4 13-2 2.5 (5.5) 0.16 (0.35) 296(42.9) <10 (18) P4 13-3 2.5 (5.5) 0.16 (0.35) 383 (55.5) <10 (18) A-50 l
. = _ .
NEDO-32391, Revision C l l Table A.3-2b PANTHERS /PCC Steady State l Performance Matrix. Air-Steam Mixture Tests (Continued) ; Test Test Steam Air Inlet , Group Condition Flow
- Flow
- Pressure
- Superheat i Number Number [kg/s(Ib/s)] [kg/s(Ib/s)] [kPa (psia)] [ C( F)]
P4 13 4 2.5 (5.5) 0.16(035) 470 (68.2) <10 (18)
- P4 13-5 2.5 (5.5) 0.16(035) 560 (81.2) <10 (18)
P4 16-1 6.6 (14.5) 0.16(035) 300 (43.5) <10 (18) 4 P4 16-2 6.6 (14.5) 0.16(035) 421 (61.0) <10 (18)
<10 (18)
P4 16-3 6.6 (14.5) 0.16(035) 538 (78.0) P4 16-4 6.6 (14.5) 0.16(035) 662 (96.0) <10 (18) P4 16-5 6.6 (14.5) 0.16(035) 788 (114 3) <10 (18) P4 17 1 2.5 (5.5) 0.41(0.90) 275 (39.9) <10 (18) P4 17-2 2.5 (5.5) 0.41 (0.90) 362 (52.5) <10 (18) P4 17-3 2.5 (5.5) 0.41(0.90) 453 (65.7) <10 (18) P4 17-4 2.5 (5.5) 0.41(0.90) 520 (75.4) <10 (18) P4 17-5 2.5 (5.5) 0.41 (0.90) 606 (87.9) <10 (18) P4 19-1 5.7 (12.5) 0.41(0.90) 295 (42.8) <10 (18) P4 19-2 5.7 (12.5) 0.41(0.90) 384 (55.7) <10 (18) P4 19-3 5.7 (12.5) 0.41(0.90) 472 (68.4) <10 (18) P4 19-4 5.7 (12.5) 0.41 (0.90) 567 (82.2) <10 (18) P4 19-5 5.7 (12.5) 0.41(0.90) 665 (96.4) <10 (18) P4 22-1 1.4 (3.0) 0.86 (1.9) 198 (28.7) <10 (18) P4 22-2 1.4 (3.0) 0.86 (1.9) 261 (37.8) <10 (18) P4 22-3 1.4 (3.0) 0.86 (1.9) 322 (46.7) <10 (18) P4 22-4 1.4 (3.0) 0.86 (1.9) 389 (56.4) <10 (18) l P4 22-5 1.4(3.0) 0.86 (1.9) 463 (67.1) <10 (18) P4 25-1 6.6 (14.5) 0.86 (1.9) 330(47.9) <10 (18) P4 25-2 6.6 (14.5) 0.86 (1.9) 381 (55.2) <10 (18) l P4 25-3 6.6 (14.5) 0.86 (1.9) 451(65 4) <10 (18) P4 25-4 6.6 (14.5) 0.86 (1.9) 530(76.9) <10 (18) P4 25-5 6.6 (14.5) 0.86 (1.9) 609(883) <10(18) P5 35-1 5.0 (11.0) 0.86 (1.9) 270 (39.2) 20 (36)* P5 35-2 5.0 (11.0) 0.86 (1.9) 298 (43.2) 20(36)* P5 35-3 5.0 (11.0) 0.86 (1.9) 359 (52.1) 20 (36)* P5 35-4 5.0 (11.0) 0.86 (1.9) 436 (63.2) 20(36)* A-51
NEDO-32391, Revision C Table A.3 2b PANTHERS /PCC Steady State Performance Matrix- Air-Steam Mixture Tests (Continued) Test Test Steam Air Inlet Group Condition Flow
- Flow
- Pressure
- Superheati Number Number [kg/s(ib/r;.] [kg/s(Ib/s)] [kPa (psia)] [*C( F)]
P5 35-5 5.0 (11.0) 0.86(1.9) 499 (72.4) 20 (36)* P5 35-6
- 5.0 (11.0) 0.86(1.9) 587 (85.1) 20 (36)*
P5 36-1 5.0 (11.0) 0.86 (1.9) 263 (38.1) 20 (36)* P5 36-2 5.0 (11.0) 0.86 (IS) 341 (49.4) 20 (36)* gi P5 36-3 5.0 (11.0) 0.86 (1.9) 422(61.2) 20 (36)* P5 36-4 5.0 (11.0) 0.86(1.9) 507 (73.5) 20 (36)* P5 36-5 5.0 (11.0) 0.86 (1.9) 558 (80.9) 20 (36)* P6 1-1 0.45(1.0) 0.014 (0.030) 300(43.5) <10 (18) P6 3-1 2.5 (5.5) 0.027 (0.060) 300(43.5) <10 (18) P6 4-1 3.6 (8.0) 0.027 (0.060) 300(43.5) <10 (18) P6 5-1 5.0 (11.0) 0.027 (0.060) 301(43.6) <10 (18) P6 6-1 %7 (12.5) 0.027 (0.060) 304 (44.1) <10 (18) P6 7-1 6.6 (14.5) 0.027 (0.060) 301(43.6) <10 (18) P6 8-1 1.4 (3.0) 0.076 (0.17) 300(43.5) <10 (18) P6 10-1 5.7 (12.5) 0.076 (0.17) 308 (44.7) <10 (18) P6 11-1 6.6 (14.5) 0.076 (0.17) 308(44.7) <10 (18) P6 12-1 0.15 (1.0) 0.16(035) 300(43.5) <10 (18) P6 14-1 3.6 (8.0) 0.16 (0.35) 303(43.9) <10 (18) P6 20-1 5.0 (11.0) 0.59 (1.29) 303(43.9) <10 (18) P6 21 1 6.6 (14.5) 0.59 (1.29) 353 (51.2) <10 (18) P6 24-1 5.7 (12.5) 0.86(1.9) 352(51.0) <10 (18)
- Nominal Value t Superheat referenced to steam partial pressure.
A-52
i NEDO-32391, Revision C Table AJ-2c PANTHERS /PCC Noncondensible - Buildup Matrix Test Steam Helium Test Group Condition Flow
- Flow
- Air Flow
- Superheati Number Number [kg/s(Ib/s)] [g/s] [g/s] [ C ( F)]
) P7 50 1.4 (3.0) 0(0) 4.4 <10 (18) P7 51 5.0(11.0) 0(0) 4.4 <10 (18) l P7 52 1.4 (3.0) 0(0) 4.4 20 (36)*
- 4.4 P7 53 5.0(11.0) 0(0) 30 (54)*
P7 75 1.4 (3.0) 0.7 0(0) <10 (18) P7 76 5.0(11.0) 0.7 0(0) <10 (18) P7 77 1.4 (3.0) 1.5 4.8 <10 (18) P7 78 5.0(11.0) 1.2 4.4 <10 (18)
- Nominal Value t Superheat referenced to steam partial pressure.
4 1 Table A3-2d PANTHERS /PCC Pool Water Level Effects - Test Matrix Steam Test Group Test Condition Flow
- Air Flow
- Superheati Number Number [kg/s (Ib/s)] [kg/s (Ib/s)] [ C ( F)]
, P8 54 5.0 (11.0) 0(0) < 10 (18)
P8 55 5.0 (11.0) 0.14 (0.31) < 10 (18) ; ~ P8 56 6.6 (14.5) 0.86(1.9) < 10 (18) I
- Nominal Value t Superheat referenced to steam partial pressure.
l l A-53
NEDO-32391, Revision C Table A.3-3 PANTHERS /PCC TRACG Qualification Points Test Pre / Post Condition Test Number Analysis Data Comparison 41 Post Heat Rejection Rate : PCC Pressure Drop 43 Post Heat Rejecdon Rate . 9 Post Heat Rejection Rate Degradation Factor PCC Pressure Drop 15 Pre / Post Heat Rejection Rate Degradation Factor PCC Pressure Drop 18 Post Heat Rejection Rate Degradation Factor PCC Pressure Drop 23 Pre / Post Heat Rejection Rate Degradation Factor PCC Pressure Drop 2 Post Heat Rejection Rate Degradation Factor 17 Post Heat Rejection Rate Degradation Factor 19 Post Heat Rejection Rate Degradation Factor 22 Post Heat Rejection Rate Degradation Factor , 35 Post Heat Rejection Rate Degradation Factor 49 Post Heat Rejection Rate - l 55 Post Inlet Pressure l 51 Post Inlet Pressure l 76 Post Inlet Pressure l 78 Post Inlet Pressure A-54
NEDO-32391, Revision C Table A.3-4 Required Therrr Mydraulic Measurements -IC Test Accuracy Frequency (2 Std. (samples Measurement Units Expected Range Dev.) per sec) Pressures: Steam vessel mPa gage 0.4 - 10.34 (70 1500) 2%* 0.1
, IC inlet (psig) 0.4 - 10.34 (70- 1500) 2% 0.1 IC upper plenum 0.4 - 10.34 (70- 1500) 2% 0.1 , Differential pressures:
IC inlet /IC vent line kPa(psi) 0 - 69 (0 - 10) 2% 0.1 ICinlet/IC drainline 0 - 69 (0 -10) 2% 0.1 Upper plenum / lower plenum 0 - 69 (0 - 10) 2% 0.1 j Elbow meter taps (4) 0 -? (0 -7) 2% 0.1 Flow rates:
. Steam inlet kg/s (Ib/s) 016(035) 2% 0.1 Noncondensible inlet kg/s (Ib/s) 0 - 0.3 (0 - 0.5) 2% 0.1 IC pool makeup 1/s (gpm) 0 - 11.4 (0- 180) 5% 0.1 Temperatures:
ICinlet steam *C (*F) 157 - 314 (315 - 598) 3 (5) 0.1 IC inlet pipe (6),(leak det.) 100 - 314 (212 - 598) 3 (5) 0.1 Drain line 10 - 314 (50 -598) 3 (5) 0.1 Ventlines(2) 10 - 314 (50 -598) 3 (5) 0.1 . Steam vessel 65 - 314 (150 -598) 3 (5) 0.1 IC pool (12 places) 10 - 1N (50 - 220) 3 (5) 0.1 Pool makeup water 10 IN(50 220) 3 (5) 0.1 Pool outlet temperature 10 - IN (50 - 220) 3 (5) 0.1
- Tubes (3 @ 5 axiallocations) 10 - 314 (50 -598) 3 (5) 0.1 Waterlevels(collapsed)
IC pool m (ft) 3.5 - 5.5 (11.5 18.0) 0.03 (0.1) 0.1 Simulated RPV later(later) 0.03 (0.1) 0.1 Drain line later(later) 0.03(0.1) 0.1 Ventlines(2) later(later) 0.03 (0.1) 0.1 Other (indirect): IC heat rejection rate MWth 0 20 0.1 0.02 System heat loss M Wth 0-1 0.1 0.02 ,. * % means percent of full-scale f A-55
NEDO-32391, Revision C c Table A.3-5a PANTIIERS/IC Steady State Performance - Test Matrix Test Test Condition Group Inlet Pressure l Number No. [MPag (psig)] 2 Il 7.920 (1150) - 3 11 7.240 (1050) 4 Il 6.21 (900) : 5 II 5.52 (800) 6 Il 4.83 (700) 7 Il 4.14 (600) 1 8 Il 2.76 (400) j 9 11 1.38 (200) 10 11 0.69 (100) 11 11 0.21 (30) l Table A.3 5b PANTHERS /IC Start'up and Operation - Test Matrix Test Test Initial Inlet Initial Pool Condition No. of Group Pressure, P1 Pressure, P2 Temp. Number Cycles Number [MPag(psig)] [MPag (psig)] [( C F)] 1 2 12 9.480 (1375) 8.618 (1250) <21 (70) s i t A-56
NEDO-32391, Revision C Table A.3-5c PANTHERS /IC Noncondensible Gas Effects - Test Matrix Test Condition Test Group Initial Inlet Number Number Pressure [MPag (psig)] d l 12 13 0.48 (70) l 13 13 2.08 (300) 1 1 ! l Table A.3-5d PANTHER /IC Water Level Effects - Test ' Matrix Test C ndition Test Group Initial Inlet Number Number Pressure [MPag (psig)] l 14 I4 0.48(70) l 15 I4 2.08 (300) J A-57
NEDO-32391, Revision C I Table A.3-6 PANTHERS /IC TRACG Analysis Cases Test ! Condition Pre / Post Number Test Data Comparison 2 Post Heat Rejection Rate 6 Pre / Post Heat Rejection Rate , 11 Post Heat Rejection Rate
?
12 Post Heat Rejection Rate Inlet Pressure 13 Pre / Post Heat Rejection Rate Inlet Pressure 15 Post Heat Rejection Rate s e A-58
NEDO 32391, Revision C l l l Table A3-7 PANDA Instrumentation Summary Measurement Type Instrument Type Number Total Temperature Chromel-alumel-thermocouples 442 Pt100-Resistance thermometers 21 Thermistors (TC ref, temp.) 30 493 8 Pressure Rosemount Model 33051CA transducer 15 Rosemount Model 2088A transducer 3 Rosemount Model 1144A transducer 3 21 Pressure difference Rosemount Model 3051CD transducer 14 Rosemount Model 1151DP transducer 13 27 Level Rosemount Model 3051CD transducer 7 Rosemount Model 1151DP transducer 11 18 l Gas concentration Oxygen partial pressure probe 8 8 I Flow rate Vortex flow meter 11 Ultrasonic flow meter 3 Hot film flowmeter 1 15 Fluid phase detector Conductivity probe 9 9 Electricalpower Wattmeter 6 Electronic totalizer 1 7 Total 598 A-59 l
NEDO-32391, Revision C l Table A.3-8 Instrumentation Required for Test Si to S13 - Ident. Code - Description , MVJ1F Steam flow to PCC3 MM. BOG Air flow to PCC3 MV.P3C PCC3 condensate flow (PCC3 to GDCS) l MV.P3V PCC3 Vent flow to WW2 ML.U3 PCC3 poollevel MLRP.1 RPV level MP.llF PCC3 upper header pressure MP.RP.1 RPV pressure MP.P3V PCC3 ventline pressure l MTG.P2F.1 Air / steam temperature in steady state supply line l MTO.P3F.1 Steam temperature in steady state supply line MTL.P3C.1 PCC3 condensate temperature at GDCS inlet MTL.GRT.1 PCC3 condensate temperature in GDCS drain line MTGP3V.1 Gas temperature in PCC3 ventline MIL.P3C.2 PCC3 condensate temperature in PCC3 outlet MTL.GRT.2 PCC3 condensate temperature at RPV inlet MTG.P3V.2 Gas temperature in PCC3 vent line outlet at PCC3 many PCC3 temperature
- It is required that 30% of the pool ternperature sensors and 50% of the tube wall and fluid sensors be available.The available pool sensors must include at least one of the three lowest elevations. The available tube wall and fluid sensors must include at least 40% of the probes above and below the horizontal mid-plane of the tube bundle. Within these constraints, the test engineer has the responsibility and authority to judge whether or not sufficient PCC3 temperature sensors are operable to initiate tests. -
A-60
NEDO-32391, Revision C l 1
)
l Table A.3-9a PANDA Steady State PCC Performance Test Matrix , GIRAFFE l PANTHERS Phase 1, PANDA Steam Flow Air Flow Test Condition Step 1 . Test No. (kg/s) (kg/s) No. Test No. ]: S1 0.195 0 41 2 i S2 0.195 0.003 9 4
, S3 0.195 0.006 15 6 S4 0.195 0.016 18 8 S5 0.195 0.034* 23 10 S6 0.26 0 43 3 l S10 0.195 0.006 15 6 l Sil 0.195 0.034* 23 10 l S12 0.26 0 43 3 l S13t 0.26 0
- It may not be possible for the PANDA air supply to deliver this flow rate. If this flow rate cannot be reached, then the test will be run at the maximum air flow rate that can be reached.
t Test S13 is to be conducted with PCC pool water level at bottom of upper PCC header. All other tests are to be run with normal PCC pool level (4.5m above bottom of pool). t 4 d A-61
NEDO-32391, Revision C Table A.3-9b PANDA System Test Matrix Summary Bypass PANDA Break No. of Drywell IC Leakage Initial Test No. Type PCC Spray Operation Area Conditions Comments , M1 Test Deleted M2 MSL 1in D W 1 no no 0 SSAR Asymmetric 0 % to D W 1 2 in DW2 steam flow to :
-100% to DW1 and 2 DW2 M3 MSL 1in D W 1 no no O SSAR Base Case 50% to each 2 in DW2 Same as DW GIRAFFE /HE testH1 M4 Same as M3 1 in DW1 no no 0 SSAR Repeatability 2 in DW2 M5 Same as M3 1 in D W 1 Yes no 0 SSAR Drywellspray to 2 in DW2 initiate vacuum breaker operation M6 Same as M3 1in D W l no Yes 0 SSAR IC operation 2 in DW2 M7 Same as M3 1in D W 1 no no 0 DW and PCC Boundingcase for 2 in DW2 initially PCC start-up air-filled M8 Same as M3 1in D W 1 no no 10 times SSAR DW to WW 2 in DW2 allowable bypass leakage M9 Conditions to be defined later M10 Conditions to be defined later O
A-62
NEDO-32391, Revision C l Table A.3-10a Initial Conditions for PANDA Test M3 RPV Drywell Wetwell GDCS PCC Pools Total Pressure (kPa) 295 294 285 294 101 Air / Nit. Pressure 0 13 240 274 n/a (kPa)
,. Vapor Temperature 406 405 352 333 n/a (K)
Liquid Temperature 406 405 352 333 373 (K) Collapsed WL (m)(O 11.2 (2) 3.8 10.7 23.2 l Notes: (1) Water levels are specified relative to the top of the PANDA heater bundle. , (2) The nominal DW condition is no water. However, a small amount of spill from the RPV to the DW at the start of the test is acceptable. Table A.3-10b Initial Conditions for PANDA Test M7 RPV Drywell Wetwell GDCS PCC pools Total Pressure (kPa) 101 101 101 101 101 0 (5) 56 81 n/a Air / Nit. Pressure (kPa) Vapor Temperature (K) 373 (5) 352 333 n/a Liquid Temperature (K) 373 (5) 352 333 373 Collapsed WL (m)(1) (2) (3) 3.8 (4) 23.2
. Notes-(1) Water levels are specified relative to the top of the PANDA heater bundle.
(2) The RPV water level will be such that the swollen level rises just to the elevation of the i steamline when the heaters are ramped to their initial power level. (3) The nominal DW condition is no water. However, a small amount of spill from the RPV to the DW at the start of the test is acceptable. l (4) The GDCS water level will be set in hydrostatic equilibrium with the initial RPV level. (5) PANDA capability for producing a dry air environment in the DW is currently being investigated. A-63
NEDO-32391, Revision C l Table A.3-11 SBWR Containment Conditions at 3600 sec for Main Steamline Break LOCA level / Ring Ring 1(R=2.5sn) Ring 2(R=3.0m) Ring 3(R=4.26se) Ring 4(R.>7.8m) Ring 5(R=13.75sa) Ring 6(R=15.75m) level 9 vol=29.598 m 3 vd=13.063 m 3 vd=43A52 m3 vd=202.792 m 3 not used not used Z=24.812m dw head dw head upper dw upper dw p=293.85 kPa p=293.85 kPa p=293.85 kPa p=293.85 kPa pe=101.95 kPa pa=50.79 LPa pe=934 kPa pa- 9.66 kPa alp =1.000 alp =1.000 alp =1.000 alp =1.000 I Tv=394.56 K Tv=409.63 K Tv=439.97 K Tv=440L18 K U=n/a T1=n/a TI=n/a T1=n/a level 8 vd=85.726 m3 vd=37.720 m 3 vd=196t606 m3 vd=917.570 m3 vol=232.509 m 3 vd=465.047 ms Z=2330m epv-sim dame rpy-stm dome upper dw upper dw goes pt-up cell goes pl-up cell p=294.71 kPa p=294.70 kPa P293.89 kPa p=293.89 kPa p=293.90 kPa p=293.90 kPa ps=3.91 kPa pe=3.10 kPa pa=9.23 kPa pe=8.91 kPa pa=48.12 kPa pa-4937 kPa alp 0.908 Gevel) alp =0.909 Gevel) alp =1.000 alp =1.000 alp 0.785 Oevel) alp =0.786 Gevel) Tv=424.40 K Tv=419.81 K Tv=439.92 K Tv=437.24 K Tv=417.02 K Tv=414.92 K D=404.65 K TI=405.26 K E=n/a T!=n/a T!=335.24 K T1=333Al K Level 7 vd=20.263 m 3 vol=50.178 m3 vd=127.528 m3 vol=595.180 m 3 vol=150.817 m 3 vol=301.652 m 3 Z=19.60m rpv.sep region rpv-up dwncmr upper dw upper dw gdes pl-low cell goes pt-low cell p321.90 kPa p=321.64 kPa p=293.94 kPa 7293.94 kPa p=313.08 kPa p=313.04 kPa pe=0.00 kPa pad).00 kPa pa=10.75kPa pe=9.60 kPa pa=n/a pa=n/a alpaC302 (mixture) a!p=0.017 (mixture) alp =1.000 alp 1.000 alp =0.000 alpd).000 Tv=408.09 K Tv=408.06 K Tv=433.94 K Tv=437.75 K Tv=n/a Tv=n/a U=408.22 K Tl=407.88 K T1 = n/a D=n/a TI=330.66 K TI=333AlK 3 3 3 3 Level 6 vol=6.597 m vol=12A93 m vol=31.882 m vd=148.795 m not used not used Z=17.20m rpv-sind pipes rpv-dwncmr upper dw upper dw p=332.29 kPa p=335.26 kPa p=293.97 LPa p=293.97 kPa pe=n/a pa=n/a pa=16.54 kPa pe=15.70 kPa alp =0.000 alp 0.000 alp =1.000 alp 1.000 Tv=n/a Tv=n/a Tv=412.54 K Tv=415.65 K U=409.15 K TI=407.77 K T1=n/a Tl=n/a Level 5 vol=93.030 m 3 vol=40A36 m3 vd=244.429 m3 vol=1140.763 m3 vol=1852.985 m3 vol=852.628 m3 Z=16.60m rpv-chimney rpv-d wnmer drywell drywell ww-vap space ww.vap spam p=355.91 kPa p=358.90 kPa p294.01 kPa p=294.01 kPa p=284.79 kPa p=284.79 kPa pe=n/a pe=n/a pa=16.56 kPa pa=16.51 kPa pa=255.43 kPa pa=26235 kPa alp 0.000 alp 0.000 alp =1.000 alp =1.000 alp =1.000 alp =1.000 1 Tv=n/a Tv=n/a Tv=412.42 K Tv =412.73 K Tv=346.52 K Tv=342.06 K n=409.15 K TI=40734 K E=n/a D=n/a TI=n/a TI=n/a level 4 vol=40.448 m3 vol=17.624 m3 vol=106.273 m3 vol=495.984 m3 vd=805.646 m3 vol=370.708 m3 . Z=12.00m rpv-chimney rpvwiwncmr drywell drywell ww.vap space ww.vap space p=385.89 kPa p=388.93 kPa p=294.07 kPa p=294.07 kPa p=284.88 kPa p=284.88 kPa pe=n/a pa=n/a ps=16.55 kPa ps=16.53 kPa pa=244.15 kPa pa=248A1kPa alp 0.000 alp =0.000 alp =1.000 alp l.000 alp =0.839 0cvel) alp =0.838 0evel)
- Tv=n/a Tv m/s Tv=412.40 K Tv=412.65 K Tv=35139 K Tv=348.85 K U=409.13 K TI=407.13 K R=n/a D=n/a TI=35231 K T1=351.97 K Ievel 3 vol=44.493 m3 vol=19387 m3 vol=33.868 m3 vol=158.157 m3 vol=261.432 m3 vol=120.295 m 3 Z=10.00m rpv-chimney rpv.dwnernr dryweD drywell ww.sup pool ww-sup pod p404.97 LPa p=408.05 LPa p=294.10 kPa P294.11 kPa p=298 42 LPa p=298.44 kPa pe=n/a pe=n/a pe=1832 kPa ps=1830 kPa pa=n/a pa=n/a alp =0.000 alp 0.000 alp =1.000 alp =1.000 alp =0.000 alp =0.000 Tv=n/s Tv=n/a Tv=412.40 K Tv=412.53 K Tv=n/a Tv=n/a TI-408.49 K TI =406.82 K E=n/a TI=n/a TI=35232 K TI-352.06 K A-64
NEDO 32391, Revision C Table A.3-11 SBWR Containment Conditions at 3600 sec for Main Steamline Break LOCA a levd' Ring Ring 1(R=2.5e) Ring 2(R=3.0m) Ring 3(R=4.26m) Rlag 4(R=7.8m) Ring 5(R=13.75m) Ring 6(R=15.75m) level 2 vd=18.555 mi vd=18.233 m 3 vd=142.033 m 3 vd=26.000 m 3 vd=549.430 m3 vd=251814 m 3 Za7.80m rpv-core bypass rpv-dwncmr drywell drywell ww-sup pod ww-sup pod P429.45 kPa p=43237 kPa p294.15 kPa P294.15 kPa p=323:13 kPa p=323.73 LPa ]' pe=n/a pe=n/a pe=16.56 kPa pe=16.55 kPa pa=n/a pa=n/a j alpo.000 alp =0.000 alpi.000 alp =1.000 alp =0.000 alp =0.000 Tv=n/a Tv=n/a Tv=412.29 K Tv=412.37 K Tv=n/a Tv=n/a E=404.99 K Tl=406.36 K Tl=n/a E=n/a U=352.03 K TI=351.99 k 3 , level I vd=63.912 m 3 vd=18.078 m 8 vd=31.804 m3 vd=148.433 m not used not used 1 Z=4.65m rpv-lwr plenum rpv-lwr pienum lower dw lower dw p=466.17 kPa p=466.18 kPa p=294.22 kPa p294.22 kPa
, pa=n/a pe=n/a pe=16.54 kPa pe=16.53 kPa alp =0.000 alp =0.000 alpl.000 alpl.000 Tv=n/a Tv=n/a Tv=412.30 K Tv=412.3 K D=403.87 K T!=405.85 K E=n/a E=n/a 3 3 TEE 35 vd=187.2 m'(cell vd=233.0m (cell vd=213.0 m (cell
- 3) 2) 1) twrdw Z=-3.4m twr dw, Z=.6.8m lwr dw.Z= 10.an p294.29 kPa p=29435 kPa p305.89 kPa pe=12.87 kPa pa=9.86 kPa pe=10.94 kPa
. alp l.000 alp =1.000 alp 0.189 (level) Tv=410.8 K Tv=408.4 K Tv=405.1 K En/a TI=n/a E=405.1 K 4 3
.- 1 i
1 a . ; 1 1 A-65
l NEDO-32391, Revision C Table A.3-12 Time Derivatives of Key PANDA Initial Conditions Time Derivative PANDA Initial SBWR@ t=3600 l Parameter Condition (kPa) (Pa/sec) Ratio (1/sec) Drywell Pressure 294 0.899 3.06 x104 3.15 x 10 4 i Wetwell Pressure 285 0.899 4 Wetwell AirPartial 240 1.29 5.38 x 10 Pressure : 8.38 x 10 4 Mid DW Air Partial 13 0.11 Pressure 4 Upper DW Air 13 -1.726 -1.33 x 10 Pressure e A-66
NEDO-32391, Revision C Table A.3-13 PANDA TRACG Analysis Cases Test Pre / Post Data Number Test Comparison S1 Pre / Post Heat Rejection Rate S2 Pre / Post Heat Rejection Rate Degradation Factor
~
S3 Pre / Post Heat Rejection Rate Degradation Factor
. S4 Pre / Post Heat Rejection Rate Degradation '~
Factor S5 Pre / Post Heat Rejection Rate Degradation Factor S6 Pre / Post Heat Rejection Rate M1 DELETED M2 Pre / Post Drywell Pressure Drywellair distribution Wetwell Pressure Wetwellair distribution Drywell Temp. Wetwell Temp. Suppression PoolTemp. PCC Flows M3 Pre / Post Drywell Pressure Drywellair distribution
- Wetwell Pressure Wetwellair distribution Drywell Temp.
Wetwell Temp. PCC Flows Suppression PoolTemp. M4 Same as M3 A-67
NEDO-32391, Revision C Table A.3-13 PANDA TRACG Analysis Cases Test Pre / Post Data Number Test Comparison i M5 Pie / Post Drywell Pressure DrywcIl air distribution l Wetwell Pressure Wetwc!! air distribution Drywell Temp. : Wetwell Temp. Suppression PoolTemp. ,
~ , PCC Flows Vacuum Breaker Flows M6 Post Drywell Pressure Drywellair distribution Wetwell Pressure Wetwellair distribution Drywell Temp.
Wetwell Temp. Suppression PoolTemp. 4 PCC Flows IC Flow M7 Post Drywell Pressure Drywellair distribution Wetwell Pressure Wetwellair distribution Drywell Temp. Wetwell Temp. Suppression Pool Temp. PCC Flows M8 Post DrywellPressure Drywell air distribution , Wetwell Pressure i Wetwell air distribution Drywell Temp. . Wetwell Temp. PCC Flows Suppression PoolTemp. I.cakage Flow M9 Pre / Post To Be Determined M10 Post To Be Determined A-68
NEDO-32391, Revision C Table A.3-14 GIST Test Matrix Initial Conditions (RPV at 100 psig) No. of RPV Scram Decay LDW UDW S/P S/P WW GDCS Level Time Heat Level Press. Level Temp. Press. TestW Lines (in)W (sec)W (kW) (in) (psig) (ft) ('F) (psig) BDLB Tests: . A01 Base Case 3 347 369 89 4 13.0 67.2 105 6.5 A02 Low S/P Water 3 347 369 89 4 13.0 59.2 105. 6.5 level I A03 Maximum 4 347 369 89 4 13.0 67.2 105 6.5 GDCS Flow A04 Low RPV 3 327 369 89 4 13.0 67.2 105 6.5 Water Level A05 CRD Level 3 347 369 89 4 13.0 67.2 105 6.5 A06 Minimum 1 347 369 89 4 13.0 67.2 105 6.5 GDCS Flow A07 NoIow Press 3 347 369 89 4 13.0 67.2 105 6.5 DPVs MSLB Tests: i B01 Base Case 3 340 212 99 6 14.5 67.2 110 7.0 l B02IowPRVWater 3 320 212 99 6 14.5 67.2 110 7.0 level B03 low S/P Water 3 340 212 99 6 14.5 67.2 110 7.0 l Level 1 l B(M First Repeat 3 340 212 99 6 14.5 67.2 110 7.0 l Test ! B06 Last Repeat 3 340 212 99 6 14.5 67.2 110 7.0 Test B07 Low-Low RPV 3 300 21". 99 6 14.5 67.2 110 7.0
- Wl' B08 Accumulator 3 300 212 99 6 14.5 67.2 110 7.0 Makeup B09 Accumulator 3 286 212 99 6 14.5 67.2 110 7.0 Makeup GDLB Tests:
C01 A Base Case 2 347 373 88 5 11.5 67.2 105 7.0 C02 Max HP DPV 2 347 373 88 5 11.5 67.2 105 7.0 Area C03 Min HP DPV 2 347 373 88 5 11.5 67.2 105 7.0 Area A-69
NEDO-32391, Revision C Table A.3-14 GIST Test Matrix Initial Conditions (RPV at 100 psig) (Continued) No. of RPV Scram Decay LDW UDW S/P S/P WW GDCS Level Time licat Level Press. Level Temp. Press. 1 Test (D Lines (in)W (sec)(3) (kW) (in) (psig) (ft) (*F) (psig) l C04 High LP DPV 2 347 373 88 5 11.5 67.2 105 7.0 ; Setpt. NB Tests: D01 A Base Case 3 347 865 74 0 0.0 67.2 107 0.0 D02 Maximum 4 347 865 74 0 0.0 67.2 107 0.0 GDCS Flow . D03A App.K Decay 3 347 865 94 0 0n 67.2 107 0.0 Heat D04 Pressunzed 3 347 865 74 0 14.7 67.2 107 14.7 WW DOS High Pool 3 347 865 74 0 0.0 67.2 157 0.0 Temp. D06 Low GDCS 4 347 865 74 0 0.0 67.2 107 0.0 Injection D07 No Power 3 347 - 0 0 0 67.2 107 0.0 Notes: (1) Suffix "A"in Test Number signifies a repeat test. (2) Collapsed waterlevel relative to bottom of RPV. (3) Time since reactor scram in SBW. Used to determine decay heat. l i A-70
NEDO-32391, Revision C Table A.3-15 GIST Runs With ExistingTRACG Analysis Run Type B01 MSLB, Base Case B07 MSLB, Low Initial RPV Level C01A GDLB, Base Case A07 BDLB, No low Pressure DPVs D03A NB, Zero Containment Pressure Table A.3-16 GIRAFFE Test Matrix (Phase I Step-1) Nitrogen Partial Steam Flow Pressure Pressure Test No. Test Group Rate (kg/s) (fraction of total press.) (kPa) I l 1 Phase 1 0.02 0 300 l 2 Phase 1 0.03 0 300 l l l 3 Phase 1 0.N 0 300 l 4 Phase 1 0.03 0.01 300 l 5 Phase 1 0.02 0.02 300 l 6 Phase 1 0.03 0.02 300 l 7 Phase 1 0.M 0.02 300 l 8 Phase 1 0.03 0.05 300 l 9 Phase 1 0.02 0.10 300 . l 10 Phase 1 0.03 0.10 300 l l 11 Phase 1 0.N 0.10 300 l 12 Phase 1 0.03 0.02 200 l 13 Phase 1 0.03 0.02 400 4 A-71
NEDO-32391, Revision C Table A.3-17 GIRAFFE / Helium Integral Systems Test Matrix Drywell Initial Partial Pressures (kPa) (i2kPa) Helium GIRAFFE Test Injection Rate No. (Kg/sec) Nitrogen Steam Helium H1 0 13 281 0 H2 0 0 281 13 H3 0 13 214 67 H4 0.00027 13 281 0 Table A.3-18 GIRAFFE / Helium Base Case (HI) Initial Conditions Parameter Value Tolerance RPV Pmssure (kPa) 293 i6 kPa l Initial Heater Power (kW) 93 1kW l RPV Water Level (m)* 12.0 0.150 m Drywell Pressure (kPa) 294 14 kPa Wetwell Pmssure (kPa) 285 i4 kPa Wetwell Nitrogen Pressure (kPa) 240 i4 kPa GDCS Gas Space Pressum (kPa) 294 4 kPa GDCS Nitrogen Pressum (kPa) 274 14 kPa Suppression Pool 7'emperature (K) 352 12 K ., PCC Pool Temperature (K) 373 12 K GDCS Pool Temperature (K) 333 i2 K - i GDCS Pool Level * (m) l Suppression Pool Level * (m) 3.25 i0.075 m PCC Pool Collapsed Water Level * (m) 23.2 i0.075 m l PCC Vent Line Submergence (m) 0.95 i0.075 m
- Referenced to the Top of Active Fuel (TAF) t GDCS poollevel should be positioned in hydrostatic equilibrium with the RPV level (including an appropriate adjustment for temperature difference).
A-72
\
NEDO-32391, Revision C l 1 Table A.319 GIRAFFE / Helium " Tie-Back" Initial Conditions Parameter Value Tolerance RPV Pressure (kPa) 189 i6 kPa RPV Collapsed Water Level (m)* 9.1 10150 m Initial Heater Power (kW) 96 ilkW Drywell Total Pmssure (kPa) 188 i4 kPa
; Drywell Nitrogen Partial Pressure (kPa) 53 4 kPa Drywell Steam Partial Pressure (kPa) 135 4 kPa Wetwell Pressure (kPa) 174 i4 kPa Wetwell Nitrogen Pressure (kPa) 164 4 kPa GDCS Pool Gas Space Total Pressure (kPa) 188 4 kPa GDCS Pool Gas Space Nitrogen Partial 151 i4 kPa Pressure (kPa)
Suppression Pool Temperature (K) 326 12 K PCC Pool Temperature (K) 373 i2 K GDCS Pool Temperature (K) 350 i2 K GDCS Pool Level * (m) 14.1 i0.075 m , l Suppression Pool Level *(m) 3.2 i0.075 m ) PCC Pool Collapsed Water Level * (m) 23.2 10.075 m PCC Vent Line Submergence (m) 0.90 10.075 m I
- Referenced to the TAF.
A-73
NEDO-32391, Revision C Table A.3 20 GIRAFFE / Helium Test T2 Initial Conditions Parameter Value Tolerance RPV Pressure (kPa) 267 6 kPa l Initial Heater Power (kW) 93 1Kw RPV Water Level (m)* 12.0 0.150 m l , DrywellPressure(kPa) 266 i4 kPa DrywellNitrogen Pressure (kPa) 38 4 kPa
~
~
WetwellPressure (kPa) 257 4 kPa l ', WetwellNitrogen Pressure (kPa) 212 4 kPa i GDCS Gas Space Pressure (kPa) 266 i4 kPa GDCS Nitrogen Pressure (kPa) 246 i4 kPa
- Suppression PoolTemperature (K) 352 12 K PCC PoolTemperature (K) 373 i2 K GDCS PoolTemperature (K) 333 12 K i
- GDCS Poollevel(m) l Suppression PoolLevel* (m) 3.25 0.075 m 1 PCC Collapsed Water Level *(m) 23.2 0.075 m l PCC Vent Line Submergence (m) 0.95 0.075 m
.
- Referenced to the TAF.
4 t GDCS pool level should be positioned in hydrostatic equilibrium with the RPV level (including an appropriate adjustment for temperature difference. i 1 Table A.3-2I GIRAFFE / SIT Test Matrix Test Break Single Failure IC/PCCS on? ., GS1 GDL DPV No i GS2 GDL DPV Yes . GS3 BDL DPV Yes GS4 GDL GDCS Yes GDL = Gravity Drain Line BDL = Bottom Drain Line DPV = Depressurization Valve GDCS = GDCS Injection Valve A-74
NEDO-32391, Revision C Table A.3-22 Test GS1 Initial Conditions The information previously provided in this table is in the Test Specification. 4 3 t i a 4 Table A.3 23 Basis for GIRAFFFJSIT Test Conditions Option IC/PCC Objective Break Failure Operation Test ID 2 Worst Break / Single Failure GDL DPV No GS1
- Combination Benefit ofIC/PCC GDL DPV No GS1 :
and GDL DPV Yes GS2
~
Slow Water Level Recovery GDL GDCS Yes GS4 ; l Fast Water Level Recovery BDL DPV Yes GS3 l Case showing GDCS void GDL DPV Yes GS2 quenching and break flow GDL DPV No GS1 ! depressurizing drywell l l A-75
NEDO-32391, Revision C Table A.3-24 PANTHERS PCC StructuralInstrumentation Quantity at Total each Measure-Measurement / Location No. of Positions Position ments Direction (s) Acceleration: Steam distributor 1 3 3 X,Y,Z : Mid-lengthof tube 5 2 10 X, Y Upper header cover 1 3 3 X,Y,Z Displacement: Inlet /headerjunction 1 2 2 X, Z Steam distributor 1 1 1 Z Lower header suppon 2 1 2 Y Total Strain: axial Inlet elbow 1 2 2 Z Inlet /headerjunction 1 2 2 Z Upper header /tubejunction 5 1 or 2 7 Z Tube /lowerheaderjunction 3 1 3 X, Y Lower header 2 2 4 Z, X Lower header cover 1 2 2 X, Z Upper header 2 4 8 X, Z Upper header cover 1 4 4 Y Upper headercover bolts 3 1 or 2 5 Y Lower headercover bolts 3 1 or 2 5 X, Z Drain /lowerheaderjunction 1 2 2 Z Lower header suppons 1 2 2 Pennanent strain: Inlet /headerjunction 1 1 1 Z Upper header /tubejunction 3 1 3 Z Lower header /drainjunction 1 2 2 Z Temperature: l Steamline 2 1 1 1 . Temperature: l Inlet /headerjunction 1 1 1 Upper header /tubejunction 3 1 3 - Tube /lowerheaderjunction 3 1 3 Lower header 2 1 2 Lower headercover 1 1 1 Upper header 2 2 4 Upper headercover 1 2 2 Drain / lower headerjunction 1 1 1 A-76
NEDO-32391, Revision C Table A.3-25 PANTHERS PCC Component Demonstration Test Matrix Maximum Number of Maximum Temperature Cycle Duration Cycle Type Cycles Pressure (kPa) (Deg C) (Min.) LOCA 10 379 Saturation 30 I Pneumatic Test 300 758 Ambient 2 t Table A.3-26 PANTHERS /PCC LOCA Cycle Time History
- PCC Inlet Pressure Time to Reach
[kPag (psig)] Pressure (Sec) 175 (25.4) start 249 (36.1) <30 261 (37.8) <65 379 (55) <30 minutes
- ;
- The unit is initially pressurized with air at ambient conditions.
A-77
- NEDO-32391, Revision C l Table A.3-27 PANTHERS IC Structural Instrumentation Quantity at No. of each Measurement / Location Positions Position Total Measurements Acceleration-Mid-length of tube 5 2 10
Drainline curve 1 3 3 Lower header cover 1 1 1 Upperheader cover 1 5 3 , Displacement: Steam distributor 1 1 1 Drain /lowerheaderjunction 1 1 1 Steam pipe lower zone 1 1 21 Total Strain: Inlet / upper headerjunction 1 6 6 Upperheader/tubejunction 5 1 or 2 7 Mid-length of tube 3 1 3 Tube / lower headerjunction 3 1 3 14wer header 2 2 4 Lowerheader cover 1 2 2 Upperheader 2 4 8 Upperheader cover 1 4 4 Drain /lowerheaderjunction 1 4 4 Drainline curve 1 2 2 Drainline/ drain tube 1 4 4 Upper header cover bolts 3 2 or 1 5 Lowerheader cover bolts 3 2 or 1 5 Guard pipe / distributor 1 3 3 Support 1 2 2 Upperheader near support 1 4 4 Permanent strain: Inlet /headerjunction 1 3 3 Upper header /tubejunction 3 1 3 Lowerheader/drainjunction 1 1 2 Temperature: Guard pipe / distributor 1 1 1 Inlet pipe / upper header 2 2 4 , Upperheader/tubejunction 3 1 3 Tube / lower headerjunction 3 1 3 Lower header 2 1 2 Upper header 2 2 4 Drainline bend 1 1 1 Upperheader cover 1 2 2 Lower header cover 1 1 1 r A-78
NEDO-32391, Revision C l Table A.3-28 PANTHERS IC Component Demonstration Test Matrix l Test Initial Inlet Inlet Pressure, Inlet Pressure, Cond. No. of Cycle Pressure, P1 P2 P3 Initial Pool No. Cycles Type [MPag (psig)] [MPag (psig)] [MPag (psig)] Temp. *C( F) l 1 1 2 9.480 (1375) 8.618 (1250) N/A <21 (70) l 16 20 5 8.618 (1250) 8.618 (1250) N/A <32 (90) l 17 5 6 8.618 (1250) N/A N/A <32 (90) l 18 1 7 8.618 (1250) 9.480 (1375) 8.618 (1250) <32 (90) 't i A-79
f NEDO-32391, Revision C I Table A.3-29 Comparison of Non-Dimensional Parameters Between SBWR and CRIEPI Non-Dimenr%nal Parameters Physical Meaning Full Power Condition An example of startup (7.2 MPa) condition (0.1 MPa) Reactor Test Reactor Test Facility Facility , Froude Number F, gravity to fluid inertia 4 4 0.058 0.053 10.5x10 7.6x10 ratio l Loss cocf. in channel 4 pressure loss coefficient 3.4 2.7 6.9 5.7 - l l Imss coef. at channelinlette,in 50 30 50 30 Loss coef.at chimney exit 31 21 31 21 (separatorloss)(,,o Phase change number N pch quantity of heating in 3.7 3.7 11.6 13.1 channel l Subcool number Neub channelinlet subcooling 0.58 0.58 9.0 9.0 l Flashing parameter Nr quantity of flashing 0.057 0.036 67 46 Ratio of vapor density toliquid density ratio 0.052 0.052 6.2x104 6.2x10 4 one R,,i Ratio of vapor density at channel 1.01 1.01 2.01 1.63 inlet to that of dome pressure Pa2At Nondimensional downcomer parameters depending 1.05 1.11 1.05 1.11 cross sectional am A d.e on the test facility shape Nondimensional chimney cross 2.59 2.47 2.59 2.47 sectionalarea A, Nondimensional chimney length 3.34 3.38 3.34 3.38 1 Nondimensional drift velocity vgi relative velocit, 0.138 0.183 1.32 1.97 between vapor phase , and liquid phase Arbitrary condensation parameter subcooled boiling 0.62 0.52 0.035 0.029 Ho . l Peclet number PE 120000 590000 13500 6000 Thermodynamic equilibrium 4 4
-0.047 -0.089 -3.45x10 3.81x10 quality at void departure point xd A-80
{ I l NEDO-32391, Revision C l i i o,
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h PCC capaaty - decay heat
@ Reduced non-condensables Top LOCA Vent l l 2e Submergence au l h Reduced non-condensables l A w a g l 4 Low pressure limit vacuum breaker opens l
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I s-tm h] Vacuum Breaker l g g g $f& k 5 Setpoint g l l l I I I I V Drywell PCC Hx PCC Hx Suppression PCC Vent inlet Outlet Pool Oudet Surface Distance Along Flowpath i Figure A.3-3 PCC Heat Exchanger Operational Modes
NEDO-32391, Revision C 1. 0.9 - 23 3 22 Ay X 24 m 25 0.7 y
- TEST GROUP P, g + TEST GROUP P3 6 0.6 m TESTGROUP P4 A TEST GROUP PS x 20 x 21 x TEST GROUP P6 0.5 -
8 1'0.4 g - I 1 m 19 4 18 0.3 - BREA 0.2 - MSL BREAK x ,2 ,, x ,4 .,, N -,e o 0 1_ . 2 35 yvvfM /,[4 I 1 2 3 ~4 5 6 7 STEAM MASS FLOW (kg/sec) Figure A.3-4 Comparison of PANTHERS /PCC Steam-Air Test Range to SBWR Conditions A-84
l NEDO-32391, Revision C , l 09 - e @
., x 24 .=
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NEDO-32391, Revision C Figure A.3-12 Deleted A-92
I NEDO-32391, Revision C PCC,S,TE_A,D,Y,STA3,S,UPPLY TUBE WALL / GAS TEMPS i IC PCC PCC PCC POOL TEMPS 4XO - 1- 2 3- X2 OTHER 3 IDENTICAL FOR ALL 3X \ Y 1 1 1 X1 CONDENSERS FOUR CONDENSERS 2XA[2" ": X4 0x1 2X b:3 [M INCLUDES 1 GAS TEMP SX :: ;; M X1 X5 O X3 AT EACH LEVEL 3X .n.)I NX4 O X2 I T: 4)' X2 4X O X1 IC-DRAIN j l PCC 3 VENT IC-SUPPLY
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NEDO-32391, Revision C IC PCC 3 STEADY STATE SUPPLY LINE TUBE WALUGAS TEMPS --g MTG.P2F.1 AIR SUPPLY INCLUDES 1 GAS TEMP UNE MTG.P3F.1 b AT EACH LE'KL POOL TEMPS 4X X2
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NEDO-32391, Revision C T TEMPERATURE . STEAM B P T P PRESSURE , DP DIFFERENTIAL n +_ 1611D OUTER INNER f PRESSURE SURFACE SURFACE 2B x 1
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. NEDO-32391, Revision C TYPE 6 ,
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NEDO-32391, Revision C
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4 NEDO-32391, Revision C ATTACHMENT A1 - TABLE OF CONTENTS FOR TEST AND ANALYSIS DOCUMENTS Apparent Test Results
- Brief report on each test
- Tables and plots of key measurements
- Identification of any non-conformances related to test results
!. Data Transmittal Report 1.0 Introduction
- General description and purpose of tests
- Purpose of report 2.0 Objectives
- General Objectives
- Specific Objectives (Note: General Objectives are given in TAPD, Appendix A, for each of the tests) 3.0 Test Facility Description
- Detailed description of facility layout
- Scaling study (Note: Facility descriptions will be from the Test Specification and/or Test Plan and l Procedures documents. Scaling is addressed in Reference 32.)
4.0 Instrumentation
- Instmment type and characteristics
- Calibration
~ 5.0 Data Acquisition System
- Hardware configuration
- t
- Data Reduction
- Software 6.0 Test Matrix
- Grouped by type of test 7.0 Test Results
- Grouped by type of test 8.0 Conclusions
- Adequacy of test data
- Applicability to test objectives l A1-1
l NEDO-32391, Revision C 9.0 References Appendices
- A. Instrument List (Type of instrument, number of insvument, measurement, and range)
B. Modified and Failed Instruments
- Listed by test C. Facility Characterization Tests
- Pressure drop tests .
- Heat loss tests D. Error Analysis
- Maximum error of measurement E. Data Records
- Format of Data Tapes Data Analysis Report 1.0 Introduction e General description and purpose of tests
- Purpose of report 2.0 Objectives
- General Objectives
- Specific Objectives (Note: General Objectives are given in TAPD, Appendix A, for each of the tests) 3.0 Test Analysis
- Grouped by type of test
- Description of test conditions
- Analysis of test results (Note: Framework of test results analysis is given in the " Test Matrix and Data Analysis sections of the TAPD," Appendix A)
- Discussion of observed phenomena .
4.0 Conclusions
- Adequacy of test data
- Applicability to test objectives 5.0 References l Al-2
l NEDO-32391, Revision C i Preliminary Validation Results 1.0 Introduction
- General description and purpose of tests
- What tests will be used for assessment
- How data will be used 2.0 Brief description of Test Facility and Test Matrix
- Referenced to appropriate test reports 3.0 Applicability of data to SBWR e Range of relevant parameters / scaling groups compared to SBWR 4.0 TRACG model and nodalization
- Noding used and basis
- Any modifications for post-test analysis
- Justification for difference in nodalization vs. SBWR nodalization,if any
- Discussion of new models,if any 5.0 Test Simulation
- Choice of tests to be simulated
- Procedure for simulation, including initial and boundary conditions ,
6.0 Qualification results - data vs. predictions l
- Comparisons between data and TRACG results j
- Plots and discussion of key parameters for each test 7.0 Results of Assessment
- Adequacy of TRACG models
- Implications for SBWR calculations 8.0 References l TRACG Qualification For SBWR Abstmet 1.0 Introduction
- ' 1.1 Relationship to CSAU Process 2.0 Qualification Strategy 2.1 Assessment Matrix 2.2 Coverage of PIRT Phenomena 3.0 Separate Effects Tests
- Refer to previous report NEDE-32177P 4.0 Component Performance Tests l Al-3
l NEDO-32391, Revision C 4.1 PANTHERS PCC Performance 4.2 PANTHERS IC Performance 4.3 PANDA PCC Performance 4.4 Suppression Pool Stratification in Blowdown Tests 4.5 SLCS AccumulatorPerformance 5.0 Integral System Tests 5.1 GIST tests (from NEDE-32177P) ' 5 't GIRAFFE / Helium Tests S.3 GIRAFFE / SIT 5.4 1/6 Scale Bomn Mixing , 5.5 CRIEPI Geysering Tests 5.6 ' PSTF MARK III Containment Response 5.7 4T / MARK II Containment Response 5.8 Dodewaard Stanup 5.9 PANDA Transient Tests M2 - M10 6.0 SBWR Plant Nodalization 6.1 Reactor Vessel 6.2 Containment 7.0 Determination of Model Uncertainties and Bias 7.1 Application to LOCA/ECCS and transients 7.2 Application to containment 8.0 Conclusion 8.1 Adequacy of TRACG models 8.2 Implications for SBWR calculations e i l A l-4
l NEDO-32391, Revision C l APPENDIX B - SCALING APPLICABILITY B.1 Scaling Summary f The details of Scaling of SBWR Related Tests is provided in NEDC-32288 [32), which presents a scaling study applicable to the SBWR-related tests. The scope of the study includes: (a) a description of the scaling philosophy used for the GIST, GIRAFFE, PANDA, 'g PANTHERS, and single-tube condensation-heat-transfer tests which have been, or will be, conducted in support of the SBWR program; i (b) the description of a set of scaling laws which are applicable to the SBWR-related test ;
- j. facilities; and j (c) an evaluation of the test facilities with respect to the proper scaling of the important pheomena and processes identified in the SBWR Phenomena Identification and Ranking l Table (PIRT). !
i l The study is fundamentally motivated by the need to demonstrate that the experimental ! observations from the test programs are representative of SBWR behavior. This includes an ; !, identification of any distortions in the representation of the phenomena and the manner in which ! these distortions can be considered when the experimental data are used for computer code 1
- qualification or the development of computer code models.
The Hierarchical Two-Tier Scaling (H2TS) methodology developed by the US Nuclear i Regulatory Commission (US NRC) is applied to the extent practical throughout the study.
< Several scaling considerations addressed by H2TS are automatically satisfied in the SBWR-j related experiments where, in all cases, the fluids and their thermodynamic states are i prototypical. The various scaling issues are addressed, as appropriate, by either the top-down or i bottom-up methodologies embodied in H2TS. The top-down scaling technique, as appued to l generic containment-relatedprocesses, leads to a familiar set of scaling laws with a system scale
- for power, volume, horizontal area in volumes, and mass flow rate, and 1
- 1 scaling for pressure j differences, elevations, and vent submergences.
- The scaling of SBWR system components in relation to specific highly-ranked phenomena i and processes is conducted according to the bottom-up H2TS methodology. This includes i consideration of the following
- thermal plumes, mixing and stratification; heat and mass transfer
!, at liquid-gas interfaces; the heat capacity of structures and heat losses; scaling of the vents; and i heat and mass transfer in the condensers used for decay heat removal. Finally, the scaling
, approach followed in designing the various SBWR-related facilities is reviewed in relation to the ] test objectives. The data collected from these facilities are used in the qualification of the system code TRACG.
l l i
; B-1 +
NEDO-32391, Revision C APPENDIX C - TRACG INTERACTION STUDIES
! C.1 Introduction i
If a LOCA were actually to occur in an SBWR, several of the limiting assumptions used in the licensing analysis may not (in fact, probably will not) apply. In particular, not all power may be lost, and non-safety grade systems and safety grade systems that are not engineered safety
; features (ESF) may be available to support accident management. This Appendix investigates 1 interactions between active and non-ESF systems with the safety systems designed to operate e during the LOCA, to determine if adverse effects due to interactions could result in conditions worse than the case if the non-ESF systems had not been available. The figure-of-merit used to
- l measure the effect of system interactions inside the reactor vessel is the two-phase level inside
,. the chimney. Outside the vessel, the containment pressure and temperature are used. These i studies are an extension of earlier work described in the SSAR which examined the effect of break location on the LOCA and the use of non-ESF systems to prevent core damage.
l The TRACG code has been used for these studies. For interactions affecting the primary system response (inside the vessel) the TRACG input model for LOCA analysis was used. This
, input model provides a detailed representation of the reactor core, vessel internals and associated systems, but a less detailed representation of the containment. For interactions which may affect the containment response (outside the vessel) the TRACG input model used for containment response was used. This input model provides a more detailed representation of the containment ,
and its systems, but a'less detailed pressure vessel model. Both input models have been benchmarked to assure that they predict similar global response for the pressure vessel and containment. Accident scenarios used for the study are similar to those used for LOCA licensing analysis, but additional systems are made available. The use of any additional systems is guided by the SBWR emergency procedure guilelines (EPGs). , C.2 Scenario De6nition for Interaction Studies l The systems selected for the study were those that would likely be available and could produce adverse interactions with the ESF systems. Systems that would clearly benefit the ] , system response were not considered. For example, with power and the feedwater system
. available, vessel inventory could be controlled and there would be no threat of core damage and no need for the passive systems. He Reactor Water Cleanup (RWCU) System is another beneficial system. It removes water from the vessel, cools it, and returns it through the feedwater line. For all but a feedwater line break, it provides heat removal capability in addition
,- to the passive systems. The exception is for a feedwater line break, where operation of the l RWCU System could reduce vessel inventory. This potentially adverse interaction is considered in the study. i For the several break locations which were analyzed, three conditions on the power availability were considered:
- 1. Loss of all AC power, except that provided from inverters he ESF systems, such as GDCS, ADS, and PCCS, are assumed to operate as designed. 3 1
' C-1 1 1 i
l NEDO-32391, Revision C
- 2. On-site diesel generator power available CRD, RWCU and FAPCS are available in addition to the above systems.
i 3 Normal auxiliary power available Feedwater system is available in addition to the above systems. The first condition is the basis used for the LOCA licensing analysis, and the results provide a measure of the system performance for the other conditions where additional systems are available. The first condition also provides an opportunity to examine system interactions , between those safety systems expected to be available during the design basis accident. For all conditions, the ESF systems were assumed to operate as designed. C.3 Primary System Interaction Studies The primary system interactions study investigated the effects of non-ESF systems on the vessel downcomer level and chimney level response. Several break locations were considered. C.4 Containment Interaction Studies The containment system interaction studies investigated interactions between ESF systems, and interactions of ESF systems with other systems which could be available for containment cooling without a loss of power. C.5 Summary of Interaction Studies The system interactions included in this study were those considered most likely to occur when some form of external power was available and which were not clearly beneficial to the operation of the ESF systems. 6 C-2}}