ML20195E939
| ML20195E939 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 08/31/1998 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20138L695 | List: |
| References | |
| WCAP-14253, WCAP-14253-R01, WCAP-14253-R1, NUDOCS 9811190163 | |
| Download: ML20195E939 (400) | |
Text
{{#Wiki_filter:_ . _ . _ _ . _ _ . . _ _ _ . . _ _ . _ . _ _ _ . - . _ _ _ _ . . . . . _ . . . . _ _ . . , WESTINGHOUSE NON-PROPRIETARY CLASS 3 ( WCAP-14253 Revision 1 4 AP600 LOW-PRESSURE INTEGRAL SYSTEMS TEST AT OREGON STATE UNIVERSITY FINAL DATA REPORT AUGUST 1998 i i
- o i
i WESTINGHOUSE ELECTRIC COMPANY Nuclear Technology Division P. O. Box 355 Pittsburgh, Pennsylvania 15230-355 C 1998 Westiny, house Electric Company All 7dghts Reserved 9911190163 981109 7 PDR ADOCK 05200003i A PDRj c:\l 536wRev l\l 536w.non:I b-091198 REVISION 1
FINAL DATA REPORT TABLE OF CONTENTS Section Title P_aage
SUMMARY
l ACKNOWLEDGMENTS 2
1.0 INTRODUCTION
1-1 1.1 Background 1.1-1 1.2 Pre-Operational Test Objectives 1.2 1 1.2.1 Cold Pre-Operational Tests 1.2-1 1.2.2 Hot Pre-Operational Tests 1.2-1 1.3 Matrix Test Objectives 1.3 1 2.0 TEST FACILITY DESCRIPTION 2-1 2.1 Overall Facility Description 2.1-1 2.1.1 Reactor Coolant System 2.1-1 2.1.2 Steam Generator System 2.1-1 2.1.3 Passive Core Cooling System 2.1-2 2.1.4 Automatic Depressurization System 2.1-3 g 2.1.5 Lower Containment Sump 2.1-3
., @ 2.1.6 Normal Residual Heat Removal System r.nd Chemical jij and Volume Control System 2.1-4
- g. 2.1.7 Break and ADS Measurement System 2.1-4 J' 2.1.8 Orifices and Nozzles 2.1-6 2.2 Facility Scaling 2.2-1 2.2.1 Methodology 2.2 1 2.2.2 Facility Scaling Parameters 2.2-2 r
2.2.3 Mass / Energy Balances 2.2-3 2.3 Facility Component Description 2.3-1 2.3.1 Reactor Vessel 2.3-1 2.3.2 Rod Bundle 2.3-2 2.3.3 Reactor Internals 2.3-4 2.3.4 Hot-Leg Piping 2.3-5 2.3.5 Cold-Leg Piping 2.3-6 2.3.6 Pressuriter Surge Line 2.3-7 2.3.7 Pressurizer 2.3-8 2.3.8 Steam Generators 2.3-9 2.3.9 Reactor Coolant Pumps 2.3-11 2.3.10 Accumulators 2.3-11 2.3.11 Core Makeup Tanks 2.3-12 2.3.12 In-Containment Refueling Water Storage Tank 2.3-14 o:\l536wRevl\l536w.rmn:ltWJ90398 iii REVISION 1 _ __ , . . . _ .-__J
FINAL DATA REPORT ', TABLE OF CONTENTS (Continued) Tide Page Section Safety Injection Lines 2.3-15 2.3.13 Containment Sumps 2.3-15 2.3.14 Automatic Depressurization System, Stages 1-3 2.3-16 2.3.15 Automatic Depressurization System, Stage 4 2.3-17 2.3.16 Nonsafety Injection Systems 2.3-18 2.3.17 Passive Residual Heat Removal 2.3-19 2.3.18 2.3.19 Break Simulators 2.3-20 Break and ADS Measurement System (BAMS) 2.3-21 2.3.20 2.3.21 Test Support Systems 2.3-22 2.4-1 2.4 Instrumentation General Information on Instrumentation 2.4-1 2.4.1 Calibration Methods and Standards 2.4-6 2.4.2 Phenomena Affecting Readings 2.4-11 2.4.3 Data Acquisition System 2.5-1 2.5 System Hardware 2.5-1 2.5.1 DAS Architecture 2.5-1 2.5.2 Software 2.5-1 2.5.3 2.5.4 LabVIEW Description 2.5-2 2.5.5 Sequence-of-Events Log 2.5-2 Test Facility Control System 2.6-1 2.6 Operator Panel 2.6-1 2.6.1 2.6.2 Test Signal or Safety Signal (S Signal) 2.6-2 2.6.3 CVS Pump and Discharge Valve Control 2.6-3 2.'6.4 RNS Pump Control 2.6-4 2.6.5 IRWST Valve Control 2.6-5 2.6.6 Main Feed Purap and Discharge Valve Control 2.6-5 2.6.7 Pressurizer Pressure Control 2.6-5 2.6.8 RCP Gland Seal Cooling System Control 2.6-6 2.6.9 CMT Valve Control 2.6-6 2.6.10 CMT Steam Trap Isolation Valves 2.6-6 2.6.11 RCP Control 2.6-7 2.6.12 Reactor Heater Control 2.6-7 2.6.13 Passive Heat Removal 2.6-8 2.6.14 Condensate Return Pump Control 2.6-9 2.6.15 Reactor Heater Sheath High-Temperature Trip 2.6-9 2.6.16 Automatic Depressurization System Control 2.6-9 2.6.17 Steam Generator-1 Level Control 2.6-10 2.6.18 Steam Generator-1 Main Steam Valve 2.6-11 2.6.19 Steam Generator-2 Control 2.6-11 jy REVISION 1 oA1536wRevl\l536w.non lb 090398
FINAL DATA Rrroar TABLE OF CONTENTS (Continued) Section ILtle Eass 2.6.20 Steam Generator-2 Main Steam Valve 2.6-12 2.6.21 Main Steam Control Valve Control 2.6-12 2.6.22 Large-Break BAMS Control 2.6-12 2.7 Pre-Test Operation - 2.7-1 2.8 Drawings 2.8-1 3.0 DATA REDUCTION 3-1 3.1 Introduction 3-1 3.2 Test Validation 3.2-1 3.3 Pre-Operational Tests 3.3-1 3.4 Matrix Tests 3.4-1 3.5 . Instrumentation Error Analysis 3.5-1 3.5.1 General 3.5-1 3.5.2 Defm' itions 3.5-3 3.5.3 - Results 3.5-3 3.6 Zero-Time Shift File Correction . 3.6-1 3.6.1 Test Data Collection Timing 3.6-1 Q 3.6.2 Time Correction Method 3.6-1 4.0 PRE-OPERATIONAL TEST RESULTS 4-1 4.1 Cold Volume Determinations 4.1-1 4.1.1 Accumulator Voluine Test 4.1-1 4.1.2 CMT Volume Test 4.1-4 4.1.3 - Pressurizer Volume Test 4.1-8 4.1.4 IRWST Volume Test 4.1-9 4.1.5 Primary and Secondary Sump Tank Volume Test 4.1-10 4.1.6 SG-1 and SG-2 Secondary-Side Volume Test 4.1-12 4.1.7 ADS and BAMS Moisture Separators Volume Test 4.1-14 4.1.8 Reactor Vessel Volume Test 4.1-15 4.2 Pressure Drop Determination 4.2-1 4.2.1 Background Information 4.2-1 4.2.2 Test Procedu e, Instrumentation, and Results 4.2-4 4.2.3 RCP Flow Test 4.2-7 4.2.4 CMT Injection Flow Test 4.2-9 4.2.5 Accumulator Injection Flow. Test 4.2-15 4.2.6 IRWST Injection Flow Test 4.2-18 4.2.7 Primary Sump Tank Injection Flow Test 4.2-24 4.2.8 Cold-Leg Balance Line Injection Flow Test 4.2-26 L 4.2.9 ADS 1-3 Flow Test - 4.2-28 cA15hltevl\l5W.non:Ibe9039s y REVISION 1
FINAL. DATA REPORT TABLE OF CONTENTS (Continued) Title Page Section 4.2.10 Normal Residual Heat Removal Flow Belance 4.2-32 4.3 HS01 Ambient Heat Losses 4.3 1 4.3.1 Ambient Heat Loss Data at 100 F (Test Procedure Step 4.1.3) 4.3-2 4.3.2 CMT Cooldown 4.3-2 4.3.3 IRWST Cooldown 4.3-2 4.3.4 Conclusion 4.3-2 5.0 MATRIX TESTS RESULTS 5-1 5.1 Cold-Leg Breaks with a Single Failure 5.1-1 5.1.1 2-In. Cold-Leg Break (Matrix Test SB01) 5.1.1-1 5.1.2 Test Repeatability (Matrix Test SB18 Comparison with Matrix Test SB01) 5.1.2-1 5.1.3 Effect of Backpressure (Matrix Text SB19 Comparison with Matrix Tests SB01 and SB18) 5.1.3-1 5.1.4 Effect of a Larger Break Size (Matrix Test SB21 Comparison with Matrix Tests SB01 and SB18) 5.1.4- 1 5.1.5 Effect of a Smaller Break Size (Matrix Test SB23 Comparison with Matrix Tests SB01 and SB18) 5.1.5-1 5.1.6 Effect of an Intermediate Break Size (Matrix Test SB05 Comparison with Matrix Tests SB01 and SB18) 5.1.6-1 5.2 Cold-Leg Breaks with Operation of Nonsafety Systems 5.2-1 5.2.1 Reference 2-In. Cold-Leg Break (Matrix Test SB04) 5.2.1-1 5.2.2 Effect of a Smaller Break Size (Matrix Test SB24 Comparison with Matrix Test SB04) 5.2.2-1 5.3 Core Makeup Tank / Cold-Leg Balance Line Breaks 5.3-1 5.3.1 Reference Double-Ended Guillotine Line Break (Matrix Test SB10) 5.3.1-1 5.3.2 Effect of a Smaller Break Size (Matrix Test SB09 Comparison with Matrix Test SB10) 5.3.2-1 5.4 Direct Vessel Injection Line Breaks 5.4-1 l 5.4.1 Reference Double-Ended Guillotine Line Break (Matrix Test SB12) 5.4.1 -1 ! 5.4.2 Effect of a Smaller Break Size (Matrix Test SB13 Comparison with Matrix Test SB12) 5.4.2-1 5.4.3 Effect of Additional Failures (Matrix Test SB28 Comparison with Matrix Test SB12) 5.4.3-1 5.5 Automatic Depressurization System Impact 5.5-1 5.5.1 Inadvertent ADS Actuation (Matrix Test SBl4) 5.5.1-1 l 5.5.2 Multiple ADS Failures (Matrix Test SB26) 5.5.2-1 5.6 Inadvertent S Signal (Matrix Test SB31) 5.6-1 5.6.1 System Configuration and Initial Conditions 5.6-1 oA1536wRevi\l536w.non:Ib-090398 yi REVISION 1
.. _ . _ . ~ _ _ _ _ - _ . . _ . _ _ . - _ _ . _ . - _ . . . _ _ _ _ _ _ _ . _ _ . . _ . _ _ . . _ . . _
FINA1. DATA REPORT D ' TABLE OF CONTENTS (Continued) l s-*. um t 5.6.2 Inoperable Instmments 5.6-2 5.6.3 Sequence of Events 5.6-2 ! 5.6.4 --Test Results and Evaluation 5.6-3 h '5.6.5 ' Mass' Balance 5.6-4 : I: 5.6.6 Conclusions 5.6-5 5.7 Hot-Leg Break (Matrix Test SB15) 5.7-1 5.7.11 System Configuration and Initial Conditions 5.7-1 5.7.2 Inoperable Instruments 5.7-2 l 5.7.3 Sequence of Events 5.7-3
.5.7.4 Test Results and Evaluation 5.7-5 5.7.5 ' Component Responses 5.7-8 5.7.6 Mass Balance 5.7-16 i 5.7.7 Conclusions 5.7-16 i 6.0 MATRIX TEST GROUP COMPARISONS 6-1 6.1 Effect of 2-In. Break Location (Matrix Tests SB13 and SB15 Comparison with Matrix Test SB01) .6.1-1
( 6.1.1 Influence of Break Location on Break Flow Rates - 6.1-1 6.1.2 Influence of Break Location on ADS Actuation 6.1 6.1.3 Influence of Break Location on Downcomer Levels .6.1-3 I 6.1.4 Influence of Break Location on Core Levels 6.1-4 l 6.2 Effects of Nonsafety Systems (Matrix Test SB04 Comparison with Matrix Test SB01) 6.2-1 i i 7.0 OTHER TEST OBSERVATIONS 7-1 7.1 Condensation Events 7.1-1 7.2 CMT Temperature Measurement 7.2-1 ; 7.2.1- Matrix Test SB13 (U0113) Observation and Evaluation 7.2-1 7.2.2 Matrix Test SB12 (U0ll2) Observation and Evaluation 7.2-4 7.2.3 Matrix Test SB01 (U0001) Observation and Evaluation 7.2-5 7.2.4 Matrix Test SBIO (U0110) Observation and Evaluation 7.2-6 7.2.5 Summary 7.2-6 f 8.0 ; REFERENCES 8-1 O V l\ . o:M5hRevlu5h.non:lt@90398 yji REVISION 1
FINAt. DATA REPORT TABLE OF CONTENTS (Continued) Section Title _P_ ate e APPENDICES A-1 A DATA REDUCTION METHODS AND VALIDATION PROCESS B-1
,B DATA ACCEPTANCE RESULTS C-1 C INSTRUMENTATION DATA BASE D-1 D DATA ERROR ANALYSIS E-1 E MASS BALANCE F-1 F DECAY HEAT COMPARISONS G-1 G PIPING AND INSTRUMENTATION DIAGRAMS H-1 H KEY FACILITY DRAWINGS I-l I DATA FILES O
i REVISION 1 oA1536wRevl\l536w.non:Ib4)9039s viii
FINAL DATA REPORT O
. LIST OF TABLES
.Tattlt Ildt P. Bat 1.3 1 .OSU Matrix Test Summary 1.3-2 2.2 Summary of System Scaling Results for the 1/4-Length Scale Model Primary Loop 2.2-4 2.3 1 Rod Bundle Characteristics 2.3-25 2.3-2 Insulation Applications 2.3 26 2.6-1 Programmable Controller Sumtr.ary 2.6-16 2.6-2 Process Control System Components 2.6-19
- 3.2-1 Overall Acceptance Criteria 3.2-3 3.2-2 Critical Instrument List 3.2-4 3.6-1 Data Files - Correction of Zero Time 3.6-4 4.1-1 ACC-1 Volume Raw Test 4.1-17 4.1-2 ACC-1 Volume Test Results Versus Design 4.1 18 4.1-3 ACC-2 Volume Raw Test 4.1-19 4.1-4 ACC-2 Volume Test Results Versus Design 4.1-20 4.1-5 CMT-1 Volume Test Data 4.1-21 4.1 CMT-1 Volume Test Data 4.1-22 4.1-7 Comparison of Test Results with Design Values 4.1-23
. .lg 4.1-8 Pressurizer Volume Test Raw Data 4.1 24 4.1-9 Summary of Test Results 4.1-25
' _4.1-10 IRWST Volume Data . 4.1-26 4.1-11 Summary of IRWST Volume 4.1-27 4.1-12 Primary Sump Volume Data 4.1-28 4.1-13 ' Secondary Sump Volume Data 4.1-29
' 4.1-14 SG-1 Secondary-Side Volume Data 4.1-30 4.1-15 SG-2 Secondary-Side Volume Data 4.1-31 4.1-16 Summary of SG Secondary-Side Volume -4.1-32 4.1-17 ADS 1-3 Separator Volume Data 4.1-33 4.1-18 ADS 4-1 Sepamtor Volume Data' 4.1-33 4.1 ADS 4-2 Separator Volume Data 4.1-34
'4.1-20 Break Separator Volume Data 4.1-35 4.1-21' Reactor Vessel Volume Data 4.1-36
'4.1-22 Reactor Vessel Volume Summary 4.1-38 4.2-1 Summary of Reactor Vessel and Primary Loop Instrumentation Used in Flow Tests 4.2-34
,4.2-2 Reactor Vessel Test Data Summary - First Flow Test Series 4.2-38 L4.2 Reactor Vessel Test Data Summary - Third Flow Test Series 4.2-40
, 4.2 Comparison of OSU-F-01 and OSU-F-02 Data for Reactor Vessel 4.2-43 n
Qf I-o' :\l5hRevl\l5han:ll>49039s ix REVISION l ____-____-___-___2
FINAI, DATA REPORT
- LIST OF TABLES (Continued)
Table Title Page 4.2-5 RCP Total Developed Head Summary 4.2-44 . 4.2-6 Summary of Line Resistance for Reactor Vessel and Primary Loops 4.2-45 l 4.2-7 CMT-1 and CMT-2 Injection Test Raw Data 4.2-52 l 4.2-8 ACC-1 and ACC-2 Injection Test Raw Data 4.2-53 l 4.2-9 IRWST-1 and IRWST-2 Injection Test Raw Data 4.2-55 4.2-10 Primary Sump Tank Injection Flow Test - Test Data Summary 4.2-56 l 4.2 11 CMT to Cold-Leg Balance Line Injection Flow Test - Test Data Summary 4.2-57 4.2-12 ADS 1-3 Flow Test - Test Raw Data Summary 4.2-58 4.2-13 Comparison of Calculated and Measured Pressure Drop for ADS 1-3 Lines 4.2-59 4.2-14 Line Resistance FIJD+K From Point 0 to Point 3 4.2-60 4.2-15 RNS Injection Data 4.2-61 l' 4.2-16 Comparison of Test Results for Injection Lines 4.2-62 4.3-1 Raw Data File Identification and Description 4.3-3 ! 4.3-2 Failed Instrumentation 4.3-4 4.3-3 Instrumentation Outside Test Boundary But Affected By CMT Cooldown 4.3-5 4.3-4 Instrumentation Outside Test Bouridary But Affected By IRWST Cooldown 4.3-6
- 5.1.1 -1 Matrix Test SB01 Initial Conditias 5.1.1-30 l 5.1.1-2 Matrix Test SB01 Inoperable Instruments / Invalid Data Channels 5.1.1-32 5.1.1-3 Matrix Test SB01 Sequence of Events 5.1.1-35 5.1.2-1 Matrix Test SB18 Initial Conditions 5.1.2 12 5.1.2-2 Matrix Test SB18 Inoperable Instruments / Invalid Data Channels 5.1.2-14 l 5.1.2-3 Matrix Test SB18 Sequence of Events 5.1.2-16 l 5.1.2-4 Data Recorded in SB18 Test Log 5.1.2-25 5.1.3-1 Matrix Test SB19 Initial Conditions 5.1.3-14 5.1.3-2 Matrix Test SB19 Inoperable Instruments / Invalid Data Channels 5.1.3-16 L 5.1.3-3 Matrix Test SB19 Sequence of Events 5.1.3-18 l 5.1.4-1 Matrix Test SB21 Initial Conditions 5.1.4-16 l 5.1.4-2 Matrix Test SB21 Inoperable Instruments / Invalid Data Channels 5.1.4-18 5.1.4-3 Matrix Test SB21 Sequence of Events 5.1.4-21 5.1.5-1 Matrix Test SB23 Initial Conditions 5.1.5-7 l 5.1.5-2 Matrix Test SB23 Inoperable Instruments /Invai.'d Data Channels 5.1.5-9 5.1.5-3 Matrix Test SB23 Sequence of Events 5.1.5 12 5.1.6-1 Matrix Test SB05 Initial Conditions 5.1.6-9 5.1.6-2 Matrix Test SB05 Inoperable Instruments / Invalid Data Channels 5.1.6-11 5.1.6-3 Matrix Test SB05 Sequence of Events 5.1.6-13 5.2.1-1 Matrix Test SB04 Initial Conditions 52 1-2.
5.2.1-2 Matrix Test SB04 Inoperable Instruments / Invalid Data Channels 5.2.1-24 9 c:U536wRevluS36w.non:lb-090398 x REVISION 1
I FINA1. DATA REPORT j LIST OF TABLES (Continued) )
- m =
l 5.2.1-3 Matrix Test SB04 Sequence of Events 5.2.1-27 j 5.2.2-1 Matrix Test SB24 Initial Conditions 5.2.2-9 5.2.2-2 Matrix Test SB24 Inoperable Instruments / Invalid Data Chaimels 5.2.2-11 5.2.2-3 Matrix Test SB24 Sequence of Events 5.2.2-13 l 5.3.1-1 Matrix Test SB10 Initial Conditions 5.3.1-35 5.3.1-2 Matrix Test SB10 Inoperable Instruments / Invalid Data Channels 5.3.1 37 1 5.3.1-3 Matrix Test SBIO Sequence of Events 3.3.1 5.3.2-1 Matrix Test SB09 Initial Conditions 5.3.2-14 5.3.2-2 Matrix Test SB09 Inoperable Instruments / Invalid Data Channels 5.3.2-16 5.3.2-3 Matrix Test SB09 Sequence of Events 5.3.2-19 5.4.1-1 Matrix Test SB12 Initial Conditions 5.4.1-34 5.4.1-2 Matrix Test SB12 inoperable Instruments / Invalid Data Channels 5.4.1-36 5.4.1-3 Matrix Test SB12 Sequence of Events 5.4.1-39 5.4.1-4 Temporary Test Thermocouples 5.4.1-47 5.4.2-1 Matrix Test SB13 Initial Conditions 5.4.2-24 5.4.2-2 Matrix Test SB13 Inoperable Instruments / Invalid Data Channels 5.4.2-26 5.4.2-3 Matrix Test SB13 Sequence of Events 5.4.2-29 Q 5.4.3-1' Matrix Test SB28 Initial Conditions 5.4.3-16 5.4.3-2 Matrix Test SB28 Inoperable Instruments / Invalid Data Channels 6.4.3-18 5.4.3-3 Matrix Test SB26 Sequence of Events 5.4.3-A 5.4.3-4 TeiWporary Test Thermocouples 5.4.3-28 5.5.1-1 Matrix Test SBl4 Initial Conditic:u 5.5.1-34 5.5.1-2 Matrix Test SB14 Inoperable lastruments/ Invalid Data Channels 5.5.1-36 5.5.1-3 Matrix Test SB14 Sequence of Events 5.5.1-39 5.5.2-1 Matrix Test SB26 Initial Cor.ditions 5.5.2-18 5.5.2-2 Matrix Test SB26 Inoperable Instruments / Invalid Data Channels 5.5.2-20 5.5.2-3 Matrix Test SB26 Sequence of Events 5.5.2-22 5.6-1 Matrix Test SB31 Initial Conditions 5.6-6 5.6.2 Matrix Test SB31 Inoperable Instruments / Invalid Data Channels 5.6-8 5.6.3 Matrix Test SB31 Sequence of Events 5.6-10 5.7-1 Matrix Test SB15 Initial Conditions 5.7-17 5.7.2 Matrix Test SB15 Inoperable Instruments / Invalid Data Channels 5.7-19
~
5.7.3 Matrix Test SB15 Sequence of Events 5.7-22 6.1-1 Sequence of Events Comparison for Matrix Tests with 2-in. Breaks at Different Locations 6.1-5 6.2.1-1 Sequence of Events for EtI cts of Nonsafety Systems on 2-in. Bn.aks 62-3 7.2-1 Matrix Test SB13 CMT-2 FuM Temperature Condition Changes Summary 7.2-7 O 17.2-2 7.2 Matrix Test SB13 CMT-1 Fnid Temperature Condition Changes Summary CMT Supe heating Parameters Summary 7.2-7 7.2-8 oA15hRevi\l5h.rm:Ib-090398 xi REVISION 1
FINAL DATA REPORT l LIST OF FIGURES P_ age Figure Title Reactor Ve',sel 2.1-8 2.1-1 2.1-9 2.1-2 IRWST and Reactor Vessel Primary ' Jump Tank and Break Separator 2.1-10 2.1-3 2.1-11 2.1-4 Upper level (Reactor Vessel Cover in Foreground) l 2.1-12 ! 2.1-5 Isometr'c Drawing of OSU Test Facility Simplif.ed Flow Diagram of the OSU Test Facility 2.1-13 2.1-6 General Scaling Methodology 2.2-5 ! '2.2-1 RCP Performance Head Versus Flow 2.3-27 2.3-1 Flow Schematic for the ADS 2.3 28 l 2.3-2 CVS Pump Head Versus Flow 2.3-29 2.3-3 RNS Pump Head Versus Flow 2.3-30 l 2.3-4 Electrical One-Line Diagram 2.3-31 2.3-5 DAS Hardware 2.5-4 2.5-1 2.5-2 DAS Architecture 2.5-5 2.5-3 DAS Hierarchy 2.5-6 2.6-1 Photograph of Operator Panel 2.6-20 2.6-2 Drawing of Operator Panel 2.6-21 3.2-1 Data Documentation Steps 3.2-9 3.4-1 Steps in OSU Data Processing 3.4-4 4.1-1 Schematic of Accumulator Volume Test Setup 4.1-39 4.1-2 Schematic of CMT Test Setup 4.1-40 ( 1.1-3 CMT-1 Volume versus Height from Bottom 4.1-41 4.1-4 CMT-2 Volume versus Height from Bottom 4.1-42 4.1-5 Schematic of Pressurizer Test Setup 4.1-43 l 4.1-6 IRWST Test Setup 4.1-44 4.1-7 Schematic of Containment Sump Test Setup 4.1-45 4.1-8 Schematic of SG Secondary-Side Volume Test Setup 4.l a 4.1-9 Schematic of ADS and BAMS Separator Test Setup 4. i , - 4.1-10 Schematic of ADS and BAMS Separator Test Setup 4.1-48 4.2-1 CMT-1 Injection Test Flow Path 4.2-64 4.2-2 CMT-2 Injection Test Flow Path 4.2-65 4.2-3 CMT-1 Injection Line Pressure Drop Versus Flow Square 4.2-66 4.2-4 CMT-1 Injection Line Pressure Drop Versus Flow Square 4.2-66 4.2-5 CMT-1 Injection Line Pressure Drop Versus Flow Square 4.2-67 4.2-6 CMT-2 Injection Line Pressure Drop Versus Flow Squam 4.2-67 4.2-7 CMT-2 Injection Line Pmssure Drop Versus Flow Square 4.2-68 4.2-8 CMT-2 Injection Liu Prrssure Drop Versus Flow Square 4.2-68 4.2-9 CMT-1 Total Line Emc.re Versus Flow Rate Square 4.2-69 4.2-10 CMT-2 Total Line Pressure Versas Flow Rate Square 4.2-69 c:u536wRevl\l536w.non:Ib-090398 xii REVIS17N 1 i
, .- -. _ -.__m_ _ _ _ _ _ . . _ - ~ _ _ _ _ . _ _ ..-._-. __ _ _ . _ _ . _ FINAL DATA REPORT LIST OF FIGURES (Continued) m m t- , 4.2 11 ACC-1 Injection Test Flow Path 4.2-70
, 4.2 ACC-2 Injection Test Flow Path 4.2 71 4.2-13 ACC-1 Injection Line Pressure Drop Versus Flow Square 4.2-72
. 4.2 14 ACC-1 Injection Line Pressure Drop Versus Flow Square 4.2-72 4.2-15 ' -ACC-1 Injection Line Pressure Drop Versus Flow Square 4.2 73 4.2-16 ACC-2 Injection Line Pn:ssure Drop Versus Flow Square 4.2 73
'4.2 17 ACC-2 Injection Line Pressure Drop Versus Flow Square 4.2-74 4.2-18 ACC-2 Injection Line Pressure Drop Versus Flow Square 4.2-74 4.2-19 ACC-1 Injection Line Pressure Drop Versus Flow Square 4.2-75 4.2-20 ACC-2 Injection Line Pressure Drop Versus Flow Square 4.2-75 4.2 21 IRWST Injection Test Flow Path 4.2-76 4.2-22 IRWST-1 Injection Line Pressure Drop Versus Flow Rate 4.2-77 4.2-23 IRWST-1 Injection Line Pressure Drop Versus Flow Rate 4.2 77 4.2 . IRWST-1 Injection Line Pressure Drop Versus Flow Rate 4.2-78 4.2-25 IRWST-2 Injection Line Pressure Drop Versus Flow Rate 4.2-78 4.2 26 IRWST-2 Injection Line Pressure Drop Versus Flow Rate 4.2-79 4.2 IRWST-2 Injection Line Pressure Drop Versus Flow Rate 4.2-79 j 4.2-28 Primary Sump Ttnk Injection Test Flow Path 4.2-80
.4.2-29 Primary Sump Tank Injection Pressure Drop Versus Flow Rate 4.2 81 4.2-30 Primary Sump Tank Injection Pressure Drop Versus Flow Rate 4.2-81 4.2-31 CMT-1 to CL-3 Balance Line Injection Test Flow Path 4.2-82 4.2-32 CMT-2 to CL-1 Balance Line Injection Test Flow Path 4.2-83 4.2 CMT-1 to CL-3 Balance Line Injection Pressure Drop Versus Flow Rate 4.2-84 4.2-34 CMT-1 to CL-3 Balance Line Injection Pressure Drop Versus Flow Rate 4.2-84 4.2-35 ADS 1-3 Injection Test Flow Path 4.2-85 4.2-36 ADS 1-3 Test Level Comparisons 4.2-86 4.2-37 Pressure Drop Via ADS-1 4.2-87 4.2-38 - Pressure Drop Via ADS-2 4.2 87 A.2-39 Pressure Drop Via ADS 3 4.2-88 4.2-40 Pressure Drop Via ADS-1 4.2-88 4.2-41 Predicted Pressure Drop Versus Flow Rate Square 4.2-89 4.2 Pressure Drop Via ADS-1 4.2-89 4.2-43 Pressure Drop Via ADS-2 4.2-90 4.2-44 Pressure Drop Vis ADS-3 4.2 90 4.2-45 RNS Injetion Test Flow Path 4.2-91 Matrix Test SB01-5.1.1-1 Primary Loop and Break Piping Layout 5.1.1-44 5.1.1-2 Primary Loop and Break Pipe Arrangement 5.1.1-45 oA1536wRevi\l536w.nortib-090398 xiii REVISION 1 a
FINAI. DATA REPORT I l l LIST OF FIGURES (Continued) Figure Title f.!yte Reactor and Downcomer Annulus Steam Percent 5.1.1-46 5.1.1-3 Reactor Core Steam Percent 5.1.1-47 5.1.1-4 5.1.1-5 Pressurizer and Surge Line Levels 5.1.1-48 5.1.1-6 CMT-1 and CMT-2 Levels 5.1.1-49 5.1.17 SG-1 Tube Levels- 5.1.1-50 5.1.1-8 SG-2 Tube Levels 5.1.1-51 5.1.1-9 SG-1 Channel Head Levels 5.1.1-52 5.1.1 10 SG-2 Channel Head Levels 5.1.1-53 5.1.1-11 Hot-Leg Levels 5.1.1-54 5.1.1-12 Reactor Core Levels 5.1.1-55 5.1.1-13 HL-1 Temperatures 5.1.1-56 5.1.1-14 HL-2 Temperatures 5.1.1-57 5.1.1-15 Reactor and Downcomer Annulus Wide-Range Levels 5.1.1-58 5.1.1-16 Accumulator and CMT Injection Flows 5.1.1-59 5.1.1-17 Total DVI Flow 5.1.1-60 5.1.1-18 Accumulator Injection Line Temperatures 5.1.1-61 5.1.1-19 Upper Head DPs '>.1 1-62 5.1.1-20 DVI Nozzle Temperatures 5.1.1-63 5.1.1-21 Pressurizer Heater Temperature 5.1.1-64 5.1.1-22 Reactor and Downcomer Annulus Wide-Range Levels 5.1.1-65 5.1.1-23 IRWST/ Primary Sump Injection Temperatures 5.1.1-66 5.1.1-24 Pressurizer Temmature and kW 5.1.1-67 5.1.1-25 IRWST Ovenlow and Associated Pressures 5.1.1-68 5.1.1-26 IRWST, Sump, and Break Separator Levels 5.1.1-69 5.1.1-27 Pressurizer Temperatures 5.1.1-70 5.1.1-28 Separator Loop Seal Flows 5.1.1-71 5.1.1-29 ADS 1-3 Pressures 5.1.1-72 5.1.1-30 CMT-1 and Reactor Vessel Parameters during CMT Reflood and Subsequent Draindown 5.1.1-73 5.1.1-31 CMT-2 and Reactor Vessel Parameters during CMT Reflood and Subsequent Draindown 5.1.1-74 5.1.1-32 CMT-2 Fluid Temperatures 5.1.1-75 5.1.1-33 CMT-1 Fluid Temperatures 5.1.1-76 5.1.1-34 CMT-1, CMT-2, and IRWST Levels 5.1.1-77 5.1.1-35 IRWST, Sump, and Break Separator Levels 5.1.1-78 5.1.1-36 Separator Loop Seal Flows 5.1.1-79 5.1.1-37 IRWST and Primary Sump Flows 5.1.1-80 5.1.1-38 Downcomer Annulus Temperatures at 0 degrees Azimuth 5.1.1-81 5.1.1-39 Total DVI Flow 5.1.1-82 oA1536wRevluS36w.non:ltM)90398 xiv REVISION 1
l l FINAL DATA REPORT
.q LIST OF FIGURES (Continued)
Elanar.t Titis Esat i 5.1.1-40 . PRHR HX Temperatures 5.1.1-83 ~l
- 5.1.1-41 PRHR HX Short-Tube and Long-Tube Temperatures 5.1.1-84 5.1.1-42 CL-2 Temperatures 5.1.1-85 5.1.1-43 CL-4 Temperatures 5.1.1-86 5.1.1-44 Reactor Heater Temperatures @ 46 in. - Top of Core 5.1.1-87 5.1.1-45 Primary and Secondary Pressures 5.1.1-88
. 5.1.1-46 Upper-Plenum and Upper-Head Temperatures 5.1.1-89 5.1.1-47 Upper-Head and Downcomer Temperatures 5.1.1 90 5.1.1-48 IRWST and Primary Sump Flows 5.1.1-91 i 5.1.1 Upper Head DPs 5.1.1-92 5.1.1-50 CMT-1 Levelfremperature vs. Time 5.1.1-93 5.1.1-51 CMT-2 Level / Temperature vs. Time 5.1.1-94 5.1.1-52 Cold-Leg Levels 5.1.1-95 5.1.1-53 CL-1 Temperatures 5.1.1-%
.15.1.1-54 CL-3 Temperatures 5.1.1-97 l 5.1.1 ADS 1-3 Flow DPs 5.1.1-98 5.1.1-56 Reactor /HL-2/SG-2 Channel Head Levels 5.1.1-99 Q - 5.1.1-57 5.1.1-58 Reactor /HL-2/SG-2 Channel Head Steam Percent Pressurizer and Surge Line Levels 5.1.1-100 5.1.1-101 5.1.1;59 - Pressurizer and Surge Line Steam Percent 5.1.1-102 5.1.1-60 DVI Flows 5.1.1 103 5.1.1-61 ADS 1-3 Liquid and Steam Flows 5.1.1-104 5.1.1-62 ' Break Separator Liquid and Steam Flows 5.1.1 105 5.1.1-63 Pressurizer and Reactor Pressures 5.1.1-106 5.1.1-64 Accumulator Levels 5.1.1-107
-5.1.1-65 ADS 1-3 Separator Level 5.1.1-108 5.1.1-66 - PRHR HX Flows 5.1.1 109
.5.1.1-67 PRHR HX Levels 5.1.1-110 5.1.1-68 PRHR HX Levels - 5.1.1-111 5.1.1-69 IRWST Overflow and Associated Pressures 5.1.1-112 5.1.1-70 IRWST Short-Rod and Sparger Tip Temperatures 5.1.1 113 5.1.1-71 IRWST Long-Rod Top-Half Temperatures 5.1.1 114 5.1.1-72 IRWST Long-Rod Bottom-Half Temperatures 5.1.1-115 5.1.1-73 BAMS Pressures 5.1.1-116 5.1.1-74 BAMS Pressures 5.1.1-117 5.1.1-74x BAMS Pressures 5.1.1-118 5.1.1-75 BAMS Header Steam Flows 5.1.1-119 5.1.1 76 Separator Steam Flows 5.1.1-120 oA15hRevl\l$h.non:Ib-090398 xy REVISION 1
.a
FINrt. DATA REPORT LIST OF FIGURES (Continued) Figure Title .P_ age CMT-1 Wide-Range and Balance Line Levels 5.1.1-121 5.1.1-77 5.1.1-78 CMT-1/ Reactor Vessel /CL-3 Pressures 5.1.1-122 5.1.1-79 CMT-1 Inlet Temperature 5.1.1-123 5.1.1-80 CMT-1 Temperatures 5.1.1-124 Matrix Text SB18 Comparison with Matrix Test SB01 5.1.2 1 Primary Loop and Break Piping Layout 5.1.2-26 5.1.2-2 Primary Loop and Break Pipe Arrangement 5.1.2-27 5.1.2-3 Reactor and Downcomer Annulus Steam Percent 5.1.2-28 5.1.2-4 Reactor Core Steam Percent 5.1.2-29 5.1.2-5 Pressurizer and Surge Line Levels 5.1.2-30 5.1.2-6 CMT-1 and CMT-2 Levels 5.1.2-31 5.1.2-7 SG-1 Tube Levels 5.1.2-32 5.1.2-8 SG-2 Tube Levels 5.1.2-33 5.1.2-9 SG-1 Channel Head Levels 5.1.2-34 5.1.2-10 SG-2 Channel Head Levels 5.1.2-35 5.1.2-11 Hot-Leg Levels 5.1.2-36 5.1.2-12 Reactor Core Levels 5.1.2 37 5.1.2-13 HL-1 Temperatures 5.1.2-38 5.1.2-14 HL-2 Temperatures 5.1.2-39 5.1.2-15 Reactor and Downcomer Annulus Wide-Range Levels 5.1.2-40 5.1.2-16 Accumulator and CMT Injection Flows 5.1.2-41 5.1.2-17 Total DVI Flow 5.1.2-42 5.1.2-18 Accumulator Injection Line Temperatures 5.1.2-43 5.1.2 19 Upper Head DPs 5.1.2-44 5.1.2 20 DVI Nozzle Temperatures 5.1.2-45 5.1.2-21 Pressurizer Heater Temperature 5.1.2-46 5.1.2-22 Reactor and Downcomer Annulus Wide-Range Levels 3.1.247 5.1.2-23 IRWST/ Primary Sump Injection Temperatures 5.1.2-48 l 5.1.2-24 Pressurizer Temperature and kW 5.1.2-49 5.1.2-25 IRWST Overflow and Associated Pressures 5.1.2-50 l 5.1.2-26 IRWST, Sump, and Break Separator Levels 5.1.2-51 5.1.2-27 Pressurizer Temperatures 5.1.2-52 5.1.2-28 Separator Loop Seal Flows 5.1.2-53 5.1.2-29 ADS 1-3 Pressures 5.1.2-54 5.1.2-30 CMT-1 and Reactor Vessel Parameters during CMT Reflood and subsequent Draindown 5.1.2-55 l O o:\l 536w Rev i\l 536w.norr i b-090398 xvi REVISION 1
FINAL IIATA REPOILT l l LIST OF FIGURES (Continued)
)-
EigEe Title hge 5.1.2-31 CMT-2 and Reactor Vessel Parameters during l CMT Reflood and subsequent Draindown 5.1.2-56 5.1.2-32 CMT-2 Fluid Temperatures 5.1.2-57 5.1.2-33 CMT-1 Fluid Temperatures 5.1.2-58
'5.1.2-34 CMT-1, CMT-2, and IRWST Levels 5.1.2-59 5.1.2-35 IRWST, Sump, and Break Separator Levels 5.1.2-60 5.1.2 36 Separator Loop Seal Flows 5.1.2-61 5.1.2-37 IRWST and Primary Sump Flows 5.1.?-62 5.1.2-38 Downcomer Annulus Temperatures at 0 degrees Azimuth 5.1.? .63 5.1.2-39 Total DVI Flow 5.1.4 04 5.1.2-40 PRHR HX Temperatures 5.1.2-65 5.1.2-41 PRHR HX Short-Tube and Long-Tube Temperatures 5.1.2-66 5.1.2-42 CL-2 Temperatures 5.1.2 67 5.1.2-43 CL-4 Temperatures 5.1.2-68 5.1.2-44' Reactor Heater Temperatures @ 46 in. - Top of Core 5.1.2-69 5.1.2-45 Primary and Secondary Pressures 5.1.2-70 5.1.2-46 Upper-Plenum and Upper-Head Temperatures 5.1.2-71
( 5.1.2-47_ 5.1.2-48 Upper-Head and Downcomer Temperatures 5.1.2-72 IRWST and Primary Sump Flows 5.1.2-73 5.1.2-49 Upper-Head DPs - 5.1.2-74
-5.1.2-50 CMT .1 Level /remperature vs. Time 5.1.2-75 5.1.2-51 CMT-2 Level /remperature vs. Time 5.1.2-76 5.1.2-52 Cold Leg Levels 5.1.2-77 5.1.2-53 CL-1 Temperatures 5.1.2-78 I 1
5.1.2-54' CL-3 Temperatures 5.1.2-79 i 5.1.2-55 ADS 1-3 Flow DPs 5.1.2-80 5.1.2-56 Reactor /HL-2/SG-2 Channel Head Levels 5.1.2-81 5.1.2-57 Reactor /HL2/SG2 Channel Head Steam Percent 5.1.2 82 5.1.2 58 Pressurizer and Surge Line Levels 5.1.2-83 l 5.1.2-59 Pressurizer and Surge Line Steam Percent 5.1.2-84 l 5.1.2-60 DVI Flows 5.1.2-85 i 5.1.2-61 ADS 1-3 Liquid and Steam Flows 5.1.2-86 i 5.1.2-62 Break Separator Liquid and Steam Flows 5.1.2-87 l 5.1.2-63 Pressurizer and Reactor Pressures 5.1.2-88 . 5.1.2-64 Accumulator Levels 5.1.2-89 5.1.2-65 ADS 1-3 Separator Level 5.1.2-90 l _ 5.1.2-66 PRHR HX Flows 5.1.2-91 5.1.2 PRHR HX Levels 5.1.2-92 U ' I o:\l5hRevi\l5h.non lb 090398 xvii REVISION 1
FINAt. DATA REPORT LIST OF FIGURES (Continued) Figure Title P_ age PRHR HX Levels 5.1.2-93 5.1.2-68 IRWST Overflow and Associated Pressures 5.1.2-94 5.1.2-69 IRWST Short Rod and Sparger Tip Temperatures 5.1.2-95 5.1.2-70 IRWST Long Rod Top Half Temperatures 5.1.2-96 5.1.2-71 IRWST Long Rod Bottom Half Temperatures 5.1.2-97 5.1.2-72 BAMS Pressures 5.1.2-98 5.1.2-73 BAMS Pressures 5.1.2-99 5.1.2-74 BAMS Pressures 5.1.2-100 5.1.2-74x BAMS Header Steam Flows 5.1.2-101 5.1.2-75 Separator Steam Flows 5.1.2-102 5.1.2-76 CMT-1 Wide-Range and Balance Line Levels 5.1.2-103 5.1.2-77 CMT-1/ Reactor Vessel /CL-3 Pressures 5.1.2-104 5.1.2-78 CMT-1 Inlet Temperature 5.1.2-105 5.1.2-79 CMT-1 Temperatures 5.1.2-106 5.1.2-80 i Matrix Test SB19 Comparison with Matrix Test SB01 5.1.3-1 Primary Loop and Break Piping Layout 5.1.3-26 Primary Loop and Break Pipe Arrangement 5.1.3-27 . 5.1.3-2 Reactor and Downcomer Annulus Steam Percent 5.1.3-28 5.1.3-3 Reactor Core Steam Percent 5.1.3-29 5.1.3-4 5.1.3-5 Pressurizer and Surge Line Levels 5.1.3-30 CMT-1 and CMT-2 Levels 5.1.3-31 5.1.3-6 SG-1 Tube Levels 5.1.3-32 5.1.3-7 SG 2 Tube Levels 5.1.3-33 5.1.3-8 SG 1 Channel Head Levels 5.1.3-34 5.1.3-9 SG-2 Channel Head Levels 5.1.3-35 5.1.3-10 5.1.3-11 Hot-Leg Levels 5.1.3-36 Reactor Core Levels 5.1.3-37 5.1.3-12 HL-1 Temperatures 5.1.3-38 5.1.3-13 5.1.3-14 HL-2 Temperatures 5.1.3-39 5.1.3-15 Reactor and Downcomer Annulus Wide-Range Levels 5.1.3-40 Accumulator and CMT Injection Flows 5.1.3-41 5.1.3-16 5.1.3-17 Total DVI Flow 5.1.3-42 Accumulator Injection Line Temperatures 5.1.3-43 5.1.3-18 Upper-Head DPs 5.1.3-44 5.1.3-19 DVI Nozzle Temperatures 5.1.3-45 5.1.3-20 5.1.3-21 Pressurizer Heater Temperature 5.1.3-46 5.1.3-22 Reactor and Downcomer Annulus Wide-Range Levels 5.1.3-47 xvijj REVISION 1 oA1536wRevi\l536w.non:lb-o90398
FmAt DATA RaronT O LIST OF FIGURES (Continued) Elant .T.111e E.ast 5.1.3-23 IRWST/ Primary Sump Injection Temperatures 5.1.3-48 5.1.3-24 Pressurizer Temperature and kW 5.1.3-49 5.1.3-25 IRWST Overflow and Associated Pressures 5.1.3-50
'5.1.3-26 IRWST, Sump, and Break Separator Levels 5.1.3-51 5.1.3-27 Pressurizer Temperatures 5.1.3-52 5.1.3-28 Separator Loop Seal Flows 5.1.3-53 5.1.3-29 ADS 1-3 Pressures 5.1.3-54 5.1.3-30 CMT-1 and Reactor Vessel Parameters during CMT Reflood and subsequent Draindown 5.1.3-55 5.1.3-31 CMT-2 and Reactor Vessel Parameters during CMT Reflood and subsequent Draindown 5.1.3-56 5.1.3-32 CMT 2 Fluid Temperatures 5.1.3-57 5.1.3-33 CMT-1 Fluid Temperatures 5.1.3-58 5.1.3-34 CMT-1, CMT-2, and IRWST Levels 5.1.3-59
. 5.1.3-35 IRWST, Sump, and Break Separator Levels 5.1.3-60 5.1.3-36 Separator Loop Seal Flows 5.1.3-61 s 5.1.3-37 IRWST and Primary Sump Flows 5.1.3-62 5.1.3-38 Downcomer Annulus Temperatures at 0 degrees Azimuth 5.1.3-63 5.1.3-39 - Total DVI Flow 5.1.3-64 5.1.3-40 PRHR HX Temperatures 5.1.3-65 5.1.3-41 PRHR HX Short-Tube and Long-Tube Temperatures 5.1.3-66 5.1.3-42 CL-2 Temperatures 5.1.3-67 5.1.3-43 CL-4 Temperatures . 5.1.3-68 5.1.3-44 Reactor Heater Temperatures @ 46 in. - Top of Core 5.1.3-69 5.1.3-45 Primary and Secondary Pressures 5.1.3-70 5.1.3-46 Upper-Plenum and Upper-Head Temperatures 5.1.3-71 5.1.3-47 Upper Head and Downcomer Temperatures 5.1.3-72 5.1.3-48 IRWST and Primary Sump Flows 5.1.3-73 5.1.3-49 Upper-Head DPs 5.1.3-74 5.1.3-50 CMT-1 Level /remperature vs. Time 5.1.3-75 5.1.3-51 CMT-2 Level / Temperature vs. Time 5.1.3-76 5.1.3 Cold Leg Levels 5.1.3-77
. 5.1.3-53 CL-1 Temperatures 5.1.3-78
.5.1.3-54 ~ CL-3 Temperatures 5.1.3-79 5.1.3-55 ADS 1-3 Flow DPs 5.1.3-80 5.1.3-56 Reactor /HL-2/SG-2 Channel Head Levels 5.1.3-81 5.1.3-57 Reactor /HL-2/SG-2 Channel Head Steam Percent 5.1.3-82
/Q 5.1.3 58 Pressurizer and Surge Line Levels 5.1.3-83 V
L oA1536wRevnl536w.noscibe90398 xix REVISION 1
FINAI. DATA REPOIT LIST OF FIGURES (Continued) Title Eage Figure 5.1.3-84 5.1.3-59 Pressurizer and Surge Line Steam Percent 5.1.3-85 5.1.3-60 DVI Flows 5.1.3-86 5.1.3-61 ADS 1-3 Liquid and Steam Flows 5.1.3-87 5.1.3-62 Break Separator Liquid and Steam Flows 5.1.3-88 5.1.3-63 Pressurizer and Reactor Pressures 5.1.3-89 5.1.3-64 Accumulator Levels 5.1.3-90 5.1.3-65 ADS 1-3 Separator Level 5.1.3-91 5.1.3-66 PRHR HX Flows 5.1.3-92 5.1.3-67 PRHR HX Levels 5.1.3-93 5.1.3-68 PRHR HX Levels 5.1.3-94 5.1.3-69 1RWST Overflow and associated Pressures IRWST Short-Rod and Sparger Tip Temperatures 5.1.3-95 5.1.3-70 5.1.3-96 5.1.3-71 IRWST Long-Rod Top Half Temperatures 5.1.3-97 5.1.3-72 IRWST Long-Rod Bottom Half Temperatures 5.1.3-98 5.1.3-73 BAMS Pressures 5.1.3-99 5.1.3-74 BAMS Pressures 5.1.3-100 5.1.3-74x BAMS Pressures 5.1.3-101 5.1.3-75 BAMS Header Steam Flows 5.1.3-102 5.1.3-76 Separator Steam Flows CMT-1 Wide-Range and Balance Line Levels 5.1.3-103 5.1.3-77 5.1.3-104 5.1.3-78 CMT-1/ Reactor Vessel /CL-3 Pressures CMT-1 Inlet Temperature 5.1.3-105 5.1.3-79 5.1.3-106 5.1.3-80 CMT-1 Temperatures Matrix Test SB19 Comparison with Matrix Test SB01 5.1.4-30 5.1.4-1 Primary Loop and Break Piping Layout 5.1.4-31 5.1.4-2 Primary Loop and Break Pipe Arrangement 5.1.4-32 5.1.4-3 Reactor and Downcomer Annulus Steam Percent 5.1.4-33 5.1.4-4 Reactor Core Steam Percent Pressurizer and Surge Line Levels 5.1.4-34 5.1.4-5 5.1.4-35 5.1.4-6 CMT-1 and CMT-2 Levels 5.1.4-36 5.1.4-7 SG-1 Tube Levels 5.1.4-37 5.1.4-8 SG-2 Tube Levels 5.1.4-38 5.1.4-9 SG-1 Channel Head Levels 5.1.4-39 5.1.4-10 SG-2 Channel Head Levels 5.1.4-40 5.1.4-11 Hot-Leg Levels 5.1.4-41 5.1.4-12 Reactor Core Levels 3.1.4-42 5.1.4-13 HL-1 Temperatures l REVISIOil 1 oA15%wRevi\l536w.non:Ib-090398 xx
FINAL DATA REPORT !. LIST OF FIGURES (Continued) f.lEllLt .Thlt P.BILt 5.1.4-14 HL-2 Temperatures 5.1.4-43 5.1.4-15 Reactor and Downcomer Annulus Wide-Range Levels 5.1.4-44 i 5.1.4 16 Accumulator and CMT Injection Flows 5.1.4-45 5.1.4-17 Total DVI Flow 5.1.4-46 5.1.4 Accumulator Injection Line Temperatures 5.1.4-47 5.1.4 19 Upper-Head DPs 5.1.4-48 5.1.4-20 DVI Nozzle Temperatures 5.1.4-49
- 5.1.4-21 Pressurizer Heater Temperature . 5.1.4-50
[ 5.1.4 Reactor and Downcomer Annulus Wide-Range Levels 5.1.4-51 j 5.1.4-23 IRWST/ Primary Sump Injection Temperatures 5.1.4-52 5.1.4-24 Pressurizer Temperature and kW 5.1.4-53 5.1.4-25 IRWST Overflow and Associated Pressures 5.1.4-54 I 5.1.4-26 IRWST, Sump, and Break Separator Levels 5.1.4-55 5.1.4-27 Pressurizer Temperatures 5.1.4-56 5.1.4-28 Separator Loop Seal Flows 5.1.4-57 5.1.4-29 ADS 1-3 Pressures 5.1.4-58 5.1.4-30 CMT-1 and Reactor Vessel Parameters during CMT Reflood and subsequent Draindown 5.1.4-59 j 5.1.4-31 CMT-2 and Reactor Vessel Parameters during } CMT Reflood and subsequent Draindown 5.1.4-60 5.1.4-32 CMT-2 Fluid Temperatures 5.1.4-61 5.1.4-33 CMT-1 Fluid Temperatures 5.1.4-62 5.1.4-34 CMT-1, CMT-2, and IRWST Levels 5.1.4-63 5.1.4-35 IRWST, Sump, and Break Separator Levels 5.1.4-64 5.1.4-36 Separator Loop Seal Flows 5.1.4-65
- 5.1.4 37-IRWST and Primary Sump Flows 5.1.4-66 5.1,4-38 Downcomer Annulus Temperatures at 0 degrees Azimuth 5.1.4-67 5.1.4-39 Total DVI Flow 5.1.4-68 5.1.4-40 PRHR HX Temperatures 5.1.4-69 5.1.4-41 PRHR HX Short-Tube and Long-Tube Temperatures 5.1.4-70 5.1.4-42 CL-2 Temperatures 5.1.4-71 5.1.4-43 CL-4 Temperatures 5.1.4-72 5.1.4-44 Reactor Heater Temperatures @ 46 in. - Top of Core 5.1.4-73 5.1.4 Primary and Secondary Pressures 5.1.4-74 5.1.4-46 Upper-Plenum and Upper-Head Temperatures 5.1.4-75 5.1.4-47 Upper-Head and Downcomer Temperatures 5.1.4-76 5.1.4-48 IRWST and Primary Sump Flows 5.1.4-77 ./O 5.1.4-49 Upper-Head DPs 5.1.4-78 d
4 c:\l5hRevi\l5h.non lb4)90398 xxj REVISION 1
FINAL DATA REPORT LIST OF FIGURES (Continued) Eage Figure Title 5.1.4-79 5.1,4 50 CMT-1 Levelfremperature vs. Time 5.1.4-80 5.1.4-51 CMT-2 Levelfremperature vs. Time 5.1.4-81 5.1.4-52 Cold-Leg Levels 5.1.4-82 5.1.4-53 CL-1 Temperatures 5.1.4 83 5.1.4-54 CL-3 Temperatures 5.1.4-84 5.1.4-55 ADS 1-3 Flow DPs 5.1.4-85 5.1.4-56 Reactor /HL-2/SG-2 Channel Head Levels 5.1.4-86 5.1.4-57 Reactor /HL-2/SG-2 Channel Head Steam Percent 5.1.4-87 5.1.4-58 Pressurizer and Surge Line Levels 5.1.4-88 5.1.4-59 Pressurizer and Surge Line Steam Percent 5.1.4 89 5.1.4-60 DVI Flows 5.1.4-90 5.1.4-61 ADS 1-3 Liquid and Steam Flows 5.1.4-91 5.1.4-62 Break Separator Liquid and Steam Flows 5.1.4-92 5.1.4-63 Pressurizer and Reactor Pressures 5.1.4-93 5.1.4-64 Accumulator Levels 5.1.4-94 5.1.4-65 ADS 1-3 Separator Level 5.1.4-95 5.1.4-66 PRHR HX Flows 5.1.4-96 5.1.4-67 PRHR HX Levels 5.1.4-97 5.1.4-68 PRHR HX Levels 5.1.4-98 5.1.4-69 IRWST Overflow and Associated Pressures 5.1.4-99 5.1.4-70 IRWST Short-Rod and Sparger Tip Temperatures 5.1.4-100 5.1.4-71 IRWST Long-Rod Top Half Temperatures 5.1.4-101 5.1.4-72 IRWST Long-Rod Bottom Half Temperatures 5.1.4-102 5.1.4-73 BAMS Pressures 5.1.4-103 5.1.4-74 BAMS Pressures 5.1.4-104 5.1.4 74x BAMS Pressures 5.1.4-105 5.1.4-75 BAMS Header Steam Flows 5.1.4-106 5.1.4-76 Separator Steam Flows 5.1.4-107 5.1.4-77 CMT-1 Wide-Range and Balance Line Levels 5.1.4-108 5.1.4-78 CMT-1/ Reactor Vessel /CL-3 Pressures 5.1.4-109 5.1.4-79 CMT-1 Inlet Temperature 5.1.4 110 5.1.4-80 CMT-1 Temperatures Steam Generator Primary Side Differential Pressures 5.1.4-111 5.1.4-81 Steam Generator Wide-Range Levels 5.1.4-112 5.1.4 82 SG-1 U-Tube Wall Temperatures 5.1.4-113 5.1.4 83 5.1.4-114 5.1.4-84 SG-2 U-Tube Wall Temperatures SG 1 Steam and Downcomer Fluid Temperatures 5.1.4-115 5.1.4-85
)
REVISION 1 o:\l 536w Rev l\l 536w.non: l tM)90398 Xxii
1 FINAL DATA REPORT i (O LIST OF FIGURES (Continued) l V l f.lEILES I!M.s P. BILE l 5.1.4-86 SG-2 Steam and Downcomer Fluid Temperatures 5.1.4 116 Matrix Test SB23 Comparison with Matrix Test SB01 l 5.1.5-l a Primary Loop and Break Pipe Arrangement 5.1.5-20 5.1.5-lb Primary Loop and Break Pipe Arrangement 5.1.5-21 5.1.5-2 Reactor Upper Head Pressure 5.1.5-22 l 5.1.5-3 Reactor and Downcomer Annulus Steam Percent 5.1.5-23 5.1.5-4 Upper-Plenum Steam Percent 5.1.5-24 5.1.5-5 Pressurizer and Surge Line Levels 5.1.5-25 5.1.5-6 CMT-1 and CMT-2 Levels 5.1.5-26 5.1.5-7 SG-1 Tube Levels 5.1.5-27 5.1.5-8 SG-2 Tube Levels 5.1.5-28 5.1.5-9 SG-1 Channel Head Levels 5.1.5-29 5.1.5-10 SG-2 Channel Head Levels 5.1.5-30 5.1.5-11 Hot-Leg Levels 5.1.5-31 5.1.5-12 Upper-Plenum Levels 5.1.5-32 g 5.1.5-12x Upper-Plenum Levels 5.1.5-33 l y) 5.1.5-13 HL-1 Temperatures 5.1.5-34 5.1.5-14 HL-2 Temperatures 5.1.5-35 5.1.5-15 Reactor and Downcomer Annulus Wide-Range Levels 5.1.5-36 5.1.5-16 Accumulator and CMT Injection Flows 5.1.5-37 5.1.5-17 Total DVI Flow 5.1.5-38 5.1.5-18 Accumulator Injection Line Temperatures 5.1.5-39 5.1.5-19 Upper }{ead DPs 5.1.5-40 5.1.5-20 DVI Nozzle Temperatures 5.1.5-41 5.1.5-21 Pressurizer Heater Temperature 5.1.5-42 5.1.5-22 Reactor and Downcomer Annulus Wide-Range Levels 5.1.5-43 5.1.5-23 IRWST/ Primary Sump Injection Temperatures 5.1.5-44 5.1.5-24 Pressurizer Temperature and kW 5.1.5-45 5.1.5-25 IRWST Overflow and Associated Pressures 5.1.5-46 5.1.5-26 IRWST, Sump, and Break Separator Levels 5.1.5-47 5.1.5-27 Pressurizer Temperatures 5.1.5-48 l 5.1.5-28 Separator Loop Seal Flows 5.1.5-49 5.1.5-29 ADS 1-3 Pressures 5.1.5-50 5.1.5-30 CMT-1 and Reactor Vessel Parameters during CMT Reflood and subsequent Draindown 5.1.5-51 5.1.5-31 CMT-2 and Reactor Vessel Parameters during CMT Reflood and subsequent Draindown 5.1.5-52 c:\l 536wRev l\1536w.non:I b-090398 xxiii REVISION 1
FINAL DATA REPORT LIST OF FIGURES (Continued) Figure Title Page 5.1.5-32 CMT-2 Fluid Temperatures 5.1.5-53 5.1.5-33 CMT-1 Fluid Temperatures 5.1.5-54 5.1.5-34 CMT-1, CMT-2, and IRWST Levels 5.1.5-55 5.1.5-35 IRWST, Sump, and Break Separator Levels 5.1.5-56 5.1.5-36 Separator Loop Seal Flows 5.1.5-57 5.1.5-37 IRWST and Primary Sump Flows 5.1.5 58 5.1.5-38 Downcomer Annulus Temperatures at 0 degrees Azimuth 5.1.5-59 5.1.5-39 Total DVI Flow 5.1.5-60 5.1.5-40 PRHR HX Temperatures 5.1.5-61 5.1.5-41 PRHR HX Short-Tube and Long-Tube Temperatures 5.1.5-62 5.1.5-42 CL-2 Temperatures 5.1.5-63 5.1.5-43 CL-4 Temperatures 5.1.5-64 5.1.5-44 Reactor Heater Temperatures @ 46 in. - Top of Core 5.1.5-65 5.1.5-44x Reactor Heater Temperatures @ 46 in. - Top of Core 5.1.5-66 5.1.5-45 Primary and Secondary Pressures 5.1.5-67 5.1.5-46 Upper-Plenum and Upper-Head Temperatures 5.1.5-68 5.1.5-47 Upper-Head and Downcomer Temperatures 5.1.5-69 5.1.5-48 IRWST and Primary Sump Flows 5.1.5-70 5.1.5-49 Upper Head DPs 5.1.5-71 5.1.5-50 CMT-1 Level / Temperature vs. Time 5.1.5-72 5.1.5-51 CMT-2 Level / Temperature vs. Time 5.1.5-73 5.1.5-52 Cold-Leg Levels 5.1.5-74 5.L5 53 CL-1 Temperatures 5.1.5-75 5.1.5-54 CL-3 Temperatures 5.1.5-76 5.1.5-55 ADS 1-3 Flow DPs 5.1.5-77 5.1.5-56 Reactor /HL-2/SG-2 Channel Head Levels 5.1.5-78 5.1.5-57 Reactor /HL-2/SG-2 Channel Head Steam Percent 5.1.5-79 5.1.5-58 Pressurizer and Surge Line Levels 5.1.5-80 5.1.5-59 Pressurizer and Surge Line Steam Percent 5.1.5-81 5.1.5-60 DVI Flows 5.1.5-82 5.1.5-61 ADS 1-3 Liquid and Steam Flows 5.1.5-83 5.1.5-62 Break Separator Liquid and Steam Flows 5.1.5-84 5.1.5-63 Pressurizer and Reactor Pressures 5.1.5-85 5.1.5-64 Accumulator Levels 5.1.5-86 5.1.5-65 ADS 1-3 Separator Level 5.1.5-87 5.1.5-66 PRHR HX Flows 5.1.5-88 5.1.5-67 PRHR HX Levels 5.1.5-89 5.1.5-68 PRHR HX Levels 5.1.5-90 o:\l536wRevi\l536w.non:lt> 090398 xxjy REVISION 1 l
FINA1, DATA REroIT i 1
} LIST OF FIGURES (Continued)
ElEll.Ct 11.01 f881 5.1.5-69 IRWST Overflow and Associated Pressures 5.1.5-91
-5.1.5-70 IRWST Short-Rod and Sparger Tip Temperatures 5.1.5-92 5.1.5 IRWST Long-Rod Top-Half Temperatures 5.1.5-93 5.1.5-72 IRWST Long-Rod Bottom-Half Temperatures 5.1.5-94 i 5.1.5-73 BAMS Pressures 5.1.5-95 l 5.1.5-74 BAMS Pressures 5.1.5-% ) . 5.1.5-75 BAMS Header Steam Flows 5.1.5 97 5.1.5-76 Separator Steam Flows 5.1.5-98 5.1.5-77 CMT-1 Wide-Range and Balance Line Levels 5.1.5 5.1.5-78 CMT-1/ Reactor Vessel /CL-3 Pressures 5.1.5-100 5.1.5-79 CMT-1 Inlet Temperature 5.1.5-101 5.1.5-80 CMT-1 Temperatures . 5.1.5-102 ,
l 5.1.5-81 Reactor Core Temperatures 5.1.5-103 5,1,5 Reactor Core Temperatures 5.1.5-104 5.1.5-83 SG 1 Fluid Temperatures in Tubes 5.1.5-105 5.1.5-84 SG-2 Fluid Temperatures in Tubes 5.1.5 106 Matrix Test SB05 Comparison with Matrix Test SB01 5.1.6-l a Primary Loop and Break Pipe Arrangement 5.1.6-22 5.1.6-lb Primary Loop and Break Pipe Arrangement 5.1.6-23 1 5.1.6-2 Reactor Upper Head Pressure 5.1.6-24 I 5.1.6-3 Reactor and Downcomer Annulus Steam Percent 5.1.6-25 5.1.6-4 Upper-Plenum Steam Percent 5.1.6-26 5.1.6-5 Pressurizer and Surge Line Levels 5.1.6-27
'5.1.6-6 CMT-1 and CMT-2 Levels 5.1.6-28 5.1.6-6x CMT-1 and CMT-2 Levels 5.1.6-29 5.1.6-7 SG-1 Tube Levels 5.1.6-30 5.1.6-8 SG-2 Tube Levels 5.1.6-31 5.1.6-9 SG-1 Channel Head Levels 5.1.6-32 5.1.6-10 SG 2 Channel Head Levels 5.1.6-33
] 5.1.6-11 Hot-Leg Levels 5.1.6-34 5.1.6-12 Upper-Plenum Levels 5.1.6 5.1.6 HL-1 Temperatures 5.1.6-36 5.1.6-14 HL-2 Temperatures 5.1.6-37 5.1.6-15 Reactor and Downcomer Annulus Wide-Range Levels - 5.1.6-38 5.1.6-16 Accumulator and CMT Injection Flows 5.1.6-39
'5.1.6-17 Total DVI Flow 5.1.6-40 ~
5.1.6-18 Accumulator Injection Line Temperatures 5.1.6-41 < t
- oA15hRevl\l5h.non:Ib-090398 xxy REVISION 1
FINAL DATA REPORT I l 1 l LIST OF FIGURES (Continued) Title P.agg g 5.1.6-42 5.1.6-19 Upper Head DPs 5.1.6-43 5.1.6-20 DVI Nozzle Temperatures Pressurizer Heater Temperature 5.1.6-44 5.1.6-21 Reactor and Downcomer Annulus Wide-Range Levels 5.1.6-45 5.1.6-22 IRWST/ Primary Sump Injection Temperatures 5.1.6-46 5.1.6-23 Pressurizer Temperature and kW 5.1.6-47 5.1.6-24 5.1.6-48 5.1.6-25 -IRWST Overflow and Associated Pressures IRWST, Samp, and Break Separator Levels 5.1.6-49 5.1.6-26 Pressurizer Temperatures 5.1.6-50 5.1.6-27 Separator Loop Seal Flows 5.1.6-51 5.1.6-28 5.1.6-52 5.1.6-29 ADS 1-3 Pressures 5.1.6-30 CMT-1 and Reactor Vessel Parameters during CMT Reflood and subsequent Draindown 5.1.6-53 5.1.6-31 CMT-2 and Reactor Vessel Parameters during CMT Reflood and subsequent Draindown 5.1.6-54 5.1.6-32 CMT-2 Fluid Temperatures 5.1.6-55 5.1.6-33 CMT-1 Fluid Temperatures 5.1.6-54 5.1.6-56 5.1.6-34 CMT-1, CMT-2, and IRWST Levels 5.1.6-57 5.1.6-33 IRWST, Sump, and Break Separator Levels 5.1.6-58 l 5.1.6-36 Separator Loop Seal Flows 5.1.6-59 l 5.1.6-60 l 5.1.6-37 IRWST and Primary Sump Flows 5.1.6-38 Downcomer Annulus Temperatures at 0 degrees Azimuth 5.1.6-61 5.1.6-39 Total DVI Flow 5.1.6-62 5.1.6-40 PRHR HX Temperatures 5.1.6-63 l 5.1.6-41 PRHR HX Short-Tube and Long-Tube Temperatures 5.1.6-64 5.1.6-65 ( 5.1.6-42 CL-2 Temperatures 5.1.6-43 CL-4 Temperatures 5.1.6-66 l 5.1.6-44 Reactor Heater Temperatures @ 46 in. - Top of Core 5.1.6-67 5.1.6-45 Primary and Secondary Pressures 5.1.6-68 5.1.6-46 Upper-Plenum and Upper-Head Temperatures 5.1.6-69 L 5.1.6-47 Upper-Head and Downcomer Temperatures 5.1.6-70 5.1.6-48 IRWST and Primary Sump Flows 5.1.6-71 5.1.6-49 Upper Head DPs 5.1.6-72 5.1.6-50 CMT-1 Levelfremperature vs. Time 5.1.6-73 5.1.6-51 CMT-2 Level / Temperature vs. Time 5.1.6-74 5.1.6-52 Cold-Leg Levels 5.1.6-75 5.1.6-53 CL-1 Temperatures 5.1.6-76 c:\l536wRevl\l536w.non:1b-(D0398 xxvi REVISION 1
l FINAL. DATA REPORT l LIST OF FIGURES (Continued) v f.isnt .T.itit P.ast 5.1.6-54 CL-3 Temperatun;s 5.1.6-77 5.1.6-55 ADS 1-3 Flow I'Ps 5.1.6-78 5.1.6-56 Reactor /HL-2/SG-2 Channel Head Levels 5.1.6-79 5.1.6-57 Reactor /HL-2/SG-2 Channel Head Steam Percent 5.1.6-80 5.1.6-58 Pressurizer and Surge Line Levels 5.1.6-81 5.1.6-59 Pressurizer and Surge Line Steam Percent 5.1.6-82 5.1.6-60 DVI Flows 5.1.6-83 5.1.6-61 ADS 1-3 Liquid and Steam Flows 5.1.6-84 5.1.6-62 Break Separator Liquid and Steam Flows 5.1.6-85
- l. 5.1.6-63 Pressurizer and Reactor Pressures 5.1.6-86 ,
I 5.1.6-64 Accumulator Levels 5.1.6-87 I 5.1.6-65 ADS 1-3 Separator Level 5.1.6-88 ! 5.1.6-66 PRHR HX Flows 5.1.6-89 l 5.1.6-67 PRHR HX Levels 5.1.6-90 5.1.6-68 PRHR HX Levels 5.1.6-91 5.1.6-69 IRWST Overflow and Associated Pressures 5.1.6-92 g 5.1.6-70 IRWST Short-Rod and Sparger Tip Temperatures 5.1.6-93 5.1.6-71 IRWST Long-Rod and Top-Half Temperatures 5.1.6-94
'5.1.6-72 l l IRWST Long-Rod and Bottom-Half Temperatures 5.1.6-95
~
5.1.6-73 BAMS Pressures 5.1.6-96 l j 5.1.6-74 BAMS Pressures 5.1.6-97 5.1.6-74x BAMS Pressures - 5.1.6-98 5.1.6-75 BAMS Header Steam Flows 5.1.6-99 ! 5.1.6-76 Separator Steam Flows 5.1.6-100 I 5.1.6-77 CMT-1 Wide-Range and Balance Line Levels 5.1.6 101 5.1.6-78 CMT 1/ Reactor Vessel /CL-3 Pressures 5.1.6-102 5.1.6-79 CMT-1 Inlet Temperature 5.1.6-103 5.1.6-80 CMT-1 Temperatures' 5.1.6 104 I Matrix Test SB04 i 5.2.1-l a Primary Loop and Break Piping Layout 5.2.1-33 5.2.1-l b Primary Loop and Break Pipe Arrangement 5.2.1-34 5.2.12 Separator Loop Seal Flows 5.2.1-35 5.2.1-2x Separator Loop Seal Flows 5.2.1-36 5.2.1-3. BAMS Header Steam Flows 5.2.1-37 1
.5.2.1-4 Reactor and Downcomer Annulus Wide-Range Levels 5.2.1-38 l 5.2.1-4x Reactor and Downcomer Annulus Wide-Range Levels 5.2.1-39 5.2.1-5 Reactor and DVI Pressures 5.2.1-40 l
c:\1536wRev i\l 536w.non:1 b-090398 xxvii REVISION 1 l
FINAL DATA REPORT LIST OF FIGURES (Continued) P. age i F_lgure Title 5.2.1-41 5.2.1-6 Pressurizer Pressures Pressurizer and Surge Line Levels 5.2.1-42 5.2.1-7 Pressurizer and Surge Line Levels 5.2.1-43 5.2.1-7x 5.2.1-44 5.2.1-8 RNS and CVS Flows 5.2.1-45 5.2.1-8x RNS and CVS Flows 5.2.1-46 5.2.1-9 Reactor Core Levels 5.2.1-47 5.2.1-9x Reactor Core Levels Core Heater Temperature 5.2.1-48 5.2.1-10 Core Heater Temperature 5.2.1-49 5.2.1-10x keector Core Temperatures 5.2.1-50 5.2.1-11 Reactor Cere Temperatures 5.2.1-51 5.2.1-11 x 5.2.1-52 5.2.1-12 Hot-Leg Levels Hot-Leg Levels 5.2.1-53 5.2.1-12x HL-1 Temperatures 5.2.1-54 5.2.1-13 5.2.1-55 l 5.2.1-13 x HL-1 Temperatures HL-2 Temperatures 5.2.1-56 5.2.1-14 5,2.1-57 5.2.1 -14x HL-2 Temperatures SG-1 Tube Levels 5.2.1-58 5.2.1-15 5.2.1-59 5.2.1 -15x SG-1 Tube Levels SG-2 Tube Levels 5.2.1-60 5.2.1-16 5.2.1-61 5.2.1-16x SG-2 Tube Levels SG-1 Channel Head Levels 5.2.1-62 5.2.1-17 5.2.1-63 5.2.1 -17x SG-1 Channel Head Levels 5.2.1-64 5.2.1-18 SG-2 Channel Head Levels 5.2.1-65 5.2.1-18x SG-2 Channel Head Levels Accumulator Flow Rates 5.2.1-66 5.2.1-19 5.2.1-67 5.2.1-19x Accumulator Flow Rates Hot-Leg and Cold-Leg Pressures 5.2.1-68 5.2.1-20 Accumulator and CMT Pressures 5.2.1-69 5.2.1 21 Accumulator Levels 5.2.1-70 5.2.1-22 CMT-1 and CMT-2 Levels 5.2.1-71 5.2.1-23 CMT-1 and CMT-2 Levels 5.2.1-72 5.2.1-23x Upper Head DPs 5.2.1-73 5.2.1-24 Upper Head DPs 5.2.1-74 5.2.1-24x Upper-Plenum and Upper-Head Temperatures 5.2.1-75 5.2.1-25 Upper-Plenum and Upper-Head Temperatures 5.2.1-76 5.2.1-25x Downcomer Annulus Temperatures 5.2.1-77 5.2.1-26 Downcomer Annulus Temperatures 5.2.1-78 5.2.1 -26x xxviij REVISION 1 oA1536wRevi\l536w.non:Ib-090398
- . - . - _ - _ _ _ - - - . - _ ~ . . . . - . . - - . - . . . . - - . . . - - . - . _ . - - - . . - .
1 FINAL DATA Rt. roar ( LIST OF FIGURES (Continued) Elsis.ts I!!!s Esas 5.2.1-27 Reactor Core Temperatures 5.2.1 79 I 5.2.1-27x Reactor Core Temperatures 5.2.1-80 ! 5.2.1-28 Reactor Core Temperatures 5.2.1-81 ; 5.2.1-28x Reactor Core Temperatures 5.2.1-82 5.2.1-29 Reactor Core Temperatures 5.2.1-83 1 5.2.1-29x Reactor Core Temperatures 5.2.1-84 5.2.1-30 Top of Core Temperature Profile 5.2.1-85 5.2.1-30x Top of Core Temperature Profile 5.2.1-86 5.2.1-31 CMT Flows 5.2.1-87 5.2.1-32 CMT Cold-Leg and Balance Line Levels 5.2.1-88 I 5.2.1-33 CMT-1 Fluid Temperatures 5.2.1 5.2.1-34 CMT-1 Wall Temperatures - 5.2.1-90 5.2.1-35 CMT-2 Fluid Temperatures 5.2.1-91 5.2.1-36 CMT-2 Wall Temperatures 5.2.1-92 5.2.1-37 PRHR HX Flows 5.2.1-93 5.2.1-37x PRHR HX Flows 5.2.1-94 5.2.1-38 PRHR HX Levels 5.2.1-95 Q 5.2.1-38x 5.2.1-39 PRHR HX Levels PRHR HX Temperatures 5.2.1-96 5.2.1-97 5.2.1-39x PRHR HX Temperatures 5.2.1-98 5.2.1-40 Steam Generator Steam Flows 5.2.1-99 5.2.1-41 SG-1 Fluid Temperatures in Tubes 5.2.1-100 5.2.1-41 x SG-1 Fluid Temperatures in Tubes 5.2.1-101 5.2.1-42 SG-2 Fluid Temperatures in Tubes 5.2.1-102 . 5.2.1 -42x SG-2 Fluid Temperatures in Tubes 5.2.1-103 4 5.2.1-43 SG-1 Primary and Secondary Temperature 5.2.1-104 5.2.1-44 SG-2 Primary and Secondary Temperature - 5.2.1-105 5.2.1-45 Primary and Secondary Pressures 5.2.1-1 % 5.2.1-46 Cold-Leg Levels 5.2.1-107 5.2.1-46x Cold-Leg Levels 5.2.1-108 5.2.1-47 CL-1 Temperatures 5.2.1-109 5.2.1-47x CL-1 Temperatures 5.2.1-110 5.2.1 CL-2 Temperatures 5.2.1-111 5.2.1-48x CL-2 Temperatures 5.2.1-112 5.2.1-49 CL-3 Temperatures 5.2.1-113
~
5.2.1-49x CL-3 Temperatures 5.2.1-114 5.2.1-50 CL-4 Temperatures 5.2.1-115 5.2.1-50x CL-4 Temperatures 5.2.1-116 oA15hRevi$15W.non:Ib 09039s xxix REVISION 1
FINAt. DATA REPORT LIST OF FIGURES (Continued) Figure Title Page 5.2.1-51 IRWST Temperatures 5.2.1-117 5.2.1-51x IRWST Temperatures 5.2.1-118 5.2.1-52 Break Pressure 5.2.1-119 5.2.1-53 Separator Steam Flows 5.2.1-120 5.2.1-53x Separator Steam Flows 5.2.1-121 Matrix Test SB24 Comparison with Matrix Test SB04 5.2.2-l a Primary Loop and Break Pipe Arrangemtnt 5.2.2-16 5.2.2-l b Primary Loop and Break Pipe Arrangemvat 5.2.2-17 5.2.2-2 Separator Loop Seal Flows 5.2.2-18 5.2.2-3 BAMS Header Steam Flows 5.2.2-19 5.2.2-4 Reactor and Downcomer Annulus Wide-Range Levels 5.2.2-20 5.2.2-5 Reactor and DVI Pressures 5.2.2-21 5.2.2-6 Pressurizer Pressures 5.2.2-22 5.2.2-7 Pressurizer and Surge Line Levels 5.2.2-23 5.2.2-8 RNS and CVS Flows 5.2.2-24 5.2.2-9 Reactor Core Levels 5.2.2-25 5.2.2-10 Core Heater Temperatures 5.2.2 26 5.2.2-11 Reactor Core Temperatures 5.2.2-27 5.2.2-12 Hot-Leg Levels 5.2.2-28 5.2.2-13 HL-1 Temperatures 5.2.2-29 5.2.2-14 HL-2 Temperatures 5.2.2-30 5.2.2-15 SG-1 Levels 5.2.2-31 5.2.2-16 SG-2 Tube Levels 5.2.2-32 5.2.2 17 SG-1 Channel Head Levels 5.2.2-33 5.2.2-18 SG-2 Channel Head Levels 5.2.2-34 5.2.2-19 Accumulator Flow Rates 5.2.2-35 5.2.2-20 Hot-Leg and Cold-Leg Pressures 5.2.2-36 5.2.2-21 Accumulator and CMT Pressures 5.2.2-37 5.2.2-22 Accumulator Levels 5;2.2-38 5.2.2-23 CMT-1 and CMT-2 Levels 5.2.2-39 5.2.2 24 Upper Head DPs 5.2.2-40 5.2.2-25 Upper-Plenum and Upper-Head Temperatures 5.2.2-41 5.2.2-26 Downcorr..:t Annulus Temperatures at 180 degrees Azirnuth 5.2.2-42 5.2.2-27 Reactor u Temperatures 5.2.2-43 5.2.2-28 Reactor Core Temperatures 5.2.2-44 5.2.2-29 Reactor Core Temperatures 5.2.2-45 5.2.2-30 Top of Core Temperature Profile 5.2.2-46 o:\l536wRevl\l536w.non:Ib.o90398 xxx REVISION 1
FINAL. DATA REPORT i I O LIST OF FIGURES (Continued)
\d Figure Title P_ age 5.2.2-31 CMT Flows 5.2.2-47 5.2.2-32 CMT - Cold Leg Balance Line Level 5.2.2-48 5.2.2-33 CMT-1 Fluid Temperatures 5.2.2-49 5.2.2-34 CMT-1 Wall Temperatures 5.2.2-50 l 5.2.2-35 CMT-2 Fluid Temperatures 5.2.2-51 5.2.2-36 CMT-2 Wall Temperatures 5.2.2 52 5.2.2-37 PRHR HX Flows 5.2.2-53 5.2.2-38 PR.HR HX Levels 5.2.2-54 5.2.2-39 PRHR HX Temperatures 5.2.2 55 5.2.2-40 Steam Generator Steam Flows 5.2.2-56 5.2.2-41 SG-1 Fluid Temperatures in Tubes 5.2.2-57 5.2.2 42 SG-2 Fluid Temperatures in Tubes 5.2.2-58 5.2.2-43 SG-2 Primary and Secondary Temperature 5.2.2-59 5.2.2-44 SG-2 Primary and Secondary Temperature 5.2.2-60 5.2.2-45 Primary and Secondary Pressures 5.2.2-61 5.2.2-46 Cold-Leg Levels 5.2.2-62 5.2.2-47 CL-1 Temperatures 5.2.2-63 5.2.2-48 CL-2 Temperatures 5.2.2-64 5.2.2-49 CL-3 Temperatures 5.2.2 65 5.2.2-50 CL-4 Temperatures 5.2.2-66 5.2.2-51 IRWST Temperatures 5.2.2-67 5.2.2-52 Break Pressure 5.2.2-68 5.2.2 53 Steam Generator Secondary Leveh 5.2.2-69 Matrix Test SB10 5.3.1-1 ADS-4 to Separator Pipe Arrangeme.it for Matrix Test SB10 5.3.1-48 5.3.1-2 CMT-1 Balance Line DEG Break Pipe Arrangement 5.3.1-49 5.3.1-3 Break DPs 5.3.1-50 5.3.1-4 Break Pressure 5.3.1-51 5.3.1-5 Reactor and Downcomer Annulus Wide-Range Levels 5.3.1-52 5.3.1-6 Reactor Core Levels 5.3.1-53 5.3.1-7 Pressurizer and Surge Line Levels 5.3.1-54 5.3.1-8 CMT-1 and CMT-2 Levels 5.3.1-55 5.3.1-9 Reactor and DVI Pressures 5.3.1-56 5.3.1-10 Pressurizer Pressures 5.3.1-57 5.3.1-11 Accumulator and CMT Injection Flows 5.3.1-58 5.3.1-12 PRHR HX Flows 5.3.1-59 O
d 5.3.1-13 Cold-Leg Levels 5.3.1-60 oA1536wRevi\1536w.rmn:Ib-090398 xxxi REVISION l
FINAL DATA REPORT LIST OF FIGURES (Continued)
.I.' age Figure Title 5 3.1-61 5.3.1-14 Hot Leg Levels 5.3.1-62 5.3.1-15 Hot-Leg and Cold-Leg Pressures 5.1 1-63 5.3.1-16 SG 1 Levels 5 3.l-64 5.3.1-17 SG-1 Channel Head Levels 5.3.1 65 5.3.1-18 Steam Generator DPs and Levels 5.3.1-66 5.3.1-19 SG 1 Fluid Temperatures in Tubes 5.3.1-67 5.3.1-20 Primary / Secondary Pressure Comparison 5.3.1-68 5.3.1 21 Primary / Secondary Temperature Comparison 5.3.1-69 5.3.1-22 Upper-Head and Downcomer Temperatures 5.3.1-70 5.3.1-23 HL-1 Temperatures 5.3.1-71 5.3.1-24 HL-2 Temperatures 5.3.1-72 5.3.1-25 ADS 1-3 Valve Actuation 5.3.1-73 5.3.1-26 Separator Steam Flows 5.3.1-74 5.3.1-27 Separator Loop Seal Flows 5.3.1-75 5.3.1-28 Upper-Head DPs 5.3.1-76 5.3.1 29 Upper-Head DPs 5.3.1-77 5.3.1-30 Total DVI Flow IRWST and Primary Sump Flows 5.3.1-78 5.3.1 31 5.3.1-79 5.3.1-32 Upper-Plenum and Upper-Head Temperatures Reactor Core Temperatures 5.3.1-80 5.3.1 33 5.3.1-81 5.3.1-34 CL-1 Temperatures 5.3.1-82 5.3.1-35 CL-2 Temperatures 5.3.1-83 5.3.1-36 CL-3 Temperatures 5.3.1-84 5.3.1-37 CL-4 Temperatures Accumulator and CMT Injection Flows 5.3.1-85 5.3.1-38 5.3.1-86 5.3.1-39 CMT.I and CMT-2 Levels Reactor and Downcomer Annulus Wide-Range Levels 5.3.1-87 5.3.1-40 5.3.1-88 5.3.1-41 Total DVI Flow 5.3.1-89 5.3.1-42 Upper-Plenum and Upper-Head Temperatures IRWST and Primary Sump Flows 5.3.1-90 5.3.1-43 5.3.1-91 5.3.1-44 Upper-Head and Downcomer Temperatures 5.3.1-92 5.3.1-45 Separator Loop Seal Flows Reactor Core Temperatures 5.3.1-93 5.3.1-46 5.3.1-47 Reactor Heater Temperatures @ 46 in, from 5.3.1-94 Reactor Vessel Bottom 5.3.1-48 Reactor Heater Temperatures @ 46 in. from 5.3.1-95 Reactor Vessel Bottom 1RWST, Sump, and Break Separator Levels 5.3.1-96 5.3.1-49 REVISION 1 o:\l536*RevluS36w.non:lb.090398 xxxii
FmA1. DATA REPORT LIST OF FIGURES (Continued)
- m -
-5.3.1-50 1RWST, Sump, and Break Separator Levels 5.3.1-97 5.3.1-51 BAMS Header Steam Flows 5.3.1-98 5.3.1-52 . IRWST and Primary Sump Steam Flows 5.3.1-99 5.3.1-53 CMT 2 Temperature Profile 5.3.1-100 5.3.1-54 CMT-2 Temperature Profile 5.3.1-101 5.3.1-55' CMT-2 Upper Dome Temperatures 5.3.1-102 5.3.1-56 CMT-2 Upper Dome Temperatures 5.3.1-103 5.3.1-57 CMT-2 Fluid Temperature Distribution 5.3.1-104 5.3.1-58 CMT-2 Wall Temperature Distribution 5.3.1-105 5.3.1-59 CMT 2 Upper Dome Fluid / Wall Interface 5.3.1-106 5.3.1-60 CMT-1 Temperatures 5.3.1-107 5.3.1-61 CMT-1 Temperatures 5.3.1-108 5.3.1-62 Accumulator Levels 5.3.1-109 5.3.1-63 Pressurizer Temperature Profile 5.3.1-110 5.3.1 Steam Percentage Conditions at Reactor Vessel, SG-2, and HL-2 5.3.1-111 5.3.1-65 Steam Percentage Conditions for Pressurizer and Surge Line 5.3.1-112 5.3.1-66 . ADS 1-3 Separator Discharge Flows 5.3.1-113 Og 5.3.1-67 Pressurizer and Surge Line Levels 5.3.1-114 5.3.1-68 PRHR HX Temperatures 5.3.1-115 5.3.1-69 PRHR HX Temperatures 5.3.1-116 5.3.1-70 PRHR HX Levels 5.3.1-117 5.3.1-71 PRHR HX Levels 5.3.1-118 5.3.1-72 PRHR HX Flows 5.3.1-119 5.3.1-73 Upper-Head, CL-1, and CMT-2 Temperatures 5.3.1-120 5.3.1-74 Upper-Head, CL-1, and CMT-2 Temperatures 5.3.1-121 5 3.1-75 . Upper-Head, CL-1, and CMT-2 Temperatures 5.3.1-122 5.3.1-76 Hot-Leg and Upper-Head Temperatures 5.3.1-123 5.3.1-77 Comparison of Hot-Leg and Upper-Head Temperature 5.3.1 124 5.3.1-78 Comparison of Hot-Leg and Upper-Head Temperatures 5.3.1-125 5.3.1-79 Comparison of Hot-Leg and Upper-Head Temperatures 5.3.1-126
-5.3.1-80 Downcomer Annulus Temperatures Between HUDVI Elevations 5.3.1 127 5.3.1-81 Cold-Leg Levels 5.3.1-128 5.3.1-82 Cold-Leg Temperatures 5.3.1-129
- 5.3.1-83 ~ Hot-Leg Levels 5.3.1-130 5.3.1 84 Hot-Leg Temperatures 5.3.1 131 5.3.1-85 IRWST Temperatures 5.3.1-132 n
\j oA1536wRevi\1536w.non:lt4NO398 xxxiij REVISION 1
FINAL DfaTA REPORT LIST OF FIGURES (Continued) Figure Title Page 5.3.1-86 IRWST Temperatures 5.3.1-133 5.3.1-87 Core Steam Percentage 5.3.1-134 Matrix Test SB10 5.3.2-1 CMT-1 Balance-Line Break Pipe Arrangement for Matrix Test SB09 5.3.2-28 5.3.2-2 CMT-12-Inch Balance-Line Break Pipe Arrangement 5.3.2-29 5.3.23 Break DPs 5.3.2-30 5.3.2-4 Break Pressure 5.3.2-31 5.3.2-5 Reactor and Downcomer Annulus Wide-Range Levels 5.3.2-32 5.3.2-6 Reactor Core Levels 5.3.2-33 5.3.2-7 Pressurizer and Surge Line Levels 5.3.2-34 5.3.2-8 CMT-1 and CMT-2 Levels 5.3.2-35 5.3.2-9 Reactor and DVI Pressures 5.3.2-36 5.3.2 10 Pressurizer Pressures 5.3.2-37 5.3.2-11 Accumulator and CMT Injection Flows 5.3.2-38 5.3.2-12 PRHR HX Flows 5.3.2-39 5.3.2-13 Cold-Leg Levels 5.3.2-40 5.3.2-14 Hot-Leg Levels 5.3.2-41 5.3.2-15 Hot-Leg and Cold-Leg Pressures 5.3.2-42 5.3.2-16 SG-1 Levels 5.3.2-43 5.3.2-17 SG-1 Channel Head Levels 5.3.2-44 5.3.2-18 Steam Generator DPs and Levels 5.3.2-45 5.3.2-19 SG 1 Fluid Temperatures in Tubes 5.3.2-46 5.3.2-20 Primary / Secondary Pressure Comparison 5.3.2-47 5.3.2-21 Primary / Secondary Temperature Comparison 5.3.2-48 5.3.2-22 Upper-Head and Downcomer Temperatures 5.3.2-49 5.3.2-23 HL-1 Temperatures 5.3.2-50 5.3.2-24 HL-2 Temperatures 5.3.2-51 5.3.2-25 ADS 1-3 Valve Actuation 5.3.2-52 5.3.2-26 Separator Steam Flows 5.3.2-53 5.3.2-27 Separator Loop Seal Flows 5.3.2-54 5.3.2-28 Upper-Head DPs 5.3.2-55 5.3.2 29 Upper-Head DPs 5.3.2-56 5.3.2-30 Total DVI Flow 5.3.2-57 5.3.2-31 IRWST and Primary Sump Flows 5.3.2-58 5.3.2-32 Upper-Plenum and Upper-Head Temperatures 5.3.2-59 5.3.2-33 Reactor Core Temperatures 5.3.2-60 5.3.2-34 CL-1 Temperatures 5.3.2-61 5.3.2-35 CL-2 Tempemtures 5.3.2-62 c:U536wRevi\l536w.non:lb-090398 XXXiv REVISION 1
~ Fmrt. DATA REPC:T J' i LIST OF FIGURES (Continued) n.- m e-f' 5.3.2-36 CL-3 Temperatures 5.3.2-63
;. 5.3.2-37 CL-4 Temperatures 5.3.2-64 5.3.2 38 Accumulator and CMT Injection Flows 5.3.2-65
. 5.3.2-39 CMT-1 and CMT-2 Levels 5.3.2-66 5.3.2-40 Reactor and Downcomer Annulus Wide-Range Levels 5.3.2-67 j- -5.3.2-41 Total DVI Flow 5.3.2-68 5.3.2-42 Upper-Plenum and Upper-Head Temperatures 5.3.2-69 j' 5.3.2-43 IRWST and Primary Sump Flows 5.3.2-70 5.3.2-44 Upper-Head and Downcomer Temperatures 5.3.2-71 5.3.2-45 Separator Loop Seal Flows 5.3.2-72
- 5.3.2-46 Reactor Core Temperatures 5.3.2-73
-5.3.2-47 Reactor Heater Temperatures @ 46 in. from Reactor Vessel Bottom 5.3.2-74 5.3.2-48 Reactor Heater Temperatures @ 46 in from Reactor Vessel Bottom 5.3.2-75 5.3.2-49 IRWST, Sump, and Ereak Separator Levels 5.3.2-76 4 5.3.2-50 IRWST, Sump, and Break Separator Levels 5.3.2-77 5.3.2-51 BAMS Header Steam Flows 5.3.2-78
. - s 5.3.2-52 IRWST and Primary Sump Steam Flows 5.3.2-79
+
5.3.2 CMT-2 Temperature Profile 5.3.2-80 5.3.2-54 ~ CMT-2 Temperature Profile
- 5.3.2-81 )
[; 5.3.2-55 CMT-2 Upper Dome Temperatures 5.3.2-82 1 5.3.2-56 CMT-2 Upper Dome Temperatures 5.3.2-83 1 5.3.2-57 CMT-2 Fluid Temperature Distribution 5.3.2-84 4
-5.3.2-58 CMT-2 Wall Temperature Distribution 5.3.2-86 5.3.2-59 CMT-2 Upper-Dome Fluid / Wall Interface 5.3.2-87 5.3.2-60 CMT-1 Temperatures 5.3.2-89
. 5.3.2-61 CMT-1 Temperatures 5.3.2-90 5.3.2-62 Accumulator Levels 5.3.2-91 5.3.2-63' Pressurizer Temperature Profile 5.3.2 92 j' 5.3.2-64 Steam Percentag: Conditions at Reactor Vessel, SG-2, and HL-2 5.3.2-93 , 5.3.2-65 Steam Percentage Conditions for Pressurizer and Surge Line 5.3.2-94
.5.3.2-66 ADS 1-3 Separator Discharge Flows 5.3.2 95 5.3.2-67 Pressurizer and Surge Line Levels 5.3.2-%
J 5.3.2 PRHR HX Temperatures 5.3.2-97 5.3.2-69 PRHR HX Temperatures 5.3.2-98 5.3.2-70 PRHR HX Levels 5.3.2-99 5.3.2-71 PRHR HX Levels 5.3.2-100 5.3.2-72 PRHR HX Flows 5.3.2-101 O d 5.3.2-73 5.3.2 Upper-Head, CL-1, and CMT-2 Temperatures Upper-Head, CL-1, and CMT-2 Temperatures 5.3.2-102 5.3.2-103 oA15hRevlu5h.non:ll>09039s xxxy REVISION 1
r -- i I FmAI DATA REroar l l LIST OF FIGURES (Continued) Figure Title Me f 5.3.2 104 5.3.2-75 Upper-Head, CL-1, and CMT-2 Temperatures Hot-Leg and Upper-Head Temperatures 5.3.2-105 l 5.3.2-76 Comparison of Hot-Leg and Upper-Head Temperature 5.3.2-107 5.3.2-77 Comparison of Hot-Leg and Upper-Head Temperatures 5.3.2-109 5.3.2-78 Comparison of Hot-Leg and Upper-Head Temperatures 5.3.2-111 5.3.2-79 Downcomer Annulus Temperatures Between HUDVI Elevations 5.3.2-113 5.3.2-80 Cold-Leg Levels 5.3.2-115 5.3.2-81 Cold-Leg Temperatures 5.3.2-116 5.3.2-82 Hot-Leg Levels 5.3.2-117 5.3.2-83 5.3.2-84 Hot-Leg Temperatures 5.3.2-118 5.3.2-85 IRWST Temperatures 5.3.2-119 5.3.2-86 IRWST Temperatures 5.3.2-120 5.3.2-87 Reactor Pressure and Core Level Comparison 5.3.2-121 5.3.2-88 Reactor Pressure and Core Level Comparison 5.3.2-122 5.3.2-89 Reactor Pressure and Core Level Comparison 5.3.2 123 5.3.2-90 Reactor Downcomer Level Comparison 5.3.2-124 5.3.2-91 Reactor Downcomer Level Comparison 5.3.2-125 5.3.2-92 Reactor Downcomer Level Comparison 5.3.2-126 5.3.2-93 Netflow for Test SB09 For 0-2000 Seconds 5.3.2-127 5.3.2-94 Netflow for SB10 for 0-2000 Seconds 5.3.2-128 5.3.2-95 Netflow and Core Level Comparison for Test SB10 vs SB09 5.3.2 129 5.3.2-96 Netflow and Core Level Comparison for Test SBIO vs SB09 5.3.2-130 5.3.2-97 Netflow and Core Level Comparison for Test SBIO vs SB09 5.3.2-131 5.3.2-98 Reactor Pressure, CMT-1, and ACC-1 Flow Comparison 5.3.2-132 5.3.2-99 Reactor Pressure, CMT-2, and ACC-2 Flow Comparison 5.3.2-133 5.3.2-100 Reactor Pressure r.nd CMT Level Comparison 5.3.2-134 5.3.2-101 Reactor Vessel Response 5.3.2-135 5.3.2-102 Reactor Vessel Response 5.3.2-136 Matrix Test SB12 5.4.1-1 Primary Loop and Break Pipe Arrangement 5.4.1-48 5.4.1-2 ADS 1-3 and Break Separator Liquid Flows 5.4.1-49 5.4.1-3 Break Separator and BAMS Steam Flows 5.4.1-50 5.4.1-4 ACC-1 and CMT-1 Injection Flows 5.4.1-51 5.4.1-5 CMT-1 and CMT-2 Wide-Range Levels 5.4.1-52 5.4.1-6 ACC-1 and ACC-2 Levels 5.4.1-53 5.4.1-7 Main Steam Header and Reactor Pressures 5.4.1-54 5.4.1-8 ACC-2 and CMT-2 Injection Flows 5.4.1-55 5.4.1-9 CMT-1 and CMT-2 Balance Line Levels 5.4.1-56 o.\l 536w Rev i\l 536w.norr i b-090398 XXXvi REVISION 1
l l FINA1. DATA REromT LIST OF FIGURES (Condnued) l EiE9Et .TMs fant
'5.4.1-10 PRHR HX Flows 5.4.1-57 5.4.1-11' PRHR J-IX Inlet and Outlet Temperatures 5.4.1-58 l 5.4.1-12 SG-1 Tube Temperatures 5.4.1-59 5.4.1-13 SG-2 Tube Temperatures 5.4.1-60 5.4.1-14 Wide-l'.ange Downcomer Levels 5.4.1-61 5.4.1-15 Pressurt. Upstream of Break Valves 5.4.1-62 5.4.1-16 Upper-P1 num and Upper-Head Steam Percent 5.4.1-63 ;
5.4.1-17 Hot-Leg Panum and Elbow Steam Percent 5.4.1-64 5.4.1-18 ADS 1-3 Separator Liquid and Steam Flows 5.4.1-65 l
-5.4.1-19 Reactor Pressure 5.4.1-66 5.4.1-20 ADS Separators' Steam Flows 5.4.1-67 5.4.1-21 Effect of Pressure on CMT-2, ACC-2 Flows 5.4.1-68 5.4.1-22 Heater Temperatures-Top of Core 5.4.1-69 l 5.4.1-23 Reactor Wide-Range Level 5.4.1-70 5.4.1-24 ADS 4-1 and 4-2 Separator Liquid Flows 5.4.1 71 5.4.1-25 IRWST and Primary Sump Injection Flows 5.4.1-72 5.4.1-26 F/mtor and Downcomer Wide-Range Levels 5.4.1 73 l \ 5.4.1-27 IRWST, Break Separator, and Sump Levels - 5.4.1-74 5.4.1-28 ADS 4-1 and 4-2 Separator Liquid Flows and Levels 5.4.1-75 5.4.1-29 CMT-1 Flows and Levels 5.4.1-76 ,
5.4.1-30 CMT-1 and Cold-Leg Pressures 5.4.1-77 5.4.1-31 Plessurizer and Surge Line Steam Percent 5.4.1-78 5.4.1 Pressurizer /RCS and Pressurizer / ADS 1-3 Separator Diff Pressures 5.4.1-79 5.4.1-33 PRHR HX Inlet and Outlet Temperatures 5.4.1-80 5.4.1-34 PRHR HX Inlet and Outlet Temperatures 5.4.1-81 5.4.1-35 PRHR HX Tube and Channel Head Levels 5.4.1-82 5.4.1-36 HL-2 Minus SG-2 Tube Pressure Difference 5.4.1-83 5.4.1-37 PRHR HX Inlet, Outlet, and Tube Temperatures 5.4.1-84 5.4.1-38 PRHR HX Inlet, Outlet, and Tube Temperatures 5.4.1 85 5.4.1-39 Cold-Leg Temperatures-Top of Pipe 5.4.1-86 5.4.1 Cold-Leg Temperatures-Bottom on Pipe 5.4.1-87 5.4.1-41 Downcomer Level Top'of Cold Leg to DVI Elevation 5.4.1 88
-5.4.1-42 Cold-Leg Wall Differential Temperatures 5.4.1-89 5.4.1-43 SG-1 Tube Temperatures 5.4.1-90 5.4.1-44' SG-2 Tube Temperatures 5.4.1-91 5.4.1-45 Upper Support Plate and Bypass Hole DP's 5.4.1-92
- 5.4.1-46 IRWST Temperatures 5.4.1-93 5.4.1-47 IRWST and Primary Sump Exhaust Steam Flows 5.4.1-94 o:\l5hRevl\l5W.non:lb-090398 xxxyjj REVISION 1
4 FINAL DATA REPORT LIST OF FIGURES (Continued) Figure Title Eage Downcomer Wide-Range Level 5.4.1-95 5.4.1-48 5.4.1-49 Downcomer Fluid Temperatures at Hot-Leg Elevation 5.4.1-96 Downcomer Fluid Temperatures 5.4.1-97 5.4.1-50 Downcomer Fluid Temperatures 5.4.1-98 5.4.1-51 5.4.1-52 Downcomer Fluid Temperatures 5.4.1-99 5.4.1-53 Downcomer Fluid Temperatures 5.4.1-100 5.4.1-54 Axial Temperature Profile at Center of Core 5.4.1-101 5.4.1-55 Heater Temperatures at Top of Core 5.4.1-102 5.4.1-56 Narrow-Range and Wide-Range Core Levels 5.4.1-103 5.4.1-57 Narrow-Range and Wide-Range Core Levels 5.4.1-104 5.4.1-58 Core Steam Percent 5.4.1-105 5.4.1-59 Core Steam Percent 5.4.1-106 5.4.1-60 Differential Pressure of Upper Support Plant and Bypass Holes 5.4.1-107 5.4.1-61 Upper-Plenum, Upper-Head, Upper-Downcomer Temperatures 5.4.1-108 l I 5.4.1-62 SG-1 Tube Levels 5.4.1-109 5.4.1-63 SG-2 Tube Levels 5.4.1 110 5.4.1-64 SG-1 Channel Head Levels 5.4.1-111 5.4.1-65 SG-2 Channel Head Levels 5.4.1-112 5.4.1-66 SG-1. SG-2 Primary and Secondary Pressures 5.4.1-113 5.4.1-67 CMT-2 and Cold-Leg Pressures 5.4.1-114 5.4.1-68 CMT-2 Levels and Flows 5.4.1-115 5.4.1-69 CMT Long-Rod Thermocouple Temperatures 5.4.1-116 5.4.1-70 CMT-1 Fluid Thermocouple Temperatures 5.4.1-117 5.4.1-71 CMT-1 Inside-Wall Temperatums 5.4.1-118 Matrix Test SB13 5.4.2-1 Primary Loop and Break Piping Layout 5.4.2-37 5.4.2-2 ADS 1-3 and Break Separator Liquid Flows 5.4.2-38 l 5.4.2-3 Break Separator and BAMS Steam Flows 5.4.2-39 5.4.2-4 ACC-1 and CMT-1 Injection Flows 5.4.2-40 5.4.2-4x ACC-1 and CMT-1 Injection Flows 5.4.2-41 l 5.4.2-5 CMT-1 and CMT-2 Wide-Range Levels 5.4.2-42 l 5.4.2-5x CMT-1 and CMT-2 Wide-Range Levels 5.4.2-43 5.4.2-6 ACC-1 and ACC-2 Levels 5.4.2-44 5.4.2-7 Main Steam Header and Reactor Pressures 5.4.2-45 5.4.2-8 ACC-2 and CMT-2 Injection Flows 5.4.2-46 5.4.2-8x ACC-2 and CMT-2 Injection Flows 5.4.2-47 o:\l 536w Rev l\l 536w.non:I b-090398 xxxviii REVISION 1
FINAL DATA REMMT
.( LIST OF FIGURES (Continued)
= m 1.
5.4.2-9 CMT-1 and CMT-2 Balance Line Levels 5.4.2-48 5.4.2-9x CMT-1 and CMT-2 Balance Line Levels 5.4.2-49 5.4.2-10 PRHR HX Flows 5.4.2-50 5.4.2-11 PRHR HX Inlet and Outlet Temperatures 5.4.2-51
- 5.4.2-12 SG-1 Tube Temperatures 5.4.2-52 5.4.2-13 SG-2 Tube Temperatures 5.4.2-53 5.4.2-14 Wide Range Down:omer Levels 5.4.2-54 5.4.2-15 Pressure Upstream of Break Valves 5.4.2-55 5.4.2 16 Upper-Plenum and Upper-Head Steam Percent 5.4.2-56 5.4.2-17 Hot-Leg Plenum and Elbow Steam Percent 5.4.2-57 5 '.2-18 ADS 1-3 Separator Liquid and Steam Flows 5.4.2 58 1 .2-19 Reactor Pressure 5.4.2-59
-5.4.2-20 ADS Separators' Steam Flows 5.4.2-60 i 5.4.2-21 Effect of Pressure on CMT-2, ACC-2 Flows 5.4.2-61 5.4.2-22 Heat:r Temperatures--Top of Core 5.4.2-62 5.4.2-23 Reactor Wide-Range Level 5.4.2-63 n 5.4.2-24 ADS 4-1 and 4-2 Separator Liquid Flows 5.4.2-64 Q 5.4.2-24x 5.4.2-25 ADS 4-1 and 4-2 Separator Liquid Flows IRWST and Primary Sump Injection Flows 5.4.2-65 5.4.2-66 5.4.2-25x IRWST and Primary Sump Injection Flows 5.4.2-67 5.4.2-26 Reactor and Downcomer Wide-Range Levels 5.4.2-68 5.4.2-27 IRWST, Break Separator, and Sump Irvels 5.4.2-69 5.4.2-28 ADS 4-1 and 4 2 Separator Liquid Flows and Levels 5.4.2-70 5.4.2-29 CMT-1 Flows and Levels 5.4.2-71 5.4.2 30 ~ CMT-1 and Cold-Leg Pressures 5.4.2-72 5.4.2-30x CMT-1 and Cold-Leg Pressures 5.4.2-73 5.4.2-31 Pressurizer and Surge Line Steam Percent 5.4.2-74 5.4.2-31x Pressurizer and Surge Line Steam Percent 5.4.2-75 5.4.2-32 Pressurizer /RCS and Pressurizer / ADS 1-3 Separator Diff Pressures 5.4.2-76 5.4.2-33 PRHR HX Inlet and Outlet Temperatures 5.4.2-77 5.4.2-34. PRHR HX Inlet and Outlet Temperatures 5.4.2-78 5.4.2-35 PRHR HX Tube and Channel Head Levels 5.4.2-79 5.4.2-36 HL-2 Minus SG 2 Tube Pressure Difference 5.4.2-80
~ 5.4.2-37 PRHR HX Inlet, Outlet, and Tube Temperatures 5.4.2-81 5.4.2-38 PRHR HX Inlet, Outlet, and Tube Temperatures 5.4.2-82
' 5.4.2-39 Cold-Leg Temperatures--Top of Pipe 5.4.2-83
/ 5.4.2-40 Cold-Leg Temperatures--Bottom on Pipe 5.4.2-84 k 5.4.2-41 Downcomer Level Top of Cold Leg to DVI Elevation 5.4.2-85 o:\l536wRevi\l536w.non:IM90398 xxxix REVISION 1 l
L
FmAt, DATA REPORT LIST OF FIGURES (Continued) Title B!ge Figure Cold-Leg Walt DiB~erential Temperatures 5.4.2-86 5 4.2-42 SG-1 Tube Temperatures 5.4.2 87 5.4.2-43 SG-2 Tube Temperatures 5.4.2 88 5.4.2-44 Upper Support Pla'e and Bypass Hole DP's 5.4.2-89 5.4.2-45 IRWST Temperatur es 5.4.2-90 5.4.2-46 IRWST and Primary Sump Exhaust Steam Flows 5.4.2-91 5.4.2-47 Downcomer Wide-Rac.ge Level 5.4.2-92 5.4.2-48 5.4.2-49 Downcomer Fluid Temperatures at Hot-Leg Elevation 5.4.2-93 5.4.2-50 Downcom,-r Fluid Temperatures 5.4.2-94 5.4.2-51 Downcomer Fidd Temperatures 5.4.2-95 5.4.2-52 Downcomer Fluid Temperatures 5.4.2-96 '5.4.2-53 Downcomer Fluid Temperatures 5.4.2-97 5.4.2-54 Axial Temperature Profile at Center of Core 5.4.2-98 Heater Temperatures at Top of Core 5,4.2-99 5.4.2-55 5.4.2-56 Narrow-Range and Wide-Range Core Levels 5.4.2-100 5.4.2-57 Nanow-Range and Wide-Range Core Levels 5.4.2-101 5.4.2-58 Core Steam Percent 5.4.2-102 5.4.2-59 Core Steam Percent 5.4.2-103 5A2-60 Differential Pressure of Upper Suppoit Plant and Bypass Holes 5.4.2-104 5.4.2-61 Upper-Plenum, Upper-Head, Upper-Downcomer Temperatures 5.4.2-105 5.4.2-62 SG-1 Tube Levels 5.4.2 106 5.4.2-63 SG-2 Tube Levels 5.4.2-107 5.4.2-64 SG-1 Channel Head Levels 5.4.2-108 5.4.2-65 SG-2 Channel Head Levels 5.4.2-109 5.4.2-66 SG-1, SG-2 Primary and Secondary Pressures 5.4.2-110 5.4.2-67 CMT-2 and Cold-Leg Pressurec 5.4.2-111 5.4.2-68 CMT-2 Levels and Flows 5.4.2-112 5.4.2-69 CMT Long-Rod Thermocouple iemperatures 5.4.2-113 5.4.2-70 CMT-1 Fluid Thermocouple Temperatures 5.4.2-114 5.4.2-71 CMT-1 Inside-Wall Temperatures 5.4.2-115 5.4.2-72 Calculated Break Flow from Reactor Vessel 5.4.2-116 Matrix Test SB28 Comparison with Matrix Test SB12 5.4.3-1 Primary Loop and Break Pipe Arrangement 5.4.3-29 5.4.3-2 ADS 1-3 and Break Separator Liquid Flows 5.4.3-30 5.4.3-3 Break Separator and BAMS Steam Flows 5.4.3-31 5.4.3-4 ACC-1 and CMT-1 Injection Flows 5.4.3-32 5.4.3-5 CMT-1 and SMT-2 Wide-Range Levels 5.4.3-33 0:\l 536w Rev l\l 536w.non: l t>.090398 x1 REVISION 1
1 FINAL DATA REPORT j l l l i LIST OF FIGURES (Continued) rians m ran ) 5.4.3-6 ACC-1 and ACC-2 Levels 5.4.3-34 5.4.3-7 Main Steam Header and Reactor Pressures 5.4.3 35 5.4.3 8 ACC-2 and CMT-2 Injection Flows 5.4.3 36 5.4.39 CMT-1 and CMT-2 Balance Line Levels 5.4.3-37 5.4.3-10 PRHR HX Flows 5.4.3-38 5.4.3-11 PRHR Inlet and Outlet Temperatures 5.4.3-39 5.4.3-12 SG-1 Tube Temperatures 5.4.3-40 5.4.3-13 SG 2 Tube Temperatures 5.4.3-41 5.4.3-14 Wide-Range Downcomer Levels 5.4.3-42 5.4.3-15 Pressure Upstream of Break Valves 5.4.3-43 5.4.3-16 Upper Plenum and Upper-Head Steam Percent 5.4.3-44 5.4.3-17 Hot-Leg Plenum and Elbow Steam Percent 5.4.3-45
. 5.4.3-18 ADS 1-3 Separator Liquid and Steam Flows 5.4.3-46 5.4.3-19 Reactor Pressure 5.4.3-47 5.4.3-20 ADS Separators' Steam Flows 5.4.3-48 5.4.3-21 Effect of Pressure on CMT-1, ACC-1, Flows 5.4.3-49
- 5.4.3-22 Heater Temperatures - Top of Core 5.4.3-50
( 5.4.3-23
- 5.4.3-24 Reactor Wide-Range Level 5.4.3-51 ADS 4-1 and 4-2 Separator Liquid Flows 5.4.3-52 5.4.3-25 IRWST and Primary Sump Injection Flows 5.4.3-53 ;
5.4.3-26 Reactor and Downcomer Wide-Range Levels 5.4.3 54 5.4.3-27 IRWST, Break Separator, and Sump Levels 5.4.3 55 5.4.3-28 ADS 4-1 and 4-2 Separator Liquid Flows 5.4.3-56 5.4.3-29 CMT-1 Flows and Levels 5.4.3-57 5.4.3 30 CMT-1 and Cold-Leg Pressures- 5.4.3-58 5.4.3-31 Pressurizer and Surge Line Steam Percent 5.4.3-59 5.4.3 31x Pressurizer and Surge Line Steam Percer;t 5.4.3-60 5.4.3 32 Pressurizer and Surge Line Steam Percent 5.4.3-61 5.4.3-33 Pressurizer /RCS and Pressurizer ADS 1-3 Separator Differential Pressures 5.4.3-62 5.4.3-34 PRHR HX Inlet and Outlet Temperatures 5.4.3-63 5.4.3-35. ' PRHR HX Tube and Channel Head Levels 5.4.3-64 5.4.3-36 SG-2 Tube /HL-2 Pressure Difference 5.4.3-65
= 5.4.3-37 PRHR HX Inlet, Outlet, and Tube Temperatures 5.4.3-66 5.4.3-38 ' PRHR HX Inlet, Outlet, and Tube Temperatures -5.4.3-67 5.4.3-39 Cold-Leg Temperatures-Top of Pipe 5.4.3-68 5.4.3-40 Cold-Leg Temperatures-Bottom of Pipe 5.4.3.69 5.4.3-41 Downcomer Level Top of Cold Leg to DVI Dmti,n 5.4.3-70 5.4.3-42 Cold-Leg Walt Differential Temperatures 5.4.3-71 o:\l536wRevi\l536w.non:lt> 090398 xlj RF. VISION l -, ~ , . _ _ .
FINA1. DATA REPORT LIST OF FIGURES (Continued) Figure Title page 5.4.2-43 SG 1 Tube Temperatures 5.4.3-72 5.4.3-44 SG-2 Tube Temperatures 5.4.3-73 5.4.3-45 Upper Support Plate and Bypass Hole DP's 5.4.3-74 5.4.3-46 IRWST Temperatures 5.4.3-75 5.4.3-47 IRWST and Primary Sump Exhaust Steam Flows 5.4.3-76 5.4.3 48 Downcomer Wide-Range Level 5.4.3-77 5.4.3-49 Downcomer Fluid Temperatures at Hot-Leg Elevation 5.4.3-78 5.4.3-50 Downcomer Fluid Temperatures 5.4.3-79 5.4.3-51 Downcomer Fluid Temperatures 5.4.3-80 5.4.3-52 Downcomer Fluid Temperatures 5.4.3-81 5.4.3-53 Downcomer Field Temperatures 5.4.3-82 5.4.3-54 Axial Temperature Profile at Center of Core 5.4.3 83 5.4.3-55 Heater Temperatures at Top of Core 5.4.3 84 5.4.3-56 Narrow-Range and Wide-Range Core Levels 5.4.3-85 5.4.3-57 Narrow Range and Wide-Range Core Levels 5.4.3-86 5.4.3-58 Core Steam Percent 5.4.3-87 5.4.3-59 Core Steam Percent 5.4.3-88 5.4.3-60 Differential Pressure of Upper Support Plate and Bypass Holes 5.4.3-89 5.4.3-61 Upper-Plenum, Upper-Head, Upper-Downcomer Temperatures 5.4.3.90 5.4.3-62 SG-1 Tube Levels 5.4.3-91 5.4.3-63 SG-2 Tube Levels 5.4.3-92 5.4.3-64 SG 1 Channel Head Levels 5.4.3-93 5.4.3-65 SG-2 Channel Head Levels 5.4.3-94 5.4.3-66 SG-1, SG-2 Primary and Secondary Pressures 5.4.3-95 5.4.3-67 CMT-2 and Cold-Leg Pressures 5.4.3-96 5.4.3-68 CMT-2 Levels and Flows 5.4.3-97 5.4.3-69 CMT-1 Long-Rod Thermocouple Temperatures 5.4.3-98 5.4.3-70 CMT-1 Fluid Thermocouple Temperatures 5.4.3-99 5.4.3 71 CMT-1 Inside-Wall Temperatures 5.4.3-100 Matrix Test SB14 5.5.1-1 ADS 1-3 Valve Actuation 5.5.1-47 5.5.1-2 Separator Loop Seal Flows and IRWST Overflow 5.5.1-48 5.5.1-3 Separator Steam Flows 5.5.1-49 5.5.1-4 Reactor and DVI Pressures 5.5.1-50 5.5.1-5 Pressurizer Pressures 5.5.1-51 5.5.1-6 Reactor and Downcomer Anr ulos Wide-Range Levels 5.5.1-52 5.5.1 -7 Reactor Core Levels 5.5.1-53 5.5.1-8 PRHR HX Initial Flow 5.5.1-54 o:\l 536wRev i\l 536w.non:I b-090398 xlij REVISION 1 __j
.- .~ - - . - - . . -. - . - - . - . . . . . - -_ - .~.~. - .. - - - .-. - - - . - .
FmAt. DATA REroni Q LIST OF FIGURES (Continued) D Elants Illis East 5.5.1-9 PRHR HX Flows 5.5.1 55 5.5.1-10 Pressurizer and Surge Line Levels 5.5.1-56
' 5.5.1-11 Accumulator and CMT Iniection Flows 5.5.1-57
, 5.5.1-12 CMT 1 and CMT-2 Levels 5.5.1-58 5.5.1-13 Cold-Leg Levels 5.5.1 59 5.5.1 14 Hot-Leg Levels 5.5.1-60
. 5.5.1-15 -SG 1 Levels 5.5.1-61 5.5.1-16 - SG 1 Channel Head Levels 5.5.1-62 5.5.1-17 SG-1 Fluid Temperatures in Tubes ' 5.5.1-63 i 5.5.1-18 Steam Generator DPs and Levels 5.5.1-64 5.5.1-19 Primary / Secondary Pressure Comparison 5.5.1-65 5.5.1 20 Primary / Secondary Temperature Comparison 5.5.1-66 5.5.1-21 Upper-Head and Downcomer Temperatures 5.5.1-67 5.5.1-22 HL-1 Temperatures 5.5.1-68.
5.5.1-23 HL-2 Temperatures 5.5.1-69 5.5.1-24 Total DVI Flow 5.5.1-70
- 5.5.1 25 Steam Percentage Conditions at SG-2 5.5.1-71 s 5.5.1-26 l Steam Percentage Conditions at SG-2 and HL-2 5.5.1-72 5.5.1 27 Upper-Head DPs - 5.5.1-73 5.5.1-28 Upper Head DPs 5.5.1-74 5.5.1-29 Upper-Plenum and Upper-Head Temperatures 5.5.1-75 J
.5.5.1-30 IRWST and Primary Sump Flows 5.5.1-76 5.5.1-31 Reactor Core Temperatures 5.5.1-77 5.5.1-32 .CL-1 Temperatures 5.5.1-78 5.5.1-33 CL-2 Temperatures 5.5.1-79 5.5.1 CL-3 Temperatures 5.5.1-80 5.5.1-35 CL-4 Temperatures 5.5.1-81 5.5.1 CMT-1 and CMT-2 Levels 5.5.1-82 5.5.1-37 Accumulator and CMT Injection Flows 5.5.1-83 5.5.1-38 Reactor and Downcomer Annulus Wide-Range Levels 5.5.1-84 5.5.1-39 Separator Loop Seal Flows 5.5.1 5.5.1-40 Total DVI Flow 5.5.1-86
;5.5.1-41 Reactor Core Temperatures 5.5.1-87 5.5.1-42 Reactor Core Temperatures 5.5.1-88 5.5.1-43 . Upper Head and Downcomer Temperatures 5.5.1-89 5.5.1-44 = IRWST and Primary Sump Flows 5.5.1-90 5.5.1-45 Reactor Heater Temperatures @ 46 in. from O
sj-Reactor Vessel Bottom (Top of Core) 5.5.1-91 o:\l5hRevl\15W.non:Ib 090398 xljji REVISION 1
FINAL, DATA REPORT LIST OF FIGURES (Continued) Title Eage Figure 5.5.1-46 Reactor Heater Temperatures @ 46 in from Reactor Vessel Bottom (Top of Core) 5.5.1-92 Steam Percentage Conditions in Lower Core Region 5.5.1-93 5.5.1-47 Steam Percentage Conditions in Lower Core Region 5.5.1-94 5.5.1-48 1RWST, Sump, and Break Separator Levels 5.5.1-95 5.5.1-49 IRWST, Sump, and Break Separator Levels 5.5.1-96 5.5.1-50 BAMS Header Steam Flows 5.5.1-97 5.5.1-51 5.5.1-52 IRWST and Primary Sump Steam Flows 5.5.1-98 5.5.1-53 CMT-2 Temperature Profile 5.5.1-99 5.5.1-54 CMT-2 Temperature Profile 5.5.1-100 5.5.1-55 CMT-2 Upper-Dome Temperatures 5.5.1-101 5.5.1-56 CMT-2 Upper-Dome Temperatures 5.5.1-102 5.5.1-57 CMT-2 Fluid Temperature Distribution 5.5.1-103 5.5.1 58 CMT-2 Wall Temperature Distribution 5.5.1-104 5.5.1-59 CMT-2 Upper-Dome Fluid / Wall Interface 5.5.1-105 5.5.1-60 CMT-1 Temperatures 5.5.1-106 5.5.1-61 CMT-1 Temperatures 5.5.1-107 5.5.1-62 Accumulator Levels 5.5.1-108 5.5.1-63 Steam Percentage Conditions for Pressurizer and Surge Line 5.5.1-109 5.5.1-64 Pressurizer Temperature Profile 5.5.1-110 5.5.1 65 ADS 1-3 Separator Discharge Flows 5.5.1-111 5.5.1-66 Pmssurizer and Surge Line Levels 5.5.1-112 5.5.1-67 PRHR HX Temperatures 5.5.1-113 5.5.1-68 PRHR HX Temperatures 5.5.1-114 5.5.1-69 PRHR HX Levels 5.5.1 115 5.5.1-70 PRHR HX Levels 5.5.1-116 5.5.1-71 PRHR HX Flows 5.5.1-117 5.5.1-72 Upper-Head, CL-1, and CMT-2 Temperatures 5.5.1-118 5.5.1-73 Upper-Head, CL-1, and CMT-2 Temperatures 5.5.1-119 5.5.1-74 Upper-Head, CL-1, and CMT-2 Temperatures 5.5.1-120 5.5.1-75 Hot-Leg and Upper-Head Temperature 5.5.1-121 5.5.1-76 Hot-Leg and Upper-Head Temperatures 5.5.1-122 5.5.1-77 Hot-Leg and Upper-Head Temperatures 5.5.1-123 5.5.1-78 Hot-Leg and Upper-Head Temperatures 5.5.1-124 5.5.1-79 Downcomer Annulus Temperatures between Hot-Leg /DVI Elevations 5.5.1-125 5.5.1 80 Cold-Leg Levels 5.5.1-126 5.5.1-81 Cold-Leg Temperatures 5.5.1-127 5.5.1-82 Hot-Leg Levels 5.5.1-128 5.5.1-83 Hot-Leg Temperatures 5.5.1-129 o:\l 536w Rev l\l 536w.non:I b-090398 xliy REVISION 1
FINAL, DATA REPORT
)
i 1 LIST OF FIGURES (Continued)
\
flEllu I.ldt ' f.BER 5.5.1-84 IRWST Temperatures 5.5.1-130 1 5.5.1 85 IRWST Temperatures 5.5.1-131 5.5.1-86 Transition from IRWST to Primary Sump Injection 5.5.1-132 5.5.1 87 Reactor Vessel Response 5.5.1-133 5.5.1-88 Reactor Vessel Response 5.5.1-134 5.5.2-1 ADS 1-3 Valve Actuation 5.5.2-29 l 5.5.2-2 Separator Loop Seal Flows and IRWST Overflow 5.5.2-30 I 5.5.2-3 Separator Steam Flows 5.5.2-31 5.5.2-4 Reactor and DVI Pressures 5.5.2-32 5.5.2-5 Pressurizer Pressures 5.5.2-33 5.5.2-6 Reactor and Downcomer Annulus Wide-Range Levels 5.5.2-34 5.5.2-7 Reactor Core Levels 5.5.2 35 5.5.2-8 PRHR HX Initial Flow 5.5.2-36 5.5.2-9 PRHR HX Flows 5.5.2-37 5.5.2 10 Pressurizer and Surge Line Levels 5.5.2-38 5.5.2 11 Accumulator and CMT Injection Flows 5.5.2-39 O, 5.5.2 12 CMT-1 and CMT-2 Levels 5.5.2-40 5.5.2-13 Cold-Leg Levels 5.5.2-41 5.5.2-14 Hot-Leg Levels 5.5.2-42 5.5.2-15 SG-1 Levels 5.5.2-43 5.5.2-16 SG-1 Channel Head Levels 5.5.2-44 5.5.2-17 SG-1 Fluid Temperatures in Tubes 5.5.2-45 5.5.2-18. Steam Generator DPs and Levels 5.5.2-46 5.5.2-19 Primary / Secondary Pressure Comparison 5.5.2-47 5.5.2-20 Primary / Secondary Temperature Comparison 5.5.2-48 5.5.2-21 Upper-Head and Downcomer Temperatures 5.5.2-49 5.5.2 22 HL-1 Temperatures 5.5.2-50 5.5.2 23 HL-2 Temperatures 5.5.2-51 5.5.2-24 Total DVI Flow 5.5.2 52 . 5.5.2-25 Steam Percentage Conditions at Reactor Vessel and SG-2 5.5.2-53
.5.5.2 26 Steam Percentage Conditions at SG-2 and HL-2 5.5.2 54
;5.5.2-27 Upper Head DPs 5.5.2-55 5.5.2-28 Upper-Head DPs 5.5.2-56 5.5.2-29 Upper-Plenum and Upper Head Temperatures 5.5.2 57 5.5.2-30 IRWST and Primary Sump Flows 5.5.2-58 5.5.2-31 Reactor Core Temperatures 5.5.2-59
'5.5.2-32 CL-1 Temperatures 5.5.2-60 5.5.2-33 CL-2 Temperatures 5.5.2-61 o;\l536wRevl\l536w.non It490398 x]y REVISION 1
FINAL DATA REPORT LIST OF FIGURES (Continued) Title Page Figure CL-3 Temperatures 5.5.2-62 5.5.2-34 CL-4 Temperatures 5.5.2-63 5.5.2-35 CMT-1 and CMT-2 Levels 5.5.2-64 5.5.2-36 Accumulator and CMT Injection Flows 5.5.2-65 5.5.2-37 Reactor and Downcomer Annulus Wide-Range Levels 5.5.2-66 5.5.2 38 Separator Loop Seal Flows 5.5.2-67 5.5.2-39 Total DVI Flow 5.5.2-68 5.5.2-40 Reactor Core Temperatures 5.5.2-69 5.5.2-41 5.5.2-42 Upper-Plenum and Upper -Head Temperatures 5.5.2-70 Upper-Head and Downcomer Temperatures 5.5.2-71 5.5.2-43 IRWST and Primary Sump Flows 5.5.2-72 5.5.2-44 5.5.2-45 Reactor Heater Temperatures @ 46 in. from Reactor Vessel Bottom (Top of Core) 5.5.2-73 5.5.2-46 Reactor Heater Temperatures @ 46 in. from Reactor Vessel Bottom (Top of Core) 5.5.2-74 Steam Percentage Conditions in Lower Core Region 5.5.2-75 5.5.2-47 5.5.2-48 Steam Percentage Conditions in Lower Core Region 5.5.2-76 5.5.2-49 IRWST, Sump, and Break Separator Levels 5.5.2-77 5.5.2-50 IRWST, Sump, and Break Separator Levels 5.5.2-78 5.5.2-51 BAMS Header Steam Flows 5.5.2-79 5.5.2-52 IRWST and Primary Sump Steam Flows 5.5.2-80 5.5.2-53 CMT-2 Temperature Profile 5.5.2-81 5.5.2-54 CMT-2 Temperature Profile 5.5.2-82 5.5.2-55 CMT-2 Upper-Dome Temperatures 5.5.2-83 5.5.2-56 CMT-2 Upper-Dome Temperatures 5.5.2-84 5.5.2-57 CMT-2 Fluid Temperature Distribution (Sheet 1 of 2) 5.5.2-85 5.5.2-57 CMT-2 Fluid Temperature Distribution (Sheet 2 of 2) 5.5.2 86 5.5.2-58 CMT-2 Wall Temperature Distribution 5.5.2-87 5.5.2-59 CMT-2 Upper-Dome Fluid / Wall Interface 5.5.2-88 5.5.2-60 CMT-1 Temperatures 5.5.2-89 5.5.2-61 CMT-1 Temperatures 5.5.2-90 5.5.2-62 Accumulator Levels 5.5.2-91 5.5.2-63 Steam Percentage Conditions for Pressurizer and Surge Line 5.5.2 92 5.5.2-64 Pressurizer Temperature Profile 5.5.2-93 5.5.2-65 ADS 1-3 Separator Discharge Flows 5.5.2-94 Pressurizer and Surge Line Levels 5.5.2-95 5.5.2-66 5.S.2-67 PRHR HX Temperatures 5.5.2-96 PRHR HX Temperatures 5.5.2-97 5.5.2-68 PRHR HX Levels 5.5.2-98 , 5.5.2-69 i Xlvi REVISION 1 o:\l536wRevi\l536w.non:lb490398
~
FINAL DATA RErostr LIST OF FIGURES (Continued) Figure IJdt East 5.5.2-70 PRHR HX Levels 5.5.2-99 5.5.2-71 PRHR HX Flows 5.5.2-100 5.5.2-72 Upper-Head, CL-1, and CMT-2 Temperatures 5.5.2-101 5.5.2-73 ' Upper-Head, CL-1, and CMT-2 Temperatures 5.5.2-102 5.5.2-74 Upper-Head, CL-1, and CMT-2 Temperatures 5.5.2-103 5.5.2-75 Hot-Leg and Upper-Head Temperature (Sheet 1 of 2) 5.5.2-104 5.5.2-75 Hot-Leg and Upper-Head Temperature (Sheet 2 of 2) 5.5.2 105 5.5.2-76 Comparison of Hot-Leg and Upper-Head Temperatures 5.5.2-106 5.5.2-77 Comparison of Hot-Leg and Upper-Head Temperatures 5.5.2-107
. 5.5.2-78 Comparison of Hot-Leg and Upper-Head Temperatures 5.5.2-108 5.5.2-79 Downcomer Annulus Temperatures between Hot-Leg /DVI Elevations 5.5.2-109 5.5.2 80 Cold-Leg Levels 5.5.2-110 5.5.2-81 Cold-Leg Temperatures 5.5.2-111 5.5.2-82 - Hot-Leg Levels 5.5.2-112 5.5.2-83 Hot-Leg Temperatures 5.5.2-113 5.5.2-84 IRWST Temperatures 5.5.2-114 5.5.2-85 IRWST Temperatures 5.5.2-115 O 5.5.2-86 5.5.2-87 Separator Steam Flows Reactor Pressure and Core Level Comparison 5.5.2-116 5.5.2-117
~ 5.5.2-88 Reactor Pressure and Cose Level Comparison 5.5.2-118 5.5.2-89 Reactor Pressure and Core Level Comparison 5.5.2-119 5.5.2-90 Reactor Downcomer Level Comparison 5.5.2-120 5.5.2-91 Reactor Downcomer Level Comparison 5.5.2-121 5.5.2-92 Reactor Downcomer Level Comparison 5.5.2-122.
5.5.2-93 Net Flow for Test SB26 from 0 - 2000 Seconds 5.5.2-123 5.5.2-94 Net Flow for Test SB14 from 0 to 2000 Seconds 5.5.2-124 5.5.2-95 Net Flow and Core Level Comparison for Test SB14 vs SB26 5.5.2-125 5.5.2-% Net Flow and Core Level Comparison for Test SB14 vs SB26 5.5.2-126 5.5.2-97 Net Flow and Core Level Comparison for Test SB14 vs SB26 5.5.2-127 5.5.2-98 Reactor Pressure, CMT-1, and ACC-1 Flow Comparison 5.5.2-128 5.5.2 Reactor Pressure, CMT-2, and ACC-2 Flow C irrparison .5.5.2-129 5.5.2-100 Reactor Vessel Response 5.5.2-130 5.5.2-101 Reactor Vessel Response 5.5.2-131 5.5.2-102- Reactor Pmssure and IRWST Overflow Comparison 5.5.2-132 5.5.2-103 Reactor Pressure and IRWST Overflow Comparison 5.5.2-133 5.5.2 104 Reactor Pressure and IRWST Injection Flow Comparison 5.5.2-134 5.5.2-105 Reactor Pressure and IRWST Injection Flow Comparison 5.5.2-135 C 5.5.2-106 Reactor Pressure and IRWST Injection Flow Comparison 5.5.2-136 1 - 5.5.2-107 Reactor Pressure and PRHR HX Flow Comparison 5.5.2-137 o:u5hRevnl5W.non:lt@90398 xivii REVISION 1
FINAL DATA REPORT LIST OF FIGURES (Continued) Figure Title Page Matrix Test SB31 5.6-1 Reactor Core Temperatures 5.6-17 5.6-2 Reactor and DVI Pressures 5.6-18 5.6-3 Accumulator and CMT Injection Flows 5.6-19 5.6-4 CMT-1 Temperatures 5.6-20 5.6-5 CMT-2 Temperatures 5.6-21 5.6-6 PRHR HX Flows 5.6-22 5.6-7 PRHR HX Temperatures 5.6-23 5.6-8 IRWST Temperatures 5.6-24 5.6-9 Core Levels 5.6-25 5.6-10 Upper-Plenum Levels 5.6-26 5.6-11 Reactor Heater Temperatures @ 46 in. from Reactor Vessel Bottom 5.6-27 5.6-12 Core Heater Temperatures @ 46 in. from Reactor Vessel Bottom 5.6-28 5.6-13 Reactor and Downcomer Annulus Wide-Range Levels 5.6-29 5.6-14 Cold-Leg Levels 5.6-30 5.6-15 Hot-Leg Levels 5.6-31 5.6-16 SG-1 Levels 5.6-32 5.6-17 SG-1 Channel Head Levels 5.6-33 5.6-18 SG-2 Tube Levels 5.6-34 5.6-19 SG-2 Channel Head Levels 5.6-35 5.6-20 CMT-1 and CMT-2 Levels 5.6-36 5.6-21 Accumulator and CMT Pressures 5.6-37 5.6-22 Accumulator Levels 5.6-38 5.6-23 Pressurizer and Surge Line Levels 5.6-39 5.6-24 Upper-Plenum and Upper-Head Temperatures 5.6-40 5.6-25 Upper Head and Downcomer Temperatures 5.6-41 5.6-26 CL-1 Temperatures 5.6-42 5.6-27 CL-2 Temperatures 5.6-43 5.6-28 CL-3 Temperatures 5.6-44 5.6-29 CL-4 Temperatures 5.6-45 5.6-30 HL-1 Temperatures 5.6-46 5.6-31 HL-2 Temperatures 5.6-47 5.6-32 SG 1 Fluid Temperatures in Tubes 5.6-48 5.6-33 SG-2 Fluid Temperatures in Tubes 5.6-49 5.6-34 Primary and Secondary Pressures 5.6-50 e o:\l536wRevluS36w.non:Ib-090398 xjyjij REVISION 1
. . ~ , . - - - - ._, -.-.-.- -. ... - - .. _ . .-.-. -. .-. - _ _ .- - -
i
- i. FmAL DATA REPORT
/ LIST OF FIGURES (Continued)
( 1 , EsEt li9se f.att t i Matrix Test SB15 4 5.7-l a Primary Loop and Break Pipe Arrangement (Sh. I of 2) 5.7-31 ! j 5.7-lb Primary Loop and Break Pipe Arrangement, Side View (Sh. 2 of 2) 5.7-32 l 5.7-2 Accumulator and CMT Injection Flows 5.7-33 l 5.7-2x Accumulator and CMT Injection Flows 5.7-34 5.7-3: Total DVI Flow' 5.7-35 5.7-3x Total DVI Flow 5.7-36 5.7-4 PRHR HX Flows 5.7-37 5.7-4x -PRHR HX Flows 5.7-38 5.7 Separator Flows 5.7-39 5.7-5x Separator Flows 5.7-40 5.7-6 IRWST Overflow 5.7-41 5.7-7 Core Levels 5.7-42 5.7-8 Separator Steam Flows - 5.7-43 5.7-9 IRWST and Primary Sump Steam Flows 5.7-44 5.7-10 BAMS Header Steam Flows 5.7-45 , p 5.7-11 Upper-Head DPs 5.7-46 I
.5.7-11x Upper-Head DPs 5.7-47 5.7-12 Cold-Leg Line DPs 5.7-48 5.7-12x Cold-Leg Line DPs 5.7-49 5.7-13 Break DPs 5.7-50 5.7-14 Balance Line DPs 5.7-51 5.7-14x Balance Line DPs 5.7-52 5.7-15 Reactor and Pressurizer Heater Power 5.7-53 5.7-15x Reactor and Pressurizer Heater Power 5.7-54 5.7-16 Reactor and Downcomer Annulus Wide-Range Levels 5.7-55 5.7-16x Reactor and Downcomer Annulus Wide-Range Levels 5.7-56 5.7-17 Upper-Plenum Levels 5.7-57 5.7-17x Upper Plenum Levels 5.7-58 5.7-18' Cold-Leg Levels 5.7-59 5.7-18x Cold-Leg Levels 5.7-60 5.7-19 Hot-Leg Levels 5.7-61 5.7-19x Hot-Leg Levels 5.7-62 5.7-20 SG-1 Levels 5.7-63 5.7-21 SG-1 Channel Head Levels 5.7-64 5.7-21x ' SG-1 Channel Head Levels 5.7-65 5.7-22 SG-2 Tube Levels' 5.7-66
- (^\
\
5.7-23 5.7-23x SG-2 Channel Head Levels 5.7-67 SG-2 Channel Head Levels 5.7-68 o:\l 536wRev i\l 536w. mon:l b-090398 xdx REVISION 1 4
FINAL DATA REroRT LIST OF FIGURES (Continued) Title Page, Figure 5.7-69 5.7-24 Accumulator Levels 5.7-70 5.7 ,3 CMT-1 and CMT-2 Levels 5.7-71 5.7-25x CMT-1 and CMT-2 Levels Pressurizer and Surge Line Levels 5.7-72 5.7-26 Pressurizer and Surge Line Levels 5.7-73 5.7-26x 5.7-74 5.7 27 Separator Levels 5.7-75 5.7-27x Separator Levels IRWST, Sump, and Break Separator Levels 5.7-76 5.7-28 IRWST, Sump, and Break Separator Levels 5.7-77 5.7-28x 5.7-78 5.7-29 PRHR HX Levels 5.7-79 5.7-29x PRHR HX Levels Hot-Leg and Cold-Leg Pressures 5.7-80 5.7-30 5.7-81 5.7-31 Reactor and DVI Pressures 5.7-82 5.7-32 Break Pressure 5.7-83 5.7-33 Accumulator and CMT Pressures 5.7-84 5.7-34 Pressurizer Pressures 5.7-85 5.7-35 Upper-Head and Downcomer Temperatures Upper Head and Downcomer Temperatures 5.7 86 5.7-35 x 5.7-87 5.7-36 CL-1 Temperatures 5.7-88 5.7-36x CL-1 Temperatures 5.7-89 5.7-37 CL-2 Temperatures CL-2 Temperatures 5.7-90 5.7-37x 5.7-91 5.7-38 CL-3 Temperatures CL 3 Temperatures 5.7-92 5.7-38x CL-4 Temperatures 5.7-93 5.7-39 5.7-94 5.7-39x CL-4 Temperatures HL-1 Temperatures 5.7-95 5.7-40 HL-1 Temperatures 5.7-96 5.7-40x HL-2 Temperatures 5.7-97 5.7-41 HL-2 Temperatures 5.7-98 5.7-41 x Downcomer Annulus Temperatures at 180 degrees Azimuth 5.7-99 5.7-42 Downcomer Annulus Temperatures at 180 degrees Azimuth 5.7-100 5.7-42x Downcomer Annulus Temperatures at 0 degrees Azimuth 5.7-101 5.7-43 Downcomer Annulus Temperatures at 0 degrees Azimuth 5.7-102 5.7-43x SG-1 Fluid Temperatures in Tubes 5.7-103 5.7-44 SG-1 Fluid Temperatures in Tubes 5.7-104 5.'7-44x SG-2 Fluid Temperatures in Tubes 5.7-105 5.7-45 SG-2 Fluid Temperatures in Tubes 5.7-106 5.7-45x REVISION 1 c:\1536wRev i\l $36w.non:l b-090398 ] e . _ .
.._ . -._.__.____.m._._-_ --.- . _ - _ . _ . _ _ _ _ _ . _
FmAt. DATA REPORT i /"3 LIST OF FIGURES (Continued) U naus uns ran
-5.7-46 CMT-1 Temperatures 5.7-107 5.7-46x CMT-1 Temperatures 5.7-108 5.7-47 CMT-2 Temperatures 5.7-109 5.7-47x ~ CMT-2 Temperatures
' 5.7-110 _5.7-48 IRWST Temperatures 5.7-111 ] 5.7-48x IRWST Temperatures ' 5.7-112 5.7 PRHR HX Temperatures 5.7-113 5.7-49x PRHR HX Temperatures 5.7-114 5.7-50 Reactor Core Temperatures 5.7-115 5.7-50x ' Reactor Core Temperatures
- 5.7-116 !
5.7-51 Upper-Plenum and Upper-Head Temperatures 5.7-117 5.7-51x Upper-Plenum and Upper-Head Temperatures 5.7-118 i 5.7-52 SG-1 and SG-2 Primary Side DPs 5.7-119 l 5.7-53 Primary and Secondary Pressures 5.7-120 { 5.7-53x Primary and Secondary Pressures 5.7-121 1 5.7-54 IRWST and Primary Sump Flows 5.7-122 _- 5.7-55 Reactor Heater Temperatures @ 46 in. from Reactor Vessel Bottom 5.7-123
- s ). 5.7-55x - Reactor Heater Temperatures @ 46 in, from Reactor Vessel Bottom 5.7-124 1 5.7-56 IRWST Flow Rates 5.7-125
'6.1-1 Comparison of Break Flow 6.1-15 6.1-2 - Comparison of Break Flow 6.1-16 6.1-3 . Comparison of Primary Sump Weight 6.1-17 l 6.1-4 Comparison of Primary Sump Weight 6.1-18 I 6.1-5 Comparison of CMT-1 Injection Flows 6.1-19 6.1 Comparison of CMT-1 Injection Flows 6.1-20
' 6.1-7 Comparison of CMT-2 Injection Flows 6.1-21 !
6.1-8 Comparison of CMT-2 Injection Flows 6.1-22 6.1-9 Comparison of CMT-1 and CMT-2 Flows 6.1-23 6.1-10 Comparison of CMT-1 and CMT-2 Flows 6.1-24 6.1-11 Comparison of CMT-1 and CMT-2 Flows 6.1-25
/ 6,1-12 Comparison of CMT-1 and CMT-2 Flows 6.1-26 6.1-13 CMT Levels vs Primary Sump Weight 6.1-27 6.1-14 CMT Levels vs Primary Sump Weight 6.1-28 6.1 15 CMT Levels vs Primary Sump Weight 6.1-29 6.1-16 '- - CMT Levels vs Primary Sump Weight 6.1-30 6.1-17 Comparison of Downcomer Level 6.1-31 /~' '6.1-18 Comparison of Downcomer Level V) 6.1-19 Comparison of Downcomer Level 6.1-32 6.1-33 c:\l$hRevi\l5h.n=lb 090398 li ' REVISION 1
FINAL DATA REPORT i LIST OF FIGURES (Continued) Figure Title f.ag I 6.1-20 Comparison of Downcomer Level 6.1-34 6.1-21 Comparison of Break Flow 6.1-35 6.1-22 Comparison of Lower Core Levels 6.1-36 6.1-23 Comparison of Lower Core Levels 6.1-37 6.1-24 Comparison of Upper Core Levels 6.1-38 6.1-25 Comparison of Upper Core Levels 6.1-39 7.2-1 CMT-2 Through-the-Wall Temperatures - 20 Percent Volume (14 in.) 7.2-9 7.2-2 CMT-1 Through-the-Wall Temperatures - 20 Percent Volume (14 in.) 7.2-10 7.2-3 CMT-2 Through-the-Wall Temperatures - 50 Percent Volume (29 in.) 7.2-11 7.2-4 CMT-1 Through-the-Wall Temperatures - 50 Percent Volume (29 in.) 7.2-12 7.2-5 CMT-2 Through-the-Wall Temperatures - 75 Percent Volume (41 in.) 7.2-13 7.2-6 CMT-1 Through-the-Wall Temperatures - 75 Percent Volume (41 in.) 7.2-14 7.2-7 CMT-2 Temperature-Fluid TCs at the Same Elevation 7.2-15 7.2-8 CMT-1 Temperature-Fluid TCs at the Same Elevation 7.2-16 1 7.2-9 CMT-2 Temperature-Fluid TCs at the Same Elevation 7.2-17 7.2-10 CMT-1 Temperature-Fluid TCs at the Same Elevation 7.2-18 7.2-11 CMT-2 Comparison of Fluid Temperatures at the Same Elevation 7.2-19 7.2-12 CMT-1 Comparison of Fluid Temperatures at the Same Elevation 7.2-20 l 7.2-13 CMT-2 Comparison of Level Versus Pressure 7.2-21 l 7.2-14 CMT-1 Comparison of Level Versus Pressure 7.2-22 7.2-15 CMT-2 Through-the-Wall Temperatures - 20 Percent Volume (14 in.) 7.2-23 7.2-16 CMT-1 Through-the-Wall Temperatures - 20 Percent Volume (14 in.) 7.2-24 7.2-17 CMT-2 Through-the-Wall Temperatures - 50 Percent Volume (29 in.) 7.2-25 1 7.2-18 CMT-1 Through-the-Wall Temperatures - 50 Percent Volume (29 in.) 7.2-26 7.2-19 CMT-2 Through-the-Wall Temperatures - 75 Percent Volume (41 in.) 7.2-27 7.2-20 CMT-1 Through-the-Wall Temperatures - 75 Percent Volume (41 in.) 7.2-28 7.2-21 CMT-2 Temperature-Fluid TCs at the Same Elevation 7.2-29 7.2-22 CMT-1 Temperature-Fluid TCs at the Same Elevation 7.2-30 7.2-23 CMT-2 Temperature-Fluid TCs at the Same Elevation 7.2-31 7.2-24 CMT-1 Temperature-Fluid TCs at the Same Elevation 7.2-32 7.2-25 CMT-2 Comparison of Fluid Temperatures at the Same Elevation 7.2-33 7.2-26 CMT-1 Comparison of Fluid Temperatures at the Same Elevation 7.2-34 7.2-27 CMT-2 Comparison of Level Versus Pressure 7.2-35 7.2-28 CMT-1 Comparison of Level Versus Pressure 7.2-36 7.2-29 CMT-2 Through-the-Wall Temperatures - 20 Percent Volume (14 in.) 7.2-37 7.2-30 CMT-1 Through-the-Wall Temperatures - 20 Percent Volume (14 in.) 7.2-38 7.2-31 CMT-2 Through-the-Wall Temperatures - 50 Percent Volume (29 in.) 7.2-39 7.2-32 CMT-1 Through-the-Wall Temperatures - 50 Percent Volume (29 in.) 7.2-40 o:\l536wRevN536w.non:Ib-090398 ]ji REVISION 1
FINrt. DATA REPORT l 1 1 N 0 LIST OF FIGURES (Continued) , Finure Title ! Eage 7.2-33 CMT-2 Through-the-Wall Temperatures - 75 Percent Volume (41 in.) 7.2-41 7.2-34 CMT-1 Through-the-Wall Temperatures - 75 Percent Volume (41 in.) 7.2-42 7.2-35 { CMT-2 Temperature-Fluid TCs at the Same Elevation 7.2-43 1 7.2-36 CMT-1 Temperature-Fluid TCs at the Same Elevation 7.2-44 7.2-37 CMT-2 Temperature-Fluid TCs at the Same Elevation 7.2-45 7.2-38 CMT-1 Temperature-Fluid TCs at the Same Elevation 7.2-46 i 7.2-39 CMT-2 Comparison of Fluid Temperatures at the Same Elevation 7.2-47
- 7.2-40 CMT-1 Comparison of Fluid Temperatures at the Same Elevation 7.2-48 l 7.2-41 CMT-2 Comparison of Level Versus Pressure 7.2-49 7.2-42 CMT-1 Comparison of Level Versus Pressure 7.2 50 7.2-43 CMT-2 Through-the-Wall Temperatures - 20 Percent Volume (14 in.) 7.2-51 l 7.2-44 CMT-1 Through-the-Wall Temperatures - 20 Percent Volume (14 in.) 7.2-52
) 7.2-45 CMT-2 Through-the-Wall Temperatures - 50 Percent Volume (29 in.) 7.2 53 I 7.2-46 CMT-1 Through-the-Wall Temperatures - 50 Percent Volume (29 in.) 7.2-54 7.2-47 CMT-2 "Ihrough-the-Wall Temperatures - 75 Percent Volume (41 in.) 7.2-55 7.2-48 CMT-2 Through-the-Wall Temperatures - 75 Percent Volume (41 in.) 7.2-56 7.2-49 CMT-2 Temperature-Fluid TCs at the Same Elevation 7.2-57 7.2-50 CMT-1 Temperature-Fluid TCs at the Same Elevation 7.2-58 7.2-51 CMT-2 Temperature-Fluid TCs at the Same Elevation 7.2-59 7.2-52 CMT-1 Temperature-Fluid TCs at the Same Elevation 7.2-60 7.2 53 CMT-2 Comparison of Fluid Temperatures at the Same Elevation 7.2-61 7.2-54 CMT-1 Comparison of Fluid Temperatures at the Same Elevation 7.2-62 7.2-55 CMT-2 Comparison of Level Versus Pressure 7.2-63 7.2-56 CMT-1 Comparison of Level Versus Pressure 7.2-64 l l l l l l l l C/ 0:\l5%wRevl\l5%w.non:lb490398 ]jji REVISION I
FINAL DATA REPORT ACRONYMS ACC accumulator ADS automatic depressurization system APEX Advanced Plant Experiment test facility at OSU ASME American Society of Mechanical Engineers BAMS break and ADS measurement system CCT condensate collection tank CD ROM compact disk read-only memory CMT core makeup tank CRP condensate return pump CVS chemical and volume control system DAS data acquisition system DEG double-ended guillotine DP differential pressure DVI direct vessel injection FMM magnetic flow meter GSM general scaling methodology H2TS hierarchical two-tiered scaling analysis HPS heated phase switch HX heat exchanger 1RWST in-containment refueling water storage tank LAN local area network LBLOCA large-break loss-of-coolant accident LCS lower containment sump LDP level LOCA loss-of-coolant accident LRGMS large main steam MSS main steam system NSS nonsafety systems OSHA Occupational Safety and Health Administration OSU Oregon State University PC- personal computer PCS passive containment cooling system PIRT phenomena identification ranking table PPIRT plausible phenomena identification ranking table PQP project quality plan PRHR passive residual heat removal 17I' pressure transducer PWR pressurized water reactor PXS passive core cooling system PZR pressurizer RCP reactor coolant pump RCS reactor cooling system RNS normal residual heat removal system RV reactor vessel SASM severe accident scaling methodology SBLOCA small-break loss-of-coolant accident SCR silicon-controlled rectifier SG steam generator SGS steam generator system VI virtual instrumentation o:\l536wRevl\l5%w.non:Ib-o90398 liy REVISION 1
FINAL DATA REPO%T p
SUMMARY
b Westinghouse Electric Corporation and the Nuclear Engineering Department of Oregon State University (OSU) have designed and constructed a 1/4-scale model of the AP600 plant that includes the reactor coolant system (RCS), steam gr.:nerators (SGs), passive core cooling system (PXS), automatic depressurization system (ADS), and nonsafety injection systems-such as partial normal residual heat removal system (RNS) and partial chemical and volume control system (CVS)-- in the Radiation Center at OSU in Corvallis, Oregon. This facility was used to perform tests of the AP600 passive safety systems in a reduced size and at lower temperatuits and pressures for validation of the safety analysis codes. The test facility, fabricated completely from austenitic stainless steel, was designed for normal operation at 450 F and 400 psig, and was scaled using the hierarchical, two-tiered scaling analysis (H2TS) method developed by the U.S. Nuclear Regulatory Commission (NRC).* Simulated piping breaks were tested in the hot leg, cold leg, pressure balance line between the cold leg and the core makeup tank (CMT), and the direct vessel injection (DVI) line. Decay heat that scales to 3 percent of the full power (about 2 minutes after shutdown) was supplied by electrically heated rods in the reactor vessel. Simulated accidents were programmed by the control system to proceed automatically. About 850 data channels were recorded by the data acquisition system (DAS) for each test and downloaded to compact disks for subsequent data reduction and plotting. Data from the test facility were transmitted to the Weninghouse Energy Center for reduction and review. To the extent that instrument indications provided an understanding of the system response to a transient, the system response was defm' ed. Final analysis of system behavior will be part of AP600 Low-Pressure Integral Systems Test at Oregon State University, Test Analysis Report, WCAP-14292.* Conclusions from the test program are discussed with each test in Section 5 of this report. O U oA1536wRevi\l536w.non:Ib-o81298 1 REVISION 1
FINAL DATA REPCCT ACKNOWLEDGMENTS l ne authors express their appreciation for the extensive discussions and inputs obtained from the key designer of the test facility, Mr. L. K. Lau, and the developer of the scaling analyses, Dr. J. N. Reyes. The Westinghouse engineers, Mike Carter, Jerry Schlaman, Ralph Ferrell, and team leader Carl Dumsday, who performed these tests, and their very patient families deserve special recognition. Each of them lived "on-the-road" during the final phases of construction, facility check-out, pre-operational test, and formal test program. With the cooperation of the Oregon State University (OSU) test facility operators, the facility was in check-out or testing 24 hours a day,7 days a week for 11 months. Sue Fanto, Tim Andreychek, Larry Hochreiter, and Mike Roidt provided invaluable consultation in developing an understanding of the facility performance characteristics and served as final reviewers of the report. The word-processing team members, led by Denise Kephart, deserve special thanks not only for their skill but for their cooperation, long hours, and enthusiasm. O O o.\l536mRevl\l536w.non:lt481298 2 REVISION l g
_ , _ . _ _ _ _ _ _ _ . _ . . . . _ _ . _ _ _ . . . ~ . _ . _ _ . _ , _ . . _ _ . _ _ . _ _ _ - l FmAL DATA REPORT I
1.0 INTRODUCTION
%e low-pressure integral systems test facility at Oregon State Usersity (OSU) is a scale model of the AP600 reactor used to evaluate the thermal-hydraulic characteristics of the AP600 during various accident conditions. Westinghouse performed a matrix of tests at OSU, as part of an integrated test plan (described in AP600 Test and Analysis Plan for Design Certification, WCAP-14141*), to provide data to verify the capability of the analytical methods used to predict the response of the AP600 integrated passive safety systems. Experiments were designed to simulate small-break loss-of-coolant accidents (SBLOCAs), the greatest challenge to the passive safety systems. The purpose of this report is to provide final verification of test results by comparing instrument indications to plausible phenomena expected or demonstrated by the test facility and to document the data results of each test used for validation of the safety codes applied to the AP600 plant.
Requirements for the test program are detailed in, Long-Term Cooling Test Specification, ! WCAP-13234.* The test program consisted of a series of pre-operational tests performed to l characterize the plant in both cold and hot conditions, followed by a series of tests performed to supply test data during the facility's response to a break in the primary system. He pre-operational tests confirmed system volumes, line resistances, and heat losses through the pressure boundary. In j addition, a complete check-out was performed on the facility's instrumentation, associated instrument i channels, and the data acquisition system. His check-out included testing of all control interlocks, l O V trips, and alarms. In the matdx tests, breaks were simulated in the primary cold leg, direct vessel injection (DVI) line, and core makeup tank-1 (CMT-1) balance line. Additional tests were performed to evaluate break size, containment backpressure, nonsafety systems, and break location. The j reference loss-of-coolant accident (LOCA), a 2-in, break in a cold leg, was performed twice to show test repeatability. An inadvertent automatic depressurization system (ADS) actuation and an l inadvertent safety systems actuation (S) signal leading to ADS actuation were also evaluated. Testing was performed at OSU from June through September,1994, with a combined test staff of l OSU and Westinghouse Electric Corporation engineers. Tests were run according to specific l procedures developed by the Westinghouse Safety and Licensing Group and the Test Engineering
- Group at the Energy Center in Pittsburgh, Pennsylvania, in consultation with the Onsite Test Group.
Test results were recorded in the data acquisition system (DAS) and were subject to several reviews. The initial review took place at the OSU test facility to determine if the test met the acceptance criteria or had to be repeated. The Day-of-Test Report was issued by the Onsite Test Group to the Test Engineering Group at the Energy Center documenting the initial data review and delineating the acceptability of the test. Additional reviews of the data were conducted at the Energy Center, in an expedited fashion, to make the data available to the Nuclear Regulatory Commission (NRC) so that it , could evaluate the test results. The data were released in the preliminary form of Quick Look Reports ! for each test. d o:\l5hRevnl536w.non lb-081298 - }.1 REVISION 1 l
O O-O
FINAi, DATA REroRT l A 1.1 Background U l The low-pressure,1/4-height integral systems tests were conducted at the Corvallis campus of OSU. Scaling studies indicated that a scaled, low-pressure test facility could be constructed to capture the thermal-hydraulic phenomena of interest for the lower pressure characteristics of the AP600. The OSU test facility is a new facility that was constructed specifically to investigate the AP600 passive system characteristics. The facility design accurately models the detail of the AP600 geometry, including the primary system, pipe routings, and layout for the passive safety systems. The primary system consists of one hot leg (HL) and two cold legs (CLs) with two active pumps and steam generator (SG) for each of the two loops. There are two CMTs. each connected to a cold leg of _ one primary loop. The pressurizer is connected to the other primary loop, as in the AP600 plant design Gas-driven accumulators are connected to the DVI lines. The discharge lines from a CMT l and one of the two in-containment refueling water storage tank (IRWST) and reactor sump lines are connected to each DVI line. The two independent lines of each stage of ADS 1,2, and 3 are modeled by one line containing an orifice. The two-phase flow from the ADS 1-3 is separated in a swirl-vane t separator, and the liquid and vapor flows are measured to obtain the total flow rate. The separated flow streams are then recombined and discharged into the IRWST through a sparger. 'Ihus, the mass and energy flows from the ADS into the IRWST are preserved. The time period for simulation included not only IRWST injections, but also IRWST draining and sump injection in order to simulate the long-term cooling mode of the AP600. The time scale for the ! OSU test facility is approximately one-half; that is, the sequence of events occurred approximately twice as fast as it would in the AP600. To model the long-term cooling aspects of the transients, the two-phase flow from the break was separated in a swirl-vane separator, and the liquid and vapor portions of the total flow were measured. l The liquid fraction of the flow was discharged to the reactor sump, as in the AP600 plant. The vapor ! was discharged to the atmosphere, and the equivalent liquid flow was capable of being added to the IRWST and sump to simulate the condensate return from the passive containment. A similar approach was also used for the two ADS-4 valves on the hot legs. The two-phase flow was separated in a swirl-vane separator, the two stream flows were measured, the liquid phase was discharged into the reactor sump while the vapor phase was discharged to the atmosphere, and the liquid equivalent was capable of being added to the IRWST and sump. The IRWST, reactor sump, and separators could be pressurized to simulate the containment pressurization following a postulated LOCA. A multi-tube passive residual heat removal heat exchanger (PRHR HX) was located in the IRWST. The HX used the same C-tube design as the AP600 and two tubes were instrumented with l thermocouples and flow meters to obtain wall heat fluxes during the tests. Data from fluid ! thermocouples, wall thermocouples, and level transmitters were used to determine HX performance. l The IRWST had vertical rods containing fluid thermocouples located at various radial and axial 3 o:\l5%wRevl\l5%w.non:lb-081298 ],].] REVISION I l' - . . .. - - , ._ .
FINAL DATA REPORT locations to determine the degree of mixing in the tank and assess the temperature of the liquid in the IRWST as it was delivered to the reactor vessel. The reactor vessel for the OSU tests included a 3-ft. heated core simulator consisting of 481-in. diameter heater rods. The heater rods had a top-skewed power shape. There were wall thermocouples swaged inside some of the heater rods to measure the heater rod temperature and five fluid l thermocouple rods in the heater rod bundle to measure the axial and radial coolant temperature distribution. The scaled flow area in the core and the flow area in the vessel upper plenum wele preserved. There were simulated reactor internals in the upper plenum to preserve the flow area and the scaled coolant volume. The reactor vessel included an annular downcomer, into which the four cold legs and the two DVI lines were connected. The two hot legs penetrated the reactor annulus and connected with the loops. The AP600 reactor vessel neutron reflector was simulated using a ceramic liner to reduce the metal heat release to the coolant. The OSU test facility had approximately 600 kW of electrical power available. This corresponds to about 3 percent decay heat. The OSU experiments examined the passive safety system response for the SBLOCA and the large-break loss-of-coolant accident (LBLOCA) transitions into long-term cooling. A range of SBLOCAs was simulated at different locations on the primary system-such as the cold-leg, hot-leg, CMT-1 balance line, and DVI line. The break orientation, at either the top or bottom of the cold leg, was also examined. Selected tests continued into the long-term cooling, post-accident mode during which the passive safety injection was from the reactor sump and the IRWST. A larger-break, post-accident, long-term cooling situation was also simulated. O 0:\l 536w Rev I\l 536w.non: 1 b-081298 ],1 2 REVISION 1
i l FmAL DATA REPORT l l l 1.2 Pre-Operational Test Objectives l Pre-operational tests were performed to provide an understanding of facility control and operating characteristics, to confirm design features essential to scaling, and to ensure that the instruments and l DAS were performing as expected. Formal tests wem conducted while the plant was in cold I l conditions to measure system pressure drops and volume of the components. Pressure drops in the test l facility were adjasted to their desired scale AP600 values by using orifice plates, when required. Pre-operational testing was performed in the hot condition to characterize system heat losses. Results l of the pre-operational tests are presented in Section 4. I 1.2.1 Cold Pre-Operational Tests i Volume determination tests were performed in January and February,1994 for the accumulators, CMTs, pressurizer, IRWST, sumps, SG secondary sides, and reactor vessel to compare the actual volumes with the calculated volumes. Volume determination was made by filling the vessels with water and then measuring the weight of the water in the vessel. The weights were determined by draining the water into a container and then placing the container on a precision scale, or by using calibrated load cells under the larger tanks. Flow test /line resistance determination tests were performed in Febmary and September,1994. The l O objective of the first set of tests was to measure line resistance far the accumulator lines, IRWST lines, and sump injection lines for a given flow rate. The pressure drop in CMT injection and ADS 1-3 lines was measured over a range of flows. Resistance of the normal residual heat removal system (RNS) injection lines was measured to ensure that they were within 10 percent of each other. The reactor coolant pump head was measured for full flow and pump coastdown conditions. Flow tests were repeated in September to obtain additional information on the pressure drop around the primary l circuit and PRHR HX. l l 1.2.2 Hot Pre-Operational Tests 1 Three separate and formal hot functional tests were performed. The objectives of the tests were to ensure proper operation of the equipment prior to the formal matrix test program and to provide data necessary to document temperature characteristics of the system. The first hot functional test (OSU-HS01) measured the steady-state heat loss, natural-circulation flow, and forced-flow characteristics. These data were used to verify the AP600 thermal-hydraulic computer codes (the test ! results are repoited in Subsection 4.3). The objective of the second test (OSU-HS-02) was to verify the measuring capability of the break and ADS measurement system (BAMS) and the control of the i ADS. The third hot functional test (OSU-HS03), inadvertent ADS-1 Actuation, was a rehearsal for the formal matrix test program. (Data from OSU-HS02 and HS03 are not used in AP600 safety analysis computer code validation and are not included in this report.)
%d l
l l oA1536wRevnl536w.non:lb-o81298 1.2-1 REVISION 1 l
FINAL DATA REPO2T OSU HS01 was performed to determine surface heat losses from the system at 100*,200*,300*,and 400*F; characterize PRHR under natural circulation and forced cooling; characterize the primary cooling system state at 100 kW,300 kW,500 kW, and 600 kW; and characterize the CMT natural convection characteristics. O O c:\l 536w Rev l\1536w.non: I b-081298 1.2-2 Pav:SION 1
_m . _ _ _ _ _. . _ . __- . - _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ - . _ ~ . _ _ _ _ _ . _ _ FMAL DATA REPORT j l q 1.3 Matrix Test Objectives ne objective of the Matrix Test Program was to obtain test data for a series of AP600 simulated transients, varying the test parameters on a one-at-a-time basis, and to overlap the test program at SPES-2, a full-height, full-pressure test facility. Tests at OSU were specifically designed to evaluate i the passive safety system response to a range of SBLOCAs, during the transition into long-term cooling. He reference LOCA, Matrix Test SB01, is a 2-in. cold-leg break with a simulated failure of one of the two ADS-4 valves in one ADS-4 line. De effect of break location was determined by also i testing a break in the DVI line (Matrix Test SB12), a break in the CMT-1 balance line (Matrix Test SB10), and a break in hot leg-2 (Matrix Test SB15). The impact of a larger break size was evaluated
- comparing the results of SB01 with SB21, a 4-in. cold-leg break test case. Matrix Tests SB05 and SB23 were performed to demonstrate the impact of smaller break sizes. !
l De reference 2-in cold-leg LOCA, Matrix Test SB01, was duplicated in Matrix Test SB18, the last matrix test performed to demonstrate test facility repeatability over a period of time. Matrix j Test SB19 repeated SB01 with the objective of showing the effect of high containment backpressure. Possible interactions of the nonsafety systems with the passive safety systems were addressed by l performing Matrix Test SB04, a repeat of Matrix Test SB01, with the nonsafety systems running. The l ! impact of a smaller break size was evaluated in Matrix Test SB24. ([] v Additional tests were performed to evaluate the effects of a CMT/ cold leg balance line breaker of a smaller size (Matrix Test SB09) and a DVI break of a smaller size (Matrix Test SB13). A DVI l double-ended break with additional failures was evaluated in Matrix Test SB28. 1 Several tests were performed to show the effect of transients not involving breaks. Matrix Test SB14 ^ tested the impact of an inadvertent ADS actuation, Matrix Test SB26 evaluated multiple ADS failures, t and Matrix Test SB31 tested an inadvertent S signal scenario. I A summary of the matrix tests used in the AP600 thermal-hydraulic code validation is provided in ! Table 1.3-1. De pertinent facility conditions for each test are also noted. There were additional ! matrix tests performed and omitted from this report because of their similarity to other tests. Those j test results were issued in Quick Look Reports. i l a } a. { i oA1536wRevlM536w.uon:It>481298 1.3-1 REVISION 1
U-G" TABLE 1.3-1 2 OSU MATRIX TEST SU51NIARY 0 a F Break l Test No. Sire and Location PRIIR IIX CVS Pump RNS Pump ADS 4-1 (111 1) ADS 4-2 (III 2) Comments SB01 2-in. Cle3 On Off Off 50 percent flow 100 percent flow Failure of one of two lines in Q area in AP600 area in AP600 ADS 4-1; reference cold-leg break a bottom of cold leg case (Ch1T side) SB04 2-in Cl 3 On On On 50 percent flow 100 percent flow Same as SB01 except safety and bottom of cold leg area in AP600 area in AP600 nonsafety system interaction (Ch1T side) 1-in. Cl 3 On Off Off 50 percent flow 100 percent flow Same as SB01 except break size SB05 bottom of cold leg area in AP600 area in AP600 change (CNIT side) On Off Off 50 percent flow 100 percent flow Same as SB01 except different P SBfG 2-in. Cle3 to Ch1T-1 area in AP600 Area in AP600 break location; asymmetric [ balance line behavior of Ch1Ts SBIO DEG Cl 3 to Ch1T-1 On Off Off 50 percent flow 100 percent flow Limiting break on balance line; area in AP600 area in AP60;; asymmetric behavior of ChtTs; balance line failure of one of two lines in ADS 4-1 DEG DVI-l On Off Off 100 percent flow 100 percent flow Limiting break on DVI line; SB12 line break area in area in AP600 Failure of one of two lines of AP600 ADS-1 and ADS-3 2-in. DVI-I On Off Off 50 percent flow 100 percent Dow Same as SB01 except different SBl3 line break area in AP600 area in AP600 break location SBl4 Inadvertent ADS On Off Off 50 percent flow 100 percent flow No-break case with one failure of y area in AP600 area in AP600 two lines in ADS 4-1 2 (no break) SBIS 2-in. III 2 On Off Off 50 percent flow 100 percent flow Same as SB01 except break ( e bottom of pipe area in AP600 area in AP600 location ,:=!! E 3 4 O O O
.) b V
i o 1 s i f { TABLE 1.3-1 (Continued) i y OSU MATRIX TEST
SUMMARY
.E Break l
Test N Siae and Location PRHR HX CVS RNS ADS 4-1 ADS 4-2 Posep Pump (HI 1) (HI 2) Comunents ' h
- SB18 2-in. Cl 3 On Off Off 50 percent flow 100 percent flow Repeat test of SB01; confirm 3 bottom of cold leg area in AP600 area in AP600 behavior of system and (CMT side) instrumentation SB19 2-in. Cl 3 On Off Off 50 percent flow 100 percent flow Same as SB01 except containment bottom of cold leg area in AP600 area in AP600 backpressure simulated '
(CMT side) SB21 4-in. top of and 4-in. On Off Off 50 percent flow 100 percent flow Same as SB01 except larger break bottom of Cl 3 area in AP600 area in AP600 size; largest break size simulated in (CMT side) matrix tests i f SB23 1/2-in. Cl 3 On Off Off 50 percent flow 100 percent flow Same as SB01 except smaller break l [ bottom of cold leg area in AP600 area in AP600 size
& (CMT side) :
SB24 1/2-in. Cl 3 On On On 50 percent flow 100 percent flow Safety and nonsafety system : bottom of cold leg area in AP600 area in AP600 interaction; single failure l
' (CMT side) ,
SB26 Inadvertent ADS with Off Off Off N/A N/A No-break PRA case with ADS-1 i multiple isolate isolate isolated and failure of one of two l failures (no break) this line this line lines of ADS 4-1; PRHR HX [ isolated SB28 DEG DVI-I Off Off Off N/A N/A Limiting break on DVI line weh [ line break isolate isolate ADS 4-1, ADS 4-2,~ accumulator-1, ; this line this line and PRHX HX isolated. [ SB31 Spurious On Off Off 50 percent flow 100 percent flow Failure of one of two lines in [ S signal area in AP600 area in AP600 ADS 4-1 21 I (no break) $ e- i e i A h DEG - double-ended guillotine , i PRA - probabilistic risk assessment m 3 - 4 l i I
.)
FINAL DATA Rzroar 2.0 TEST FACILITY DESCRIPTION 1he Otegon State University (OSU) test facility is designed for operation at 400 psig and 450*F, in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Pressure Vessel Codes, Section VIII, Pressure Piping B31 (ANSI /ASME B31.1)*; Occupational Health and Safety Administration (OSHA) Standards; and Oregon State Fire Protection Codes. In this section, the overall test facility and each component are described. O
-O c:\l5hRevl\l5h-imn:lt461298 2-1 REVISION 1
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l FINAL DATA REPORT 2.1 Overall Facility Description The OSU test facility is a scaled model of the AP600 reactor coolant system (RCS), steam generator system (SGS), passive core cooling system (PXS), automatic depressurization system (ADS), lower containment sump (LCS), chemical and volume control system (CVS), and normal residual heat removal system (RNS). In addition, the facility is capable of simulating the AP600 passive i containment cooling system (PCS) condensate return process. Figures 2.1-1 through 2.1-4 are photographs of the completed facility. Figure 2.1-5 is an isometric drawing of the test facility, and Figure 2.1-6 is a simplified flow diagram of the test facility. Appendix G contains detailed piping and instmmentation diagrams (P& ids) for the test facility and its systems. The facility accurately reflects the AP600 geometry, including the piping routings. All components and piping are fabricated from austenitic stainless steel. Flanged, gasketed connectors are used throughout the test facility. The relative locations of all tanks and vessels-such as the in-containment refueling water storage tank (IRWST), core makeup tanks (CMTs), and accumulators-are properly modeled both horizontally and vertically. The facility uses a unique break and ADS measurement system (BAMS) to measure two-phase break and ADS flow. j l i 2.1.1 Reactor Coolant System The RCS is composed of a reactor vessel, which has electrically heated rods to simulate the decay heat in the reactor core, and two primary loops. Each primary loop consists of two cold-leg pipes and one hot-leg pipe connecting a steam generator (SG) to the reactor vessel. A reactor coolant pump (RCP) on each cold leg takes suction from the SG channel head (downstream of the SG U-tubes) and discharges it into the downcomer region of the reactor vessel. A pressurizer with an electric heater is connected to one of the two hot legs through surge-line piping. The surge line is formed to a spiral - geometry to accommodate piping expansion loads at the pressurizer during thermal transients. The top of the pressurizer is connected to the ADS 1-3 line. An ADS-4 line is connected to each hot leg. The reactor vessel also contains two direct vessel injection (DVI) nozzles that connect to the DVI lines of the passive core cooling system (PXS). A flow venturi is incorporated in each DVI nozzle to limit the loss of inventory from the reactor vessel in the event of a double-ended DVI line break. Each hot leg provides a connection to one of the two ADS-4 lines. Detailed descriptions of RCS components are discussed in Subsection 2.3. 2.1.2 Steam Generator System This test models the primary side of the SGS with two SGs, one per primary loop. A simulated feedwater line is used for each SG to maintain proper SG secondary water level. The steam produced in each SG is measured and exhausted to the atmosphere through a common diffuser and stack. Proper AP600 SGS operations during loss-of-coolant accident (LOCA) transients are simulated. In the AP600, a drop in primary water level produces a safety systems actuation (S) signal, which shuts the oil 5%wRevl\l5%w.l.non.lb.081298 2.1-1 REVISION 1
FiNAs. DATA REPORT feedwater supply and maintains SG secondary side pressure at its proper value. De test control logic simula:es the response of the AP600 by providing an S signal at a fixed time following a break. Consequentiy, the proper initial conditions for the transient behavior of the facility are provided. A detailed description of the SGS is provided in Subsection 2.3. 2.1.3 Passive Core Cooling System I The PXS consists of two CMTs, two accumulators, one IRWST, and one passive residual heat removal hert exchanger (PRHR HX). The test facility simulates the AP600 IRWST with a cylindrical tank with praperly scaled water volume and height. He IRWST is located above the reactor core; two injection. !!res connect to the two DVI lines -one per DVI line. Each IRWST injection line also communicates with the sump tank with interconnecting piping and isolation valves. Two additional conditiens are simulated in the test model: venting of the IRWST to the containment and water overflow:ng the IRWST tu the sump. Two CMrs and rwo accumulators are used in the PXS. One CMT and one accumulator are connected to one DV! Ik.e; the other CMT and accumulator are connected to the second DVI line. Both DVI lines enter the downcomer of the reactor vessel at the DVI nozzles. The PRHR HX is located inside l the IRWST, using IRWST water as the cold reservoir. The inlet of the PRHR HX is connected to the t ' pressurizer side hot leg, via a tee at the ADS-4 line, and the outlet is connected to the SG channel head at the cold-leg side. Since the inlet is hot and the outlet is cold, water is circulated through this system by natural convection. The water volume and elevation of each CMT are properly scaled and modeled. They are elevated above the reactor vessel and the DVI lines. A line connecting the top of each CMT to its cold leg provides pressure balance between the RCS and the CMT. Therefore, the CMT injects cooling water by its own elevation head. The accumulators are also modeled with proper volume and height. However, they are pre-pressurized and, therefore, inject when RCS pressure is below the preselected accumulator pressure. The PXS and the ADS provide adequate reactor core cooling for the complete range of LOCAs. In the event of a LOCA, the CMTs inject ambient water to the reactor vessel when the injection isolation valve opens. The accumulators also inject water when RCS pressure drops below the pre-set accumulator pressure. As the CMT water level drops, the ADS-1 through ADS-4 isolation valves open sequentially to depressurize the RCS. The opening of the ADS-4 valves reduces the RCS pressure to equal containment pressure; hence, the IRWST injects by its own elevation head. Finally, the sump injects water to the reactor vessel when enough elevation head is established. Accurate, direct measurement of the two-phase flows vented from the system (simulated breaks, ADS Dows) are difficult and expensive. A unique system, the BAMS, was specifically designed to measure these two-phase Dows. The BAMS is based on separating the two-phase Dows into individual single-phase Dow streams that can be accurately measured with conventional instrumentation. O\ l REVISION 1 c:\l 536wRev i\l 536w. l .non: l b.081298 2.1-2
FINAI. DATA REPORT 2.1.4 Automatic Depressurization System he AP600 uses four stages of valves to depressurize the RCS. De first three stages of the ADS are provided through connections to the pressurizer, ne three stages are arranged in parallel with each stage containing two lines and each line containing an isolation valve and control valve. The fourth stage of ADS contains four separate lines. Two lines have a common connection to HL-1, and two lines have a common connection to HL-2. Similar to ADS 1-3, each line contains an isolation valve and control valve. l The OSU test facility uses only one line of valves to model the ADS 1-3 stages of AP600. This is done using removable flow nozzles to match the scaled flow characteristics of either one or two lines i of valves. De first , second , and third-stage lines of the ADS split into parallel lines from one connection off the pressurizer. Each line includes a pneumatically operated, full port ball valve and a flow nozzle. He ball valve simulates the isolation valve in the AP600, and the flow nozzle simulates the flow control valve in the AP600. Two sets of flow nozzles are used to simulate sin;;le- or double-line operation, allowing proper flow area scaling. He discharge lines from the ADS 13 valves are joined into one line that is connected to the ADS 1-3 separator. Dese valves are opened by the test logic controller. Once the ADS valves are opened by the test logic controller, RCS pressure drops and the flow flashes into two-phase flow. This two-phase D flow is separated using a swirl-vane separator, and the liquid and vapor flows are measured to obtain U the ADS total flow for mass and energy balance analysis. The separated flow streams are then recombined and discharged into the IRWST through a sparger. Thus, the mass flow and energy flow from ADS 1-3 into the IRWST are preserved. . He OSU test facility uses one ADS-4 line connected to the top of each hot leg. Each line contains a pneumatically operated, full port ball valve acting as the ADS-4 isolation valve and a flow nozzle simulating the flow area in the AP600. Two sets of flow nozzles are used in the test: one simulates 100-percent flow area, and the other simulates 50-percent flow area. For those tests which require a complete failure of the ADS-4 lines on one hot leg, the ADS-4 line is closed. In the AP6'JO plant,
. when the ADS-4 isolation valves open, the flow is directed inside the containment. In the test, the ADS-4 discharge flows to the ADS-4 separators, where the steam and liquid flows are separated. The steam flow is measured and exhausted to the atmosphere. The liquid flow is measured and directed to the primary sump tank. This two-phase flow measuring scheme is part of the BAMS, which is discussed in more detail in Subsection 2.1.7.
2.1.5 Lower Containment Sump The LCS in the AP600 consists of two volumes-normally flooded and normally nonflooded. De normally flooded volume consists of those compartments which collect liquid break flow and ADS
\ flow. For example, the compartments that house the reactor vessel or the SGs are normally flooded.
The normally nonflooded volume includes those compartments which do not collect any liquid flow. ouswaeviusw.i.non:ib.os 29s 2.1-3 REVISION 1
FINAL DATA REPORT l i The only communication path between the normally flooded volume and normally r;onlooded volume is at the top of these compartments, which is called the curb. In the test, a cylindrical tank (pnmary 9l sump tank) is used to model the normally flooded volume. The normally nonflooded volume is modeled with another cylindrical tank, identified as the secondary sump tank. These two tanks are connected with a line at a level simulating the curb level in the AP600. He primary sump tank is designed t.o properly scaled water volume and height. 1; includes sump injection lines that inject water into the DVI lines. These injection lines also cemmunicate with the IRWST injection lines at the properly scaled elevation and locations. He overflow from the IRWST is also collected in this tank, simulating the overflow path in the AP600. The secondary sump is also designed to properly scaled water volume and height. It is connected to the primary sump tank by a short length of 6-in. Sch. 40 pipe. His pipe is very short in order to minimize flow resistance, since the flooded and nonflooded AP600 containment volumes are only separated by the curb. The pipe also has a flange joint with a weir in between. The height of this weir models the curb level in the AP600. Detailed descriptions of LCS components are included in Subsection 2.3. 2.1.6 Normal Residual Heat Removal System and Chemical and Volume Control System In the AP600, the RNS is used to provide nonsafety cooling water injection to the reactor core. In this case, the RNS pump takes suction from the IRWST and discharges it into the DVI lines. *he delivered flow rate is a function of RCS pressure. This process and its time-dependent flow are modeled in the test. During testing, the RNS pump takes suction from the IRWST at the properly scaled location and elevation, and it discharges the flow to both DVI lines at properly scaled locations. These two lines are balanced so that equal flow can be delivered to each DVI line. The pressure-dependent flow in the AP600 is also modeled and automatically controlled by the proponional, integral, derivative (PID) controller. Subsection 2.3.17 provides more details of the RNS components. The makeup line in the AP600 CVS is modeled in the test. This line contains a pump taking suction from the feed ste sge tank and discharging to the SG-2 (presstirizer side) channel head at the cold-leg l side. The makeup flow is scaled from the AP600 makeup flow rate as a function of RCS pressure and is controlled automatically by the process controller. 2.1.7 Break and A3S Measurement System The mass and energy of the test facility, both on individual components and the overall system, must be maintained in order to properly scale the long-term cooling phenomenon in the AP600. To do this, the flow rate at various locations and equipment must be known. For those locations where single-phase flow exists, the flow rate measurement is relatively simple and reliable. However, there are some locations and equipment in the test facility with two-phase flows. Since direct measurement l of two-phase flow is not practical and is extremely expensive, an indirect method is used-the BAMS. oA1536wRevlM36w-l.non:lb-081298 2.1-4 REVISION 1
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w FINA1, DATA REPORT The BAMS is uniquely designed for the test facility to indirectly measure two-phase flow and energy. This system uses separators to separate the two-phase flow into single-phase liquid and single-phase steam flows for direct flow rate and temperature measurements. This system also measures all break flows; that is, LOCA operations and inadvertent ADS operations. The BAMS consists of steam-liquid separators and the interconnecting pipes and valves to the various break sources, the phmary sump tank, the ADS 1-3 lines, and the main steam header. 2.1.7.1 ADS 13 Separator and Pipe Route One separator is dedicated as the ADS 1-3 moisture separator; it has one inlet and two outlets. Two-phase flow (steam and water) from the ADS 1-3 lines enters the ADS 1-3 separator, where the , r. team is separated from the mixture. The steam flows out of one outlet while the liquid drains down the other. These two lines recombine at some distance downstream and discharge into the IRWST via ! the sparger located inside the IRWST. To prevent the steam from blowing through the liquid drain, a
- liquid loop seal is incorporated to the liquid drain line of the separator. Also, the steam and liquid lines are carefully sized so that at full flow, the pressure drop from the steam outlet to the recombined common point is smaller than that from the liquid drain outlet to the same recombined point. This ensures that the steam outlet pressure is less than the liquid drain line outlet pressure; hence, steam can only exit the dedicated steam line, where it is measured by a vortex flow meter and fluid thermocouples.
t' The ADS 1-3 separator, the steam line, the liquid line, and the recombined line are all insulated to I minimize heat loss to the atmosphere. Furthermore, both the separator tank and the steam line are heat-traced to maintain a temperature of approximately 200 F to minimize nonprototypical steam condensation. Consequently, prototypical quality is preserved. The liquid loop seal is prefilled with hot water at approximately 180*F prior to actual testing. All these features ensure negligible energy flow to the atmosphere and proper energy transfer directly to the IRWST, as in the AP600. 2.1.7.2 ADS-4 Separators and Pipe Route Two ADS-4 separators are used-one for each ADS-4 line. The separator connecting to the ADS-4 line from hot leg-1 (IIL-1 on the CMT side) is the ADS 41 separator. The separator connecting to the ADS-4 line from HL-2 (pressurizer side) is the ADS 4-2 separator. The ADS 4-2 separator is sized to perform two functions-it serves to separate two-phase ADS-4 flow and separate break flow for certain cases. The ADS 4-1 separator is designed to separate two-phase ADS-4 flow only, and it can handle 150 percent of normal ADS-4 flow.
' Each ADS-4 seprator separates the two-phase mixture into single-phase steam and single-phase liquid for flow rate, pressure, and temperature measurements. The steam exits the top outlet nozzle while the s liquid drains at the bottom outlet. A loop seal is used in the liquid drain line to prevent steam blowdown through the liquid line. The steam line connects to a common steam header, and the liquid oA15hRevluS36w-I.non:Ib-081298 '2,1 5 REVISION 1
l'INAL DATA REPORT line connects to the primary sump tank. These connections simulate the ADS-4 operation process in the AP600, where the steam flow rises to the containment wall and liquid drains to the sump. The ADS-4 separators, the liquid lines, and steam lines are all insulated to minimize condensation and heat loss. Also, the steam lines and separators are all heat-traced to maintain a temperature of approximately 200*F, and the liquid line loop seal is prefilled with hot water of approximately 180*F prior to actual testing. 9 2.1.7.3 Break Separator and Pipe Route The following break simulations are tested:
- Small break at bottom of DVI line
- Double-ended break at DVI line
- Small break at bottom of cold-leg /CMT balance line
- Double-ended break at cold-leg /CMT balance line
- Small break at bottom of cold leg
- Small break at top of cold leg
- Smalli ak at bottom of hot leg The break separator is designed to receive two-phase break flow from the break source and separate steam and liquid for single-phase flow pressure and temperature measurement. The separator and steam lines are heat-traced, and the loop seal is prefilled with hot water to minimize heat loss and nonprototypical condensation.
In the event of a double-ended CMT-1 balance line break simulation, the break separator receives break flow from the cold-leg side of the break. The ADS 4-2 separator receiws break flow from the CMT side of the break, and the ADS 4-1 separator receives all ADS-4 flow. When a double-ended DVI line break is simulated, the break separator receives the flow from the reactor vessel side of the break. The sump receives the break flow from the DVI line. 2.1.8 Orlfices and Nozzles Ori' ices or flow nozzles are used in critical lines to scale the line resistances in the OSU test facility to j l the AP600. The bases for these devices are discussed in this section, and details of the orifice and l nozzle design and placement can be found in Reference 7. I 2.1.8.1 ADS 1-3 Flow Nozzles < Proper scaling of the flow through the ADS 1-3 lines requires that the quality of the flow be considered in sizing the nozzles. Since the mass flow through the nozzle depends on the flow quality, 2.1 6 REVISION 1 o:\l5%wRevl\l5%w l.non:lb-081238
.)
. - . . . - - . . - . . _ - . - ~ -- - _ - . . - . . _ - . - . - . . . - . - - - - -
l FINAt. DATA REPC'rT O two venturi-type flow nozzles are used. Sharp-edged orifice plates are not suitable for this application ! because their discharge coefficients vary with the Reynolds number, leading to wide ranges in mass l flow as the quality changes. The flow nozzles were sized as follows: l l One set of flow nozzles for single-line simulation with small-break LOCA (SBLOCA) One set of flow nozzles for double-line simulation with SBLOCA One set of the nozzles for single-line simulation with large-break LOCA (LBLOCA) j One set of flow nozzles for double-line simulation with LBLOCA 2.1.8.2 ADS-4 Flow Nozzles l l l Two types of flow nozzles are used in the test. The first type simulates a 50-percent flow area of one AP600 ADS-4 stage, and the second type simulates a 100-percent flow area of one AP600 ADS-4 stage. ' Fluid similarity is the scaling basis, j The scaling analysis requires that each line be properly scaled to have proper fluid similarity. For all scaling criteria, the pressure drop scaling ratio is an important criterion. The pressure drop ratio is [ ]'6 Any line that does not meet this requirement must be fitted with an orifice plate to bring the pressure drop ratio to [ ] 2.1.8.3 Direct Vessel Injection Venturi
- N 1 The DVI venturi scaling criteria are the same as break hole (venturi) scaling criteria. This is because
{ the DVI venturi is used to rers the break flow out of the reactor vessel. The geometry of the DVI venturi is the same as the AP600, and the throat diameter and the IJD ratio are properly scaled. j I l i l a .- o i l
- o:\l5hRevi\l5h l.non:Ib.081298 2.1-7 REVISION I Ii .. ; ,_
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FINAL DATA REPORT 2
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l FINAL DATA REPORT l lO x y x 77 V"pn. p* h [ s
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FmAr, DATA REPORT j 2.2 Facility Scaling U Re design basis of the OSU test facility is the scaling analysis of the AP600, presented in AP600 l low Pressure Integral Systems Test at Oregon State University, Facility Scaling Repon, WCAP-14270.M i The scaling analysis guides the design of the test facility so that 'he thermal-hydraulic performance of the test facility properly simulates the phenomena imponant to the passive safety features of the AP600. The test facility is designed with reduced dimensions and lower tengratures and pressures than the actual reactor. He scaling methodology and the overall results are provihi in this section. l 2.2.1 Methodology To ensure that the scaling objectives were met in an organized and traceable manner, a general scaling methodology (GSM) for the OSU test facility was developed. The model development methodology can be found in Facility Scaling Report.M A flow diagram describing the GSM is presented in Figure 2.2-1. The first step is to specify the experimental objectives. The experimental objectives define the types of tests to be performed in order to respond to specific licensing and design needs. He experimental objectives determine the general modes of operation to be simulated in the test facility, j The second step is to develop plausible phenomena identification ranking tables (PPIRTs). The nature (V~T of scaling prohibits exact similitude between the AP600 and the test facility operating conditions. As a result, the design and operation of the test facility are based on simulating the processes most important to passive safety system performance and long-term cooling. The function of the PPIRTs is to identify the key thermal-hydraulic phenomena to be scaled in the context of LOCA transients. Many of the phenomena of importance to AP600 LOCA behavior were already identified by existing phenomena identification ranking tables (PIRTs). However, some of the AP600 modes of operation
- were not verified. Therefore, additional thermal-hydraulic phenomena of importance were identified j j and included in the PPIRTs. Hence, PPIRTs were used rather than PIRTs. i l
I
- l. De third step is to perform a scaling analysis for each of the modes of operation specified by the l- experimental objectives and further defined by the PPIRTs. The hierarchical, two-tiered scaling I analysis (H2TS) developed by the U.S. NRCW was selected for the scaling analysis of this facility, j
Detailed discussion of the application of this method for the OSU test facility is also provided in l Facility Scaling Report.@ \
. The fourth step is to use the scaling analysis results to develop a set of characteristic time ratios 4 (di.nensionless x groups) and similarity criteria for each mode of operation. Because it is impossible to identically satisfy all of the similarity criteria simultaneously, the set included only those criteria which had to be satisfied in order to scale the most imponant phenomena identified by the PPIRT.
O ! v o%p60(A15Mw.la.non:ll41298 2.2-1 REVISION 1 F
FINAL DATA REPORT Step five is an evaluation of the scaling criteria to determine if the scale model geometry, initial conditions, or operating conditions would introduce significant scaling distonions. Distortions an: also evaluated relative to other modes of operation. 2.2.2 Facility Scaling Parameters The height scaling ratio was set at 1:4, and the diameter scaling ratio at 1:6.93. These ratios were based on the objective of minimizing power requirements while maximizing height and maintaining sufficient system volume to properly model loop pressure drop and three-dimensional flow in the downcomer, core, and plenum regions. The important factors that were considered in determining the height scaling ratio were: l
- Minimum diameter ratio was satisfied so that skin friction pressure drop requirements could be met easily with commercially available pipe and drawn tubing. ;
1
- Diameter choice was consistent with two-phase scaling and flow regime transitions.
1
- Fluid volume requirements were reasonable (e.g., IRWST volume ~3000 gal). i
= Power requirements were reasonable (-600 kW).
- Time scale made long-term cooling test duration reasonable (Tg = 0.5).
- UD ratio indicated that multi-dimensional flow effects would scale well under fluid propeny similitude (UD = 1.73).
- Elevation was sufficient such that differential pressure measurements between hot and cold legs were well within instrument capability.
- Construction and material costs were economical.
The scaling ratio for the piping was selected to ensure that the frictional losses due to piping roughness did not bias the buoyancy effects that determine natural convection rates. Analysis of the buoyancy-friction balance equation resulted in a minimum diameter ratio of [ ]a.b(3) l Therefore, a diameter ratio of approximately [ Ja.b was selected ecause it was greater than the minimum and could be obtained with commercial pipe sizes. Also, thermal effects of this size piping (that is, heat losses and heat storage effects) could be modeled. Once the length and diameter scaling ratios were determined, the dimensions of the test facility could be geometrically scaled. Table 2.2-1 summarizes the scaling ratios for the test facility. o:\l 536* Rev i\l 536w. l a.non: I b-082698 2.2-2 REVISION l
FINA1. DATA RtronT l l 2.2.3 Mass / Energy Balances i ( Since accurate measurements of the components of a two-phase flow mixture are difficult and the equipment to do these measurements is very expensive, two-phase flow was separated into its single phases and the individual single-phase flows were measured with conventional instmmentation. Therefore, all two-phase streams vented from the RCS are measured using conventional vapor-liquid separation devices. Where required to simulate AP600 systems, the resultant single-phase flows are recombined before being returned to the system. In other cases, the steam is vented and hot water from an auxiliary storage tank is injected to match the mass of the steam that would have condensed had the steam been released into the AP600 containment. It should be noted that the heat transport processes in the containment are not modeled in this test facility; however, the condensate return process is modeled. This approach permits accurate measurement of the two-phase flows released frem the RCS with simple, relatively inexpensive components. Thermal-hydraulic similitude is maintained either by recombining the single-phase streams or by makeup of hot water for vented steam that would have been condensed. ( o:\l 536w Rev l\l 536w. l a.non: I b-081798 2.2-3 REVISION 1
FINAL DATA REroar i TABLE 2.21
SUMMARY
OF SYSTEM SCALING RESULTS FOR THE 1/4. LENGTH SCALE MODEL PRIMARY LOOP Geometry Length scaling ratio 1:4 System diameter scaling ratio 1:6.93 Area scaling ratio 1:48 Volume scaling ratio 1:192 Flow Velocity scaling ratio 1:2 Mass flow rate scaling ratio 1:% Residence Time Time scaling ratio 1:2 Power Power scaling ratio 1:96 Power density scaling ratio 2:1 Model Power Requirements Percent of total power 5% 4% 3%* 2% 97.00 77.60 58.20 38.80 AP600 decay power (MW) 1009.58 807.66 605.75 403.83 Model power (kW) Note:
- Nominal power used to simulate the decay heat in the model was 600 kW.
O ! 2.2-4 REVISION 1 o:\l536wRevi\l536w la.non:lb-081798 g
FINAt. DATA REPORT i i O (1) Specify Experimental Objectives
,r (2) Develop PPIRTS
( ir i i Perform Scaling Analysis for j (3) Operational Modes 1 through N l l
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Operational Mode .___ Operational Mode . _. Operational Mode
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- (4) Develop n Groups (4) l (4) and Similarity *--
Criteria l l ir ir l \ (5) (5) Determine (5) Significant Scaling ------ Distortions l No (6) Develop (6) (6) System Design i Specifications i 4 l (7) Test Facility Design
- Specifications and O/A :
, , , , , Critical Attributes Figure 2.2-1 General Scaling Methodology
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FINAI. DATA REPORT i l l l p 2.3 Facility Component Description L The system components are described in this section. l 2.3.1 Reactor Vessel The reactor vessel is a right-circular cylinder with a flanged hemispherical upper head and a flanged flat bottom. It models the following irgions:
- Lower plenum
. Core
.- Upper plenum e Upper head The lower plenum region is the region below the lower core plate of the reactor core. De net water volume in the AP600 lower region is modeled in the test. Since the reactor vessel inner diameter (I.D.) is determined by the proper scaling of upper-plenum volume, core region volume,' and special downcomer region scaling criteria, the lower-plenum region cross-sectional area is slightly distorted on the scaling basis.
[ 2.3.1.1 Vessel Design / Dimensions ( All components of the reactor vessel are fabricated of Type 304 stainless steel. The reactor vessel, designed to Section VIII of the ASME Pmssure Vessel Code,W consists of a [ ]C diameter, right-l circular cylinder made by rolling and welding [ ]C thick stainless steel plate, with a flanged j l hemispherical top and a flanged flat plate bottom. Both the bottom and top flanges are [
]C flanges, which are machined to fit the outside diameter (O.D.) of the shell and are seal- l
~~ welded. To failitate assembly, the reactor vessel is fabricated in two sections, a [ ;
i ~ ]C lower section. Rese two sections are joined by welded [ ]C slip- 1 on flanges, machined to fit the O.D. of the shell. De vessel is fitted with four cold-leg flanged nozzles (300 lb), two hot-leg flanged nozzles (300 lb), and DVI flanged nozzles (300 lb). De bottom flange carries an endplate through which the heater tubes penetrate for connection to the electric power o source. De bottom flange is fitted with a 1/2-in. thick Teflonm insert to minimize the effect of the heat capacity of this flange. A 1/8-in. thick stainless steel plate retains the Teflonm insert in this ;
- flange.
2.3.1.2 ' Reactor Vessel Instrumentation L hirty-three Type K thermocouples (1/4 in. diameter, Type 304 stainless steel sheath) art, installed in !- ' the reactor vessel to measure both wall and fluid temperature. Twenty-one differential pressure
; transmitters are connected to the various fluid volumes (upper plenum, lower plenum, downcomer, i
L L l: o:Wism.1b.noititrosi29s 2.3-1 REVISION 1 l
FINAL DATA REPOIT etc.) to measure liquid levels, and eleven differential pressure transmitters measure pressure gradients. Specific instmment identifications and locations are shown in Appendix G, the P&lDs. : 2.3.2 Rod Bundle ne rod bundle fumishes heat to the primary coolant in the reactor vessel to simulate the decay heat released by the AP600 following a plant trip or plant shutdown. The power scaling ratio for the rod bundle is 1:96.1, and the power density scaling ratio is 2:1. Since the maximum electrical power available at the test facility is 700 kW, the rod bundle at this power and with the power scaling factor of 1:96.1 simulates the AP600 decay heat at [ )* of full core power. The AP600 decay power at 3.7 percent occurs about 70 seconds after the reactor is shut down; therefore, the heat input from the core during this short period is not fully modeled. During the tests, most of the reactor transients were performed with the heater power at 600 kW, which is equivalent to a decay heat power of about 3 percent of full power. He rod bundle heaters were programmed to follow the calculated decay heat power starting at 600 kW. The heater rod bundle controller was programmed to follow one of the two possible decay heat power input algorithms defined below, depending on the test configuration:
- For 0 $ time < 140 seconds l
Power (kW) = 600 kW For tim > 140 seconds 6% Power (kW) = ... [1 + (C.0021 ' Time - 140)))a2848
- For 0 s time < 300 seconds Power (kW) = 600 kW For time > 300 seconds 600 Power (kW) =
[1 + (0.01021 (Time - 300))]a2848 2.3.2.1 Rod Bundle /lleater Description The rod bundle consists of 48 Type 304 stainless steel clad heaters, each with a maximum power of 15 kW. Overall rod bundle characteristics are shown in Table 2.3-1. The 1-in, diameter heater rods are 60-in. in overall length, with a 36-in. heated length and a wall thickness of 0.061-in. Power is skewed to provide higher heat fluxes at the top of the core by means of the resistance heater coil winding density. The power produced in each 6-in. incremental length of each heater rod (beginning at the bottom; that is the end closest to the electrical and thermocouple leads) is shown in the following: REVISION 1 o:\ap600\l5%w-lb.non:Ib-081298 2.3-2
FINAL DATA REPORT I Heated Length Power Heater Wire Size Heater Length Section (in.) (W) AWG No. (in.) 1 6 623 31 55 1/2 2 6 1758 28 39 1/4 l 3 6 2697 27 32-1/2 4 6 3373 26 32-1/2 5 6 3644 26 30-1/2 6 6 2905 26 37-3/4 Total 36 15,000 A section of the heater with a reduced diameter of 0.5-in. passes through the lower flange and is sealed by a Swagelok fitting. The following materials are used in the heated section with their available pertinent physical properties: 1 Thermal Conductivity Specific Heat Electrical Insulation (Blu/'F in.-ft.2) (Btu /lb *F) /~'Nl 80% boron nitride 5 95 0.306 20% magnesium oxide Resistance wire - - 80% Ni 20% Cr (ASTM B-344) Conductor - - 10 Ga nickel (ASTM B-160/E-39) 2.3.2.2 Rod Bundle Instrumentation Type K thermocouples are installed in two groups of the heater rods; one group of heater rods does not have thermocouples. These groups and the location of the thermocouples are shown in the following: Location of Thermocouples (Inches from Lower End of Heater Group No. of Heaters Heated Zone) H-A 32 None N H-B 10 15,21,27,33 H-C 6 3,9,21,27 oMp600\l536w Ib.nou:lb-081298 2.3-3 REVISION 1
i l FINAL DATA REPORT l l The leads for these thermocouples exit the rod at the lower end, together with the electrical power leads, through the Swagelok pressure seal. These thermocouples are provided as overtemperature protectors and are also recorded by the DAS. I Core coolant temperatures am measured by five thermocouple rods, each carrying at least three type K thermocouples. 233 Reactor Internals l The reactor internals have several functions related to core suppon and fluid flow. Specifically, the l j following components of the reactor internals provide these functions: l
- Core barrel -- Separates core flow from downcomer
. L ower core plate -- Supports the fuel rods and distributes the flow l
= Upper core plate - Retains fuel rods in the axial position and provides support for the upper in :mals
- Upper support plate -- Supports upper head components i
= Upper intemals -- Provide guide tubes for the insertion of the i'i-core instrumentation and l control rods
- = Downcomer -- Provides a flow path for cold-leg fluid from the cold-leg nozzles to the core inlet plenum 233.1 Reactor Internals Design / Description Core Barrel The core barrel is composed of two sections of [ ]C Sch. 30, type 304 stainless steel pipe [
]* flanged together at the center. The top section is [ ]C and the bottom section is [ ]* long. The top barrel plate is a ring [ ]C O.D., welded to the top of the core barrel. This ring is .cealed to the I.D. of the reactor vessel by two radial 0-rings installed on O.D.s, thus permittia; .intical expansion while limiting leakage. This plate includes 101/4-in. tapped holes that can be blocked with screws to adjust the leakage flow for the downcomer volume to the upper plenum.
During cold testing, it was found that all of these holes were needed to obtain the desired [ ]C bypass flow. The loss coefficients (K) measured for one tapped hole were [ Ja.b (upward flow) and [ Ja.b (downward flow). o:\ap600(1536w-lb.non:lt>481298 2.3-4 REVISION 1 ( l
. . _= ._ . _ _ . _ _ _ . _ _ _ _ . _ _ _ _ . . _ _ . _ _ _ _ _ _ _ _ _ _ _
i , FINAI., DATA REronT De bottom of the lower core barrel is welded to the lower core suppon plate. He lower core suppon plate, which is { ] thick, is drilled with [ ] and [ ] diameter holes for the heater rods. His plate is supponed from the reactor vessel on four pads, 90 degrees apan, on Belleville washers retained by a plunger that passes through the lower core suppon plate and the support pads.
- De core barrel is sized to maintain the total core volume scaling ratio at 1:192 and the length ratio of 1:4.
t Reflectors De reflectors are simulated by cast ceramic inside a cylindrical shell that fits inside the barrel. A stainless steel liner is welded to the shell to provide the stepped cruciform flow volume for the simulated core. He reflectors are fabricated in two sections, separated horizontally at the simulated core midplane to facilitate assembly. He reflector sections are secured by four tie rods that pass
- through the two reflector sections and connect the upper reflector plate with the lower core suppon plate.
Grid Rinn ( A gridded support for the heater rods consists of 0.109-in. thick by 0.19-in. high bars welded to form an egg crate. Four bolts are used to attach the grid ring to the core barrel at the joint between the upper and lower core barrel. Unoer Internals
. De upper intemals are simulated by 41 guide tubes, each consisting of a rod [ ]
in diameter with a 1 in. long threaded lower end. De threaded section screws into the upper core suppon plate. He guide tube top consists of a [
]
2.3.4 Hot-Leg Piping , Hot-leg piping provides a flow path from the outlet nozzle of the reactor vessel to the inlet to the SG.
- It also provides connections to the pressurizer surge line and to the fourth stage of the ADS and the PRHR.
Each primary loop has an identical hot leg made of [ ] Sch. 40 stainless steel pipe with an I.D. of { ] The piping is a horizontal, straight run with an upward bend, [ ]
- between the reactor vessel and the SG. The venical rise of the hot leg (centerline of the horizontal Q
V run to the centerline of the SG nozzle) is [ vessel nozzle.
] A flanged spoolpiece is provided at the reactor o:W1536w.lb.non:ll41298 2.3-5 REVISION 1
FINAL DATA REPORT l Each hot leg is provided with the following instrumentation, j
= One differential pressure transmitter j O to 25 in. of water
- One heat flux meter ;
- O to 10 Btc!hr.-ft.2 l
- One heated thermocouple (fluid phase) 70* to 550 F
- Two differential pressure transmitters (liquid level)
O to 20 in. of water
- O to 25 in. of water l
. One pressure transducer
- O to 600 psig 1
- Two thermocouples One 40' to 450 F
- One 40 to 550 F
! 2.3.5 Cold Leg Piping l The cold leg provides the coolant flow conduit from the outlet coolant nozzles of the SG to the inlet
- coolant nozzles of the reactor vessel. It also provides a connection for the line to the CMT.
l
- Each of the two reactor coolant loops has two cold legs, each made of 3-1/2 in. Sch. 40 stainless steel pipe. The cold legs, which are entirely horizontal, are fitted with 300-lb,45-degree flanged elbows at l each end and a horizontal spoolpiece with 300-1b flanges. One end of each cold leg is connected to the discharge flange of the RCP, and the other end is connected to the coolant inlet flange of the teactor vessel.
The following instmmentation is provided for each of the four cold legs:
- Two differential pressure transmitters O to 25 in. of water
- One magnetic flow meter (with transmitter; removed after the first matrix test due to repeated mechanical failure)
O to 250 gpm c:%pootA1536w.lb.non:lt>O81298 2.3-6 REVISION 1
FINAL DATA RzronT s -
- One heat flux meter i
! 3 - O to 10 Bru/hr.-ft.2 t -.
- One heated thermocouple (liquid phase) l 70* to 550*F i'.-
One differential pressure transmitter (liquid level) O to 25 in. of water i
~
- One thermocouple (fluid temperature) ~
70* to 450*F
- One thermocouple (metal temperatum) 70* to 550 F !
'2.3.6 Pressurizer Surge Line i
The pressurizer surge line provides the flow path between the RCS and the pressurizer to transmit the pressures from the pressurizer to the flow system and to transfer fluid during volume changes. The pressurizer surge line consists of 3.5-in. Sch. 40, type 304 stainless steel piping that connects HL-2 with the bottom of the pressurizer. The line is provided with two sets of 300-lb flanges to facilitate x assembly. The piping is arranged with six 90-degne bends to form a full 360-degree loop with a vertical line leading to the bottom of the pmssurizer. l The pressurizer surge line is fitted with the following instrumentation:
- Six differential pressure transmitters (liquid level)
-- Three 0 to 20 in. of water
- One 0 to 5 in, of water One 0 to 10 in. of water -
- ' One O to 40 in, of water
* , One differential pressure transmitter O to 1 in. of water
- One heat flux meter
- O to 100 Btu /hr.-ft.2 l
- Three heated thermocouples (fluid) i -
Two O to 100 F
/ --
One 70 to 500*F i
. onsp600Lt5h-Ib non:Ib-081298 2.3-7 REVISION l
FINA1, DATA REPORT
. Two pressure transmitters
- One 0 to 500 psig
- One O to 400 psig
. Two thermocouples (fluid) 70 to 450*F e One thermocouple meter 70 to 450 F 2.3.7 Pressurizer The pressurizer provides the pressure control for the RCS. Pressurizer pressure is maintained at the required level by modulating the power input to the pressurizer heater. The control pressure is transmitted to HL-2 through the pressurizer surge line. The pressurizer also must provide degassing for the primary coolant.
The pressurizer consists of a shell of [ ]a type 304 stainless steel pipe with [ ]a Sch. 40 welded pipe caps at each end and an overall length of [ ]* The 3-1/2 in. pressurizer surge line is butt-welded to the bottom weld cap. Four electrically heated 1-in. diameter rods, with an active length of [ ]C are installed through the bottom weld cap and are sealed by Swagelok fittings. Each heater rod is rated at [ ]C at full power. A [ ]' Sch. 80 line with a 90-degree elbow is provided on the top weld cap near the outer diameter for connection to the ADS. A 1-in. diameter vent line is connected through a blind 300-Ib flange with a high-pressure seal located at the center of the top weld cap. The vent line is controlled by a 3/4-in. electrically operated valve with a manually operated globe shut-off valve. l The following instrumentation is installed in the pressurizer: l
. One heat flux meter
! - O to 100 Btu /hr.-ft.2 l j
. One electric power meter / transmitter l - O to 15 kW
= One diffelential pressure transmitter (liquid) l -
O to 120 in. of water l O l l o:\ap60(A1536w.Ib.non:lt481298 2.3-8 REVISION 1 r
FINAL, DATA Raroar
- One pressure transmitter
,300 to 400 psig--
=. One pressure transmitter O to 500 psig e - Two thermocouples (fluid)
- 70* to 450*F
= Three thermocouples (wall) 70* to 450*F e Four heater thermocouples 70* to 450*F
-2.3.8 Steam Generators De SGs transfer heat from the primary coolant and generate steam from the secondary coolant. The SGs of the test model must meet the following additional functional requirements:
' Each of the test model SGs must be capable of removing 360 kW of thermal energy from the
' k]/ primary fluid while single-phase natural circulation conditiens exist on the primary side. This represents a combined energy removal rate for both SGs equivident to 3.8 percent of scaled decay power, which is the maximum rated power of the core heaters. De actual maximum heat rejection by the SGs is slightly less than the core heater input because of system heat losses.
To properly model the initial conditions of SBLOCA scenaries, the test model SGs must be capable of removing the core energy deposited in the primary fluid while operating with secondary-side pressures that are very close to primary-side pressures. This requires that the tube surface area be sufficiently large to permit heat transfer with small temperature differences between the primary and secondary sides. Each SG is fabricated of Type 304 stainless steel and is insta' led with its axis vertically oriented. The shell is [ - ]C in length, and the total length including axial nozzles
-is[ ]C J A hemispherical head, [ ]* I.D. and 3/8-in. minimum thickness, is attached at the bottom by a flange. A conical section at the top enlarges the I.D. to [ ]C to facilitate vapor separation. Steam is discharged from a [ ]C diameter, flanged nozzle located axially at the center of the top semi-elliptical head, [ ]C in O.D., [ ]C minimum thickness. An air-operated ball valve is installed in the steam discharge line. This valve is controlled by a logic controller to maintain the appropriate pressure in the steam side during the test transients.
I _ o:watsw. ibm =Iw129s 2.3-9 REVISION 1
FINAL DATA REPORT The tube bundle consists of [ ]C U-tubes, [ ]C average length,11/16-in. O.D.,0.040-in. wall welded into a 2-in. thick flange at both the front and back faces. A shroud of [ ]a thick Type 304 stainless steel plate surrounds the tubes and is welded to [ ]C support rods that attach it to the tube sheet. The tube sheet is captured between the flanges of the lower head and is sealed by gaskets on both faces. Moisture is separated by a chevron-type device manufactured by Dyna-Therm . This separator is located in the expanded diameter, upper section of the SG. Feedwater is supplied to the shell side of the SG through a [ ]C diameter line connected to an
]C diameter spray ring made of , ]C Sch. 80 pipe supported inside the vessel by brackets.
[ The hot leg is connected to the SG by a [ ]C nozzle connected to the lower head and supplies hot coolant to the tubes. Primary coolant is discharged from the tube-side h-ader through two [ ]C diameter nozzles. Upon an S signal, the motor-operated feedwater isolation globe valve is closed. The following instrumentation measures the important characteristics on the primary (tube) side of each SG:
- Two differential pressure transmitters
- O to 3 in. of water
- Seven differential pressure transmitters (liquid level)
One O to 30 in. of water Two O to 115 in. of water Two O to 110 in. of water l
- Two O to 20 in. of water l
- Five thermocouples (wall) 40 to 450 F
- Six thermocouples (fluid)
- 40 to 450 F e One pressure transmitter
- O to 500 psig
- One heat flux meter
- O to 100 Btu /hr.-ft.2 The following instrumentation is provided to measure the secondary-side conditions during steady-state or transient operations:
- Two differential pressure transmitters (liquid level)
- One 0 to 33 in. of water g
- One O to 135 in, of water T i
l vAap60(A1536w.lb.non:lt@81298 2.3-10 REVISION 1
FINAt. DATA REPORT l I
- Four thermocouples (fluid)
( - 40' to 450*F e Two heat flux meters , O to 100 Btu /hr.-ft.2 l 2.3.9 Reactor Coolant Pumps The RCPs provide the mechanical energy to circulate the primary coolant at its required flow rate through the RCS. Extended coastdown of the pump is not required. I There are four RCPs installed in the OSU test facility, one close-coupled to each of the cold primary coolant outlet nozzles from the lower head of each SG. 'Ihese pumps are mounted vertically with the motor in the downward direction. The pump (which is a model QPHT5-4* manufactured by the Queen Pump Co., Portland, Oregon) is a centrifugal pump with an internal fluid volume of 0.306 ft.3 and is driven by a 5-hp motor. The pump impeller and volute are modeled to simulate the AP600 pump components. During the tests, the coastdown of this pump was so rapid that the flow became immeasurably low as soon as the power was interrupted. The pump head-versus-flow correlation is shown in Figure 2.3-1. /] NJ The following instrumentation is provided for each of the pump seal cooling systems: One flow meter (visual) O to 5 gpm
. One thermocouple (fluid) 40 to 450 F Flow meters and pressure devices that monitor and record pump performance are included with the cold-leg piping instrumentation.
2.3.10 Accumulators The accumulators provide automatic, passive injection of water into the RCS following a loss of primary coolant. One accumulator tank is connected to each of the DVI lines. The two accumulator tanks are identical, except that the coolant discharge line for the second tank is about [ f further below the bottom of the tank than the discharge line for the first tank. The tanks are designed to the ASME Pressure Vessel Code, Section VIII, Division 1.(5) /~% b o:\npemishIb.non: b-osi29s 2.3-11 REVISION 1
FINA1. DATA REPORT Each tank consists of a [ ]C type 304 stainless steel pipe. The top head
]C stainless steel pipe cap, and the bottom is a [ ]c thick plate that is a welded [
is welded to the barrel. Each tank is supported from the floor by a stand comprised of 61.3 in. long, 20-in. diameter carbon steel pipe, welded to 1/2-in. thick carbon steel base, 26-in. O.D. by 18-in.1.D. He base ring is anchored to the floor, and the upper end of the support is welded to the bottom plate of the tank. The centrally located discharge line,1-1/4 in. Sch.160, is fitted with a 90-degree elbow and exits horizontally through a cut-out in the tank support. He vent / nitrogen inlet line enters the bottom of the tank through a Swagelok seal. The 1/2 in., 0.045 in. wall line is suppoited by a 6-in. long guide tube welded to the inside of the tank, with its center 9 in. from the barrel top head weld. The height of this line can be adjusted by loosening the Swagelok, moving the tube to its new location, and resealing the Swagelok. Teflon ferrules are used for this Swagelok seal to facilitate this adjustable length. However, during the test program, the height of this line remained fixed at 36.8 in. from the inside surface of the bottom flange. Each of the accumulators is equipped with the following instrumentation:
- One differential pressure transmitter (liquid level) 0 to 55 in. of water
- One magnetic flow meter with transmitter O to 40 gpm
- One heat flux meter
- O to 100 Btu /hr.-ft.2
- One pressure indicator O to 500 psig
= One pressure meter O to 400 psig a Three thermocouples 40 to 450 F 2.3.11 Core Makeup Tanks The CMT provides a volume of water maintained at cold-leg pressure by a balance line so that the water flows by gravity into the cold leg of the RCS on a loss of coolant. Each CMT is a cylindrical vessel oriented vertically and designed for operation at 400 F and 400 psi, according to the ASME Pressure Vessel Code, Section VIII.W The vessel consists of a [
owenis36w-ib.non:ib-08:29s 2.3-12 REVISION 1
FINAL DATA REPORT De cold-leg balance line, which is made of [ ]a pipe, enters the CMT through the steam distributor, De steam distributor consists of a [ ]* long with a welded plug at the bottom. De stub is drilled with [ ]8 diameter through-holes from a distance [ ]* above the end plug for a length of ( ]*. De steam distributor is retained between the faces of a [ ]a 300-lb flange joint with 0-ring seals installed at the top of the CMT. A vent line consisting of [ ]a is welded to the top head. A [ ]a O.D. nipple is welded to the center of the bottom head for connection to the DVI piping. Each CMT has the following instmmentation: Four differential pressure transmitters (liquid level) One 0 to 12 in. of water One 0 to 40 in. of water One 0 to 20 in. of water One 0 to 65 in. of water a Three heat flux meters O to 100 Btu /hr.-ft.2 O (V = One pressure transmitter O to 500 psig a 26 thermocouples (fluid) 40' to 450 F
= Three thermocouples (heat flux meter) 40* to 450 F
}
= 28 thermocouples (wall) 40* to 450 F In addition to the instrumentation on the CMTs, the following sensors are located on the cold-leg pressure balance .'ine:
- One differential pressure transmitter
- O to 30 in, of water 1
- One magnetic flow meter (with transmitter)
-[ -
O to 15 gpm ! s oW6axishib.non:ibai29s 2.3-13 REVISION 1
FINAL DATA REPORT
. One heat flux meter
- O to 100 Btu /hr.-ft.2
- Three heated thermocouples (fluid phase)
- Two 0* to 100*F
- One 0* to 500 F
= One differential pressure transmitter (liquid level)
- O to 95 in. of water
= Two thermocouples (fluid)
- 40' to 450*F
= One ther _ocouple (heat flux meter)
- 40' to 450 F e One visual temperature indicator 40* to 450*F 2.3.12 In Containment Refueling Water Storage Tank In the AP600, the IRWST supplies water to fill the refueling cavity during refueling and stores water during normal plant operation. The IRWST also supplies water for emergency core cooling during a LOCA after the RCS has been depressurized and the water from the CMTs and accumulators has been exhausted. It also serves as the heat sink for PRHR during a normal shutdown. In the test model, the injection during loss-of-coolant events and the PRHR operation are modeled and investigated.
The IRWST is fabricated entirely of Type 304 stainless steel and is designed to meet the ASME Section VIII, Division I codeW requirements of [ JC The vertically oriented tank consists of a cylindrical section, [ ]c It is made of [ ]C thick plate, rolled and welded, with a 2:1 elliptical top head, [ ]C thick. The bottom head is a [
]C The top ar d bottom differ because these components were available at the time the tank was fabricated. The bottom head is filled with ceramic covered with a
[ ]C to simulate the flat bottom of the AP600 IRWST. A 30-in.,150-lb blind flange is welded to the top head to provide a manway for access to the tank intemals. Steam vented from the ADS is condensed in the IRWST through a sparger. This component is discussed with the ADS in Subsection 2.1.4. Two discharge lines are provided from the bottom head to the IRWST. The line to DVI-1 is [
]C and the line to the DVI-2 is [ ]C Two 1-1/2 in.,150-lb flanges o:W1536w.lb.non:lb481298 2.3-14 REVISION 1
FINAt. DATA Rsront I 1:
/" are welded to the bottom head to mount two tubes that house thermocouples for internal fluid V] '
temperature measurements. b De PRHR HX is mounted inside the IRWST, and the inlet / outlet piping connections are made through welded 3 in.,150-lb flanges. Details of the PRHR HX are described in Subsection 2.3.18. l The following instrumentation measures the significant parameters of the IRWST during the transients being investigated:
- Eree heat flux meters
- O to 100 Btu /hr.-ft.
)
= . Four load cells 1
i - O to 11,000 lbm '
*- One differential pn:ssure transmitter (level)
- O to 150 in. of water
- One pressure transmitter 0 to 100 psig
- 19 thermocouples (fluid)
- 40' to 450 F 1
l 2.3.13 Safety Injection Lines he safety injection lines provide flow conduits for the water that is injected directly into the reactor vessel from the CMTs, accumulators, and IRWST. The instrumentation for the safety injection lines is summarized in the following: Instrumentation - CMT ACC IRWST DVI Differential pressure transmitter 1 1 2 Magnetic flow meter 1 1 2 2 Magnetic flow meter transmitter 1 1 2 2 "Ihermocouple (fluid) 3 2 4 --- 2.3.14 Containment Sumps
; The LCS recirculates water from the containment sump through the reactor vessel to provide passive cooling for removal of decay heat. Recirculation starts when the water level in the LCS has reached o:w15h ib.non:ib-ost29s - 2.3 15 REVISION I
FINAL DATA REPORT the elevation of the DVI. At this level, the recirculation line is filled and a density driving force exists as boiling in the core reduces the density of the liquid in the core compared to the cooler water in the sump. Steam is released to the containment thmugh the ADS-4 valves on the hot legs and is condensed in the containment, eventually n:tuming to the sump. ne model containment sumps simulate the volume in which water can collect from the leakage during hypothetical LOCAs. For most break locations, the AP600 reactor sump would fill and recirculation through the LCS would be initiated as soon as the water level reached the DVI line. However, some breaks could occur in compartments that are hydraulically isolated from the reactor sump, and those compartments must flood above a given elevation before the reactor sump begins to fill. This How behavior is modeled by a primary sump tank that represented the reactor sump and a secondary sump j tank that simulated the isolated compartments, 2.3.15 Automatic Depressurization System, Stages 1-3 De ADS consists of four stages. Stages I through 3 are described in this section. The function of the ADS is to reduce the pressure in the RCS by venting steam in a controlled manner to permit injection of cooling water from the CMTs, accumulators, and IRWST. i ADS 1-3, shown schematically in Figure 2.3-2, is connected to the top of the pressurizer through a [ ]C Sch. 80 pipe. Separate branch lines are provided for each of the valves, and flanged nozzles are installed in each branch to permit adjustment of the flow resistance in each branch. A [ ]C relief valve is tied into this line before the line is reduced for each of the ADS valves. The ADS-1 valve, which opens at the highest pressure, is a [ ]C valve joined by [ ]C Sch.80 piping to the [ ]C ADS 2 line from the [ ]C ADS-3 line. The combined line from the ADS 1 and 3 valves is expanded to [ ] tubing with a [ ]c wall thickness by a welded reducer. The [ ]C line carrying the ADS-2 valve is connected to the [ ]C tubing. The ADS valves are pneumatically operated ball valves; each valve is programmed to open at descending pressures. Two phase flows vented through the three valves are piped to a vapor liquid separator. The liquid and steam mass flow rates are individually measured and then recombined and flow to the ADS 1-3 sparger through a [ ]C pipe. Flow from the ADS enters the IRWST through a [ ]C 300-lb flanged connection and is dispersed at about midlevel in the tank by a sparger. The sparger consists of a [ ]C inlet line, which is expanded to [ ]C Sch. 40 at the hub. [ Je sparger arms are connected to the hub in a crucifonn arrangement. Each sparger arm has l an active length of about [ ]C which is drilled with [ ]C holes [ ]C in diameter. The following instmmentation is installed in the ADS 1-3: !
- Three flow meters (differential pressure)
Two O to 60 in. of water l One 0 to 325 in. of water o:\ap60(A15%w.lb.non:lb-081298 2.3-16 REVISION 1
FINA1, DATA F8 Ar{ p * .Two magnetic flow meters V - O to 60 gpm
. One heat flux meter O to 100 Bru/hr.-ft.2) e Bree heated thermocouples (fluid)
Two O to 100*F One 0* to 500*F
=
One differential pressure transmitter (liquid level)
- O to 140 in, of water
- Two pressure transmitters O to 500 psig
- Three thermocouples 40" to 450 F e _ One thermocouple (heat flux meter)
(' .- 40* to 450"F
\
2.3.16 ' Automatic Depressurization System, Stage 4 The function of ADS-4 is to reduce RCS pressure near containment pressure. RCS pressure must be near the containment pressure in order for the IRWST water to be injected. Two ADS-4 vent lines are provided, each connected to one of the hot legs. Each line, which models two lines in the AP600, consists of [ ]' which is connected to the hot leg through a flanged tee. The [ ]c pneumatically operated ball valve programmed to open after the ADS 1-3 valves open. After passing through a flanged flow nozzle used to adjust the line resistance, the two-phase flow enters a separator. The liquid from the separator flows to the primary sump; the steam flow is measured and then exhausted to the atmosphere. The following instrumentation is provided for each ADS-4 line and separator:
=' One vortex flow meter-O to 2000 scfm
[~ *- One differential pressure transmitter (liquid level)
\ -
O to 90 in. of water ownsh-ib.non;ib-os129s 2.3-17 REVISION 1
FINAL DATA REPORT
- One pressure transmitter
- O to 100 psig
- Two pressure transmitters
- O to 500 psig
- Three thermocouples
- 40* to 450 F
= One thermocouple (heat flux meter) 40* to 450*F 2.3.17 Nonsafety Injection Systems Two nonsafety injection systems, the CVS and the RNS, simulate the systems' operation during transient tests. Descriptions of these systems are provided in this section.
In the test facility, the CVS provides injection of additional feedwater into the SG channel head. In the AP600, this syste.n is used to adjust the RCS water chemistry and maintain the system liquid volume. In the test facility, the RNS injects demineralized water into the CMT/DVI line, simulating properties of the AP600 RNS. The CVS pump inlet is connected to the SG main feed header by 3/4-in. pipe 'Ihe pump, model 3333, type CB5-45, manufactured by Gould, is a multistage centrifugal pump with a 5-hp motor. Figure 2.3-3 shows the head-flow correlation for this pump. It discharges through a 3/4-in. pipe and check valve to an expanded section of 1-in. x 0.87-in. I.D. tubing with a motor-operated ball valve to the channel head of SG 2. A 1-in. diameter branch line with a manually operated ball valve is connected to the feed line or SG-1. The RNS pump (Grundfos Pumps Corp., Clovis, CA, series C, model CR4-100N) receives flow from the main feed line through a 2-in. pipe. Flow from the pu.np discharge is piped through 1-in. pipe and 1-1/4 in. diameter (1-in. I.D.) tubing with a 1-in. check valve and a 1-in. pneumatically operated ball valve. The pump performance curve is illustrated in Figure 2.3-4. A 2-in. diameter line is also connected to the discharge line from the pump to the IRWST injection line. This line is equipped with a 2-in. manually operated ball valve. 1 REVISIGN 1 oAnp600\l536w-lb.non:Ib.081298 2.3-18
FINAI, DATA REPCaT The following instmmentation is provided for these systems: V Instrument CVS RNS Magnetic flow meter 0 to 8 Fpm 0 to 60 gpm Visual pressure indication 0 to 600 psig 0 to 300 psig Pressure transmitter 0 to 500 psig 0 to 250 psig Temperature 40* to 450'F 40' to 450 F 2.3.18 Passive Residual Heat Removal The PRHR HX removes decay heat from the core absolutely passively during an emergency shutdown in which heat cannot be rejected through the SGs or RNS. He AP600 has two 100-percent capacity PRHR HXs, each capable of removing 2 percent of the core power using natural circulation. The C-type tube HXs are located in the IRWST and can operate at full system pressure. A single PRHR H% 6 o.mued in the OSU test facility. This loop consists of piping from the reactor p vessel to the C-tube HX mounted inside the IRWST. Heat is transferred from the C-tubes to the water () in the IRWST by conduction and natural convection. Flow from HL-2 enters the C-tube HX through a 1-1/2 in. Sch. 80 line with a normally open ball valve and a magnetic flow meter. The cooled liquid flows from the bottom header of the C-tube HX through 1-1/2-in. tubing (1.26-in. I.D.) with a normally closed pneumatically operated ball valve and magnetic flow meter to the channel head of SG-2. A drain line with a normally closed needle valve is connected to the condenser drain header. He following instrumentation is provided for the PRHR: Two magnetic flow meters (with transmitter) O to 15 gpm a Two heat flux meters O to 100 Bru/hr.-ft.2 Three heated thermocouples (fluid)
- Two 0* to 100*F One 0* to 500*F e
Two differential pressure transmitters (liquid level) O One 0 to 10 in. of water One O to 70 in of water oAmp60tA1536w-lb.non:Ib-081298 2.3-19 REVISION I ( .
FmAL DATA REPORT l
= Nine thermocouples (fluid)
- 40* to 450 F
- Two thermocouples (heat flux meter)
- 40* to 450 F
- One visual temperature indicator
[ }*
- Eight thermocouples (wall)
{ l' 2.3.19 Break Simulators The break simulators provide controlled leakar,es that simulate failures of RCS piping in the AP600. Each break simulation for the hot leg and cold leg consists of a flanged spoolpiece installed in the line in which the break is to be simulated. A line with a pneumatically actuated ball valve is connected to the spoolpiece and conducts the leakage flow to the BAMS, which is described in Subsection 2.3.20. There are four break simulation locations in the test facility-one at the hot leg, one at the cold leg, one at the DVI line, and one at the CMT-1 balance line. All break locations are capable of simulating a single-ended break of a desired size. In addition, the DVI line break and the CMT-1 balance line break can also simulate double-ended guillotine (DEG) piping failures. The flow out of each of these breaks is typically two phase. The break separator is equipped with four inlets and two outlets. Only one of these four inlets is used at a time, and their elevations are important. There are three inlets located below the curb / overflow level-one at cold-leg elevation for cold-leg bitak simulation, one at hot-leg elevation for hot-leg break simulation, and one at DVI elevation for DVI break simulation. For example, the cold-leg break inlet must be at the same elevation as the cold-leg break location. If it is located at a higher elevation, it will create backpressure at the break source (that is, break hole at the cold leg) and subsequently will change the thermal-hydraulic characteristics of the break flow. Similarly, the hot-leg break inlet is located at the hot-leg break hole elevation, and the DVI break inlet is located at the DVI break elevation. The fourth inlet of the break separator is located at a level between the curb / overflow level and the lowest break above the curb / overflow level. This inlet is used for the CMT-1 balance line break. This arrangement allows one inlet for several break locations above the curb / overflow level without introducing improper backpressure on the break source. The break separator also has a loop seal at the liquid drain line to prevent steam from blowing out of the liquid drain line and ensure valid liquid flow measurements. o$ap600(1536w.ib.non:Ib481298 2.3-20 REVISION 1
FmAt. DATA REPORT l l Instrumentation for the break system is included in the discussion of the BAMS. ne steam outlet line is directed to the common header, as described previously. The separator, the
- loop seal lines, and the steam line are preheated and insulated to minimize heat loss to the atmosphere and to prevent steam condensation. Heating the loop seal lines also ensures that the temperature of the condensate is close to the temperature at which it would collect in the AP600 containment sump.
2.3.20 Break and ADS Measurement System (BAMS) l The BAMS accurately measures the steam and liquid flows from the four ADS stages and each break simulator being tested. The approach used to accurately measure these two-phase flows is to separate each two-phase flow stream into its liquid and vapor components and then to measure the flow rate and temperature of each single-phase flow stream. For the ADS-4 and break separator, the vapor streams are vented to the atmosphere and the liquid phases are collected in the primary sump, which simulates the containment sump in the AP600. The capability exists to pump heated water from the condensate tank into the primary sump tank at a mass flow rate equivalent to the rate of vented steam. This water simulates the flow of condensate from the steam vented into the containment that would be l condensed and would drain into the containment sump. The steam and liquid flows from the ADS 1 l_ separator are recombined and' flow into the IRWST through the sparger. ! The BAMS consists of four separators and the associated piping and instmmentation to measure the !C single-phase flows and to conduct streams to the appropriate location. The following provides the data l- for the separators, which are made of type 304 stainless steel by Wright-Austin Company, Detroit, Michigan: SEPARATOR DATA l' Length Diameter Vapor Outlet Liquid Outlet Tank Model No. (in.) (in.) Inlet Diameter Diameter ADS 13 8-in. 56 20 4in. 4in. 4 in, type RR l ADS 4 5-in. 62 20 3-1/2 in. 5-in. 31/2 in. l (two tanks) type 14R Sch.40 Sch.40 Sch.40 Break 14-in. 146 32 4-in. 10-in. Sch. 40 4-in. type 14RR Sch.40 Sch.40 l Instrumentation provided for the BAMS fluid measurements are as follows: i e Bree magnetic flow meters (liquid) }g - Two O to 45 gpm One O to 60 gpm o:W1536w-lb.non:lb-081298 2.3-21 REVISION 1 i -. . - . .. _ ~ .
FINAt. DATA REPORT a Five vortex flow meters (steam)
- One 0 to 7060 scfm
- One 0 to 12,500 scfm
- One O to 22 scfm
- One 0 to 6000 scfm
- One 0 to 11,000 scfm
. Three pressure transmitters
- O to 60 psig
= 10 thermocouples (fluid)
- 40* to 450 F
= 14 thermocouples (trace heater) 40 to 450*F 23.21 Test Support Systems Several systems that are not AP600 models are required to operate the test facility. These systems are as follows:
- Demineralized water system
. Fill-and-drain system a RCP seal cooling system
. Electrical system
. Trace heaters
- Insulation 23.21.1 Demineralized Water System City water is passed through two filters, connected in parallel, each with a differential pressure transmitter to indicate plugging. The combined flow from the filters is split into two streams, each passing through a separate and isolatable bank of demineralizers. After passing through another set of filters, the demineralized water flows to the feed storage tank. Feed lines that bypass the demineralizers are also provided to fill the RCP seal cooling system and the direct condensate tank.
All lines are 1-in stainless steel tubing with manually operated ball valves. 23.21.2 Fill-and-Drain System Demineralized water from the feed storage tank is pumped by a single feedwater pump to the SGs. The feedwater line consists of a 2-in. Sch. 80 pipe from the feedwater tank and 1-in. tubing, with a 0.87-in. I.D. from the feedwater pump to the SGs. o:\ap60(A1536w-Ib.non:lb-081298 2 3-22 REVISION 1
l FmAL DATA REPORT Fill lines to the primary and secondary sumps are connected to the main feed header upstream of the feedwater pomp. The RNS pump suction also is connected to the main feed header. 23.213 RCP Seal Cooling System Heat is removed from the seals of the four RCPs by the closed-circuit RCP seal cooling system. Water is circulated by a centrifugal pump, through a jacket surrounding each pump seal. Heat is ; rejected from the coolant by a fan-driven, water-to-air HX. An expansion tank with a liquid level l switch is mounted on the building roof. Thermocouples measure the temperature at each seal, and the
. flow to each seal is monitored by a visual flow indicator.
1 23.21.4 Electrical System ! The electrical system for the OSU test facility is shown in Figure 2.3-5. Three-phase,480-V electric power is supplied to the facility from a 1000-kVA,4160-V transformer located outside the building. ! Power is divided into five circuits, two 600-A circuits and three 200-A circuits. The 600-A circuits supply the rod bundle heaters, and the 200-A circuits power the pressurizer heaters, trace heaters, and motors. Current transformers connected to the 600-A circuits and the 200-A circuits to the pressurizer supply input to the power meters for each bank of reactor heaters and the pressurizer heaters. Redundant current transformers and power meters are provided for the reactor heaters to verify the important data and to provide instrumentation backup. Eight rod bundle heaters are connected in parallel, and three sets of heaters form a delta connection of 24 heaters connected to one of the SCR power controllers. A circuit breaker is installed in the circuit to provide a safety trip. An identical electrical system is provided for the other 24 rod bundle heaters. The four 3-1/4 kW pressurizer heaters are connected to the 200-A line, which is also equipped with a
. circuit breaker for emergency trip. Power to the heaters, which are wired in a delta arrangement, is controlled by an SCR.
I l l A 200-A line supplies power to the trace heaters. Power to the motor control panel is fumished by the ! 100-A line. 23.21.5 Trace Heaters The BAMS separators and piping (up to and including the flow meters) are heated to about 220 F to prevent condensation of the steam content, which would affect the steam / water mass ratios and the i accuracy of the energy balances being measured. 'Ihese trace heaters (rated at 20 kW/ft. and 277 V), which are arranged in 14 separate zones, each with on-off control, are Raychem Chemelex Heat . Tracing Systems, Model 20XTV2-CT
- O i o:W1536w-Ib.non:lt>081298 2.3-23 REVISION 1
\ -. - - , , _ . _
FINAL DATA REPORT For the tests, heater density was based on raising the BAMS tanks to 220*F from an initial temperature of 180 F in 4 hours, including heating water in the tanks. Piping heaters were based on preheating the piping in the BAMS to the same temperature. 2.3.21.6 Insulation Thermal insulation is installed on all tanks and piping, except the CMTs and accumulators. Four types of insulation are used: l j
= 2-in. Fiberglass, Manville Micro-lokm, k,g = 0.31 Btu-indhr.-ft.2 *F at 200 F
= l-in. Fiberglass, Manville Micro-lokm, k,g = 0.31 Btu-indhr.-ft.2 *F at 200 F l = 1-1/2 in. Polyisocyanurate Foam, Dow Plastics, k,g = 0.141 Btu-indhr.-ft.2 *F at 75 F 2
= Removable Blankets, Ceramic Fiber, Lewco Specialty Products, k,g = 0.55 Btu-inihr. ft - F at l 600 F Table 2.3-2 is a matrix listing the insulation types for the major components and piping systems.
l r 1 0 i i O o:Wism.ib.nonib48:298 2.3-24 REVISION 1
FINAL. DATA REPORT
!b TABLE 2.31 ROD BUNDLE CHARACTERISTICS Characteristic Metric English Number of heater rods 48. 48 Maximum power per rod 15. kW 51,200. Bru/hr.
Rod diameter 2.54 cm 1.00 in. Heated length 91.4 cm 36.in. Heated surface area 730.3 cm 2 113.in.2 Rod cross-sectional area 5.07 cm 2 0.785 in.2 Heated volume 46.cm 3 28.3 in.3 Total heated surface area 35,000, em 2 5420. in.2 Total heated cross-sectional area 243. cm 3 37.7 in.2 Total heated volume 22,300. cm 3 1360.in.3 O
\.J Heater rod pitch 4.01 cm 1.58 in.
Pitch / diameter ratio 1.58 1.58 Subchannel flow area 11.0 cm 2 1.7) in.2 Hydraulic diameter 5.52 cm 2.18 in. Average rod heat flux 11.5 W/cm2 Radial power peaking factor 1.31 Axial power peaking factor 1.47 liot channel factor 1.93 c:W1536w-lb.non:Ib-081298 2.3-25 REVISION 1
- t. .
FINAL DATA REPORT TABLE 2.3-2 INSULATION APPLICATIONS Insulation 2-in. Fiberglass 1-in. Fiberglass 1 1/2 in. Urethane RCS X Pressurizer X ADS 1-2 X ADS - common line X SGs X Condensate return tank X Steam line X Condensate piping X 1 IRWST X [ Primary sump X Primary sump piping X Secondary sump X , DVI system piping X l ADS separators and piping X l l l Drain collection tank X Break separator tank and X piping to primary sump l l 9 a:W1536w.lb.non:1b-081298 2.3-26 REVISION 1
FLNAL DATA REPORT O e
-k b.
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-nI
-h
-S o * ' 6' ' ' ' 6* '6' 6' od N N N d I me f Figure 23-1 RCP Performance Head Versus Flow o:W1536w.lb.non:lbul298 2.3-27 REVISION 1
t FINAL DATA RzroRT O> C
<REUEFHEADERI OSU 600902 OSI j
~~
RCS-600 HDS-606 fr-615
' f a 433 3 ORI-657 db ~-
,- n 2"' SCH BOS Tr.602
- fi HFM-6 .
l RCS-603 TVHT-7 =4
- T DP
- f3 g 1.5' SCH 805i agg j W -655 .,
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[,3 [7 i ff gg TF-605 - t 8; gg ~" "'00I
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7 '> C TF-608 I ZDNE RCS-602 y i 15' SCH 805 4
, 7
- rpP " f RCS-607 ( % mrv-602 I
I RCS-620 77
'05 L 68* {
A X Hru.sa( l PZR HEATERS 13 hw 1-3 ADS l *SEPARATDR i t I " W RV-622 6' (CH 40 H l p C L6 ff ~ ~ rvn - - - 20NE 6 vM Q 168 (set ,'*rVHr-6 $" TV-6er' p Tr + 7 e 8 W TVHT-4 s g- - 001-659 ((
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$U I RCS-619
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V 4
/ SEAL wAT[p rggg ggggCR
\3SU 600002 SHEER ! C4 Figure 2.3-2 Flow Schematic for the ADS O
WWl536w Ib.non:Ib 081298 2.3-28 REV1SJON ]
FINA1. DATA REPORT O Figure 3.18.3-1: CVS Pump Head vs. Flow. Gould, Model 3333, Type CB5-45 2000 1800 - 1600 - l l 1400 - U ' M - 1200 O 1000 - O 800 - 600 - 400 - s 200 - 1 I I I I O 2 4 6 8 10 0 Figure 2.3-3 CVS Pump Head Versus Flow c:W1536w-lb.non:Ib-081298 2.3-29 REVISION 1
FINA1. DATA REPORT O Figure 3.18.3-2: RNS Pump Head vs. Flow PSI Grundfos, Series C, Model CR4-100N 850 275 -~ 600 - 250 -- 550 - 225~~ 500 - 200-- 450 - U d 175,-~ 400 - g 150 -- 350 - O
=
q 300 - r 125 -- 0 250 - 100 -- 200 - 75 -- 150 - 50 -- 100 - s 25 -- 50 - t 0 -- O 5 10 15 20 25 30 35 40 l l l Figure 2.3-4 RNS Pump IIcad Versus Flow 9 c:Wish-ib.non:li>ost29s 2.3-30 REVISION 1
.. .. . . - - . . - . - . . . - . . .-.....-... . _ . - --..-.. _....... - ~..-_........ - . _ . - . - - - - . - . . . - ,
- 1 FINAL DATA REPORT i l 1 ;
l
- Figure 2.3-5 is not included in this nonproprietary document. I i'5-4 i
4 4-4 i-5 b ,1 . s 4 1 4 f. 4 1-1' a a a 4 ( , f' J a a i \ h-1 4 i d 3 d 1-4 4 4 4 owish-Ib.non:Ib ost:9s 2.3-31 REVISION 1
~ -.- -. . . - - - . - - - ..-.-.- _-._.- - - - . - . - . - .
FmAL DATA Ruoar 2.4 Instrumentation This section provides general information about instrumentation used at the OSU test facility and specific instrumentation anomalies that occurred during testing. All instruments referenced in this section can be found in the OSU test facility piping and instrumentation drawings (P& ids), Appendix G, and in the OSU AP600 Instrumentation Data Base, Appendix C. The first two drawings of the P& ids are a legend (OSU 600 LEG, Sh. I and 2), which provide iriformation about instrument numbering. In general, all instrument identifications contain a two- or three-letter prefix, followed by a three-digit number. The letters represent the type of instrument; the number provides the system number and sequential number af that instrument in that system. For example, PT-107 is a pressure transmitter (PT), located in the prusure vessel (100 series), and has the sequential number 7. 2.4.1 General Information on Instrumentation I This subsection contains general information about the OSU test facility instrumentation. The instrument channel designator for each instrument is provided in parentheses at the beginning of the subsection describin; .he instrument. l O i 2.4.1.1 Differential Pressure Transmitters (FDP, LDP, DP) V Differential pressure transmitters measure three different parameters and have a prefix representing their application. De transmitters measure flow (FDP), level (LDP), or differential pressure (DP). Rosemount" Model 3051 transmitters measure small differential pressures of 0 to 150 in. H2 0, and Rosemount* Model 1151 transmitters measure differential pressures greater than 150 in. H2 0. l De only application of FDPS is to measure the differential pressure across the flow orifices in the ADS-1, ADS-2, and ADS 3 lines (FDP-604, FDP-605, and FDP-606). Differential pressure l transmitters measure levels in all facility tanks, the RCS hot-leg and cold-leg pipes, SG tubes, and l PRHR HX tubes. l The high- and low-pressure sides of all differential transmitters are plumbed to a component via two 3/8-in. O.D. stainless steel tubing sense lines. The transmitters are mounted on one of several l instrumentation racks at the ground elevation of the facility, at a lower elevation than both of the sense line taps at the component. He sense lines are filled and vented via vent plugs at the transmitter and high-point vent tees with caps located on the sense lines above the sense line connection to the component. The instrument can be isolated by closing root valves on the sense lines close to the sense I line taps at the component, or by closing isolation or block valves at the manifold of the transmitter at the instrument rack. An equalizing valve is provided for each transmitter to equalize the high- and low-pressure sides of the transmitter. o:W1536wi3a.non:IM81298 2,4-] REVISION 1 s,
FINAL DATA REPoTr The bench calibration of all differential pressure transmitters is the same. The 4- to 20-mA signal output of the transmitters is input to the DAS. However, the field setup is different for LDPs than for DPs and FDPS, because LDPs are designed to measure levels and not differential pressures due tc flow head loss. For the performance of the tests, the LDPs were aligned in the field to output a minimum signal, 4 mA, when the measured level in the component was at or below the lower sense line tap or variable leg tap. Maximum output from the transmitter,20 mA, occurred when the level reached the upper sense line tap or reference leg tap. In this way, the DAS could directly interpret the output of the transmitter as level indication. He configuration file included in every test data file provides the range of each LDP, corresponding to the distance between the level taps. It should be noted that the field adjustment of LDPs did not affect the bench calibration of the transmitter because the adjustment was made in the digital space of the transmitter electronics. The LDPs were calibrated at ambient temperature. The transmitters did not mechanically or electronically temperature-compensate their output to correct for measuring levels of fluid at elevated temperatures. Therefore, level data recorded by the facility's DAS were uncompensated.
- 2.4.1.2 Pressure Transmitters (PT)
He pressure transmitters (FTs) are Rosemount* Model 1151 transmitters, identical to the differential pressure transmitters, except the low-pressure side of the transmitter senses atmospheric pressure. The transmitters are plumbed to the component by a 3/8-in. O.D. sense line. 2.4.1.3 Magnetic Flow Meters (FMM) Foxboro* magnetic flow meters (FMMs) measure liquid flow. The FMMs consist of a ceramic-lined flanged flow tube connected to a remotely-mounted transmitter. The transmitter energizes the flux-producing coils of the flow tube, which then produces voltage across a pair of electrodes proportional to the liquid flow rate in the ttabe. His flow tube voltage is measured by the transmitter and converted to a 4- to 20-mA signal measured by the DAS. The FMMs are not designed to accurately measure steam or two-phase flow. The data from the transmitters are invalid when either of these are measured. 2.4.1.4 Heated Phase Switches (HPS) Heated phase switches (HPSs) manufactured by Reotherm* measure fluid phase. There are 12 switches: one each on the cold and hot legs, CMT balance lines, PRHR HX inlet, and ADS 1-3 header. In addition, two switches are installed in the pressurizer surge line. The design of the HPS is an adaptation of a flow meter design used to measure flow rate. I l o$ap600\l536w-3a.non:Ib-081298 2.4-2 REVISION 1
- ~ . ----.- .--.--_- - . . - . - - . - . - - . . - . - - FmAL DATA REPORT An HPS consists of two elements, a transducer and electronics unit. The transducer is a single probe 5 containing two sensors inserted 1 in. into a pipe. One of the sensors is at equilibrium with the fluid; the other is located near a heater so that its temperature is slightly above that of the unheated sensor. He temperature of the sensor located near the heater is a function of fluid state. The signals from the two temperature sensors are sent to remote electronics mounted inside a cabinet. He HPS electronics process the two input temperature signals and output three signals to the DAS identified as HPS-XXX-Y, whem XXX is the unique instrument channel number assigned to the HPS and Y is 1,2, or 3. The instrument channel identified as 2 is the delta temperature between the heated and unheated temperature sensors, the channel identified as 3 is the fluid temperature measured by the unheated temperature sensor, and the channel identified as 1 indicates the phase of the fluid. During testing, the 1 instrument channel output of the HPS was a 0- to 10-volt signal that was convened by the DAS to a 0 to 100 percent indication. A low voltage, or small percentage, was designed to indicate a gas phase; a high voltage, or large percentage, was designed to indicate a liquid
. phase. When the HPS was functionally checked after installation, it operated properly, providing more than 9 volts with water in the system and providing -2 volts with the system drained and air in the system. Test results from the 3 instrument channel during matrix testing when steam and two-phase fluid were present were not conclusive. Further analysis is required to determine the accuracy and usefulness of the 1 data. The fluid temperature measurement of the 3 channel is considered accurate.
V 2.4.1.5 Heat Flux Meters (HFM, TFM) The RdF heat flux meters measure heat flux through pipe or tank walls. The small, wafer-thin instruments are glued to a pipe or tank sunace. Bree thermocouples are imbedded in to each HFM. Two thermocouples measure temperature on either side of the HFM. The thermocouple signals are measured by the DAS, and their temperature difference is converted to a heat flux using coefficients provided by the vendor. The third thermocouple measures the temperature of the surface. During testing, the heat flux calculation of the DAS was designated as a HFM data channel. The wall temperature measurement of the heat flux meter was designated as a TFM data channel. An energy balance was not performed, so the output of the HFM was not evaluated. 2.4.1.6 Vortex Flow Meters (FVM) Founeen Foxboro" vonex flow meters (FVMs) measure steam flow in the test facility. The FVMs consist of a flanged flow tube connected to remote electronics. The FVMs measure steam flow from the ADS 1-3, ADS 4-1, ADS 4-2, and break separators. In addition, they measure steam flow from the primary sump, the IRWST, and in the BAMS header. As steam flows through the flow tube, a vortex-shedding element causes vortexes to form and shed at a rate proportional to the flow velocity of the steam. The vortexes create an alternating differential 0%p600\l536w-3a.non:lb-081298 2.4-3 REVISION 1
FINAI. DATA REPORT perssure that is sensed by a detector and converted to a voltage output to remote electronics. Remote electronics convert the voltage input signal to a 4- to 20-mA signal measured by the DAS. 2.4.1.7 Load Cell Transmitters (LCT) The mass of water in the IRWST, primary sump, and secondary sump is measured by load cells mounted under the four supports of each tank. The load cells contain strain gages that are stressed by applied shear forces from the weight of the tank and its contents. The strain gage produces a millivolt output proportional to the mass of the water in the tank. The input signal from the four load cells of the tank is processed by a transmitter mounted in close proximity to the load cells. The 4- to 20-mA output of the transmitter is measured by the DAS. After the transmitter was calibrated, it measured only the weight of water in the tank. The transmitter also provided local indication of weight in the tank for use by test personnel. 2.4.1.8 Thermocouples (TF, TW, TH, TR) Thermocouples are assigned one of four instrument designations, depending on the thermocouple's application. A TF thermocouple inserted through the wall of a pipe or tank or mounted on a thermocouple rod measures fluid temperature. TW thermocouples are mounted on the inside or outside walls of a tank or pipe. TR thermocouples, unique to the reactor vessel, are mounted on vertical thermocouple to_ds installed in the reactor vessel. TH thermocouples are mounted on the heaters for the reactor vessel and the pressurizer. Thermocouple type and theimocouple diameter are specified in the OSU AP600 Instrumentation Data Base, Appendix C. The database also specifies the insertion depth of through-wall fluid thermocouples (TF). Inside wall thermocouples are mounted in a groove cut into the wall of the component and silver-soldered to keep them in place. The reactor vessel contains TH thermocouples to measure temperatures of selected heaters. Selected heater thermocouples are used as inputs to the safety shutdown of the reactor heaters to detect abnormally high temperatures. Appendix G, Dwgs. OSU 600007 and 600008 provide information on the location of the heater thermocouples. The drawings provide the orientation of the heater thermocouples in the core, as well as the mounting elevation of the thermocouples. The elevation of the thermocouples specified on the drawings is referenced to the bottom of the reactor vessel. Heater thermocouples are also mounted in heaters of the pressurizer. Appendix G, Dwg. OSU 600203 provides the location of the three pressurizer heater thermocouples. Thermocouples mounted in hollow rods (TR thermocouples) are unique to the reactor vessel. Five thermocouple rods are installed in the reactor vessel to provide radial and axial fluid temperature distributions in the heated section of the reactor vessel. Each rod contains thermocouples mounted along its entire length. Thermocouples protrude from the hollow rod and are sealed from the outside oMp600(1536w.3 anon:Ib-081298 2.4-4 REVISION l
FINAI, DATA REroRT O with silver solder. Wires for the thermocouples are routed inside the rod, through the bottom of the O rod, and then through a seal at the bottom of the reactor vessel. The orientation of the five thermocouple rods in the core is provided in Appendix G, Dwgs. OSU 600007 and 600008. As the drawings indicat; e of *he four thermocouple rods is at the center of the core and the other four thermocouples are . +. * ; in the third of four concentric rings.
%e CMTs are instrumented with numerous fluid and wall thermocouples, as indicated on Appendix G, Dwgs. OSU 600501 and 600502. Each CMT contains two short and one long thermocouple rods, or rakes, instrumented with thermocouples along its entire length. In addition, inside and outside wall thermocouples, fluid thermocouples installed I in. from the inside wall, and tank centerline thermo-couples are installed at the same elevation to measure the temperature of the fluid and walls at that elevation.
Note: Appendix G, Dwgs. OSU 600501 and 600502 incorrectly identify the orientation of the long thermocoeple rake in CMT-1 and CMT-2 as 135"az. The actual orientation is 315*az. The IRWST also contains two thermocouple rods, or rakes, as shown in Dwg. OSU 600701. The thermocouple rod located on the 45*az is mounted 3 in, from the PRHR HX tube bundle to measure the effects of HX operation. De thermocouple rake mounted at 270 az measures temperature distribution from sparger operation. O One long and one short tube of each SG are instrumented with shell-side (secondary side) wall thermocouples and tube-side (primary or RCS-side) fluid thermocouples as shown in Dwg. OSU 600301. Originally, each tube had two wall thermocouples mounted on the hot-leg side and two wall thermocouples mounted on the cold-leg side. Several thermocouples were damaged during installation. The thermocouples were inaccessible, so they were not repaired. Thus, the drawing indicates the loss ~of these thermocouples. 2.4.1.9 Signal Conditioners (SC) It is necessary to provide selected signals to DAS and control panel instiumentation at the same time. For instrumentation loops containing transmitters, two dropping resistors in series are used in the instrument loop to provide the signal to both locations. To split a thermocouple signal between the DAS and the control panel, a signal conditioner is needed. Sixteen signal conditioners provide control panel indication and control input for hot and cold leg, pressurizer vapor space, SG steam, primary and
- secondary sump, and IRWST temperatures. In addition, six signal conditioners split selected heater temperatures. De heater temperatures are used for control panel indication and as input to logic to trip all heaters when an abnormally high temperature is detected.
In Appendix C, the Instrumentation Data Base indicates that the signal conditioners are designated by the prefix SC. However, the P& ids (Appendix G) contain no instruments with an SC prefix. Inputs to the signal conditioners are either fluid thermocouples or heater thermocouples. If the input to the o:W15h3non:Ib 081298 2.4-5 REVISION 1 L-....,.. .
FINAL DA'rA REFORT SC is a fluid thermocouple (TF), the SC has the same number as the thermocouple. For example, SC-101 is the signal conditioner for CL-3 temperature. The thermocouple monitoring this temperature is TF-103. If the input to the SC is a heater thermocouple (TH), the SC number contains the entire number of the thermocouple. For example, SC-TH-101-3 is the signal conditioner for heater thermocouple TH-101-3. Data from the 22 thermocouples that are inputs to the sigrial conditioners are not listed in the test data files by their thermocouple designator (TF or TH), but by their signal conditioner designator (SC). Using the previous example, temperature data from CL-3 thermocouple TF-103 are identified as SC-103 data. 2.4.1.10 Programmable Logic Controller (PLC) The Omron PLC contains the control logic of the test facility written in a ladder logic format. The PLC is the interface between the control panel instrumentation and the equipment to provide proper cor. trol and safe operation of the equipment. All inputs to the PLC are digital, except heater temp-erature signals used to trip the reactor heaters. The PLC provides digital control such as opening and closing air-operated valves, starting and stopping pumps, providing inputs to control panel alarms, and commanding the automatic operation of equipment during a matrix test. It does not provide analog functions such as reactor heater (power) control or SG steam flow control via a motor-operated valve. Digital events in the facility are recorded by an additional software program that monitors the input and output of the PLC. 2.4.1.11 Control Panel Instrumentation The control panel instrumentation provides manual and automatic control of the equipment in the facility and operator indication of facility parameters, but does not provide input to the DAS. Subsection 2.6 discusses the details of control panel instrumentation and how it was used to operate the test facility. 2.4.2 Calibration Methods and Standards The general process used for calibration of the OSU test instruments is described below:
- The instrument was identified by manufacturer and model number.
- Accuracy, input values, and output values, considered critical characteristics of the instrument were identified.
. The instmments were receipt inspected to verify that the correct model numbers were received and no damage had occurred during shipping.
REVISION 1 owmis36w.3a.non:1b-ost29s 2.4-6
FINAL DATA REPORT
=
Critical characteristics were verified by calibrating the instrument before the beginning of O- matrix testing and by performing post-test calibration after completion of matrix testing. 1 The instruments installed in the field served a number of purposes and, therefore, had different ' calibration requirements. De instrument types were divided into the following three functional groups l based on their function and calibration requirements. j Functional Groups: Group I - Instruments used for data acquisition which were calibrated onsite. Rese instruments had an input to the DAS. Group II -- Instruments used for local indication (such as pressure indicators), local monitoring (such as pressure switches), and indicators and controllers located on the control panel. These instruments did not provide an input to the DAS.
- Group III Instruments which had no onsite calibration and had vendor-supplied calibration data sheets (e.g. heat flux meters, flow meters, or flow tubes for magnetic flow meters).
)
)
/
Valves, solenoids, limit switches, etc., not normally included in the definition of an instrument. Thermocouples which were functionally checked initially (not calibrated), could not be re-spanned or re-ranged, and required no further calibration. Calibration Periodicity Categories: Category I -- Calibration frequency of 12 months due to importance in testing and data reduction. All category I instruments either had an input to the DAS or were a functional part
. of the instrument loop.
Category II - Calibration frequency of 18 months. Although these instruments did not have a direct input to the DAS, they were important for the control and safety of the equipment. Category III -- Required no further calibration beyond functional test or vendor calibration, oW1536w.3a.non:lt481298 2,4 7 REVISION 1
l FINAL DATA REPORT , 1 l Calibration Periodicity Table Prefix Category Description Frequency (months) I Differential pressure transmitter - headloss 12 DP I Differential pressure transmitter - flow 12 FDP I Flcw meter magnetic 12 FMM FVM I Flow meter vortex 12 hai III Heat flux meter N/A HPS III Heated phase switch N/A LCT I Load cell transmitter 12 LDP I Differential pressure transmitter - level 12 PT I Pressure transmitter 12 SC I Signal conditioner - temperature 12 TF III Fluid thermocouple N/A TFM III HFM thermocouple N/A TH III Heater rod thermocouple N/A TR III Core Instrument rod thermocouple N/A TW III Wall thennocouple N/A The calibration methods used for specific instmment tyr es are described below. 2.4.2.1 Thermocouples i
- A thermocouple calibrator (Tegam Model 840) and a dry-well tester were used to check the calibration of the standard rod-type thermocouples. Rese thermocouples were inserted into the 1/4-in.,1/2-in., or 3/8-in. inserts of the dry-well tester.
The thermocouple probe supplied with the calibrator was connected to the terminals labelled thermometer input. This measured the reference temperature of the dry well. Type J, K, or T thermocouples were calibrated with this setup. Most of the thermocouples used at OSU were type K. O c:'ep6(xA1536w-3tnon:ib48t298 2.4-8 REVISION 1 l
FINA1. DATA Rr.roRT A The thermocouple being tested was checked for temperature at room temperature,245"F, and 450 F. V Between every setpoint,10 minutes were allowed for the tester to reach the setpoint and stabilize. Non-standard Germocouples, which could not be placed in the dry-well tester, were checked using a heat gun set at 250*F, These thermocouples were not checked at 450'F. 2.4.2.2 Pressure Transmitter The pressure transmitters are one-piece capacitance-based absolute-gage pressure sensors and transmitters. Pressure is converted to captaitance by the transducer. The transmitter electronics and a digital converter conven the output signal to a range of 4 to 20mA. The milliampere signal is then transmitted to an indicator or the DAS. A digital multimeter (HART Communicator Interface, Rosemount Model 268) and a pressure gauge (Transmation Model 1090) were used to check the calibration of the pressure transmitters. The pressure transmitter output range was established from 4 to 20mA and, if the measured outputs were satisfactory, a linearity check of the digital to analogue converter was performed. The linearity check of the digital to analogue convener confirmed that output requests of 0,25,50,75, and 100 percent corresponded to amplifier outputs of 4,8,12,16, and 20mA, respectively. [' The pressure transmitters were initially bench-calibrated from 0 psig to about 600 psig. Once installed, the transmitter output was a combination of system pressure and pressure of the water column. Field calibration of the pressure sensors was performed to offset the effects of the height of the sensing line water column. The pressure transmitter loop check and DAS calibration were accomplished by applying power to the transmitter and recording the DAS voltage reading at 4,12, and 20mA applied. current signals. From these voltages, the slope and Y intercept values for the instrument, required for the DAS, were calibrated. 2.4.2.3 Differential Pressure Transmitters The differential pressure transmitters were bench-calibrated in a manner similar to the pressure transmitters described previously. In addition, the differential pressure transmitters were bench-calibrated to a span greater than the physical tap-to-tap height differential of the plant. Therefore, the full tank level was bench-set for 4mA output, and the empty tank level was less than 20mA. A field calibration procedure was carried out to set the output range of the transmitters to correspond to the actual tank level by using the transmitter output, in engineering units, at the tank empty level (4mA) and tank full level (20mA). This set the output range of the transmitter to a value within the bench calibration range and was consistent with the level being measured. O V ow60m:5hAnon:IM61298 ' 2,4 9 REVISION 1
FINAL DATA REPORT 2.4.2.4 Load Cells ne load cells were vendor-calibrated and required no further calibration. A loop calibrator j (Transmation Model 1090 or equivalent) was used to calibrate the load cells. Using the loop calibrator, power was applied to the control panel and field instrumentation. The loop calibrator was l connected to the loop at the transmitter and the DAS voltage reading was recorded at 4,12, and 20mA l app. lied-current signals. From these voltages, the slope and Y intercept for the instrument, required for the DAS, were calculated. The slope and Y intercept are stored in the DAS system. The engineering units of the signal are calculated from this slope and Y intercept. 2.4.2.5 Magnetic Flow Meters The magnetic flow meter transmitters use a pulsed-DC technique to energize the flux-producing coils of the flow tube. As liquid passes through the magnetic field in the flow tube, low-level voltage pulses develop across a pair of electrodes. The voltage level is directly proportional to the average velocity of the liquid. The flow tube and its electronics were not pan of the calibration. Using a digital multimeter, the magnetic flow meter transmitter amplifier was calibrated for a 4 to 20mA output range. Transmitter loop check and DAS calibration were carried out by applying power to the transmitter and recording the DAS voltage readings at 4,12, and 20mA applied-current signals. From these voltages, tho slope and Y intercept values for the instruments, required for the DAS, were calculated. 2.4.2.6 Turbine Flow Meters A frequency calibrator (Transmation Model 1070) and a digital multimeter were used to check the calibration of the turbine flow meters. The transmitter amplifier was calibrated for a 4 to 20mA output range. The frequency (or pulse) input to the amplifier was polarity-sensitive. Negative pulses were applied to the transmitter to check the amplifier polarity. Transmitter loop check and DAS calibration of the turbine flow meters were accomplished by applying power to the transmitter and recording the DAS voltage readings at 4,12, and 20mA applied-current signals. From these voltages, the slope and Y intercept values for the instrument, required for the DAS, were calibrated. Negative pulses were applied to the transmitter to check polarity. 2.4.2.7 Vortex Flow Meters Fluid passing through the flow meter body passes a specially shaped vortex shedder which causes vortexes to altemately form and shed from the sides of the shedder at a rate proportional to the flow rate of the fluid. These shedding vortexes create an alternating differential pressure which is sensed by a detector located above the shedder. A pulsed voltage is generated by the detector and the voltage conditioned by the amplifier to produce a 4 to 20mA output signal. ohp60m1536w-3amn:Ib-081298 2,4 10 REVISION 1
FmAI, DATA REPORT O A frequency calibrator (Transmation Model 1070) and a digital multimeter were used to check the calibration of the vortex flow meters. The transmitter amplifier was calibrated for an output range of 4 to 20mA corresponding to a frequency calibrator range of 0 to 100 percent. The transmitter loop check and DAS calibration were accomplished by applying power to the transmitter and recording the DAS voltage readings at 4,12, and 20mA applied-current signals. From these voltages, the slope and Y intercept values for the instrument, required for the DAS, were calibrated. 2.4.2.8 Noncalibrated Instrumentation (IIeat Flux Meters and IIeated Phase Switches) The heat flux meters measure heat passing over the surface of the sensor attached to the vessels or pipes of interest. These devices were calibrated by the manufacturer and did not require any further calibration. 2.4.3 Phenomena Affecting Readings This subsection examines the effects of different phenomena on instrument measurements. The discussion is intended to assist in the review of data from the matrix tests. 2A.3.1 Flow and Temperature Effects on Level Differential Pressure (LDP) Transmitters The LDPs measure Huid level between the upper reference leg tap and the lower variable leg tap of a component. The LDP tubing is connected to the component to be measured so that any pressure (such as steam pressure) will act equally on the reference leg and the variable leg and, thus, have no effect on the indicated level data. The difference between the constant reference leg level and measured Huid column level creates a differential pressure that is measured by an LDP and electronically converted to a level signal. An accurate level signal is dependent on static conditions, i.e., no flow in the fluid column measured. Flow in a component creates a dynamic differential pressure due to pressure loss between the component LDP taps as fluid flows through the component. When this dynamic component of differential pressure is superimposed on the static differential pressure, the resulting transmitter signal produce invalid data. If the flow direction in the component is from the LDP variable tap (low-pressure side) towards the reference tap (high-pressure side), the dynamic and static differential pressures will be additive, creating indicated level data greater than actual level. Conversely, if flow is in the opposite direction, the indicated level data are lower than actual level. - The LDPs installed at the OSU test facility were bench calibrated at ambient temperature and had no electronic temperature compensation. The temperature in the fluid column to be measured was, in many cases, elevated above ambient conditions. When the temperature of the fluid column was elevated with respect to the temperature of the tubing between the component and the LDP, the o:\ap600tl536w-3a.non:]b-081298 2.4-11 REVISION 1
i l FINAL. DATA REPORT I I i resulting transmitter signal provided indicated level data lower than actual due to the expansion of fluid with increasing temperature. For some of the data plots in this report, a program was developed that compensates selected level data for temperature. The temperature-compensated level data on those figures have the LDP designator preceded by a C. For example, CLDP-127 is the temperature-compensated data for level channel LDP-127. None of the raw level data transmitted with this report have been temperature compensated. 2.4.3.2 Effect of Two Phase or Steam Flow on Magnetic Flow Meters (FMM) The FMMs are in-line flow instruments that measure liquid flow. As long as the flow stream is liquid solid in the forward direction, the FMMs provide accurate and valid data. During the perfonnance of the matrix tests, the LOCA caused the flow stream through some of the FMMs to become two-phase fluid and, in some cases, all steam. Also, in some instances, reverse flow occurred. The effects of two-phase fluid, steam, or gas flow on FMM data were varied and unique to the meter. Some meters indicated erroneously high flow, some zero flow, and some negative flow. As an example, the CMT/ cold-leg balance line flow meters indicated oscillating flow as the CMT transitioned from recirculation mode to draindown, and two-phase fluid or steam appeared in the balance line. Balance line flow data after the transition occurred are considered invalid. In another example, the accumulator injection flow meter indicated negative flow as the accumulator emptied and the meter was placed in a gas environment. Again, the data are considered invalid after the accumulator was empty of water. For liquid-solid reverse flow, an FMM provides negative flow data. In this case, the negative data provide an accurate indication of flow reversal, but the absolute negative value of the data would be inaccurate and should be considered invalid. As an example, negative flow data were recorded in the IRWST-1 injection line when the primary sump valves opened late in a matrix test. Although the negative flow indication is valid, the actual value of the data is not. 2.4.3.3 Effect of Backflow on Vortex Flow Meters (FVM) The FVMs are in-line flow instruments designed to measure forward steam flow. FVMs measure flow of any gas going through them and in either direction. During reverse flow, the FVMs provide indication of positive flow, but the data are considered invalid. An example of FVM positive flow indication during reverse flow occurred when the break valve opened during a matrix test. When the break valve first opened, the steam flow from the break pressurized the BAMS steam header. The header pressurization caused a backflow of steam into the ADS-4 separators until the pressures were equalized. The backflow of steam into the ADS-4 separators resulted in positive flow data. Therefore, all ADS-4 separator steam flow data prior to the ADS-4 valves opening are considered invalid. ogsxx1536w-3amn:1b-08:298 2.4-12 REVISION 1
FINAL. DATA REPORT i \ O 2.4.3.4 Effect of Two-Phase Liquid or Steam on Differential Pressure Transmitters (DP) () Installed in a Vertical Orientation The DPs are bench calibrated to a range suited for the particular application. However, if the DP is installed in a vertical orientation, its data have a zero offset when a liquid level exists below the elevation of the upper sense line's tap. In DP installations where both taps are at the same elevation, as when measuring flow differential pressure in a horizontal pipe, the static pressure from the weight of fluid above the taps is sensed by both legs of the transmitter and therefore does not affect the differential pressure measurement of the transmitter. This is true if the fluid is liquid, steam, or a two-phase mixture. However, the two sense lines of a vertically-mounted transmitter experience unequal static pressures when the liquid level of the component drops below the elevation of the upper sense line's tap. When this occurs, one side of the transmitter senses static pressure equal to the column of water in the upper sense line; the other side of the transmitter senses static pressure equal to the water column in the lower sense line, plus the static pressure from the water level in the component above the elevation of the lower tap. The maximum offset results when the liq > tid level in the component decreases to an elevation below the lower tap l
~
As an example during testing, DP-130 measured the differential pressure across the core bypass holes located in the upper core barrel flange of the reactor vessel. When the liquid level was above the upper tap of DP-130, both sides of the transmitter sensed the same static pressure due to the weight of l water above it. Thus, the differential pressure experienced by the transmitter was due to a flow head l i loss only. With a decreasing level in the reactor vessel, both sides of the transmitter sensed the same decrease in pressure due to the decreasing water column above it until the liquid level dropped below the level of the upper tap. After this, the side of the transmitter measuring the upper sense line pressure did not see any additional decrease in static pressure; it sensed only the water column in the upper sense line. Thus, the desired pressure indication of the transmitter was offset by an amount equal to the difference between the elevation of the upper tap and the water level in the component. The maximum offset was -22 in. H 20, which was equal to the distance between taps for the upper and lower sense lines. In most matrix tests, the level in the reactor quickly drained below the lower tap of DP-130. When no flow existed through the bypass holes, the data from DP 130 indicted a differential pressure of -22 in. H2 0. 2.4.3.5 Effect of Reactor Coolant System Draindown on Steam Generator U-Tube LDPs A phenomenon occurred with respect to the SG U-tube level instruments during the performance of matrix tests due to the unique LDP tubing installation. The reference and variable legs for the U-tube LDPs penetrated the SG at the tube-sheet elevation. The variable leg was routed to the bottom of the ! respective tube, and the reference leg traversed the axial length of the tube bundle, internal to the SG, ( and was connected to the top of the respective tube (Appendix H, Dwgs. 20175-D1 and 20175-D2). oAsp600\l536w 3a.non:Ib-081298 2.4-13 REVISION 1
FmA1. DATA REPORT The routing of the LDP tubing inside the SG exposed the tubing to the secondary-side temperature environment of the SG. Therefore, after break valve opening during a matrix test, when the fluid in the SG U-tubes transitioned from recirculation to draindown, uncovering the reference leg tap, the fluid in the LDP reference leg began to vaporize. The decreasing level in the reference leg provided a false high indication for U tube level. When the U-tubes were completely drained, the level data erroneously indicated that the U-tubes were partially filled. As the reference leg continued to vaporize, decreasing the reference leg level, the level data incorrectly indicated the U-tubes refilling. When the reference leg was completely voided, the differential pressure between the reference and variable legs was equal indicating full U-tubes. The U-tube temperature data confirmed that the U-tubes remained drained because the temperatures superheated and remained superheated after the U-tubes drained. It is important to understand that the level data can be interpreted to determine the time that the l U-tubes began to drain and when the U-tubes were completely drained for analysis purposes. l l l O i l 1 i l I l l 1 a:waniS36w.3a.non:1bo8 298 2.4-14 IlEVISIM 1 l
FINAL DATA REPORT l O 2.5 Data Acquisition System O The DAS receives data from the test instrumentation, records it, and prepares it for review. This section describes the DAS, including architecture, hardware, and processing software. 2.5.1 System Hardware The DAS consists of about 750 data channels that are monitored during test operations. Figure 2.5-1 illustrates system hardware. The channels are distributed among three separate Fluke He'ios Data Acquisition Units @ racks. Each rack is serially connected directly to a separate 486DX PC system. Each system PC is tied to the others by an Ethernet connection. The system software used for the local area network (LAN) is Microsoft Workgroup@ for Windows. Rack i Helios contains about 300 channels and acquires data from the rapid-responding inputs such as pressure or flow instruments. The remaining channels, primarily thermocouples, are split between rack 2 and rack 3. 2.5.2 DAS Architecture Figure 2.5 2 is a schematic of the DAS architecture. The system is initialized by the user from the user's PC with the time and the system configuration data base. On triggering, either by input from the user or by signal input at the start of a test, the Helios begins acquiring data. Data are sent from
) the Helios to the PCs for all channels with instrument inputs and are stored by the PCs. Predefined channels are processed and displayed from incoming data. Burst data are acquired at a higher rate for all channels and stored in the Helios until the end of the test. Data files are recorded on a writable CD-ROM.
2.5.3 Software DAS software is divided into five main functions: initialization, data acquisition, burst data acquisition, display channel monitoring, and data storage. Before DAS execution can begin, the system configuration file must be extracted from the system data base, formatted, and transferred to the DAS PCs. The system configuration files are distributed in three main files sorted by rack and row, where each row is an analog-to-digital (A/D) converter card. As a convenience, each row is also separated into a file that correlates with burst data acquisition. From the configuration file, a set of channels is chosen to be displayed. A file is created cor*aining parameters for the display channels to be monitored during test operations.
'Ihe initialization function prompts the user for input data, sends the channel configuration to the Helios equipment, and then waits for an acknowledgment of the channel definitions. Other configuration data sent to the Helios include burst rate, system time, and other setup parameters. This ,m function also sets up the PCs to display channel information and burst delay time.
I i LJ oAap600\l536w-3tnon:1b-081298 2.5. ] REVISION 1 l ____~
FINAL DATA REPORT The data acquisition function is designed to retrieve channel data and write it to a system disk, continuously, every 8 to 10 seconds from the start to the end of the test. The start of acquisition is triggered either by a transient signal or operator input. Operator action is required to stop acquisition. Burst data acquisition acquires data at a faster rate, but for a shorter length of time, than continuous acquisition. The burst rate is user-defined and can be from once per second to once every 10 seconds. The start of burst is input by the test operator and can begin immediately with the start of continuous acquisition or can be delayed minutes or hours. The burst data system can store up to 1900 scans for each channel. This means that at a burst rate of once per second, burst data can be acquired for about one-half hour. Burst data are stored on the Helios until the end of the test. Then the operator initiates the transfer of data from the Helios to the system PC, where the data are stored in an ASCII file. The display channel monitor function selects channel values from the continuous data stream for predefined channels. These channel values are converted from voltages into engineering units and displayed on the PC monitor. From the predefined channels, the test operator can choose up to four to be dynamically charted on the monitor. The channel display indicates an alarm condition if the channel value falls outside the alarm valee limit. He data storage function accumulates all configuration files, continuous data files, display data files, and burst data files and transfers them to a single system via the LAN. PSCRIBE software is used to write the data files onto a CD-ROM for post-test processing, as described in Section 3.0, 2.5.4 LabVIEW Description The software is written using LabVIEW for Windows, which uses a graphic programming language to create programs in block diagram form. LabVIEW is a general-purpose programming system, but it also includes libraries of functions and development tools designed specifically for data acquisition and instrument control. LabVIEW subprograms are called virtual instmments (VIs) because their i appearance and operation imitate actual instruments. VIs accept parameters from higher level VIs and 1
- have an interactive user interface and a source code equivalent. Figure 2.5-3 illustrates the AP600 DAS hierarchical VI.
The interactive user interface of a VI is called thefront panel because it simulates the panel of a l physical instrument. The front panel can contain knobs, pushbuttons, graphs, and other controls and indicators. Data are input using a mouse and keyboard. Results are viewed on the computer screen. l 1 2.5.5 Sequence-of-Events Log The DAS acquires information during any given test without reference to an event initiated by the facility control system, such as a valve opening or pump trip. The sequence of and timing of these I events are recorded on a separate PC using a software program named Intouch@ made commercially 1 available by Wonderware. A Wonderware data base was constmeted using the programmable logic osp600sism-knon:ib ost298 2.5-2 REVISION I
FmAt. DATA REPORT l ( T controller (PLC) data base as a model. Signals from the facility instrumentation that affected the PLC - program were included in the data base. While the software is running, it automatically logs every signal input to the PLC.
'l> iogging format used in the following test sections is an ASCII file generated by the software:
MM/DD/YY: HH:MM:SS:MSC PRI Tagname Value I As an example, consider air-operated valves in the facility. All of these valves have limit switches at each end of the stroke so that the PLC has a positive indication of the valve in a fully open or fully close ' .osition. Both switch signals are logged as an individual point as they go through a transition. A t' rical message for a single stroke of CMT-1 pressure balance line valve RCS-503 would be listed as fc. c.vs: 07/08/9411:04:38.5391 RCS503_Open Not Closed 07/08/9411:04:38.8461 RCS503_ Closed Closed Four groups of messages are placed in the log file, including Valve Position Log, Pump Status Log, Console Log, and Alarm Log. n Y I l O c:Wl536w-3a.non:lb-081298 2.5-3 REVISION 1
FINAL DATA REPORT Ol ! 1 i l l t 1 i\ l l 1 l 1 i l
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FINAt. DATA REPORT
/ 2.6 Test Facility Control System This section describes the manual and automatic functions of the test facility control system. The test facility control system operates in either manual or automatic mode Contro! pr.r.cl and Wal switches allow operator manual control of the equipment for the following:
- To open and close valves e
To start and stop pumps
- To control the power of heaters The automatic control mode provides signals for the following:
To automatically operate the facility at steady-state j l To initiate a simulated break and transmit automatic control signals to simulate AP600 control logic during a LOCA To control the reactor heaters to simulate decay heat production of an AP600 nuclear core
.- To operate the BAMS l p
b
- To provide automatic control of the nonsafety systerns (CVS and RNS pumps)
To provide automatic equipment operation to mitigate out-of-specification conditions in order to prevent equipment damage and injury to personnel
- To provide control of heat tracing
- To provide alarms 2.6.1 Operator Panel Figure 2.61 is a photograph of the operator panel in the test facility control room. It shows the programmable controllers and the process centrol system components listed in Tables 2.6-1 and 2.6-2, respectively. Figure 2.6-2 shows the arrangement of these components on the operator panel. The major components of the operator panel, the manufacturer, model number, function, and operation are as follows:
Temperature scanner, OmegaN CN34025-DC, indicates the temperatures of trace heaters. Q
'g
- Temperature scanner, Omega CN101K-1000F, provides an alarm if RCP cooling water exceeds the adjustable setpoint.
o:W15hknon:ib-ost29s 2.6-1 REV1510N 1
FINA1. DATA REPORT
- Power meters, Power Measurementsm 3710 ACM, indicate the power to each bank of rod bundle heaters.
- Bar-graph indicators, Universal VB-120-2-4, indicate the percent range for level parameters.
'Ihe bar-graph indicators have programmable setpoints that when reached, operate a bistable to provide input to the PLC for appropriate action.
- Panel switches and lights, Squam D , provide CLOSE/ AUTO /OPEN control for air-operated valves, pump / fan power, and STOP/RUN indicator lights.
- Alarms, Panatarmm Series 90, provide an alarm annunciator and horn.
- Process indicators, Newportm Model 82, indicate pressure, temperatures, levels, and flows.
The process indicators have programmable setpoints that when reached, operate a bistable to provide an input to the PLC for appropriate action.
- Process controllers, Fischer/ Porter 53HC3300C, process field variables.
i 2.6.2 Test Signal or Safety Signal (S signal) i The PLC is programmed to automatically operate equipment in a timed sequence when the TEST pushbutton on the control panel is pressed to initiate a matrix test. Equipment operation is described in detail in the following sections. However, the following is the sequence of operation with time zero referenced as the time when the break valve open signal is generated. The actions listed after the S signal were initiated by the S signal. 1 i Equipment Operated Time DAS triggered to start collecting data -120 seconds Break valve signal to open; start of test 0 second S signal generated 0.5 second SG pressure control reset to 335 psig 0.5 second Reactor control shift to kW mode 0.5 second Main feed pump trips 3.6 seconds Pressurizer heater trips 6.1 seconds CMT valve control (open) 6.1 seconds RCP control (trip) 8.6 seconds oAmp600\l536w.3a.non:lt481298 2.6-2 REVISION 1
_ - _ _ . _ - _ . _._.m _ __ ._ _ _ _ _ _ _ _ _ _ _ _ _ _ RNAL DATA REPORT l l Equipment Operated Time PRHR HX outlet isolation valve (open) 6.1 seconds SG feedwater valve (close) 3.6 seconds 2.6.3 CVS Pump and Discharge Valve Control ne CVS pump operates automatically when its control panel switcli is placed in the AUTO position. In automatic control, the pump starts when the pressurizer level is below its low-low level setpoint of 30 in. H 2O and stops when pressurizer level increases to 45 in. H 2 0. The CVS pump operated in automatic during matrix testing of the interactions of the nonsafety systems with the passive systems. To scale the AP600 CVS flow rate to the OSU test facility, the CVS pump curve is simulated by automatically throttling the CVS pump discharge valve, RCS-808, thereby controlling flow from the pump. A Fischer/Porterm process controller uses pressurizer pressure (PT-604) as an input to the following flow function: Flow = -0.003827954 x pressurizer pressure + 5.32547 gpm 1 In automatic control, the function provides the setpoint to the discharge valve's proportional, integral, derivative (PID) software controller. The setpoint is compared with CVS pump flow measured by flow meter FMM-801. Note: The original CVS pump flow meter was a turbine meter (FTM-801). His meter was replaced by a magnetic flow meter (FMM-801) by the time the first matrix test was performed. To prevent exceeding the design pressure drop across the CVS pump discharge valve, pump speed is controlled by the same controller that controls the discharge valve. The input function for the pump speed is the same as the input to the CVS discharge valve control and pressurizer pressure (PT-604). The speed function is: ! Speed = 3.7312 x pressurizer pressure + 2000.0 rpm When the pressurizer is at atmospheric pressure, the speed of the pump is at its minimum,2000 rpm. l The lower pump speed allows the discharge valve to throttle at a position that precludes cavitation across the valve. As pressurizer pressure increases, the speed of the pump increases. Pump protection is provided by a low-suction pressure trip of 12 in. H2 0. The low-suction pressure trip is disabled for 3 seconds after the pump starts to allow for the initial suction pressure spike during .[ pump start. The pump also trips on a low supply voltage of less than 480 vac. r-oAap60(A1536w-3a.non:Ib.081298 2.6-3 REVISION I I' l
FINAI. DATA REPORT To protect the RCS against overpressurization, the CV5 putnp trips when pressurizer pressure exceeds its high high pressure setpoint of 385 psig and is reset at 380 psig dec: easing. To restart the CVS pump after an automatic trip, the operator must place the control panel switch in the OFF position and then return it to the AUTO or RUN position. 2.6.4 RNS Pump Control The RNS pump operates automatically when its control panel switch is in the AUTO position. In automatic control, the pump starts if an S signal is present and pressurizer pressure is below its high-pressure reset point of 225 psig and stops when pressurizer pressure increases to 230 psig. The pump operated in automatic during matrix testing of the interactions of the nensafety systems with the passive systems. In addition, the RNS pump has a local controller located near the pump that can be used by the operator for fill operations or maintenance activities. Consequently, there is safety interlock so that the RNS pump can only be started from the local panel when the local pushbutton is in the STOP position and the RNS pump switch on the control panel is in the LOCAL position. To scale the AP600 RNS flow rate to the OSU test facility, the RNS pump curve is simulated by automatically throttling the discharge valve of the RNS pump (RCS-810 thereby controlling the flow from the pump. A Fischer/ Porter process controller uses pressurizer pressure (I'F-604) as input to the following flow function: Flow = 20.05 - (0.03051 x pressurizer pressure)l* gpm The function provides the setpoint to the discharge valve's PID software controller. The setpoint is compared to RNS pump flow measured by flow meter FMM-805. Note: The original RNS pump flow meter was a turbine meter (FTM-804). The meter was replaced by a magnetic flow meter (FMM-805) by the time the first matrix test was performed. Pump protection is provided by a low-suction pressure trip of 12 in. H2 0. The low-suction pressure trip is disabled for 3 seconds after the pump starts to allow for the initial suction pressure spike during pump start. The pump also trips on a low supply voltage of less than 480 vac. To protect the pump from operating at shutoff head conditions, the RNS pump trips when pressurizer pressure exceeds its high-pressure setpoint of 230 psig increasing, with a reset point of 225 psig decreasing. To restart the RNS pump after any one of the automatic trips, the operator must place the control panel switch in the OFF position and then return it to the AUTO or RUN position. awnis36w-3a.mn:1wsi29s 2.6 4 REVISION 1
FINAL. DATA REPC"AT 2.6.5 IRWST Valve Control The IRWST-1 discharge isolation valve, RCS-711, and IRWST-2 discharge isolation valve, RCS-712, are normally closed valves provided with manual and automatic control. The valves are opened in manual control by piacing their respective control panel switch in the OPEN position. When their control switches are placed in the AUTO position, the valves open when reactor vessel upper-head pressure (PT-107) decreases to less than 30 psig. The IRWST isolation valves in the AP600 design are normally open valves. The OSU test facility control logic maintains the valves closed in automatic control until reactor coolant pressure is sufficiently low to prevent back leakage through the check valves in series with the isolation valves. Conversely, the setpoint to open the valves is sufficiently high so that IRWST injection is delayed for a period of time after the valves open, so the closed valves never prevent IRWST injection flow. 2.6.6 Main Feed Pump and Discharge Valve Control The main feed pump operates automatically when its control panel switch is placed in the AUTO position. In automatic control, the pump starts and runs continuously. The pump is automatically tripped 3.1 seconds after an S signal is generated. At the beginning of each matrix test, the pump was in the AUTO position.
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Pump protection is provided by a low-suction pressure trip of 12 in. H2 0. The low-suction pressure trip is disabled for 3 seconds after the pump starts to allow for the initial suction pressure spike during pump start. The pump is also tripped on a low supply voltage of less than 480 vac. The main feed pump also trips if either SG wide-range level exceeds 117 in. H2 O and is reset at 116 in. H2 0. To restait the main feed pump after any one of the automatic trips, the operator must place the control panel switch in the OFF position and then retum it to the AUTO or RUN position. A low-pressure alarm informs the operator if feed pump discharge pressure (PT-001) decreases to less than 200 psig. The alarm alerts the operator to take action to protect the pump against run-out conditions. 2.6.7 Pressurizer Pressure Control The pressurizer controller controls power to the pressurizer heaters through an SCR heater controller, in automatic control, the pressurizer controller maintains pressurizer pressure within a band of 365 to 375 psig by controlling power to the pressurizer heaters. O owism&wIb-os129s -2.6-5 REVISION 1
FINAt. DATA REroRT In manual control, the operator adjusts the pressure setpoint on the pressurizer controller, and the controller controls the heater power to maintain the setpoint. The pressurizer heaters are disabled on high-high pressurizer pressure of 385 psig, low-low pressurizer level of 30 in. H 0, 2 SCR cabinet high temperature, or 5.6 seconds after an S signal is generated. De heaters are enabled when none of these signals is present. 2.6.8 RCP Gland Seal Cooling System Control In the AUTO mode, seals for the RCPs are cooled by a separate water cooling system that consists of a pump and a fan to cool the HX located on the roof of the OSU test facility. He cooling fan and pump, if in the AUTO position, are tumed on as soon as one-out-of-four RCPs starts. When all RCPs are turned off, the cooling pump and fan are turned off after a 15-second delay. This is important because during a matrix test, the RCPs are not operating and the heat loss through the seal cooling system is minimized by tuming the seal cooling system off. The cooling fan and pump can be tumed on in the manual mode by placing individual switches on the control panel in the RUN position. 2.6.9 CMT Valve Control Re CMT discharge isolation valves, RCS-535 and RCS-536, are normally closed valves provided with manual and automatic control. The valves are opened in manual control by placing their respective control panel switch in the OPEN position. When their control switches are placed in the AUTO position, the valves open 5.6 seconds after the S signal is generated. He CMT cold-leg balance line isolation valves, RCS-529 and RCS-530, are normally open valves. The valves are closed in manual control by placing their respective control panel switch in the CLOSED position. 2.6.10 CMT Steam Trap Isolation Valves Re. CMT steam trap valves, RCS-503 and RCS-504 are normally closed valves provided with manual and automatic control. The valves are opened in manual control by placing their respective control panel switch in the OPEN position. When their control switches are placed in the AUTO position, the valves remain closed. O 2.6-6 REVISION 1 owaats36w.3a.non:ib-os129s
NAL DATA REPORT 2.6.11 RCP Control The RCPs (RCP-1, RCP-2, RCP-3, and RCP-4) operate automatically when their control panel switches are placed in the AUTO position. In automatic control, the pumps start when pressurizer r vel is above the low low level reset point of 45 in. H O2 and stop when pressurizer level decreases to 30 in. H 20. De pumps were mn in automatic control at the beginning of each matrix test and were automatically tripped 8.1 seconds after an S signal was generated. The pump trips on the following:
- Low supply voltage of less than 480 vac
= Gland seal water temperatum (TS-001) greater than the high setpoint of 150*F
= RCP gland seal water flow low The pumps can be mn in the JOG position, a manual mode of operation. The pumps run as long as the control switch is held in the JOG position. When the switch is released, it spring-retums to the OFF position.
To restart the RCP pumps after the occurrence of pressurizer low-low level, seal water temperature high, or low voltage, the operator must place the control panel switch in the OFF position and then return it to the AUTO or JOG position. An alarm is sent to the control panel to alen the operator if the RCP seal water temperature exceeds the high-temperature setpoint (150 F) or RCP cooling water flow is low. 2.6.12 Reactor Hester Control There are two groups of heater rods in the reactor vessel. Each group is controlled by a separate SCR controller. During steady-state operation of the test facility, the heater controller maintains the average reactor hot-leg temperature at a setpoint selected by the operator. When an S signal is received and a test stans, the control signal to the heater banks simulates decay power expected in the AP600 plant scaled to the OSU test facility. This control is done by a Fischer/Poner process controller. De standard algorithm for decay power was used for all the matrix tests, except Matrix Test SB21.
. The algorithm was:
For 0 < time s 140 seconds; power (KW-101 or KW-102) = 300 kW For time > 140 seconds; power (KW-101 or KW-102) = 300/[ ]c a: W 15 h h ana:Ib a l29s 2.6-7 REVISION 1
FINAL DATA REPORT The Matrix Test SB21 decay power algorithm was: For 0 < time < 300 seconds; power (KW-101 or KW-102) = 300 kW For time > 300 seconds; power (KW 101 or KW-102) = 300/[ ]c where: B = 0.01021 C = 0.2848 Safety interlocks similar to a pressurizer heater are programmed in the PLC. Reactor heaters are tripped when the following occur:
- High sheath temperature is detected
- RCP cooling water flow is low with an S signal a RCP gland seal temperature is high (150 F) without an S signal e Emergency button is pressed a KEY switch is not operated
. Power to at least one heater bank (KW1 or KW2) is more than 396 kW e Average of the hot-leg temperature (TF-140 and TF-141) is more than 440 F
- Pressurizer pressure is greater than 385 psig
= SCR cabinet high temperature occurs a Total power (KW1 + KW2) > 756 KW Once the reactor heaters are tripped because of a safety condition of high sheath temperature, high gland seal temperature, low RCP cooling flow, or high SCR cabinet temperature, the heamr power cannot be applied unless the condition for the safety violation has cleared and the KEY switch is in the OFF position. The power can be applied by placing the KEY switch first in the OFF and then in the ON position.
A control panel alarm alerts the operator if the average temperature of HL-1 and HL-2 is higher than the high-high setpoint of 440 F. At this temperature, the heaters are disabled. 2.6.13 Passive Heat Removal The PRHR HX outlet isolation valve, RCS-804, is a normally closed valve provided with manual and automatic control. The valve is opened in manual control by placing the respective control panel switch in the OPEN position. When the control switch is placed in the AUTO position, the valve 9 owis36w 3a.non: b-08:298 2.6-8 REVISION 1
FINAL DATA REPC3T p opens when the ADS 1 valve opens because of CMT low level or 5.6 seconds after an S signal is b generated. 2.6.14 Condensate Return Pump Control The condensate return pump (CRP) is capable of transferring water from the condensate return tank (CRT) to the primary sump or to the IRWST when the CRT level (BGI 14, LDP-903) is more than the low setpoint of 10 in. The water in the CRT is maintained at a specified temperature (180 F) during testing, and an alarm is sent to the operator if this temperature exceeds the high setpoint. l The automatic function of this control was not used during the matrix test program due to the small l RCS inventory that left the facility and the rather coarse controls of the condensate retum system. 2.6.15 Reactor Heater Sheath High Temperature Trip i Reactor heater rod sheaths are monitored for any increase in temperature above a 625*F setpoint by analog boards in the PLC. The reactor heaters are disabled if any of the following exceeds the 625 F setpoint. i a n Sheath temperature of two-out-of-four rods selected (TH101-4, TH319-4, TH103-3 TH104-4) (")
- Sheath temperature of the heater rod TH507-4 Sheath temperature of the heater rod TH501-4 Or if temperature of the SCR cabinet greater than 120 F Heater sheath temperatures of the same heater rods, as described in the preceding paragraph, are auctioneered at the IND-103 panel indicator. If any one of the heater rod temperatures exceeds 600 F, the operator is alerted by an alann.
2.6.16 Automatic Depressurization System Control 2.6.16.1 ADS 1, ADS 2, and ADS 3 The ADS-1, ADS-2, and ADS-3 valves are designed to open in sequence after CMT-1 (LDP-507) or CMT-2 (LDP-502) level is below the low setpoint of 41 in. and the ADS 1, ADS-2, and ADS-3 switches are in the AUTO position. The opening sequence timing of ADS valves is given in the following. The ADS logic is similar to AP600 logic. The ADS-1 valve, RCS-601, is opened after a delay of 15 seconds, when either CMT-1 or CMT-2 level is less than or equal to the low setpoint of 41 in. and the ADS-1 switch on the control panel is in the AUTO position. (g) v o:\ap600(1536w-3tnon:lt>081298 2.6-9 REVISION 1
I FINAL. DATA REPORT The ADS-1 valve, when in the AUTO position, is also actuated when reactor vessel upper-head pressure (PT-107 IND-107) is more than the high-high setpoint of 400 psig. The valve recloses when RCS pressure decreases below 360 psig. In this mode, Qe ADS-1 valve acts like a power-operated relief valve. The ADS-2 valve, RCS-602, is opened after a delay of 62 seconds when either CMT-1 (LDP-507) or CMT-2 (LDP-502) level is less than or equal to the low setpoint of 41 in, and the ADS-2 switch on the control panel is in the AUTO position.
'Ihe ADS-3 valve, RCS-603, is opened after a delay of 122 seconds, when either CMT-1 or CMT-2 level is less than or equal to the low setpoint of 41 in, and the ADS-3 switch on the control panel is in the AUTO position.
If any one of the ADS-1, ADS-2, or ADS-3 switches is in the OPEN position, it opens that specific ADS valve. 2.6.16.2 ADS 4-1 and ADS 4-2 If the ADS 4-1 switch is in the AUTO position, valves RCS-615 and RCS-625 are opened, after a delay of 180 seconds when either CMT-1 (LDP-507) or CMT-2 (LDP-502) level is less than or equal to the low setpoint of 41 in. and when either CMT-1 (LDP-507) or CMT-2 (LDP-502) level is less than or equal to the low-low setpoint of 17.14 in. Similarly, if the ADS 4-2 switch is in the AUTO position, valves RCS-616 and RCS-626 are opened, after a delay of 180 seconds from the time when either CMT-1 (LDP-507) or CMT-2 (LDP-502) level goes below the low setpoint of 41 in, and when either CMT-1 (LDP-507) or CMT-2 (LDP-502) level is less than or equal to the low-low setpoint of 17.14 in. l If any one of the ADS 4-1 or ADS 4-2 switches is in the OPEN position, it opens the associated ADS valves. 2.6.17 Steam Generator 1 Level Control The SG-1 controller maintains level in the SG above the U-tubes but low enough so that the generator moisture separator does not flood. The controller takes input from the narrow-range SG-1 level (LDP-303) and determines if the SG feedwater regulator valve, MF-11, should open or close to match the current steam demand. Since the steam-out controllt.r maintains a relatively constant steam flow by regulating the main steam valve, MS-008, the controller's main purpose is to make up for small variations and transients as well as allowing a method for filling the generators manually, if desired. O oMr6XA1536w.3a.non:ltw081298 2.6-10 REVISION 1
. . _ _ _ _ .._._..m. _ _ _ _ . _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _
FINA1, DATA REPORT O When the S signal is received,'the controller drives the output to O percent, thereby shutting the feedwater regulator valve.
' The water level in SG-1 is temperature-compensated by the correction factor:
(TCF) = T so 1 x (0.00049) + 0.94025 where: T so 1 = SG-1 temperature TF-301 De temperature-corrected water level is used in the PID level control loop and as an input to an alarm. A' control panel alarm alerts the operator if the compensated SG-1 level (LDP-301) is less than the low setpoint of 99 in. or above the high setpoint of 115 in. H2 0. A control panel alarm alerts the operator if SG-1 steam outlet pressure (IT-301) has exceeded the high setpoint of 310 psig. He SG-1 controller display monitor shows the steam flow (FVM-001) and feed flow (FMM 001) i mismatch. O 2.6.18 Steam Generator-1 Main Steam Valve The SG-1 main steam valve, MS-001, is a normally open valve provided with manual and automatic control. De valve is closed in manual control by placing the respective control panel switch in the CLOSED position. When the ccatrol switch is placed in the AUTO position, the valve remains open. 2.6.19 Steam Generator 2 Control . He SG 2 controller maintains level in the SG above the U-tubes but low enough so that the generator moisture separator does not flood. The controller takes input from narrow-range SG-2 level (LDP-304) and determines if the SG feedwater regulator valve, MF-012, should open or close to match the current steam demand. Since the steam-out controller maintains a relatively constant steam flow by regulating the main steam valve, MS-008, the controller's main purpose is to make up for small variations and transients and to allow a method for filling the generators manually, if desired. When the S signal is received, the controller drives the output to O percent, thereby shutting the feedwater regulator valve. v. o:Wism-knon:im1298 2.6-11 REVISION 1
l FINAI DATA REPORT l l l The water level in SG-2 is temperature-compensated by the correction factor: (TCF) = Tsa-2 x (0.00049) + 0.94025 where T sc-2 = SG 2 temperature, TF-310 The temperature-corrected water level is used in the PID level control loop and as an input to an alarm. A control panel alarm alerts the operator if the compensated SG-2 level (LDP-302) is less than the low setpoint of 99 in. or above the high setpoint of 115 in. H2 0. A control panel alarm alerts the operator if the SG-2 steam outlet pressure (PT-302) has exceeded the high setpoint of 310 psig. The SG-2 controller display monitor shows the steam flow (FVM-002) and feed flow (FMM-002) mismatch. 2.6.20 Steam Generator-2 Main Steam Valve The SG-2 main steam valve, MS-002, is a normally open valve provided with manual and automatic control. The valve is closed in manual control by placing the respective control panel switch in the CLOSED position. When the control switch is placed in the AUTO position, the valve remains open. 2.6.21 Main Steam Control Valve Control , The main steam control valve controller has two functions when it is in the AUTO position. First, it l takes input from both SG pressures (PT-301, PT-302), averages them, and maintains the position of the main steam control valve, MS-008, so that the average pressure is 285 psig during nonmatrix-test operation. When an S signal is detected, the average pressure maintained by MS-008 is 335 psig. l Second, break separator pressure (PT-902) from the BAMS controller via the microlink, is processed. If break separator pressure exceeds the high-high setpoint of 40 psig, the PLC provides a signal to open the containment pressure control valve, CSS-902. l l 2.6.22 Large-Break BAMS Control The BAMS was designed to measure two-phase flow indirectly. The system uses separators to separate the two-phase flow into single-phase liquid and single-phase steam flows for direct flow rate and temperature measurements. l 9 oAap600(1536w-3a.non:Ib-081298 2.6-12 REVISION 1
.. . . - ~ - - - - - . - - - . - . - - . . . . - . . . . . - - - . . . - .
FINAt. DATA REPCI:T The BAMS controller provides sevetal process functions as described in the following: 2.6.22.1 Break Separator Steam Isolation Valve The bicak separator 8-in. steam line isolation valve, CSS-906, is a normally open valve provided with ; manual and automatic control. The valve is closed in manual control by placing the respective control j panel switch in the CLOSED position. When the control switch is placed in the AUTO position, the ! valve closes when break separator steam flow (FVM-905 and FVM-906) decreases below 5000 cfm i after the initial steam flow spike from the LOCA. To set CSS-906 for automatic operation prior to a test, the following actions have to occur.
- 1. The valve is opened by selecting the OPEN position, and then the control switch is placed in the AUTO position. ;
i
- 2. When the break valve opens, the steam flow has to go above the low flow setpoint of 5000 cfm to reset the logic.
- 3. When the steam flow decreases to 5000 cfm, the BAMS controller sends a signal to the PLC and the PLC closes the valve.
O - After the initial burst of steam, the PLC controller uses an edge-triggered function to maintain the ; valve in the open position at the beginning of the test, when the total break separator steam flow (FVM-905 + FVM-906) is below 5000 cfm. The PLC is triggered to close the valve if the steam flow decreases below the setpoint of 5000 cfm. 2.6.22.2 Containment Pressure Isolation Valve The containment 10-in. pressure isolation valve, CSS 902, is a nonnally closed valve provided with manual and automatic control. The valve is opened in manual control by placing the respective control panel switch in the OPEN position. When the control switch is placed in the AUTO position, the valve closes when break separator steam flow (FVM-901 and FVM-902) decreases below 6000'efm after the initial steam flow spike from the LOCA. To set CSS-902 for automatic operation prior to a test, the following actions have to occur:
- l. The valve is opened by selecting the OPEN position, and then the control switch is placed in the AUTO position.
O owish3.!non:tb.os1298 2.6-13 REVISION 1
FINA1. DATA RErcOT
- 2. When the break valve opens, the steam flow has to go above the low flow setpoint of 6000 cfm to reset the logic.
- 3. When the steam flow decreases to 6000 cfm, the BAMS controller sends a signal to the PLC and the PLC closes the valve.
After the initial burst of steam, the PLC uses an edge-triggered function to maintain the valve in the open position at the beginning of the test, even though the total break separator steam flow (FVM 901
+ FVM-902) is below 6000 cfm. The PLC then triggers to close the valve if the steam flow decreases below the setpoint of 6000 cfm.
2.6.22.3 Containment Pressure Control Valve The contaiament pressure isolation valve, CSS-901, simulates backpressure in a post-LOCA condition by controlling the position of the valve. During the tests, the backpressure was a function of time and was different for each of the three breaks (2-in.,4-in., or DVI break). The BAMS controller simulated containment pressure during the breaks. The automatic containment backpressure control function was only used in Matrix Test SB19. For 2-in. and 4-in. breaks, the following equations were implemented: For 0 5 t 5 555 sec.; P = 2.753 x 10-2 xt O For 555 < t s 1056 sec.; P = 20.359 - 9.152 x 10-3 xt For 1056 < t s 1556 sec.; P = 10.695 For 1556< t s 4667 sec.; P = 16.042 - 3.437 x 10'3 xt For t > 4667; P = 0 psig where P = PT-902(psig) t = time (sec.) For DVI break, the following equations are implemented: For 0 s t s 77 sec.; P = 0.2265 x t For 77 < t 5 228 sec.; P = 19.48 - 0.2265 x t For 228 < t 5 530 sec.; P = 12.29 + 0.004934 x t For 530 < t s 1306 sec.; P = 19.79 - 0.009226 x t For 1306 < t s 3741 sec.; P = 11.88 - 0.003176 x t For t > 3741; P = 0 psig osap600u536w.3a.non:ib-os1298 2.6-14 REVISION 1
FINAL DATA REPORT O where P = I"T-902 (psig) t = time (sec.) 2.6.22.4 Condensate Return Pump IRWST Control Valve The condensate return pump IRWST control valve, CSS-928, simulates post-LOCA condensate makeup flow to the IRWST. The condensate retum pump primary sump control valve, CSS-927, simulates post-LOCA condensate makeup flow to the primary sump. He condensate makeup flow was a function of time and the following functions were applied: The fraction of condensate that went to the primary sump changed over time, and the fraction that was made up to the IRWST was equal to I-primary sump bha.. "g aons showing the primary sump fraction: 0 $ t s 25 sec.; fraction = 1 255 s t 5 600 sec.; fraction = 1.32 x time-o.167 600 $ t s 291I sec.; fraction = 0.5668 - 0.0001947 x time t > 291i sec.; fraction = 0 This fraction was multiplied by the total steam flow (FVM-901 + FVM-902) that was exiting the building. In the calculation process, all units were in Ibm /sec. Condensate flow proportioning was independent of the type of break that was initiated and therefore was always the same function of time. 2.6.22.5 ADS 4-1 Separator Steam Isolation Valve The ADS 4-1 separator steam isolation valve, CSS-905, is a normally open valve provided with manual and automatic control. When in the AUTO position, the valve opens coincident with the TEST pushbutton being pressed to initiate a test.
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J o:wt536w 3 anon: thost 298 2.6-15 REVISION 1
b s 7 l TABLE 2.6-1 y l PROGRAMMABLE CONTROLLER
SUMMARY
f E Centroller Inputs outputs Trip Signals Set Point tway Alarms f (see.) { Reactor heaters Hi 2 temperature (TF-140) Current output to SCR I Avg. hot-leg temp high 440 F none co Reactor waier temp. HL-I semperature (TF-141) Current output to SCR-2 high U SCR 1 power (KW-101) SCR-1 high .sutput power >396 none SCR-2 power (KW-102) SCR-2 high output power >3% none Total output power high >756 none Energency pushburton signal (Plf) Emergency pushbutton na noe - External trips: External trip: Heater Sheath 2/4 rod sheath sernp. high TR507-4/nt-501-4 temp. high 600 me KEY switch in OFT na w
- 5 signal (sest start) S signal na none p SCR Cabinet Temperature SCR cabinet temp h.ph (TSH@l) 120 none 6 PZR pressure high4igh trip PZR pressure high-high (PT404) 385 psig none i RCP coohng water km 11ow na mone O
without S signal (FSL-001) name Gland seat water temp. high wahout S 150 F none signal (TS@l) SGI SGI wide-range level (IDP-301) SGI level high/ low (Plf) SGI level high-high (WP.301) 117 in. none SGI level highnow SGI Icvel high (LDP-301) 115 in. none SGI level highAow SGI level low (LDP-301) 99 in. mme SGI widemge level (LDP-301) Control signal to valve motor (MF-SGI steam temperature (TF-301) O!!) SGI narrow-range level (LDP-303) S signal (Plf) Close MF411 na 3.1 SGI feed flow (FTMWI) Inour to feed pump trip kgic (Ptf) SGI steam flow (FVMWI) M l SGI Pressure (IT-301) SGI Eph Pressure Alarm 310 psig none SGI High Pressure j t- ~c; O N M z - 3 4 O O 9
m s.
/m \
N \ P O io sa s - x : 1 l TABLE 2.6-1 (Continued) '
$ l PROGRAMMABLE CONTROLLER
SUMMARY
h W Centroller Inputs Outputs Trip Signals Set Point Deany Alarmes (sec.) h a SG2 SG2 wide-range level (LDP-302) SG-2 level high/ low (PLC) SGI level high4igh (LDP-302) I17 m. none SG2 level high/ low SG-1 level high (LDP-302) 115 in. SG-2 level higMow SGI level inw (l.DP-302) 99 in. U SG I wide-range level (LDP-301) Control signal so valve motor (MF-SG2 temgrratwe (TF-310) 012) SG-2 nanewege level (I.DP-304) S signal (PLC) Close MF-012 na 3.1 SG-2 steam flow (FVM@2) Inpm to feed swmp trip logic (PLC) SG2 feed flow (FMM402) l SG-2 Pressee (M-302) SG2 High Pressure Alarm 310 psig none SGI High Pressine 9 BAMS Condensate now to sump (FMM-903) Connut signal to valve motw (CSS- na none 927) ' Condensate flow to IRWST Control signal to vatve motor (CSS- na none 4 (FMM-90s) 928) S signal (PLC) an none Steam now mam (6-in line)(TVM-901) Shut signal for air-operated valve low steam flow shw signal 6000 scfm none ! Steam flow mam (10-in line) (CSS-902) (CSS-902) (FVM-902) I Break steam flow (6-in hne) Indication total break steam flow Low steam flow shut signal 5000 scfm none Indicaten total tweak (FYM-905) (IND-905) } (CSS-906) steam flow (IND-905) ' Break steam flow (8-in ime) Shur signal for ar<5aared valve l (FVM-906) (CSS-906) Break separator pressure (M-902) BAMS header pressee high4 igh >40 psig none Break separator pressine (P-P. SOC) (M-9021 high-high (processed Open CSS-902 through steam out j comroller)
*s3 Break separator pressure (PT 'A)2) Control signal to valve motor (CSS- Open CSS-901 7 ;
901) 5 signal (PLC) $ na N none g l m
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O a u l10 3 l TABLE 2.6-1 (Continued)
$ l PROGRAMMABLE CONTROLLER
SUMMARY
[ Centroller inputs Outputs Trip Signals Set Point Delay Alarms P (see-) 8
$ PZR Hester PZR temperstwe GF408) SCR output Vohage PRZ heaters PZR Temp. GF408) >446 F none PZR heaters off Widemge PZR pressure (f"T404) PZR guessure high-high (kgic) PZR Pressure (PT404) High-High >385 psig none PZR level (LDP401) (RXC) PZR level low-low (LDP401) <30 in6es none Narrow-range PZR pressure (PT402) PZR level talm External trip-SCR cabines temp. high SCR cabtnet temp high (TSH001) >l20 F none (PLC. TSH 401) External inp:
S signal (PLC) *S* Signal na none Emergency pushburton Emergency pushbutton na none KEY swnch KEY switch na none PZR temperature (TF408) PZR level bar graph (BGI4. LDP401) Low Level <40 inches none PZR level high/ low PZR level (LDP401) High level >80 inches none Panel p WW-range PZR pressure (PT404) PZR pressure high/ low (panel alarm) Low Pressure 380 psig none PZR Pressure (PLC) High Pressure 360 psig none hightlow Panet 00 CVS and RNS RNS discharge flow (FMM-a05/3) Control signal to valve motor (RCS- PZR presswe (PT404). HI >230 psig None pumps RNS pump start signal (PLQ 811) PZR presswe (PZJtC)(PT446) CVS discharge flow (ntM-80lf U Centrol signal to valve motor (RCS. PZR Pressure (PT404) High-High >385 psig none CVS pump start signal (PLO 808) PZR pressure (PT404) Speed signal to CVS gump speed controller (an the field) Saram our SG-1 pressure (PT-301) Control signal to valve control motor 310 psig none SG-2 pressure (f"T-302) (MS408) S signal (Plf) Break separator presswe flow (BAMS Control signal to valve CSS-902 Break separator presswe high-high tnp >40 psig none controller PT-902) for CSS-902 to open SG-1 steam flow (FVM401) SG-I steam flow (SGI controller) *r3 SG-2 steam flow (FVM402) SG-2 steam now (SG2 controller) ~l Total stearn f:ow (FVM403) Display $ C to
'Al 5 m z 3 N
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p N d V E E $ TABLE 2.6-2 2 PROCESS CONTROL SYSTEM COMPONFETS , 5 i Component Mfr/Model No. Number Function Temperature scanner Omega CN34025-DC 2 Indicates temperatures of trace heaters; off-on control at adjustable setpoint Temperature scanner Omega CN10lK-1000F I Provides alarm if RCP seal cooling water exceeds adjustable serpoint Power meters Power Measurements . 2 Indicate power to each bank of rod bundle heaters 3710 ACM Bar-graph indicators m Universal VB-120-2-4 14 Indicate percent of rance for the following parameters: ; ADS 1-3 separator icvel ADS 4-1 separator level ADS 4-2 separator level Break separator level CRP level p Feed storage tank level ? CMT-1 level G CMT-2 level IRWST level ACC-1 level ACC-2 level Primary sump level l Secondary sump level Pressurizer level Panel switches and lights Square D 33 Provides CLOSF/ AUTO /OPEN control for air-operated valves, pump / fan power with STOP/RUN indicator lights. Alarms Panalarmm Series 90 I Alarm annunciator with 14 individuallights and a horn. Process indicatorsW Newport Model 82 31 Indicates value of process parameters: pressure, temperature, and flow. ; r Note: 3 I (1) "Ihe bar-graph indicators had setpoints programmed in for the appropriate process function alarms, contacts, and bistables. y y (2) The process indicators had setpoints programmed in for the appropriate process function alarms, c' ontacts, and bistables. 4 i e 3 - E t t 6
FINAL DATA REPORT l l 9
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FINAL DATA Rr. roar 2.7 Pre Test Operation t. Prior to every test, a series of operations is performed to ensure that every test is performed with the facility in a defined and documented condition. A fill and-vent procedute was performed using an j approved operating procedure. Then, a valve line-up (filling the accumulators, IRWST, and loop seals to the overflow levels) was completed using an approved operating procedure. The initial accumulator bets were established by filling the accumulator until flow was observed from an overflow standpipe. The standpipe was inserted and its elevation set during cold pre-operational facility testing, which ensured that the accumulators were set at the same level for each test. The accumulators were also pressurized to their initial pressure with nitrogen. Local pressure l indicators PI-401 and PI-402 indicated the proper pressure for accumulator-1 (ACC-1) and I accumulator-2 (ACC 2), respectively. The initial level for the IRWST was also established by filling the tank with water until water was l observed flowing in its overflow line. The filling was then stopped, and the tank was allowed to drain to the level of the fill tap of the tank. Similar to the accumulators, the fill standpipe level was set
- during cold pre-operational testing, ensuring that the level was the same for all tests.
l l During these initial system preparations, a check of the DAS was made for the proper response of , level and differential pressure transmitters. During heatup, the remaining channels of the DAS were I checked for proper response to changing system parameters. The facility was brought to normal operating temperature (420 F) and normal operating pressure (370 psig) using the reactor heaters in manual mode. The secondary side was brought to normal operating i pressure (285 psig) during heatup. The secondary pressure was controlled in automatic by the secondary steam pressure controller. Prior to test initiation, the reactor controller was placed in automatic control to maintain the average temperature of the hot legs (420*F). The reactor and steam j controllers interacted as they automatically controlled RCS temperature and secondary steam pressure. I I l When all prerequisites were met, and the facility 'was at the required temperature and pressure, a final valve line-up was performed from the control panel. The final preparations were conducted using the
" Initial Conditions" section of the test procedure. As pan of these initial conditions, the pressures and !
levels of the SGs, pressure and level of the pressurizer, and hot-leg temperatures were established j within tolerances specified by the procedure. When these parameters met the initial conditions, the l control board indication was recorded, and the TEST pushbutton was pressed. Per design,2 minutes elapsed between pressing the TEST pushbutton and generating the signal to ! open the break valves. Pressing the TEST pushbutton also initiated a signal for the DAS to begin ( recording data. The three racks of the DAS began recording data 12 to 14 seconds after the TEST l pushbutton was pressed. 'Ihe delay, due te the DAS communication time, was typical and did not c A 1536wRev l\l 536w-3a.non: I b-081798 2.7-1 REVISION 1
FINAL DATA REronT l affect test resultr. All DAS channels acquired steady-state data for almost 2 minutes before the break valves opened. One minute after pressing the TEST pushbutton, the operator opened CMT-1 balance line isolation valves RCS-529 and RCS-530. These valves are normally open in the AP600 plant; however, in an attempt to establish the required initial test conditions at the top of the CMTs (less than 80 F), these valves remsuned closed until 1 minute before break initiation. Control board data for initial conditions of the SGs, pressurizer, and hot legs were recorded in the procedure. Indications for SG and pressurizer levels were density-compensated by local controllers on the control panel. The same data were monitored and recorded by the DAS after the TEST pushbutton was pressed. However, during the time when initial conditions were being established, the DAS initialized and was awaiting a trigger to start. The DAS did not start acquiring data until after the TEST pushbutton was pressed. Thus, it was necessary to use control board indications to establish initial conditions. 1 k O e O I 2,72 REVISION I oA1536wRevi\l536w-3a non:Ib481798
FINAL DATA REPORT I 2.8 Drawings Physical characteristics of the test facility are documented in sketches that were prepared for several purposes, including the following: Westinghouse-prepared sketches defined hardware and instrumentation design requirements. Vendor drawings (TIC, Wright-Austin 6 Harris 'Ihermal Transfer Products) provided component fabrication details. Westinghouse-prepared sketches prepared at OSU documaned the as-installed features of the facility piping. P& ids prepared by OSU documented the interconnection of corrponents and instrumentation are
' included in Appendix G. The major systems and component draivings are included in Facility Description Report, WCAP-14124.m A review of the test data inCicated an inconsistency between the drawings and the facility condition. P&ID 600203 shows an orifice at the bottom of the ADS 1-3 separator. This orifice (ORI-659) was not in pbce for eidici me flow or matrix tests.
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1 FINAL DATA REPORT l l g 3.0 DATA REDUCTION (J ; 3.1 Introduction The following sections describe the data reduction and validation processes used for the low-pressure integral systems tests performed at Oregon State University (OSU). The data were transmitted in ASCII format on recordable compact disks (CD-R) from the OSU test site to Westinghouse. As part 3 of the data reduction and validation process, the OSU data files were nm through several l manipulations at Westinghouse to create usable data files and plots. Various data plotting methods were used to review, validate, and present the data. l l l
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n o:W1536w-4.non:Ib-081298 3-1 REVISON 1
FINAL DATA REPORT 3.2 Test Validation N./ The OSU test facility data were reviewed and validated using a three-step process. The first step was performed at the OSU test facility immediately following the test and was documented in the Day-of-Test Report. He Day-of-Test Report and the CD-Rs were sent to Westinghouse within 1 or 2 days of the test. All of the data from one test fit onto one CD-R. The Day-of-Test Report evaluated the test from a very basic standpoint, including operability of key instruments and deviations from specified initial conditions (that is, Did the test meet the minimum acceptance criteria?). De Day-of-Test Report also documented any facility modifications or onsite test observations. Specifically, the Day-of-Test Report assessed whether the test needed to be rerun because of some significant problem observed during the performance of the test. See Appendix A for a sample of the Day-of-Test Report format. The overall test acceptance criteria are shown in Table 3.2-1. Although not an explicit part of the pre- , test acceptance criteria, an overall facility mass balance objective of 10 percent was established. The critical instruments were the minimum set of instmments identified by the safety analysis personnel to perform a transient, component-by-component mass and energy balance. Critical instruments are listed in Table 3.2-2. The second step in the data validation process was performed primarily by the test engineering A personnel at the Energy Center in Pittsburgh, Pennsylvania. This step was performed after receiving V the Day-of-Test Report and processing the CD-R. This data validation was documented in the Quick Look Report (QLR). The Quick Look Report provided a preliminary validation of all test data (that is, Did the data meet all acceptance criteria?). A standard outline was issued to and accepted by the Nuclear Regulatory Commission (NRC) prior to writing the Quick Lock Reports. De key purpose of the Quick Look Report was to issue to the NRC some " pedigree" of the data, without specifically evaluating the data for code validation purposes (reviewed but not yet validated data) shortly after the test was performed. , The Quick Look Report examined the test in more detail than the Day-of-Test Report, including items such as calculation of an overall facility mass balance; deviations from specified initial conditions (heater rod bundle power decay); and identification of any out-of-range, suspicious, or failed instruments. As part of the Quick Look Report process, different types of instruments were reviewed to verify response of other instruments. The safety analysis personnel reviewed and signed the Quick Look Reports prior to issuance. His ensured that there was an understanding of how the test performed and determined if any unusual facility responses occurred. Preliminary data files on digital audio tape (DAT) were transmitted to the NRC with the Quick Look Report. A Quick Look Report was issued for each accepted test. Several tests were judged unacceptable and were rerun until an acceptable run was accomplished. Then a Quick Look Report was issued. 1 owl 5h4non:Ib.081298 3.2-1 REVISON 1
FINAL DATA REPORT The final step included a detailed review of the transient progression, facility and component performance, and cross-test comparisons. This report provides the final assessment as to whether the respective data are acceptable for code validation purposes. Figure 3.2-1 illustrates how the data validation process progressed from the OSU test site to the Energy Center. This figure shows the building-block approach followed in reviewing, evaluating, validating, and documenting the data. Each step in the process was based on the previous step, utilizing information and knowledge from the previous step, but was generally performed by a different set of personnel. As the process progressed from left to right, the steps evolved from problem identification to problem resolution. This three-step process allowed various personnel from different disciplines to review the data prior to final publication. These steps were followed in order to provide a high level of data quality assurance. O O o%*atN-4-ib4)s1298 3.2-2 REVISON 1
FINAL DATA REPORT TABLE 3.21 i t E' OVERALL ACCEPTANCE CRITERIA 4
+
Test initial conditions shall be achieved in a specified tolerance. Setpoints shall be achieved ira an acceptable tolerance band.
+
Sufficient instrumentation shall be operational before the test (exceptions shall be approved by the
- Westinghouse test engineet).
Critical instruments not operating shall be identified to the Westinghouse test engineer before the tests. Dese instruments must be operational before and during the test, or exceptions should be approved. l A zero check of LDPs, DPs, and FDPS shall be in acceptable tolerances. He zero check was eliminated from the acceptance criteria for Category III tests. The earlier pre- ! test and post-test checks of zero shift showed acceptable variation in the readings of these instruments. Performing these checks required that each instrument be manually isolated and then
- returned to service. Based on the consistency of the readings from earlier tests and the large number of manual operations, it was decided that the risk of an instrument remaining isolated after the check j was greater than an instrument having a zero shift.
[ l' l l l I l l t 4 N' % o:W1536w-4.non:Ib-081298 3.2-3 REVISON 1 f I
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FtNAt. DATA REPCCT TABLE 3.2-2 CRITICAL INSTRUMENT LIST Channel ID Description Vessel PT-107 Pressure at top of vessel head DP-130 Pressure difference between top of vessel head and top of downcomer LDP-127 Level measurement from bottom of vessel to bottom of upper head LDP-140 Level measurement from top of downcomer to bottom of vessel TF-105 or TF-107 Fluid temperature measurement in CL-1 TF-106 or TF-108 Fluid temperature measurement in CL-2 TF-101 or TF-103 Fluid temperature measurement in CL-3 TF-102 or TF-104 Fluid temperature measurement in CL-4 TF-120 Fluid temperature at bottom of vessel upper head TF-141 or TF-143 Fluid temperature measurement in HL-1 TF-140 or TF-142 Fluid temperature measurement in HL-2 Total of four heater rod thermocouples Temperature near surface of heater rod Pressurizer PT-603 Pressurizer pressure LDP-601 Pressurizer liquid level TF-605 or TF-608 Pressurizer liquid space fluid temperature measurement O l opish-4.non:itso81298 3.2-4 REVISON 1 I f 1 m _
FINAL DATA REPORT TABLE 3 2 2 (Continued) CRITICAL INSTRUMENT LIST Channel ID Description SG 1 Primary Side FT-201 Primary-side pressure at top oflong U-tubes LDP-215 Pressure difference along hot side of SG-1 long U-tube LDP-219 Pressure difference along cold side of SG-1 long U-tube FMM-201 CL-1 mass flow rate; mynetic flow meter FMM 203 CL-3 mass flow rate; magnetic flow meter SG-1 Secondary Side PT 301 SG-1 secondary side pressure, main steam line FMM-001 Feed flow to SG-1; magnetic flow meter FVM-001 SG-1 steam flow; vortex flow meter LDP-301 SG-1 secondary side level SG 2 Primary Side FT 204 Primary-side pressure at top of long U-tubes LDP-218 Pressure difference along hot side of SG-2 long U-tube LDP-222 Pressure difference along cold side of SG-2 long U-tube FMM-202 CL-2 mass flow rate; magnetic flow meter FMM-244 CL-4 mass flow rate; magnetic flow meter SG 2 Secondary Side PT 302 SG-2 secondary-side pressure, main steam line FMM-002 Feed flow to SG-2; magnetic flow meter FVM-002 SG-2 steam flow LDP-302 SG-2 secondary-side level f owir ' n-4.non:Ib-081298 3.2-5 REVISON 1 l
FINAL DATA REPORT TABLE 3.2 2 (Continued) CRITICAL INSTRUMENT LIST Channel ID Description ACC-I IT-401 Tank pressure LDP-401 Level transducer FMM-401 Flow measurement magnetic flow meter TF-401 Liquid temperature at accumulator discharge ACC 2 I'T-402 Tank pressure LDP-402 Level transducer FMM-402 Flow measurement; magnetic flow meter TF-402 Liquid temperature at accumulator discharge CMT1 l'T-501 Tank pressure LDP-507 Tank level measurement FMM-501 CMT discharge line flow rate measurement TF 501 and TF-529 CMT fluid temperatures from long rake CMT-2 IT-502 Tank pressure LDP-502 Tank le"el measurement FMM-504 CMT discharge line flow rate measurement TF-504 and 'IT-532 CMT fluid temperatures from long rake O c:\ap6GA15.16w-4.non:lN21298 3.2-6 REVISON 1
}
FINAL, DATA REPORT TABLE 3.2 2 (Continued) CRITICAL INSTRUMENT LIST Channel ID Description IRWST/PRHR HX FMM-802 Flow rate into PRHR HX; magnetic flow meter LDP-802 Level measurement in PRHR HX FMM-804 Flow rate of PRHR HX; magnetic flow meter LDP-701 Level measurement in IRWST FMM-701 Flow measurement. IRWST to DVI line-1; magnetic flow meter FMM-702 Flow measurement, IRWST to DVI line-2; magnetic flow meter TF-701 and TF-709 Fluid temperature measurements in IRWST adjacent to PRHR HX ADS 1, ADS-2, ADS 3, and ADS-4 Actuation FDP-605 ADS-1 actuation line flow; differential pressure cell FDP-604 ADS-2 actuation line flow; differential pressure cell FDP-606 ADS-3 actuation line flow; differential pressure cell IYT-605 Pressure measurement on ADS-1, ADS-2 and ADS-3 separator FVM-601 Steam flow from ADS-1, ADS-2 and ADS-3 separator; vortex flow meter FMM-601 Liquid flow from ADS-1, ADS-2 and ADS-3 separator; magnetic flow meter FT-611 Pressure in the ADS-4 separator for the loop 1 FMM-603 Flow rate from the ADS-4 separator for the loop 1; magnetic flow meter FVM-603 Flow rate from loop 1 ADS-4 separator; vortex flow meter FT-612 Pressure in the ADS-4 separator for the loop 2 FMM-602 Flow rate from the ADS-4 separator for the loop 2; magnetic flow meter FVM-602 Flow rate from the ADS-4 separator for the loop 2; vortex flow meter l owsm4non:lb-081298 3.2-7 REVtSON 1
FINAL DATA REPORT TABLE 3.2 2 (Continued) CRITICAL INSTRUMENT LIST Channel ID Description Sump, Power, and Other Instruments FT-901 Pressure in primary sump vessel LDP-901 Level measurement in primary sump vessel LDP-902 Level measurement in secondary sump vessel PT-905 Pressure of break flow separator LDP-905 Levels measurement for break flow separator FMM-905 Flow measurement from break flow separator to primary sump; magnetic flow meter FVM-901 Flow measurement in BAMS, vortex flow meter FVM-902 Flow measurement in BAMS, vortex flow meter FVM-903 Flow measurement in BAMS, vortex flow meter FVM-905 Flow measurement from break flow separator, vortex flow meter FVM-906 Flow measurement from break flow separator, vortex meter TF-916 Fluid temperature measurement in BAMS exhaust line TF-917 Fluid temperature measurement in BAMS exhaust line KW-601 Watt meter for power to the pressurizer KW-101, KW 102, KW-103, and Watt meters for power to core simulation KW-104
)
i l
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O awams36w-4 non.tb-08:298 3.2-8 REVISON 1
FtNAt. DATA REPORT STEP 3 (V) Perform evaluation of parametric effects, i.e., break sizes Provide and review cross-plots of data between tests Perform data uncertainty analysis Perform transient and component STEP 2 evaluation Perform simple benchmark Perform system-wide transient evaluation on red bundle evaluation Perform overall mass balance Confirm overall mass balance STEP 1 Provide minimum facility design Provide adequate facility design information information Identify if test met minimum Identify if test met ALL acceptance Identify all tests performed, valid and acceptance criteria criteria invalid, and reason for validity Provide unqualified data files for Provide unqualified data files for Provide qualified data files with zero all tests acceptable tests only time shift (valid data only) Provide key data plots Review key multichannel plots, e.g., Provide key multichannel plots, e.g., HL-1 versus HL-2 HL-1 versus HL-2 (v) Identify anomalous data Identify / resolve anomalous data Identify / resolve anomalous data Provide test observations Provide test observations Provide and discuss test observations; resolve problems Provide sequence of events Compare actual and specified Resolve differences in actual and sequence of events specified sequence of events Identify procedure deviations Identify procedure deviations; briefly Resolve procedure deviations; describe describe procedure test procedure Identify facility identify facdity Identify facility configuration / modifications configuration / modifications configuration / modifications Check key initial conditions Check all initial conditions (ICs) Resolve deviations in ICs Identify inoperable instruments Identify inoperable and erratic Compare inoperable instrument lists instruments between tests Check critical instruments Check all instruments Verify all instruments Day-of-Test Report at OSU test Quick Look Report at Westinghouse Final Data Report at Westinghouse site Energy Center Energy Center f3
%/
l ( l Figure 3.2-1 Data Documentation Steps oAap600\l536w-4 non:lb-082698 3.2 9 REVISON 1
FINAI. DATA REPORT l l 3.3 Pre-Operational Tests l The pre-operational tests were segregated into cold and hot tests. The purpose of the cold pre-operational tests was to determine facility characteristics such as tank volumes and component flow resistances using cold water. The tank volumes and flow resistances were calculated directly by OSU test site personnel. He cold pre-operational test data were provided by OSU in hard-copy form only. The report issued from OSU was reviewed at Westinghouse and, subsequently, transformed into a Quick Look Report format. His Quick Look Report was issued to the NRC. See Subsections 4.1 and 4.2 for a description of results from the cold pre-operational tests. l
- l. The purpose of the hot pre-operational tests was to determine facility characteristics such as the facility heat losses and verify operation of the facility under hot operating conditions. The hot pre-operational test data were provided on CD-R. The hot pre-operational test data were generally processed and reviewed in the manner described in Subsection 3.4. The data files for the final hot pre-operational test (HS01) are provided with this report. He data files for the other two pre-operational tests (HS02 j and HS03) are not transmitted in this report, since they were not needed for code validation purposes.
See Subsection 4.3 for the description of results from the hot pre-operational tests. G 4 i l V l l l l l l l l a t !O o:Wis36 4.non:ib-ost29s 3.3-1 REVISON 1 1 l i
. - - . . . - - -- - -- - - - - - - . - . - ~ . - - - . _ - . - - . ~ . - FINA1, DATA REPORT l 3.4 Matrix Tests S
.)
All data for the matrix tests were provided by OSU on CD-Rs in 12 files. Dese 12 files contained the data for the three data acquisition system (DAS) racks with each rack segregated into two parts, burst scan and continuous scan data files. The LTCT number (run and matrix number) was incorporated into the filenames. The data provided by OSU were in English engineering units with both a clock time in HH:MM:SS and a zero test start time. Zero time was dermed as the time of break valve opening. The DAS was started about 2 minutes in advance of the test initiation; therefore, the data files included about 120 seconds of negative test time. In addition to the 12 data files, a configuration file for each of the three DAS racks was provided on the CD-R. De configuration file contained the calibration information for the instmments. A series of data reduction steps was developed and implemented to process the data. The overall data l reduction process is illustrated in Figure 3.4-1, The process followed for the Quick Look Reports was l slightly different than that followed for this report because of some issues discovered during testing and data review. The key differences included resetting the data to the correct test initiation time (zero time), merging the burst scan and continuous scan files, correcting the zero settings for some pressure transtnitters, and correcting the calibration files for several instmments. These steps are briefly described in the following three paragraphs. p The zero time assigned by the DAS at OSU was off generally by about 10 seconds; therefore, the data files were adjusted for the correct zero time. Several channels were reviewed for each instrumentation rack to ascertain the precise time that the break was initiated. The respective time correction was then added to each time stamp in each of the files. See Appendix A for more details on this data reduction step. The burst scan and continuous scan files were merged to form a hybrid data file. This hybrid file provided the most practical form for the data. The burst scan data were copied into the continuous scan data file for the overlapping time period, generally between the first 3.0 to 60 minutes of the test. 4 Thus, the hybrid file contained data with a scan rate of [ Ja.b.c for the first 30 to l 60 minutes and a [ . ]a.b.c scan rate for all data thereafter. An evaluation was performed I on the differences between the burst scan and continuous scan data. It was found that the data were very similar, except for the spikes in the burst scan data, which were recorded due to the faster scan rate. For the Quick Look Report, only the continuous scan data files were reviewed and provided to the NRC (on DAT). l l After most of the tests were completed. it was discovered at the site that several of the pressure ! transmitter zero settings were incorrectly set, and the calibration files for nine instmments were i incorrect. These were corrected in the processing of the data for this report. ! ne DAS at the CS'J test facility assigned a noninteger character to those data fields that exceeded the typical imument ranges. These nonin:cger characters could not be handled in the Westinghouse 03'jMwRevl\l5%w-4.non:Ib-081298 REVISION 1 3.4-1
1 FINAL, DATA REPORT 1 l 1 l plotting package (called NSAPlot); therefore, they were replaced with integer values. The noninteger characters were generally attributed to those occasional times when the instrument was out of range. l This would not prohibit the data from being validated. After these corrections were made to the data files, the data were plotted by NSAPlot, which is capable of plotting the data in a variety of ways. A set of single- and multiple-channel plots was produced and issued for internal Westinghouse review. The multiple-channel plots generally included data from the different loops (for example, from the four cold legs or the two hot legs) or a family of instruments from a key component like the core makeup tank (CMT). De plots also contained 20-character instrument descriptions and locations. He procemed data file was made available for use I to safety analysts at Westinghouse. However, the data (both hard-copy plots and electronic files) for the blind test (Matrix Test SB09) were controlled to make them unavailable to safety analysis personnel. I 1 After the plots of all data channels were distributed for internal Westinghouse review, two sets of calculations were produced to help validate the data. The first set calculated the aterage initial conditions for channels important to establishing facility steady-state conditions prior to test initiation, , such as pr-ssurizer pressure and hot-leg temperature. Comparisons of this report to specified initial l I conditions and allowable tolerances were performed and reported in the Quick Look Report and Section 5 as a table in the review of each matrix test. Any deviations outside the allowable tolerance , were generally quite sraall and acceptable. The deviations from the established tolerances were accepted due to the tighter-than-needed tolerances imposed in the test procedure. He restrictive test tolerances were established to achieve tests conditions that were, as much as possible, repeatable from test to test. In addition to the calculation of the initial test conditions, the actual and specified heater rod power decay curves were also generated and plotted on the same graph. These curves were reviewed and are included in Appendix F of the Quick Look Report for both power supplies (KW-101 and KW-102). The redundant power supply measurements (KW-103 and KW-104) were also plotted against the specified power decay curves. The actual sequee of events for each test was tabulated, reviewed, and compared to the expected sequence of everns in the Quick Look Report. This sequence of events included pump trip times, automatic depressurization system (ADS) actuation times, CMT low-low level times, etc. This sequence of events review provided a quantitative assessment on the test performance. An expanded sequence of events is included in this report in Section 5. A table and bar chart are included for the review of each test. As pr.rt of the Quick Look Report data review process, an overall facility mass balance was performed. This mass balance consisted of calculating the water masses in each of the major components prior to the test, calculating the water masses in each of the same components after the test, and calculating the o:\l 536w Rev lu S 36w.4.non: l tro81798 3.4-2 REVISION 1
FINAL DATA RErc:T f percent difference between the two time periods. The percent difference included only the
\ in-containment refueling water storage tank (IRWST) water injected into the system, since this water contributed directly to the cooling of the heater rod bundle. Using the water remaining in the IRWST would have distorted the calculations of the percent difference. This facility mass balance provided a quantitative measure of confidence in the response of the component differential pressure transmitters and catch tanks and overall facility performance during the test. If the overall mass balance had been significantly affected (that is, greater than 10 percent), the test would have been repeated. 'Ihe overall mass balance results are reported in Appendix E. The results of a transient mass and energy balance will be reported in the AP60u Low Pressure Integral Systems Test at Oregon State University Test .
Analysis Report, WCAP-14292.* As the last step in the Quick Look Report data review process, the fluid levels in the heater rod bundle, upper plenum, upper head, and hot legs were reviewed. These fluid levels were then evaluated with respect to the actuation and termination of the safety systems (such as accumulators, CMTs, and IRWST) to determine if the respective differential pressure transmitter response was reasonable. For example, Was there any impact of the accumulator injection on the fluid level in the heater rod bundle? This evaluation, however, did not include the effect of the density difference as the heater rod bundle cooled down. A similar evaluation of the fluid levels in the heater rod bundle is contained in this report including density compensation for differential pressure transmitter. O For this report, several additional steps were performed to validate the data, including a detailed review of 'he tiansient progression; a review of the performance of each of the major components; a detailed comparison of the different tests to ascertain effects of break size, break location, containment backpressure, and nonsafety system operation; and performance of an instrument error analysis. For this report, wherever appropriate, the saturation temperature corresponding to the component pressure (or system pressure) was plotted on temperature graphs to help understand the response of the respective thermocouples. O o:\l5hRevi\l5%w.4.non:Ib-081798 3.4-3 REVISION 1 l-l
FINAL DATA REPORT l O' Data Files on CD-R Issued by OSU
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Data Files Processed by Westinghouse
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Data Plots Produced by NSA Plot
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Data Plots and Files Issued for Review
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Data Report Written and Signed-off l Figure 3.41 Steps in OSU Data Processing O c:\l536wRevl\l536w4.non:Ib 081298 3.4-4 REVISION I L
FmAL DATA REPORT (N 3.5 Instrumentation Error Analysis The instrumentation error associated with the data from the OSU test series was derived from the equipment manufacturer's specifications, and these specifications were used to compute the error estimate for the data path. Component calibrations were performed to verify that the actual equipment performance was consistent with the manufacturer's specificaticns. The data path for the OSU instruments consists of three parts: sensor, signal conditioner, and data acquisition system (DAS). In the case of a thermocouple, a signal ccaditioner is used only when its value is also displayed on the control panel. In most cases, thermocouple signals are fed directly into
. the DAS.
'Ihe errors due to the transmission wires between the elements and power supply variations were not included in this analysis; these errors are typically very small ( 0.001 percent) in comparison to the element errors and are considered negligible.
3.5.1 General This section provides an estimate of the instrumentation error associated with the OSU test facility, Test temperatures were generally measured using type-K thermocouples. Type-K thermocouple installation is discussed in Subsection 2.4.1 by individual component. Installation of the remaining instrumentation, such as pressure transmitters and flow meters is also discussed in Subsection 2.4.1 by type of instrument. These instruments were installed consistently throughout the facility. In general, the error associated with the measurement's sensor or transducer and the signal conditioning and DAS equipment was considered in determining the total error for any given data channel. Because the transducer or sensor is normally dominant, total error and sensor error are typically very similar. System and/or component calibration data were used, when available, to calculate an estimate of the
.overall error for each data channel; multiple sources of error were combined by summing the variances for each component of error.
Individual sources of error were combined in accordance with: m E= { (E,)2 m , o:\l 5%w Rev l\l 5 Mw-4.non: I b-081798 3.5-1 REVISION 1 __ _j
FINA1. DATA REPORT where: l l E = estimate of data channel enor or total probable error E = value of i* source of error component maximum error 3 N = number of sources of error All significant bias error is assumed to be accounted for using the appropriate corrections obtained from equipment calibrations. Where appropriate, when calibration data were not available (such as the l case in which calibration was performed only to verify compliance with the manufacturer's j specifications), the manufacturer-specified error was used. In these cases, the manufacturer-specified ) error was considered to be the maximum error and assumed to be uniformly distributed over the error i interval. Since the variance of a uniformly distributed random variable over the interval -a < x < a is
" standard deviation of the maximum error was calculated from the maximum instrument e: Tors by:
3 v1 N i ! E= { (E)2 l
.M i
where: l E = standard deviation of maximum error Ei = maximum error of component i l N = number of individual sources of error l l l For the OSU test facility, the calculated errors for individual channels are tabulated in Appendix D, Tables D-1 through D-9.
~
LIST OF TAllLES Table No. IIcader Page No. 1, D-1 Errors for DPs D-3, D-4 D-2 Errors for LDPs D-5, I)-6, D-7 D 3, Errors for Pfs D-8 D-4 Errors for FMMs D-9 D-5 Errors for FVMs D-10 D-6 Errors for Thermocouples D-11, D-12. D-13, D-14, D-15, D-16, D-17, D-18, D-19, D-20 D-7 Errors for HFMs, HPs, and LC D-21, D-22 D-23, D-24 D-8 Errors for FDPS D-25 D-9 kW Errors D-26 I o:\l 536w Rev l\l 536w-4.non:l t481798 3.5-2 REVISION 1 N
FINAL DATA REPORT q- The standard deviation of the maximum error is not tabulated although it can be calculated from the calculated errors in Tables D-1 through D-9. 3.5.2 Dennitions The error analysis used standard statistical terms and methods to calculate the instrument errors. Some of these terms and their definitions are:
- Error -- Manufacturer's specified maximum deviation from mean value of the measured parameter.
- Total probable error -- Calculated error; defined as the square root of the sum of the squares of all errors. This is the expected most-probable error for a particular type of instrument and its channel.
3.5.3 Results The probable error for each of the OSU test facility instruments is provided in Appendix D, Tables D-1 through D-9 along with the individual component errors. O
\
'Ihese instruments were also calibrated upon completion of the matrix testing and, in some cases, the measured error was found to be largr than the reported probable error. This would be expected because, according to statistics, there is always a chance that a measurement will fall outside the total probable error. Therefore, the calculated probable errors are reported consistently for use in evaluating measurement uncertainty.
Based on the reported probable error, the OSU test facility instrumentation satisfied the accuracy requirements specified in the test specifications (12 percent). _ oA15hRevl\l536w-4.non:lb-081798 3.5-3 REVISION 1
- m. _ > ._ . _ _ _ _ _ _ . _ _ . . _ _. _ _ _._-- _.._ _...._.._. _ _ - . _ _ . . _
FNAL DATA REPORT i l 4 3.6 -Zero-Time Shift File Correction The test data files from the test facility DAS contained event times referenced to the opening of the
=
l break valve. Test evaluation and analysis required that a correction be made to the test files to compensate for a zero-time shift of the data. The correction was performed using data from the ! uncorrected test data files sent fmm OSU. His section describes the method used to correct the zero- !~ time shift of the data and to document the results of the correcti on. Time corrections for each test are included in Table 3.6-1. L 3.6.1 Test Data Collection Timing Test data were collected at the test facility by three computers, referred to as racks. Each computer, or l rack, communicated with and acquired data from separate portions of the DAS, data acquisition system which interfaced directly with facility instrumentation. A matrix test was initiated 120 seconds after pressing the TEST pushbutton on the control panel. When the TEST pushbutton was pressed, the i microprocessor-based timer of the facility's programmable logic controller (PLC) was started. When the timer timed out 120 seconds, the PLC generated an open signal for the break valve, which started the transient and the test. Pressing the TEST pushbutton also generated a control signal that was ! ' immediately sent to each of the three computers. This control signal, processed separately by each computer, commanded the three separate subsystems of the DAS to begin acquiring data. The signal l process time of the computer and data acquisition subsystems created a time delay between the l actuation of the TEST pushbutton and the start of data acquisition. DAS testing demonstrated the time delay was not the same for the three racks and was not a constant between tests. l l De data acquisition design assumed that the first time stamp of every test data file occurred 120 i seconds before the break valve opened. Thus, when the data was reduced, an assignment of time = 0 was given to the data obtained 120 seconds after the first time stamp, presuming data acquisition started immediately when the TEST pushbutton was pressed and the break valve opened 120 seconds later. However, due to the time delay between pressing the TEST pushbutton and the time of the first time stamp, the break valve did not open 120 seconds after the first time stamp in the test data file. A zero-time shift error was introduced because a time = 0 in the file corresponded to a real time later than the opening of the break valve. The magnitude of the test data file's zero-time shift was equal to l the delay between pressing the TEST pushbutton and the start of data acquisition by the corresponding rack. De validated test data files received from the test site contained this error in the files' time references. Although the errors in the time references were known before the final data validation was performed at the OSU test facility, a decision was made to correct the files' zero-time shifts at the l Energy Center. l l l 3.6.2 Time Correction Method l b A method was developed to correct the zero-time shift of each OSU test data file and is described in . this section. When the correction factor for a given file was determined, the time was added to each oA15%wRevl\l536w-4.non:Ib-061798 3.6-1 REVISION 1
FmA1 DATA REPORT time entry in the file. Once the corrections were completed, time = 0 in the file corresponded to the time an open signal was sent to die break valve, and a zero-time shift of the data was eliminated. Within 10 seconds after the break valve opened, the control logic created a trip of the main feed pumps, reactor coolant pumps (RCPs), pressurizer heaters, and the reactor heaters. In addition, the passive residual heat removal heat exchanger (PRHR HX) outlet valve opened and the main steam controller was reset, thereby closing the main steam control valve. In each instance, the operation of the equipment affected system parameters that were monitored by test instrumentation and recorded by the DAS. In addition to being programmed to open the break valve 120 seconds after the test pushbutton was pressed, the PLC was programmed to open the PRHR HX outlet valve, trip the pumps, and trip the heaters after a specified time delay. He operation of the equipment was monitored by a software program that recorded the inputs and outputs to the PLC. Dere were no time shifts in the data recorded for the PLC. Using the Wonderware software program (Subsection 2.5.5), which monitored the inputs and outputs of the PLC, a difference in time between the open signal of the break valve and the operation of the other equipment was determined, using the open signal of the break valve as time = 0. The uncorrected test data files from the test facility were then reviewed to determine the time a specific system parameter was affected by the equipment operation. The time recorded from the test data files was referenced to the incorrect zero-time reference. A comparison was made between the correct time from the PLC and the incorrect time from the test data files. The difference between these times provided the time added to the test data files to correct the time shift. The correction process applied to Matrix Test SB01 is provided as an example. The correction data for Matrix Test SB01, which has the filename U0001, is in Table 3.6-1. The first column is labeled Design Time Delta T (sec.). This is the time the equipment operated in the test, using the actual time an open signal was sent to the break valve as time = 0. The Parameter column is a description of the j equipment operation that creates a change in the system. The two columns, Instrument Channel and l Instrument Channel Description, are the channel nu.mber and description of the channel that monitors the change in system parameter affected by the equipment operation. The two columns are Time Stamp of Change (sec.). Review of the data plot and actual data from the instrument channel provided the time the parameter changed in value as a result of the equipment operation. The time l stamp was obtained from the test data file that was referenced to the incorrect time = 0 of the test data l file. The Earliest column is the last time stamp before the system parameter changed, and the Latest i column is the first time stamp when a change in the system parameter was noted. The last column is Time Added to File to Correct Zero Time (sec.). The entry in this column is the average of the Earliest and Latest data entries added to the design time from column 1. When the table was completed, a time correction was selected for each rack. The time selected was the largest entry in the Time Added to File to Correct Zero Time (sec.) taken to the nearest data scan. The data scan rate varied for each test, but was either once per second, once per 2 seconds, or once o:\l 536wRev i\l 536w-4 mon: l b-081798 3.6-2 Revision 1
. - - . - _ - - _. - _ -. . - - _~ - . . . . . . _ . ~ . . . . - . - . - . _ . . _ . _ . . _ . . . - -
l l FINA1. DATA REPORT A
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per 4 seconds. When a time correction was selected, it was entered in the data sheet for the appropriate rack. l The selected zero-time correction was added to each of the OSU test data files originally validated at the test facility. The new conected files had a time = 0, which corresponded to the actual time of the open signal to the break valve. The first time stamp of each data file varied between
'[ ] for the largest time correction of [ ]'b' for the smallest time correction of [ ]'6#
( o rw U oA1536wRevi\l 536w4.non:I b-081798 3.6-3 REVISION 1
FINAL DATA REPORT Table 3.6-1 on pages 3.6-4 through 3.6 20 is not included in this nonproprietary document. 1 I i G o:\l536wRevi\l536w4.non:Ib481798 3.6-4 REVISION ] l
FINA1, DATA REPORT l 4.0 Pre 4perational Test Results Pre-operational tests were performed to assure that the facility was performing and recording data in a manner consistent with the expectations for the matrix tests. Several of these tests were performed to chalacterize the physical attributes of the facility including component volume measurements, pressure drop determinations, and a series of hot functional tests. l l l i l l i l l 1 l- f ( l l L b l t i l. oA15hRevl\l5h-9.non:Ib-08129s 41 REVISION l
FINAL DATA REPC::T
.4.1 Cold Volume Determinations V
This section reports the following volume tests: L
- OSU-V-01 Accumulator Volume Test, performed on 1/11/94 (Subsection 4.1.1)
- OSU-V-02 Core Makeup Tank (CMT) Volume Test, performed on 1/14/94 and 1/16/94 (Subsection 4.1.2)
- OSU-V-03 Pressurizer Volume Test, performed 1/18/94 through 1/20/94 g (Subsection 4.1.3)
- OSU-V-04 In-Containment Refueling Water Storage Tank (IRWST) Volume Test, performed on 1/30/94 (Subsection 4.1.4) l
- OSU-V-05 Primary Sump Tank and Secondary Sump Tank Volume Test, performed on 2/4/94 and 2/5/94 (Subsection 4.1.5)
- OSU V-06 Steam Generator-1 (SG-1) and SG-2 Secondary-Side Volume Test, performed from 1/20/94 through 1/27/94 (Subsection 4.1.6)
>s l * ' OSU-V-07 ' Automatic Depressurization System (ADS) and Break and ADS Measurement System (BAMS) Moisture Separators Volume Test, performed on 2/6/94 l (Subsection 4.1.7)
- OSU-V-08 Reactor Vessel Volume Test, performed on 2/5/94 (Subsection 4.1.8)
The objectives of these tests were to measure the volumes of the major components of the test facility and to. compare these volumes with the designed volumes. All tests were performed in accordance with an approved procedure. Specific objectives and test results for each test are discussed in detail in each subsection. 4.1.1 Accumulator Volume Test The objectives of this test were to determine: i
- Total volume and gas volume of accumulator-1 (ACC-1) and ACC-2 f Note: Only the cylindrical shell contained water. The 2:1 elliptic upper head and part of the
'A . upper cylindrical shell containeA gas. N.Y oA15%wRevi\l5%w 9.non:ltN)81298 4,]-] REVISION 1
FINAt, DATA REPORT a
- Average cross-sectional area of the cylindrical section of ACC-1 and ACC-2 without the standpipe installed
- Insertion depth of the standpipe for each accumulator to verify that the water volume contained inside each accumulator was within the acceptable range [ ]a.b.c Note: Design water volume was [ la.b.c for each accumulator.
- Average cross-sectional area of the cylindrical section of each accumulator with the standpipe installed 4.1.1.1 Test Procedures Figure 4.1-1 provides a schematic of the test setup. A general description of the procedures is provided here.
Each accumulator was filled with water through its accumulator fill isolation valve using a temporary line connected to the normal residual heat removal system (RNS) pump. When the accumulator was about 50 percent full, the accumulator direct vessel injection (DVI) isolation valve was opened to remove any trapped air in the accumulator injection line and then closed. The accumulators were filled until a steady stream of water issued from a temporary vent valve installed at the accumulator pressure relief valve (temporarily removed) connection. Water temperature in the accumulator was measured. Accumulator volume was determined using a precision scale to weigh the water drained from the accumulator. The draining and weighing process was performed in steps. The measured water weight was converted to a water volume using the measured temperature to obtain the water density. Also, a temporary tygon level indicator was installed along the tank to provide a visual indication of the tank water level. The tygon level indicator measured the accumulator volume without the standpipe installed. The standpipe ensured the same water volume and level in the accumulator each time it was filled. The standpipe was inserted to a predetermined depth, and the accumulator was filled completely again. The accumulator was then drained via the standpipe, and the water drained out of the standpipe was weighed again using the precision scale. The water weight was converted to a water volume that represented the water volume drained out of the accumulator. The remaining water volume in the accumulator was calculated using the previously measured overall water volume and compared with
- the acceptance criteria. This process was repeated, as necessary, to get the proper standpipe insertion height and water volume.
l l l As an initial condition of the test, the calibration of the precision scale was checked against a certified weight. At the end of the test, the scale was checked again using the same certified weight to verify the calibration of the scale reraained valid during the test. l c:\l 536w Rev l u S36w.9.non: l o-091498 4.1 -2 REVISION I
FINAL, DATA REPORT 4.1.1.2 Inoperable Instruments l' - I During testing, LDP-401 and LDP-402 measured false water levels for ACC-1 and ACC-2, respectively. Both level sensors recorded some level of water even when the accumulators were completely drained, and they consistently recorded a higher water level as the accumulators drained. { De problem stemmed from the physical configuration of the upper-pressure tap on each accumulator. 2 The upper-pressure tap was installed with a vertical rise before it turned horizontally and was used as , . the reference leg. During calibration of the sensors, the accumulator was filled completely with water, i a and reference and variable legs were filled and vented. The pressure drop across the level differential j pressure sensor (i.e., reference-leg pressure minus variable-leg pressure) was 0, and the transmitter was l calibrated as full. The accumulator was then drained quickly and completely. De pressure drop l across the sensor was the maximum, and the transmitter was calibrated as null. This calibration would i be correct if the reference leg stayed full, and if both the reference and variable legs were free of air at all times. Since the upper tap was installed with a vertical rise at the accumulator, part or all of the l l water in the vertic : rise drained out in time and left a partial vacuum behind. His partial vacuum j reduced the reference-leg pressure and, consequently, gave a lower pressure drop across the ser.sor,
- translating to a higher level. Thus, the level measurements from LDP-401 and LDP-402 were not j used in this test.
The level measarement error mentioned previously should be minimal in actual matrix testing, since the uprer heads of the accumulators were pressurized with nitrogen gas, preventing drainage of the i water in the vertical stub of the reference leg. a The precision scale measurement and the temporary tygon level readings were raccurate and were used l to calculate the tank volume and set the standpipe insertion depth. i 4.1.1.3 Test Results and Evaluation s'
- i Table 4.1-1 lists raw test data measured for ACC-1, and Table 4.1-3 lists data mmured for ACC-2.
- Table 4.1-2 provides the calculated results using raw test data and compares the data with the design j values for ACC-1. Table 4.1-4 provides the data-versus-design values for ACC-2. The design values
' ) were obtained from Table 7-6 of AP600 Low-Pressure Integral Systems Test At Oregon State , University, Facility Scaling Reporr, WCAP-14270.@ Once the standpipes were installed inside the accumulators, they remained at the same position l throughout the entire test program. Dus, the water volume in each accumulator was the same from test to test. Any level variation measured by LDP-401 and LDP-402 was due to the instrumentation error described in Subsection 4.1.1.2. I The total volume of each accumulator was in agreement with the design volume. The water volume per inch of height in the cylindrical shell portion was in agreement with the design value. , oA15%wRevi\l5%w-9.non:ltF081298 4,1-3 REVISION 1 m w- ,y4-, -.>wwy - - -=e=, -- a ,s--r -e - -
FINAL DATA REPORT The water volutae in each accumulator with the standpipe installed was within the specified acceptable range; however, ACC-2 contained slightly more water than ACC-1. Water volume with the standpipe Ja.b.c for ACC-1. Water volume with the standpipe installed was [ Ja.b.c installed was [ for ACC-2. These water volumes should be the same from test to test since they were controlled by the standpipes. He average water volume per inch of cylindrical shell height was [ Ja b.c for both accumulators. Each accumulator had a flat bottom, and all water was contained inside the cylindrical shell. 4.1.2 CMT Volume Test ne objectives of the test wem to determine: l = Upper- and lower-head volume for CMT-1
= First- through fourth-stage ADS volumes in CMT-1
. Upper- and lower-head volume for CMT-2
- First- through fourth-stage ADS volumes in CMT-2 4.1.2.1 Test Procedures Figure 4.12 provides a schematic of the test setup. A general description of the procedures is provided here.
[ Each CMT was filled with water by the RNS pump through a temporary fill / drain valve installed at the outlet flange at the bottom of the tank. Each CMT was filled until a steady stream of water issued from its vent valve located at the top of the tank. Water temperature in the tank was measured and recorded. The required CMT volumes were determined by weighing the water drained from the tank, then converting the water weight to a water volume using the measured temperature to obtain the water density. This draining and weighing process was performed in steps. The measured volumes were compared with the design volumes. A temporary tygon level indicator was installed along the tank to provide a visual indication of the tank water level. As an initial condition to the test, the calibration of the precision scale was checked against a certified weight. At the end of the test, the scale was checked again using the same certified weight to verify the calibration of the scale remained valid during the test. 4.1.2.2 Inoperable Instruments Each CMT was measured with four level differentia; pressures, one temporary tygon hose level scale, one precision weight scale, and one thermecenp'e for water temperature measurements. The precision scale provided extremely accurate weight measurement because it was calibrated against a known certified weight. The tygon hose level scale measurement was within 0.125 in. One of the four oA15hRevl\tS36w-9 non:Ib-081298 4.1 4 REVISION 1
FINA1. DATA REPcCT e level differential pressure sensors measured the wide-range overall level; the other three measumd I narrow-range upper-head, bottom-head, and middle cylindrical-shell levels. Piping and instmmentation diagrams (P&lDs) Dwg. OSU 600501 and Dwg. OSU 600502 (Appendix G) show the arrangement of these level diffemntial pressures. Instrumentation anomalies were experienced in this test and are discussed in the following text.
~ CMT-1 Inonerable Instruments CMT-1 levels were measured by a tygon hose with a scale and by four level differential pressure sensors (LDP-501, LDP-503, LDP-505, and LDP-507). LDP-501 measured the bottom-head level; LDP-503 measured the middle cylindrical-shell level; LDP-505 measured the upper-head level; and LDP-507 measured the overall CMT-1 level.
The tygon hose and differential pressure sensors LDP-501 and LDP-507 recorded a higher-than-expected level (Table 4.1-5). For example, the tygon hose level instrument recorded [ ]a,b,c of water; LDP-501 recorded [ ]a.b.c of water; and LDP-507 recorded [ ]a.b,e of water when the water level was about [ ]a b.c All of these instruments measured about [ ]a.b.c higher than the actual levels every time a data point was obtained. De reference zero level was at the bottom of the tank. The actual cause of this shift in level measurements is not known. It is possible an air bubble was trapped at the bottom instmment line at the bottom of the tank. Then, as water filled the tank, the air bubble compressed raising the tygon hose scale readings and causing false level readings. This effect was common to the tygon hose level and level differential pressure sensors LDP-501 and LDP-507 since they shared the same bottom tap. The magnitude of actual level shift was difficult to determine since the degree of compression in the air bubble was not known. ! l LDP-505 also experienced some anomalies. The physical span of this instmment was about [ ja,b.c and it measured [- Ja..b,c-an increase of [ ]a,b,e The most likely cause for this anomaly was similar to that of the accumulator level differential pmssums, i.e., the vertical upper tap introduced measurement ermr (see Subsection 4.1.1.2 for a detailed discussion). LDP-503 was designed to measure the middle-sheli level. It covered a span of about [
]a.b,c w th respect to the bottom of the CMT) and functioned well. For example, LDP-503 measared [ la.b,e of water in the middle shell versus the design water level of
[ ]a,b.ci Table 4.1-5). The bottom tap of LDP-503 was located at about [ ]a,b.c above the bottom of the tank; therefore, when it measuicd 0 in., the water level was at [ ]a.b,e Thus,
- [. ]a.b.c tr,easured by LDP-503 corresponded to [ ]a.b.c As another example, LDP 501 measured [ ]a,b.c percent volume. This corresponded to a level of [ ]a,b.c with respect to the bottom of the tank. De design level for [ ]a,b,e percent volume was [ ]a.b.c Thus, the measurement was in excellent agreement with the design. De main reason for this agreement was that both top and bottom taps of LDP-503 were horizontal taps and free of air bubbles; therefore, none of the previously mentioned anomalies occurred.
oA15hRevl\l5h-9.non:lb-o81298 4,1 5 REVISION 1
FINAt. DATA REPORT CMT-2 Inorierable Instruments CMT-2 levels were also measured by a tygon hose with a scale and by four level differential pressure sensors (LDP-502, LDP-504, LDP-506, and LDP-508). LDP-504 measured the bottom-head level; LDP-506 measured the middle cylindrical-shell level; LDP-508 measured the upper-head level; and LDP-502 measured the overall CMT-2 level. The tygon hose and differential pressure sensors LDP-504 and LDP-502 shared the same bottom tap located at the bottom of CMT-2, and the upper tap of LDP-502 was vertical, similar to CMT-l. Ti.e tygon hose level and LDP-504 measurements agreed with the design levels very well, indicating that these lines were free of air bubbles and that the measurements were acceptable (Table 4.1-6). Differential pressure sensors LDP 502 and LDP-508 recorded slightly higher-than-expected levels (less than [ ]a.b.c higher in CMT-1). This was most likely due to the vertical tap geometry at the top. For a detailed discussion of vertical-tap-induced error, see Subsection 4.1.1.2. LDP-506 was designed to measure the middle cylindrical-shell level (similar to CMT-1). LDP-506 covered a span of about [ la.b.c-from [ ]a.b.c with respect to the bottom of the CMT. It functioned well, with the exception of one data-point measurement. Step 4.4.8H in Table 4.1-6 shows that LDP-S06 measured [ ]a.b.c This data point is believed to be a misprint for the following two reasons:
- All other measurements were in good agreement with design values
- The measurement was beyond the physical span of the instmment Again, the main reason for this agreement was that both top and bottom taps of LDP-5% were horizontal taps and free of air bubbles. 'Iherefore, no unacceptable discrepancies occurred.
4.1.2.3 Test Results and Evaluation Tables 4.1-5 and 4.1-6 list raw test data for CMT-1 and CMT-2 respectively. The following columns were added to compare with raw data:
- Design water level
- Conversion of temporary tygon hose level measurements to levels with reference to the bottom of the CMTs
- Water-specific volume derived from the ASME steam table for each measured water temperature e
oM336wRevl\t 536w-9 non:lt>.r8129s 4.1-6 REVISION 1
FINAL DATA REPORT q Raw data for level differential pressure sensors, precision scale measurements, and water temperature V measurements were listed as measured. CMT-1 Test Results and Evaluation Table 4.1-7 summarizes the test results and compares them with design values. Total tank volume measurements for CMT-1 were in agreement with design values. Test procedures clearly identified when the tanks were full or empty, and precision scale and thermoc suple measurements were extn mely accurate. Lower-head volume for CMT-1 was also in agreement with the design value, implying that the lower-head height should be in good agreement with the design lower-head height. Table 4.1-5, however, shows the measured lower-head height as [ la.b,c higher than the design value. This discrepancy was due to instrument anomaly, as described previously. Because the measured lower-head volume was in agreement with the design value, it was concluded that the CMT-1 bottom-head height should be the same as the design height. 1 Cylindrical-shell height ([ la.b.c) was comparable to design-shell height ([ Ja b,c). The zero shift due to the air bubble had little effect on this measurement, because the shell height was the l difference between level measurements of the upper- and lower-head weld joints. The zero shift
/O cancelled out.
CMT total volume, lower-head volume, lower-head height, shell volume, and shell height were in j agreement with design values. Rus, there was no reason why the upper head did not agree with the l design value. Herefore, the design tank height can be used as the real height. Figures 4.1-3 and 4.1-4 plotted the measured volume as a function of tank level. i Because of the error introduced by the vertical upper tap of LDP 507, this vertical upper tap was moved to the inlet nozzle o' the CMT, and was placed horizontally after volume testing was complete and before matrix testing began. CMT-2 Test Results and Evaluation CMT-2 measured total volume was in agreement with design volume. The upper- and lower-head volumes were slightly lower than expected; however, they were within [']a.b.c percent of the design values. Level measurements taken by the tygon hose and bottom-head level differential pressure sensors were more accurate than those in CMT-1. He discrepancy was small. The overall level differential pressun: sensor still read higher level than expected, as discussed previously. Figures 4.1-3 and 4.1-4 [] plotted the measured volume as a function of tank level. V oAl5%wRevlM5%w-9.non It481298 4,]-7 REVISION 1
FrNAL DATA REPORT Because of the error introduced by the vertical upper tap of LDP-502, this vertical upper tap was moved to the inlet nozzle of the CMT and was placed horizontally after volume testing was complete and before matrix testing began. 4.1.3 Pressurizer Volume Test The objectives of the test were to determine:
- Volume of the pressurizer upper and lower elliptic heads
- Pressurizer cylindrical volume above the pressurizer heaters
= Pressurizer cylindrical volume containing pressurizer heaters 4.13.1 Test Procedures Figure 4.1-5 provides a schematic of the test setup. A general description of the procedures is provided here.
The pressurizer was filled with water through the pressurizer drain valve until a steady stream of water issued from a thermocouple connection at the top of the tank. (The thermocouple was removed for the test.) The fill line was disconnected from the water supply source and re-routed to a plastic container into which water in the pressurizer would be drained. Tank water temperature was measured and recorded. Pressurizer volume was determined by weighing the water drained from the pressurizer with a precision scale, then converting the water weight to a volume using the measured temperature to obtain the water density. This draining and weighing process was performed in several steps. A temporary tygon level indicator was installed along the tank to provide a visual indication of the tank water level. LDP-601 was also used to measure tank water level. As an initial condition of the test, the calibration of the precision scale was checked against a certified weight. At the end of the test, the scale was checked again using the same certified weight to verify the calibration of the scale remained valid during the test. 4.13.2 Test Results and Evaluation Table 4.1-8 lists raw test data used to calculate volume in the pressurizer. Tygon hose level measurements were used to calculate water volume of the pressurizer and were in agreement with the design level. Level measurements made by LDP-601 were also in reasonable agreement with the design level. Calculated volumes at selected elevations are listed in Table 4.1-9. Design volume for the pressurizer was [ Jab,e Measured pressurizer volume of [ ]a.b.c was within [ ]a.b.c percent of the design volume. The calculated volume per inch of height in the cylindrical shell portion was [ ]a.b.c oA1536wRevl\1536w 9 mon;lb48129s 4,1 8 REVISION 1
. . _ . . . -. - - - ~ . . - - .- - .. .-
FINAL DATA REPORT ,( (- 4.1.4 IRWST Volume Test , y The objectives of the test were to determine: l
- IRWST total volume 1
IRWST overflow (to primary sump tank) water volume was between [ ]a,b c IRWST normal water volume was between [ ]a,b.c Note: The design net normal water volume was [ ]a,b,c (Table 7-10), AP600 low-Pressure Integral Systems Test at Oregon State University, Facility Scaling Report, WCAP-14270W
- IRWST minimum gas volume
- IRWST normal gas volume 4.1.4.1 Test Procedures l l
(^* Figure 4.1-6 provides a schematic of the test setup. A general description of the procedures is provided here. He IRWST filled with water through the IRWST manway at the top of the tank until the water level reached the top of the IRWST manway flange. Water temperature in the tank was measured using a thermocouple. Tank volume was determined by measuring the weight of water using calibrated load cells on the feet of the tank, converting the measured water weight to a water volume using the measured temperature to obtain the water density. Water in the IRWST was drained then. and the remaining water in the IRWST was measured. This draining process was repeated several times. A temporary tygon level indicator was installed along the tank to provide a visual indication of the tank water level. LDP-701 was also used to measure the water level in the tank. 4.1.4.2 Test Results and Evaluation i Table 4.1-10 lists raw test data used to calculate volume in the IRWST, he test data showed that LDP-701 measured a maximum level of [ ]a,b.c which was in agreement with the design distance ([ Ja.b.c), The data also showed that the incremental level changes recorded by
- LDP-701 were very close to those measured by the tygon hose level scale.
l Measured data were converted into volumes and listed in Table 4.1-11. Measured net normal water 7
/ volume in the tank (i.e., water volume from the bottom of the tank to the top of the fill overflow pipe) iU .
. oA15%wRevl\l5%w-9.nostib421298 4.1 9 REVISION 1
FINAs. DATA REPORT
-n was [ ]a.b.c percent above the design volume ([ ]a b#). This measured volume was within an acceptable range (between [ ]a.b,c),
4.1.5 Primary and Secondary Sump Tank Volume Test The objectives of the test were to determine:
. Primary sump curb elevation (so that overflow from the primary to the secondary sump occurred between [ Ja.b cy
- Primary and secondary sump total volumes
= Primary sump gas volumes 4.1.5.1 Test Procedures Figure 4.17 provides a schematic of the test setup. A general description of the procedures is provided here.
Primary Sump Tank Volume Test A blank Hange was installed at the primary / secondary sump cross-connect line. The primary sump tank was filled through the manway using water from a temporary source. The water was isolated when the level reached the top of the primary sump manway flange. Total tank volume was detennined by measuring the weight of water using calibrated load cells on the feet of the tank, then converting the measured water weight to a water volume using the measured temperature to obtain the water density. Water in the tank was drained in several steps. The remaining water weight, level, and temperature were measured for volume calculation. The required overflow volume / mass of the primary sump (i.e., volume from the bottom of the primary sump to the overflow level) were calculated based on the design volume of the tank first, then measured in the test. As water drained from the primary sump, the level corresponding to the design overflow volume / mass was marked on the primary sump overflow flange. Following the initial tank draindown, an overflow weir was placed at the level marked on the flange to set the overflow level. The draining process continued until the tank emptied. The tank was refilled (with the weir installed) until the level did not change, indicating that it was at the overflow level Tank volume / mass were verified against the required overflow volume. O oA15%wRevnl5%w-9.non:lb-081298 4,1,10 REVISION I
__ __ _ __ _ _ _ _ . _ . _ _ _ _ _ . = = _ _ _ _ - . _ _ _ ._..__ _._ ___ FINAL DATA Rzec07 A temporary tygon level indicator was installed along the tank to provide a visual indication of the O tank water level, and LDP-901 was used to measure the water level. Secondary Sumo Tank Volume Test A blank flange was installed at the primary / secondary sump cross-connect !ine. The secondary sump tank was filled through the manway using water from a temporary source. The water was isolated when the level reached the top of the secondary sump manway flange then drained via the bottom
' drain line. De volume of water remaining in the tank was determined by measuring the weight of the water using calibrated load cells on the feet of the tank, then converting the water weight to a water volume using the measured temperature to obtain the water density. This draining process was performed in several steps.
A temporary tygon level indicator was installed along the tank to provide a visual indication of the tank water level. LDP-902 was also used to measure the tank water level. 4.1.5.2 Test Results and Evaluation Tables 4.1-12 and 4.1-13 list raw data for the primary and secondary sump tanks, respectively. Two columns of information were added in each table--one includes the corresponding design level l
/^ values and the other includes the reduced tygon hose level, with respect to the bottom tap of LDP-901 I v for the primary sump tank and LDP-902 for the secondary sump tank. Design levels were obtained from Dwgs. 787-PS01 and 787-SS01 in AP600 Low Pressure Integral Systems Test at Oregon State ,
University, Facility Description Report, WCAP-14124.M l Primary sump tank levels measured by the temporary tygon hose scale were in agreement with design values. LDP-901 measurements were off from [ Ja,b.c It was discovered that LDP-901 was ! calibrated to have a span from [ la,b.e og water, and the physical span was about [ Ja,b,e Although the exact cause of this discrepancy was not known, the error was within the ; 2 percent required instrument accuracy. Secondary sump tank levels measured by both the tygon hose level scale and LDP-902 were in agrument with design values. For example, LDP-902 recorded a maximum span of [ ]a,b,c versus the design span of [ Ja.b.c The centerline of the overflow pipe was designed to be at the upper-head weld joint. The corresponding height with respect to the bottom tap was [ ]a b.c LDP 902 measured [ ]a,b.c and the tygon hose measured [ ]a,b.c i Table 8 9 in AP600 Low-Pressure Integral Systems Test at Oregon State University, Facility Scaling Report, WCAP 14270,W provides the following design water volumes:
- Primary sump water volume at simulated flood-up level ]a.b,c
[
. Secondary sump water volume at curo (overflow) level
[ ]a,b,c oA15 brevi \l5%w-9.non:ltr081298 4.1 11 REVISION 1
FINAL DATA REPORT In addition, the document specifies the primary sump tank curb (overflow) level to be [ Ja.b,e above the simulated flood-up level, resulting in a volume of [ Ja.b.c at the overflow level. The following volumes were calculated using raw test data:
- Overflow from the primary to the secondary sump occurred at [ ]a.b.c within the specified acceptance criteria of [ ]a.b.c and was the same as the design value. The corresponding level was at the middle of the upper-head weld. This level was transferred to a weir that would provide the same water volume from test to test.
- Total primary sump volume was [ ]a,b.c
- Primary sump gas volume was [ la.b.c l
- Total secondary sump volume was [ ]a.b.c
- Curb overflow level at the secondary sump tank was controlled by the weir set in the primary sump tank volume test. The corresponding level at the secondary sump tank was at the middle of the upper-head cylinder weld, and the corresponding volume was [ Ja,b.c within
[ ]a.b,c percent of design volume. In addition, the average cross-sectional area of the cylindrical portion of the primary sump tank was calculated to be [ Ja.b.c versus the design cross-sectional area of l [ ]a.b,e of height). Similarly, the average cross-sectional area of the cylindrical portion of the secondary sump tank was [ ]a.b.c of height versus the design value of [ ]a.b.c of height. 4.1.6 SG-1 and SG-2 Secondary Side Volume Test The objectives of the test wie ta determine: l I
- SG-1 upper shell volume (volume above the cylindrical section of the vessel) l
- Volume of the SG 1 cylindrical section
- Total volume of SG-1 SG-2 upper-shell volume (volume above the cylindrical section of the vessel)
- Volume of the SG-2 cylindrical section
- Total volume of SG-2 l
4.1.6.1 Test Procedures l Figure 4.1-8 provides a schematic of the test setup. A general description of the procedures is ! provided here. l 1 o A1536w Rev i\l 536w-9.non: I b-081298 4.1-12 REVISION 1 l l
FINAL DATA REPORT The secondary side of each SG was filled with water through its top steam outlet flange until water reached the top of the steam outlet flange. Water temperature in each SG was measured. SG volumes were determined using a precision scale to weigh the water drained from the SG, then converting the measured water weight to a water volume using the measured temperature to obtain the water density. The container used to receive the water drained from the SG was weighed and emptied because the water volume in the SG was much larger than the volume of the container. This process was repeated several times to obtain the desired test data. A temporary tygon level indicator was installed along the SG to provide a visual indication of the water level. LDP-301 and LDP-302 were also used to measure water level in SG-1 and SG 2, respectively. As an initial condition of the test, the calibration of the precision scale was checked against a certified weight. At the end of the test, the scale was checked again using the same certified weight to verify the calibration of the scale remained valid during the test. 4.1.6.2 Test Results and Evaluation Tables 4.1 14 and 4.1-15 list raw data for SG-1 and SG-2, respecuvely. Two columns of information were added in each table--one included the corresponding design levels and the other includes the reduced temporary tygon hose level with respect to the drain line at the bottom of each SG. Design levels were obtained from Harris Thermal Transfer Products Dwg. 20175-D1 and 20175-D2 (Appendix H). 'Ihe bottom taps of LDP-301 and LDP-302 for SG-1 and SG-2 were also located at the same elevation as the drain lines. Thus, both the level differential pressure sensors and the tygon hose shared the same reference zero. Levels obtained by both the tygon hose level scale and the level differential pressure sensors were in agreement with design values. The measured maximum range of the level differential pressure sensors were [ la b,e for SG-1 and [ la,b,e for SG-2, which compared well with the design distance of [ la,b,c The tygon hoses recorded [ ]a.b,c for the level of the top of the steam outlet flange for both SGs, indicating agreement with the design value ([146.60 in.]). The measured total volume of the secondary side of the SGs ([ ]a,bc for SG-1 and [- - ]a.b.c for SG-2) differed by [ ]a,b,c percent. In the cylindrical sections, the difference in volume was less than 1 percent ([ ]a,b,c for SG-1 and [ la.b,e for SG-2). The main discarpancy between the two SGs was that the volume in the upper-head region above the cylindrical shell ([ ]a,b,e for SG 1 and [ la.b,c for SG-2) differed by [ Ja.b c percent due to manufacturing differences in the upper shell volume; however, this upper-head region was the vapor region, and precisie . control of vapor volume was not required in the test program, Table 4.1-16 summarizes these results.
\
oA15hRevl\l536w-9.non:lb-061298 4, } .13 REVISION 1
-. . , - . - . -, - _, . - - , - - - . . - - . , _ . . . ~
FINAL DATA REPORT 4.1.7 ADS and HAMS Molsture Separators Volume Test The objectives of the test were to determine:
. Volume of the ADS 1-3 separator ([ Ja.b.c separator)
. Volume of the break separator ([ Ja.b.c separator)
. Volume of the ADS 4-1 separator ([ Ja.b.c separator)
- Volume of the ADS 4-2 separator ([ ]a.b.c separator)
'4.1.7.1 Test Procedures Figure 4.1-9 provides a schematic of the test setup. A general description of the procedures is provided here. Each separator was filled with water through a temporary fill / drain valve. For the ADS 1-3 separator, this fill line was installed at the separator flow-meter flange located at the bottom of the separator. All other separators used a fill line connected to the top flange. The ADS 1-3 separator was filled until a steady stream of water issued from a temporary vent valve installed at the top of the separator. The remaining separators were filled until the water level reached the top of the separator outlet flange. Due to the unavailability of the DAS at the time of testing, direct water temperature measurement was not recorded. Instead, the water temperature was assumed to be a constant 56*F. This was consistent with the temperature of the water in the primary and secondary sump tank volume test. Separator volume was determined using a precision scale to weigh the water that was drained from the separator, then converting the measured water weight to a water volume using the water density at 56*F. The draining and weighing process was performed in several steps. A temporary tygon level indicator was installed along each separator to provide a visual indication of the water level. As an initial condition of the test, the calibration of the precision scale was checked against a certified weight. At the end of the test, the scale was checked again using the same certified weight to verify the calibration of the scale remained valid during the test. 4.1.7.2 Test Results and Evaluation The following tables list raw test data:
- Table 4.1-17 ADS 1-3 Separator Volume Data e Table 4.1-18 ADS 4-1 Separator Volume Data
- Table 4.1 19 ADS 4-2 Separator Volume Data a Table 4.1-20 Break Separator Volume Data e
o:\l5hRevl\l536w-9.non:lb481298 4, } .14 REVISION 1
FINAL DATA REPCe.T ! .) De total water volume measured and listed in these tables includes the volume of the separator vessel itself and the water volume of the interconnecting pipes; however, the volume of the interconnecting pipes was negligible compared with the volume of the separator. Here were no specific requirements on the water volume in the separators. Measured total volume of the moisture separators was included for information only. Sizing of the separators was based on expected steam flow. Each separator met minimum relative locations of the internal separating element above the liquid surface and a specified pressure drop across the element. Volume for the ADS 1-3 separator ([ ]a.b.c separator) was measured as [ ]a.b.c volume for the large-break separator ([ ]a.b,e separator) was measured as [ ]a.b,c volume for the ADS 4-1 separator ([ ]a.b.c separator) was measured as [ ]a.b,e aid volume for the ADS 4-2 separator ([ ]a.b.c separator) was measured as [ ]a.b.c ADS 4-1 and ADS 4-2 separators were designed to be the same physically, except that the separating elements, which occupied negligible volume, were different. He identical (measured) volumes of these two separators indirectly proved that the volume test results were acceptable. 4.1.8 Reactor Vessel Volume Test The objectives of the test were to determine: O ys a Upper-head volume
- Upper-plenum volume Volume between the upper support plate and the top of the upper core plate e
Volume between the upper core plate and the top of the heater rods Volume in the top core region from the top of the heater rods to the rod bundle midpoint
=
Volume of the lower core from the rod bundle midpoint to the top of the lower core plate a Volume of the lower plenum below the core region a Total reactor vessel volume 4.1.8.1 Test Procedures Figure 4.1-10 provides a schematic of the test setup. A general description of the procedures is provided here. The RNS pump filled the reactor vessel with water through a temporary fill valve installed on the DVI-l flange next to the reactor vessel. Blank flanges were installed at the DVI-2 flange and at the cold-leg and hot-leg flanges. He vessel was filled until a steady stream of water issued from its vent valve located at the upper head. Water temperature at the bottom of the reactor vessel was measured. His temperature was the same at all locations inside the reactor vessel. i-After the reactor vessel was completely filled and the calibrations of the level differential pressure } sensors were verified, the isolation valve at the drain line was opened to drain the water to a plastic oA15h Rev nl5%w-9.non:ll>.081298 4.1 15 REVISION 1
FINAL DATA RpORT container. Draining stopped at a designated level, and the drained water was weighed using a precision scale. Water weight was converted to water volume using the measured temperature to obtain the water density. On completion of weighing process, the container was emptied and ready to collect a new batch of water. This draining and weighing process was performed several times to obtain individual volume information at the desired level. Herefore, the total weight of water in the vessel was the sum of the individual measurements. Since the drain line was located 2 in. above the bottom of the reactor vessel, some water remained inside the reactor vessel at the end of the draining process. Therefom, the total vessel volume calculation should include the water volume that remained inside the vessel following draindown. A temporary tygon level indicator was installed along the vessel to provide a visual indication of the water level. As an initial condition of the test, the calibration of the precision scale was checked against a certified weight. At the end of the test, the scale was checked again using the same certified weight to verify the calibration of the scale remained valid during the test. 4.1.8.2 Test Results and Evaluation Table 4.1-21 lists raw test data. Two columns of information were added in the table-one includes the corresponding design values and the other includes the reduced tygon hose level, with respect to the bottom tap of LDP-127 (the centerline of the drain line). Level measurements made by the level diffen ntial pressure sensors and the temporary tygon hose level scale were both in good agreement with design values. Full span measurements of LDP-127 and LDP-115 were larger than the physical distance between the upper and lower taps because they shared the common vertical upper tap. This vertical upper tap had a vertical riser that increased the calibrated span of the sensors. Data indicate that this vertical rise was about 5 in. As water drained out of the reactor vessel, the levels measured by the two level sensors agreed with the design values. He anomalies associated with the vertical tap, described in Subsection 4.2.1.1, did not occur in this test. Mass measurements made by the precision scale were accurate. The temperature was measured and recorded by a thermocouple and was also accurate. Both mass and temperature measurements were independent of level measurements and, therefore, were not affected by level measurement deviations. Thus, volumes calculated using the measured weight and temperature represented the exact volume of the reactor vessel. Table 4.1-22 lists these volumes. The water volume per inch of height varies from elevation to elevation. This value is calculated using the information obtained in this test. O oA15%. nevi \isaw.9.non:ib-ost298 4.1 16 REVISION l
FINAL DATA Rrront O Tables 4.1-1 through 4.122 on pages 4.117 through 4.138 are not included in this nonproprietary document. ' O O W' .h l
FINAL DATA REPORT b U e tor Vent
- ><: l -T 1
I 1 1 Accumulator -.- I
- 1
- I
~ I
- l +-- Temporary l Level
- g indicator 1
~
+-- Fm overdow l
. Tube (Removsd first - l and then inserted) -
I
- 1
- I
> - 1 o*--Temperature indication LDP401 or 4
--{ LDP402 }- - l DAS LevelIndication - M .!
I Plastic3 X : To DVl.1 or DVl-2 i contamer (PredsI6n scald O NO )
------- Temporary Tubing / Equipment Note: LDP-401 and DVI 1 are for accumulator 1 LDP-402 and DVl-2 are for accumulator 2 umu t
i, Figure 4.1-1 Schematic of Accumulator Volume Test Setup o:\l5%wRevi\l5Mw-9.non:IMMil298 4.1-39 REVISION 1
FINAI DATA REPORT O Blanked Balance une inlet Flange CMTVent X . . n I L e [ LDP-505 ) or LDP-508 - ! X 2!: CMT 1 or CMT 2 -l
-: *- Temporary
_s Level l Indicator ( LDP-503 ) ._, or LDP 506 l t
-l I
Note: Thermocouple rods and ~' X steam diffuser are installed. -l
-i
{ LDP-501 ) -l or LDP 504 g Temperatureindication - l X .
;: Outlet Flange
<: I
! or LDP-502 DAS Levelindication .-...... I E l..... l (- l conwner L J y,;y, (Precision Scajd
------- TemporaryTubing/ Equipment Water Supply (RNS Pump)
Note: Upper and lower heads are hemispherical i l t l Figure 4.12 Schematic of CMT Test Setup oAl536wRevl\l536w-9.non:ltW1298 4.1-40 REVISION 1 l
0 0 0 B-c s , w t
'd G
V
$ CMT-1 VOLUME VERSUS HEIGHT FROM BOTTOM 1
4
?' {
R 11.00 ' s
= 5 I .
g 10.00 ._ _ _. . _ . _. __ .- . _ #+q . .. A / g b u g.00 __ _ _7 __ _ _ . . ._. _ . _ p.x + n I- /. g 8.00 --- - -- - - r-4 ----
-f -- -- ' - --. -- ---
--Ff- t H* i i
7.00 l o 4 ---
/- -- -- + -} . - p --- -
; i c a .
i
$ E d 6.00 -j -
l - ---t-
} g - , j j '
L
~
< E ! .
i a a i i lll s 5.00 . - - -a. .. ,- ..-...-. ..- -.-_ c w o 2: , i a _ 4.00 - - - -
"r g 3.00 - -- -
o B to 2.00 SL iiI t l l
! i 5 1.00 _ . _4_ . 7_ '
, _ , , y f l' 1 i 3 i : !
. I
' l l i I "
0.00 ' Di 0 10 20 30 40 50 60 {> N z 5 e* HOGHT FROSE BOTTOM OF TANK (In.) j
=4
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FINAL DATA REPORT \ PZR Spray Line Vent Valve for ion) x l M PZR S
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i Figure 4.1-5 Schematic of Pressurizer Test Setup o3'5%wRevl\l5Mw-9.non:ll>O81298 4.1 43 REVISION 1
. -- -_ -- -= _ -- _ . - - . _ .. . - . - . . - - - . - FINAL DATA REPORT O Remove manway cover forfm and visues obeenstam oriRWarintwnsis. Measuring TW 1 I k
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and MF.701 Figure 4.1-6 IRWST Test Setup O oA15%wRevi\l5%w-9 non.it41298 4,1-44 REVISION 1
FINAL DATA REPORT l ,V O l i l 1
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FINA1, DATA REPORT O Temporary Level indicator Steam Pl Removed from
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FINAL DATA REPORT l O l l Temporary Level 7 Separator Steam Outlet Piping Indication -l Removed for Fill and Vent
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FINAt. DATA REPORT
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1 FINAL DATA REPORT l l 4.2 Pressure Drop Determination : 4.2.1 Background Information he design of the test facility started in early 1992. De design work was performed concurrently with the scaling analysis for the test facility. Although the sca'bg analysis was revised several times, the basic principles and the design criteria related to hardware design did not change. For additional details, see AP600 Low-Pressure integral Systems Test at Oregon State University, Facility Scaling Report, WCAP-14270.t6) According to the scaling analysis, the following parameters related to hardware design must be properly scaled: l AP600 physical geometry including layout and elevation head modeling Each AP600 injection line in size, length, area, volume, and physical geometry a Pressure dmp across each line
- Each tank volume and height a Critical thermal hydraulic phenomena Since the AP600 design was still evolving in 1992, the test facility design was constantly updated to reflect the revised AP600 design. Each injection line in AP600 went through the following design process:
(a) De AP600 line geometries were exactly modeled, including length, elevations, bend, number of elbows, and valves. l (b) Line losses were calculated for AP600 conditions, and the results were verified to be within an acceptable range. His step ensured that proper line geometry was modeled and proper calculation methodology was used. (c) Line characteristics for the test facility were scaled, including consideration of the operating pressure, temperatures, fluid density, flow rate, flow velocity, line length, line height, line inside diameter, flow area, and flow volume. (d) Instrumentation for each line was selected. (c) Pressure drop calculations were performed using the same methodology as in (b), and the l results were compared with AP600 pressure drops in (b). De ratio of the two pressure drops should equal the required scale ratio. An orifice plate was used to provide the proper pressure drop ratio where required. Note: A properly scaled line inside diameter was never a standard pipe or tubing size. Derefore, the next closest standard-size pipe or tubing was used. This would change
- the line flow area, volume, velocity, and pressure drop ratio. A non-ideal line would oMp60ml536w-27.ron:lt481298 4.2 1 REVISION 1
FINAt. DATA Ritront result in scaling distortion. Herefon:, (c) and (d) were iterated several times to obtain the optimal line size that provided proper pressure drop scale ratio with minimum volume distortion. The scaling analysis required that frictional loss should NOT be the major pressure drop in the line. Therefore, tubing was selected whenever possible. (f) The wall thickness of the selected line was sized using industry standards. The ASME code was used for piping and vessels; ANSI B3.1I was used for tubing. (g) Finally, each line was translated into isometric drawings for fabrication. Each AP600 line was modeled for this test facility using the previous process, and each line was flow tested to verify proper pressure drop and line resistance. De first flow test was completed on May 9,1994, and pn ssure drop data were obtained for the following lines:
. CMT-1 injection line (from bottom of CMT-1 to DVI-l flange and from DVI-l flange to reactor vessel downcomer at DVI 1)
= CMT-2 injection line (from bottom of CMT-2 to DVI-2 flange and from DVI-2 flange to reactor vessel downcomer at DVI 2)
- ACC-1 injection line (from bottom of ACC-1 to DVI-I flange and from DVI l flange to reactor vess.el downcomer at DVI-1)
. ACC-2 injection line (from bo'. tom of ACC-2 to DVI-2 flange and f om DVI 2 flange to reactor vessel downcomer at DVI 2)
- CMT-1 to CL-3 balance line (from top of CMT-1 to top of CL-3)
= CMT-2 to CL 1 balance line (from top of CMT-2 to top of CL-1)
- ADS 1-3 lines and the line from ADS 1-3 wparator through the sparger inside IRWST IRWST-1 injection line (from bottom of IRWST-1 to DVI-l flange and from DVI-l flange to reactor vessel downcomer at DVI-1)
IRWST-2 injection line (from bottom of IRWST-2 to DVI-2 flange and from DVI-2 flange to reactor vessel downcomer at DVI 2)
=
Primary sump-l injection line (from bottom of primary sump tank to DVI l flange and from DVI-l flange to reactor vessel downcomer at DVI-1) oA@l536w-27.non:Ib o81298 4,2 2 REVISION 1
FINAL DATA REPORT
=
Primary sump-2 injection line (from bottom of primary sump tank to DVI 2 flange and from s DVI-2 flange to reactor vessel downcomer at DVI-2). In addition, total developed head as a function of flow rate for RCPs
, Soon after the first series of flow tests was performed, the AP600 design was revised to add a venturi nonle at the DVI l flange next to the reactor vessel. Another identical verturi nonle was added at the DVI-2 Dange next to the reactor vessel. This change increased the pressure drop from each DVI flange to the reactor vessel downcomer. Therefore, the line sizing and pressure drop calculations were re-evaluated. 'Ihe evaluation concluded that the line size was still adequate. However, the orifice plates originally installed at the DVI lines between the CMT/ accumulator tee and the reactor vessel , downcomer (ORI-253 for DVI l and ORI-254 for DVI-2) had to be removed to compensate for the l
pressure drop increase introduced by the DVI venturi nonles. Another series of How tests was performed on May 20,1994, to measure pressure drop deviations due to the installation of DVI venturi nonles and the removal of orifice plates ORI-253 and ORI-254. A third series of flow tests was pe formed on September 27,1994, after the completion of the matrix tests. This series of flow tests measured pressure drops and line resistances on other lines not covered in the first two series of How tests. In addition, some of the pressure drop measurements related to the reactor vessel obtained in the first series of now tests were found to be questionable. These measurements were repeated in the third series of flow tests. The objectives of the flow tests were as follows: To provide flow and pressure drop data for RCP eperation l
- To determine RCP coastdown data '
To verify that CMT 1 and CMT-2 injection line pressure drops were between 63.6 and 81.6 in. H2O when injection flow was 8.3 0.3 gpm To verify that the ACC-1 and ACC-2 line pressure drops were between 277 and 346 in. H2 O when flow was 22.710.3 gpm To verify that the CMT-1 and CMT 2 cold-leg balance line pressure drops were between 4.0 and 5.4 in. H2 O when now was 4.6 0.3 gpm
=
To determine pressure drop versus flow information for the RNS injection line and verify that the two RNS injection line pressure drops were balanced to within 10 percent
- To determine pressure drop versus flow for the ADS 1-3 lines O
O o:\ap600\ l 5.16w-27.non: l b-081298 4.2-3 REVISION 1
FINA1. DATA REPORT
- To verify IRWST-1 and IRWST-2 pressure drops were between 63.6 and 81.6 in. H2O when
- injection flow was 8.3 2 0.3 gpm .
- To verify primary sump-l and primary sump-2 pressure drops were between 63.6 and 81.6 in.
H2O when injection flow was 8.3 0.3 gpm
- To detennine the pressure drop for each of the four primary loops and across individual line sections or components
- To determine the pressnre drop for the PRHR HX loop from hot-leg entrance to SG channel head 4.2.2 Test Procedure, Instrumentation, and Results This section provides a brief description of the facility initial filling procedures followed by a description of the test procedures for each line. The instrumentation used for each line and the corresponding test results are also discussed in detail.
There were no specific temperature or pressure requirements associated with the cold pre-operational flow tests. 4.2.2.1 Facility Initial Filling for the First Two Series of Flow Tests Dwg. OSU 600002 and Dwg. OSU 600206 (Appendix G) provide visual representation of facil.tv filling routes. The RNS pump was used to fill the entire primary sump tank and part of the IRWST, taking suction from the feed storage tank. The discharge line of the RNS pump was connected to the IRWST-1 injection line, which in tum directed the water to fill the primary sump tank and the IRWST tank. The filling of the primary sump tank was achieved via the bypass line of the sump-1, through the primary sump tank isolation valve (CSS-909). The piping arrangement was also used to partially fill the IRWST. This piping arrangement is shown in Dwg. OSU 6002% (Appendix G). The primary sump tank was filled to the simulated overflow level, i.e., the simulated AP600 curb level. The IRWST was partially filled to the corresponding level simultaneous'y. The RNS pump was secured once this level was obtained and CSS-909 was isolated. The primary sump tank had a large water volume because it was designed to accommodate water from all components, including IRWST and reactor vessel. Therefore, it was used as the water source during the initial system fill process. All other lines connected to the RNS pump discharge line were isolated during the initial filling process. Once the primary sump tank and partial IRWST were filled with water, the RNS pump suction line was then rerouted to take suction from the bottom of the primary sump tank and discharged to two parallel lines leading to the DVI injection lines. Because check valves upstream of the D"I lines o:\ap600(1536w-27.non:t b481298 4.2-4 REVtSION 1
FINAL DATA REPCRT (CMT and accumulator injection line check valves), the water was discharged into the reactor vessel (r- ' and its interconnecting components such as the pressurizer and the SGs. The CMTs were filled via the cold leg balance lines. All vent valves on these components were initially open and closed once water issued from the valves. De two RNS discharge lines were isolated once the primary loop, the reactor vessel, the pressurizer, the PRHR HX, the SGs and the CMTs were filled and vented. F^ill taking suction from the bottom of the primary sump tank, the RNS now discharged to fill the accumulators via a common fill and drain line. This common fill and drain line enabled filling and drainmg of the accumulators and the CMTs. The vent valves at t?se accumulators and CMTs were initially open and were closed when water issued from them. After the facility was filled, the RNS pump centinued to operate to keep the plant pressurized at ; shutoff head and the RCPs were jogged for a few seconds and the selected vent valves were open and closed to remove any trapped air in the system. The RCS was then refilled again and the RNS pump stopped. l The system was now completely filled, water solid, free of air, and ready for flow testing. 4.2.2.2 Facility Initial Filling for the Third Series of Flow Tests l p/ L. The filling sequence of the test facility for the third wries of flow tests was essentially the same as for regular matrix tests. See Subsection 2.7 for fli! 24 vent procedures. ! 4.2.2.3 Methodology of Pressure Drop Data Analysis Line losses were calculated from differential pressure data measured by differential pressure cells manufactured by Rosemount Co. Subsection 2.4 provides a detailed description of the differential pressure cells which measure the pressuir difference b etween any two taps installed in the facility. For these tests, the taps were usually located in piping or within the reactor vessel. The method of converting the measured data to line losses is based on Bernoulli's equation with line loss. The equations are defined below. Vertical Line with Different Temocratures When the flowing finid temperature is different from fluid temperature in the DP cell instrument imes, temperature compensation is necessary. The following equation can be used. DP_ loss = (P u - Pt ) + RHOp * (V i2 y 2)/(2g) + (RHOp-RHOg ) * (hl - h2) (1) o$ap60fA1536w-27.non:lb-081298 4.2-5 REVISION 1
. - . . .-,4-- r-
1 FmAt. DATA RrJ'onT where: DP_ loss = pressure loss in line between two static taps, psf (Pg - Pt) = measured pressure difference from high-pressure side of instrument sensing line to low-pressure side of line, psf RHOp = weight density of flowing fluid, Ib/ft.3 V3 = flow velocity at high-pressure side of differential pressure cell, ftisec. V2 = fl w velocity at low-pressure side of differential pressure cell, ft/sec. g = gravitational acceleration, ft/sec.2 hl = vertical distance of high-pressure side pressure tap with respect to arbitrary datum, ft. h2 = vertical distance of low-pressure side pressure tap with respect to arbitrary datum ft. RHO g = Weight density of fluid in sensing lines, Ib/ft.3 When the pipe diameters at the taps are the same, Vi = V 2, and the seccad term disappears. Vertical Line with Same Temperatures When the flowing fluid in the vertical line has the same temperature as the differential pressure cell sensing line temperature, the following equation can be used: DP_ loss = (P ig - Pg) + RHGp * (Vi 2.y2 )/(2g) (2) IIorizontal Line with Different Temperature When the flowing fluid temperature is different from the differential pressure cell tap line temperature, Equation 1 can be modified to the same equation as Equation 2. Horizontal Line with Same Temperature When the flowing fluid temperature is the same as the differential pressure cell tap line temperature, Equation 2 can also be used to calculate pressure loss in the line from point I (high-pressure side) to point 2 (low-pressure side) Note: The assumed high-pressure side is referred to as the high side of the differential pressure cell, and the low-pressure side is referred to as the low side of the differential pressure cell. Each differential pressure cell (or LDP cell) was assigned with predetermined high and low sides clearly marked in the piping and instrumentation diagrams (P&lDs) in Appendix G. The flow direction was assumed to be from the high-pressure side to the low-pressure side and the differential pressure cell was calibrated accordingly. When the pressure drop measurement tumed out to be negative, the actual flow direction was just the opposite of the presumed flow direction, or the pressure at the low side was actually o%pNXA1536w-27.non:Ib-081298 4.2-6 REVISION 1
. __m. ~ . _ _ _ _ _ . . _ _ _ _ _ _ ._ _ _
L FINAL DATA REroar p higher than the high side of the differential pressure cell. In either event, the magnitude of i V the measurements was still valid. , 4.2.3 RCP Flow Test i 4.2.3.1 Procedures Dwg. OSU 600203, Dwg. OSU 600301 (Appendix .G), and Dwg. LKL911218 (Appends / H) provide details of the test loop setup and instrumentation locations. He facility was filled and ver.ted prior to actual RCP flow testing as described previously. On completion of initial filling and venti g, all four RCPs were run to assess pressure drops and flows. All four reactor coolant pumps were then tripped and restarted several times to assess coastdown time. This procedure was performed for the first series of flow tests. The RCP total developed head as a function of flow was also measured and recorded in the first series of flow tests. The third series of flow test was performed after the initial facility fill and vent. Test data were
~ recorded at ambient temperature, 100*,200*,300 , and 400 F. At each temperature, the reactor coolant pumps were first started with the PRHR HX line isolated by closing the outlet isolation valve.
O Pressure drops and temperature around the reactor vessel and within the reactor vessel were measured V and recorded. Then the PRHR HX line was open and the pressure drops and temperature around the PRHR HX line were measured and recorded. Pressure drops around the reactor vessel and within the reactor vessel were measured and recorded during the first and last series of flow tests. Note that the first series of flow test was conducted at ambient temperature and the third series of flow tests was conducted at various flow temperatures, including ambient tempetature he following instnamentation and their designated functions were used in the tests. See Dwg. LKL911218 (Appendix H) for physical locations of these instruments. 4.2.3.2 Instrumentation
- Flow rate measurements:
- FMM-201 Magnetic flow meter located at CL-1
- FMM-202 Magnetic flow meter located at CL-2
- FMM-203 Magnetic flow meter located at CL-3
- FMM-204 Magnetic flow meter located at CL-4 Temperature and pressure drop measurements:
O V e%60 mis 36w-27.non:it41298 4.2-7 REVISION 1
FINAL DATA REPORT Table 4.2-1 provides a detailed description of instrumentation within the reactor vessel and other interconnecting major components. . 4.2.3.3 Results Tyal Developed IIcad of RCP Yhe total developed head for each RCPs was measured directly by differential pressure cells across each pump. The total developed head for each pump was measured in both the first and the third series of flow tests. Table 4.2-5 summarizes the test results. DP-203 was used to measure pressure drop across RCP-1, DP 202 for RCP-2, DP-205 for RCP-3, and DP-206 for RCP-4. Each pump was single speed centrifugal pump. The test results show that the total developed head and flows are almost the same for all four pumps at the same temperature. Comparison of test results l between the first and third series of tests shows that the test results are repeatable. This is shown in , 2 the table by the ratio of pressure drop to flow rate square (psi /gpm ). The results of the third series of flow tests also reveals that the total developed head for each RCPs decreases as the fluid temperature increases. Performance curves for the pamps supplied by Queen Pump Company of Portland On gon indicate that the pump was rated at 250 gpm at 144 in. H2O of total developed head at ambient temperature. All four pumps were operating at near runout flow during testing with flows of approximately 330 gpm and 167 in. II 20 of total developed head. However, no cavitation or net positive suction head problems were observed. RCPs Coastdown Time Coastdown times for the RCPs could not be evaluated using the DAS due to relatively long sampling rate. Testing indicated that RCPs coastdown time was about I to 3 seconds. But the fastest data sampling rate available in the DAS was about 8 seconds. Therefe e, no coastdown data were obtained in these tests. Data obtained durint matrix resting indicated coastdown time was about 2 seconds. Reactor Vessel Pressure Drops Table 4.2 I oummanas the locations cf pressure and pressure drop measurements in and around the reactor vessel. Table 4.2-2 summarizet the test data obtained during the first series of flow tests. Table 4.2-3 summarizes ne test data ol udned during the third series of flow tests. The third series of flow tests cepeated all pressure drop measurements in the first series of flow tests and more. Table 4.2-1 provioes a comparison of the instrumentation used in these two serier of flow tests. Table 4.2-4 provides a comparison of the test data of these two series of flow tests at same oMp6(XA1536w-27 mon:lt> 081298 4.2-8 REVISION 1
- .- . - - - . ~ . _ . . - - ~ . - . - . - . - . - - - - . - . - ~ _.-
l FmAi, DATA Rsroar l temperature. Data from the third series of flow tests did not agree with data from the first series of V flow tests.- It is believed that the data from the first series of flow test were not accurate. In fact, this l was one reason the third series of flow tests was performed. The test data f.om the third series of j. flow tests appear to be more realistic. . This conclusion is based on the following observations: The total developed head for each RCP recorded in the last series of flow test is in good agreement with that recorded in the first series of flow tests (Table 4.2-5). The DAS had been verified by the time the third series of tests was performed. The test results from the last series of flow test compared well with steady state matrix test results. l Some data from the first series of tests agree well with the last series of flow tests and some do not. For example, the pressure drop across the upper support plate (LDP-il4) in the first l series of flow test agrees well with the last series of flow test,27.6 in. H O versus 27.2 in. 2 j H20. However, the piessure drop from the bottom of the upper grid inside the core to the top 1 of the upper core plate (LDP-Ill) in the first series of flow tests is totally different from the last series of flow test,7.89 in. H 2O versi:s 2.33 in. H 20. The exact cause of the discrepancy is not known. It is suspected that the configuration file for the first series of flow tests was not verified properly. In fact the configuration file was being verified wl.en the first series of
\_/ flow test was conducted. Also, an improper configuration file explains why some data agree well and some do not. The electronic signals generated from the transducers appear to be correct; the problem with the first series of flow tests seems to have been caused by the data {
recording system. Using the test irsults from the last series of flow tests, the line pressure loss and line resistance within and around the reactor vessel were calculated and summarized in Table 4.2-6. Equations 1 and 2 were used to calculate the line loss. 4 Primary Loon Pressure Droos Similar to the pressure drops across the reactor vessel, the pressure loss in hot legs, cold legs and other components associated with the reactor vessel in the last series of flow tests were used to estimate the line resistance. The results are also summarized in Table 4.2-6. 4.2.4 CMT Injeetion Flow Test
~
Hydraulic characteristics of the CMT-1 and CMT-2 injection lines from the bottom of each CMT to the 'DVI tees and to the reacto- vessel downcomer via the DVI-l and DVI-2 injection nozzles are I discussed ii this section. The CMT injection flow test was performed twice in the first series of flow tests only, first with the original orifice plates and then with the DVI venturi nozzles. i c:Wi$36w-27.non:1ME1298 4,2-9 REVISION 1
FINAL DATA REPORT 4.2.4.1 Procedures Figures 4.21 and 4.2-2 provide schematics of the test arrangement for each CMT. A flow path was i set up using the RNS pump and a temporary tygon hose. The hose was connected to the service connection downstream of the RNS pump discharge to pump water from the bottom of the primary sump tank to the top of the CMT, down to the DVI tee and into the reactor vessel downcomer via the DVI nozzle. Flow was then directed to the pressurizer and ADS valves. Finally, flow was directed to the IRWST via the ADS 1-3 sparger located inside the IRWST. The temporary tygon hose was connected from the pump discharge service connection to the top of the CMT. The ADS-1 valve was opened,and the pump started. The desired flow rate was obtained by throttling RCS-806 (the isolation valve at the service connection downstream of the RNS pump discharge). The desired pressure drops were measured and recorded. He test was performed the first time with the original configuration, i.e., with orifice plates ORI-253 and ORI-254 and without DVI venturi nozzler. The test was performed the second time with the new venturi nozzles installed in each DVI line at the DVI nozzles next to the reactor vessel. Due to the increased head loss in each injection line from the venturi nozzles, it was necessary to remove orifice plates ORI-253 and ORI-254 located between the DVI tees and the DVI nozzles next to the reactor vessel. The water temperature for this test was at ambient conditions. 4.2.4.2 Instrumentation Figures 4.2-1 and 4.2-2 provide schematics of the test arrangement and locations of instrumentation. Two differential pressure indications were taken. For CMT-1, DP-501 provided the pressure drop from the bottom of CMT-1 to the DVI-l flange next to the reactor vessel and DP-128 provided the exit loss into the vessel, from the DVI-l flange to the reactor vessel downcomer. FMM-501 measured the flow rate through the DVI-1. Note that the pressure drop measured by DP-501 included the pressure drop across either orifice plate ORI-253 or the DVI venturi nozzle, but not both. Figure 4.2-1 shows this pressure drop measurement. For CMT-2, DP-502 provided the pressure drop from the bottom of CMT-2 to the DVI-2 flange next to the reactor vessel and DP-129 provided the exit loss into the reactor vessel from the DVI-2 flange to the reactor vessel downcomer. FMM-504 measured the flow rate through DVI-2. Similar to CMT-1. DP-502 included the pressure drop across either orifice plate ORI-254 or the DVI venturi nozzle, but not both. Figure 4.2-2 shows this pressure drop measurement. O ospuxntsh-27.non:iwis1298 4.2-10 REVISION 1
- .. . . - - _ .- . - ~ . . . . . - - - . . . - .-. - . -
FINAL DATA RErc:rr I 4.2.4.3 Results f ( l Each CMT injection flow path consists of two line sections. The first line section is normally referred to as the tme CMT injection section, and runs from the bottom of the CMT to the DVI tee junction l where the accumulator and the IRWST injection lines meet. De second line section is referred to as the DVI line which begins at the DVI tee junction and ends at the reactor vessel downcomer. Bis DVI line is common to the CMT, accumulator, and IRWST injection lines, and it contains either the DVI venturi nozzle or the orifice plate (ORI-253 or ORI 254). All matrix tests used the DVI venturi nozzle; orifice plates ORI-253 and ORI 254 were removed. The test data are summarized in Table 4.2-7 Data were plotted in Figures 4.2-3 through 4.2-10 as the pressure drop rate versus the square of the flow rate and fitted with linear regression methodology. He plots show excellent linear results, indicating that Darcy's equation can be used to predict pressure j drops through the line. I Line Resistance De plots show that the pressure drop is directly proportional to the square of the flow rate. This is in agreement with Darcy's equation. The slope of each curve was used to calculate the overall line ! resistance (f1/D+K) using Darcy's equation as follows: I h_L = (111D+K)*V2/(2g) (3) This equation can be re-arranged as follows: h_L = (12)*(flJD+K)*Q2 /(2
- g
- A2
- 448.862) (4) where:
h_L = pressure drop through the line, in. H2O fUD+K = line total resistance, dimensionless g = gravitational acceleration = 32.2 ft/sec.2 A = flow area, ft.2 Q = volumetric flow rate, gpm f = friction factor, dimensionless D = diameter, ft. L = length, ft. K = form loss coefficient, dimensionless Equation 5 was represented by each curve plotted in the figures. Therefore, the slope of each plot was used to calculate the total line resistance as follows: a{\ ownsh.27.non:ib-os129s 4.2-11 REVISION 1
FNAL DAT4 Rzecar 2 Slope = 12 * (f1/D+K) / (2
- g
- A2
- 448.86 )
or fUD+K = slope * (2
- g
- A2
- 448.862 ) /12 (5)
Equation 5 was used to calculate total line resistance. For example, Figure 4.2-3 shows a slope of 1.0733, using Equation 6: 2 fUD+K = (1.0733)(2)(32.2)(0.006842)2 (448.86 ) /12 f1/D+K = 54.32 The flow area for the CMT injection line is 0.006842 ft.2 (inside diameter = 1.12 in.) The line resistance for each line section was calculated and is summarized in Table 4.216. Pressure DroD ComDarison Table 4.2-16 compares the measured pressure drop with the design (predicted) pressure drop and the specified acceptable pressure drop range (defined as the acceptance criteria). The design (predicted) pressure drop was obtained mathematically using Darcy's equation and the same flow path configuration as the test article. l The following observations were made for both CMT-1 and CMT 2 injection lines:
- 1. All measured pressure drops were in agreement with Darcyi equation.
I
- 2. Both CMT-1 and CMT-2 pressure drop measurements were within the acceptable range. This applied to all tests with either the orifice plates or the DVI venturi nozzles installed.
- 3. The measured pressuie drop was in agreement with the design (predicted) value, implying that the analytic model was valid. The measured total pressure drop for the entire CMT-1 injection flow path was 72.9 in. H2O versus the design value of 67.7 in H2 O at the specified flow rate of 8.24 gpm. For the CMT-2 injection flow path, the measured pressure drop was 71.7 in.
H2O versus the design value of 67.4 in. H2O at the same flow rate as CMT-1. Therefore, the measured and design pressure drops were within 7.7 percent of each other. Note: This observation only applied to the tests with orifice plates ORI-253 and ORI-254 and without DVI venturi nozzles installed. No design (predicted) pressure drop information was available for the tests with the DVI venturi nozzles installed. Design calculations were not required since the pressure drops were within the acceptable range, owaxAlsh.27.non:lt>-081298 4.2-12 REVISION 1
FmAL DATA RzronT m 4. The measured pressure drops in both CMT lines for the case with orifice plates ORI 253 and 5
, ORI 254 but without DVI venturi nozzles installed were very close: 72.8 in. H2 O in CMT-1 and 71.7 in. H2 O in CMT within 1.6 percent of each other.
For the case with the DVI venturi nozzles installed and the orifice plates removed, the pressure drop was 77.8 in. H2 O for CMT-1 and 74.1 in. H2 O for CMT within 4.9 percent of each j other. l In either case, the conclusion was that CMT-1 and CMT-2 flow lines were well balanced. Calculation of Form Loss Coefficient K for DVI Venturi Nezzle ; The second series of flow tests measured pressure drops with the DVI venturi nozzles installed and orifice plates ORI-253 and ORI 254 removed. The tests were run for CMT-1, CMT-2, ACC-1 and ACC-2 only Since the DVI venturi nozzles were installed immediately upstream of the DVI flange (nozzles), they were also part of IRWST-1, IRWST-2, primary sump l and primary sump-2. It is i desirable to estimate the loss coefficient for the venturi nozzle alone. With the known K for the venturi nozzle, the total line resistance for IRWST-1, IRWST-2, primary sump-l and primary sump-2 can be obtained from test data. D De loss coefficient (K) for each of the DVI venturi nozzles was calculated using the followmg l
\
process:
- 1. Line resistances from the bottom of each CMT to the DVI flange (nozzle) for both CMT-1 and CMT-2 lines without orifice plates ORI 253 and ORI-254 and DVI unturi nozzles were calculated using the results from the first series of flow tests (Table 4.216).
The measured pressure drops agreed with the design (predicted) values for the first series of fL 'v tests, i.e., with orifice plates ORI-253 and ORI 254 and without DVI venturi nozzles. The analytic model was acceptable, and included ORI-253 for CMT-1 and ORI-254 for CMT-2. These orifice plates were sharp-edged and well-defined mathematically; therefore, the analytic loss coefficient (K) for the orifice plate was subtracted from the measured line resistance for the line section from the bottom of the CMT to the DVI flange (nozzle), measured by DP-501 or DP-502. Consequently, the line resistances for the CMT lines from the bottom of the CMTs to the DVI flanges, without orifice plates ORI-253 and ORI-25a or DVI venturi nozzles installed were calculated.
- 2. De loss coefficient (K) for each of the DVI venturi nozzles was calculated using test results from the second series of flow tests.
Test results for the second series of flow tests were acceptable since they met the acceptance criteria. Line resistances from the bottom of each CMT to the DVI flange (nozzle), measured owlsw.2inon:Ib-ost298 4.2-13 REVISION 1
FINAL DATA Rzrcar by DP-501 and DP-502, were subtracted from the product of step 1, resulting in the K value for the venturi nozzle.
- 3. Steps 1 and 2 were performed for CMT-1 and CMT-2. The results were used to obtain an average K for the DVI venturi nozzle.
CMT-1 DVI Venturi Nozzle Loss Coemclent Estimation Bottom of CMT-1 to DVI-l flange (filD+K) = 53.6 (With orifice plat ORI 253 only) Design K for orifice plate ORI-253 = 1.27 (From analytic model) l l Thus, fi/D+K for the line from bottom of CMT 1 to DVl-1 flange = 52.4 (Without orifice plate ORI-253 or DVI venturi nozzle) Bottom of CMT-1 to DVl-1 flange (filD+K) = 58.0 (DVI venturi nozzle only) Therefore, DVI-l venturi nozzle K = 58.0 - 52.4 = 5.6 (For CMT-1 line) CMT-2 DVI Venturi Nozzle Loss Coemcient Estimation Bottom of CMT-2 to DVI-2 flange (filD+K) = 52.5 (With orifice plate ORI-254 only) Design K for orifice plate ORI-254 = 2.02 (From analytic model)
'Ihus, fl1D+K for the line from bottom of CMT-2 to DVI 2 flange = 50.5 (Without orifice plate ORI-254 or DVI venturi nozzle)
Bottom of CMT-2 to DVI-2 flange (fi/D+K) = 55.2 (With DVI venturi nozzle only) Therefore, DVI-2 venturi nozzle K = 55.2 - 50.5 = 4.72 (For CMT-2 line) e o:wp60 mis 36w-27.non:ib.os 298 4.2 14 REVISION 1
s l FINAL DATA Rzron K for Each DVI Venturi Nozzle The K value is the same for both DVI venturi nozzles because they were identical; therefore: K venturi = (5.62+4.72)/2 = 5.17 Table 4.2-7 lists raw data for the second series of flow tests after the DVI venturi nozzles were installed and the orifice plates were removed. Flow was fully recovered when it exited into the reactor l- vessel downcomer. Test results show that the pressure drop was negligible (0 or negative psig). i These results demonstrated that the instrumentation was functioning properly and that the major pressure drops were in the lines containing the venturi nozzles. 4.2.5 Accumulator Injection Flow Test Hydraulic measurements of the ACC-1 and ACC-2 injection lines from the bottom of each accumulator to the DVI tees and to the reactor vessel downcomer via the DVI-l and DVI-2 injection nozzles are discussed in this section. The accumulator injection flow test was performed twice in the first series of flow tests only, first with the original orifice plates and then with the DVI venturi nozzles. m 4.2.5.1 Procedures Figures 4.2-11 and 4.2-12 provide schematics of the test arrangement for each accumulator injection line. In the accumulator injection flow test, a flow path was set up using the RNS pump and a temporary tygon hose connecting to the service connection downstream of the RNS pump discharge. Water was pumped from the bottom of the primary sump tank to the top of the accumulator, down to the bottom of the accumulator and the DVI tee to the reactor vessel downcomer via the DVI nozzle. Flow was then directed up to the pressurizer and ADS valves. Finally, flow was directed to the IRWST via the ADS 1-3 sparger located inside the IRWST. The temporary tygon hose was connected from the pump discharge service connection to the top of the accumulator. ADS-1 was opened, and the pump started. The desired flow rate was obtained by throttling RCS-806 (the isolation valve at the service connection downstream of the RNS pump
~ discharge). The desired pressure drops were measured and recorded.
The test was performed the first time with the original configuration, i.e., with orifice plates ORI-253 and ORI 254 and without DVI venturi nozzles. The test was performed the second time with the new ( venturi nozzles installed in each DVI line at the DVI nozzles next to the reactor vessel. Due to the increased head loss in each injection line from the venturi nozzles, it was necessary to remove orifice a:Wish 27.non:lbO81298 4.2-15 REVISION 1
FINA1, DATA Rrroar plates ORI-253 and ORI-254. In addition, the orifice plate in the ACC-1 injection line (ORI-451) was increased from 0.616 in. throat diameter to 0.66 in. throat diameter in order to adjust the line resistance so that it was within the acceptable criteria range. The water temperature for this test was at ambient conditions. 4.2.5.2 Instrumentation Figures 4.2-11 and 4.2-12 provide schematics of the test arrangement and locations of instrumentation. Two differential pressure indications were tr.L::. For ACC-1, DP-401 provided the pressure drop from the bottom of the ACC-1 to DVI l nozzle next to the reactor vessel. DP-128 provided the entrance loss into the vessel from DVI-l nozzle to the reactor vessel downcomer. FMM 401 measured the flow rate through the ACC-1 injection line. The pressure drop measured by DP-401 included the pressure drop across either orifice plate ORI-253 or the DVI venturi nozzle, but not both. For ACC-2, DP-402 provided the pressure drop from the bottom of ACC-2 to the DVI-2 nozzle next to the reactor vessel. DP-129 provided the entrance loss into the vessel from the DVI-2 nozzle to the reactor vessel downcomer. Similar to ACC-1, DP-402 included the pressure drops across either orifice plate ORI-254 or the DVI venturi nozzle, but not both. FMM-402 measured the flow rate through the ACC-2 injection line. 4.2.5.3 Results Each accumulator injection flow path consists of two line sections. The first line section runs from the bottom of the accumulator to a tee junction upstream of the DVI tee. A very short line section connects the two tee junctions. For all practical purposes, these two tee junctions are the same. The second line section is the DVI injection line described in the CMT test result section. This DVI line is common to the CMT, accumulator, and IRWST injection lircs . uni contains either the DVI venturi nozzle or the orifice plate (ORI-253 or ORI-254). Test data are summarized in Table 4.2-8. Data were plotted in Figures 4.2-13 through 4.2-18 as the pressure drop versus the square of the flow rate and fitted with linear regression methodology. The plots show excellent linear results, indicating that Darcy's equation can be used to predict pressure drops through the line. Line Resistance Line resistances (fUD+K) for the accumulator injection lines were calculated using same methodology as the CMT injection flow test. Each injection line has an inside diameter of 1.12 in. (flow area = 0.006842 ft2), c:kp600\l536w-27.non:lb-OS1298 4.2-16 REVISION 1
_-_.m . _ _ _ _ . . _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ . _ . _ _ _ _ _ _ _ . _ . . _ _ FINAt. DATA REPORT Pressure Dron Comparison
' Table 4.2-16 compares the measured pressure drop with the design (predicted) pressure drop and the specified acceptable pressure drop range (acceptance criteria). The following observations were made for both the ACC 1 and ACC-2 injection lines:
1, All measured pressure drops were in agreement with Darcy's equation.
- 2. Both ACC-1 and ACC-2 pressure drop measurements were within the acceptable range. This applied to all tests with either orifice plrtes or DVI venturi nozzles installed.
- 3. The measured pressure drop was in agreement with the design (predicted) value implying that i the analytic model is valid. The measured total pressure drop for the entire ACC-1 injection flow path was 337 in. water versus the design value of 302 in. H2 O at the specified flow rate of 22.7 gpm - within 11.25 percent of each other.
The measured pressure drop for the ACC-2 injection flow path was 324 in. H 2O versus the design value of 305 in. H2O at the flow rate of 22.7 gpm - within 6.3 peicent of each other. Note: This observation only applied to the tests with orifice plates ORI-253 and ORI-254 and without DVI venturi nozzles installed. No design (predicted) pressure drop information was available for the tests with the DVI venturi nozzles installed. Design calculations were not required since the pressure drops were within the acceptable range.
- 4. The measured pressure drops in both accumulator injection lines with orifice plates ORI-253 and ORI-254 but without DVI venturi nozzles installed were very close: 337 in. H2 O in ACC 1 and 324 in. H2 O in ACC within 4.1 percent of each other.
For the case with the DVI venturi nonles installed and the orifice plates removed, the pressure ! drop was 318 in. H2 O for ACC-1 and 345 in. H2 O for ACC within 8.6 percent of each other. In either case, the conclusion was that ACC-1 and ACC-2 flow lines were well balanced. Comparison of DVI Flanne (Nozzle) with Reactor Vessel Downcomer Loss Coefficients The exit loss coefficients from the DVI flange (nozzle) to the reactor vessel downcomer for both DVI-l and DVI-2 were calculated using the test data in Table 4.216 from ACC-1 and ACC-2 flow tests. This line did not include orifice plates ORI-253 and ORI-254, or DVI venturi nozzles. This line was common to CMT-1, CMT-2, ACC-1, ACC-2, IRWST-1, and IRWST-2 injection lines; therefon:, each test should measure the same K. o:\np600(1536w 27.non:Ib-081298 4.2-17 REVISION 1
FrNAL DATA REPORT Since both DVI 1 and DVI-2 nozzle and exit geometries were identical, the different K_ exit suggests that the flow pattems between DVl-1 and DVI-2 at the downcomer were not the same. Theoretically, the K value for sudden expansion is equal to 1. Since a channel guide prevented upward flow, the exit condition was somewhat less than a sudden expansion. The CMT injection flow test results section provides the average K_ exit for each DVI line. 4.2.6 IRWST Injection Flow Test Hydraulic measurements of the IRWST 1 and IRWST-2 injection lines were made from the bottom of the IRWST to the DVI tees and to the reactor vessel downcomer via the DVI-l and DVI-2 nozzles. The test was perfonned in the first series of flow tests only with only orifice plates ORI-253 and ORI-254 installed. Flow testing with DVI venturi nozzles installed was not performed. Since all matrix tests were performed with the DVI venturi nozzles and without orifice plates ORI-253 and ORI-254 installed, the line resistance obtained from the IRWST flow test was adjusted to account for the addition of the DVI venturi nozzles and the removal of the orifice plates. A calculation methodology is provided in the following test results section. 4.2.6.1 Procedures Figure 4.2-21 provides a schematic of the test arrangement for each IRWST injection line. Initially, the primary sump tank was drained to a level below the bottom of the IRWST, and the primary sump injection bypass lines were isolated by closing isolation valves CSS-909 and CSS-910.
'The IRWST injection lines were isolated (by closing RCS-711 and RCS-712), and the IRWST was filled to the level of the IRWST/ primary sump overflow. The reactor vessel and the primary loops were drained to a level slightly above the DVI line penetrations to the reactor vessel. This arrangement assured that the primary sump tank could not provide injection flow into the reactor vessel because the IRWST gravity head was significantly higher and, therefore, blocked the primary sump tank injection lines. Since the DVI lines were filled prior to the start of flow testing, continuous IRWST injection flow at the beginning of the draining process was assured.
RCS-711 and RCS-712 were opened to gravity-drain the water from the bottom of the IRWST to the reactor vessel through the DVI-1 and DVI-2 lines, respectively, simulating the AP600 IRWST draining process. The draining flow rate and the desired pressure drops were measured and recorded. 4.2.6.2 Instrumentation Figure 4.2-21 provides a schematic of the test arrangement and locations of the instrumentation. The IRWST-1 injection line consists of two line sections: one from the bottom of the IRWST to the DVI-l flange (nozzle) next to the reactor vessel, the other from the DVI-1 flange to the reactor vessel 4,2 18 REVISION 1 o:\ap600(1536w-27.non:Ib-o81298
FINAs. DATA REPORT g downcomer. DP-701 provided the pressure drop from the bottom of the IRWST to the DVI l flange.
\
(Orifice plate ORI 253 was installed in this line section.) DP-128 provided the entrance loss into the
' vessel from the DVI-l flange to the reactor vessel downcomer. FMM-701 measured the flow rate through the entire IRWST-1 injection line.
Similarly, the IRWST-2 injection line consisted of two line sections: one from the bottom of the IRWST to the DVI-2 flange (nozzle) next to the reactor vessel, the other from the DVI-2 flange to the reactor vessel downcomer. DP-702 provided the pressure drop from the bottom of the IRWST to DVI-2 flange. (ORI 254 was installed in this line section.) DP-129 provided the entrance loss into the vessel from the DVI-2 flange to the reactor vessel dewncomer. FMM-702 measured the flow rate through the entire IRWST-2 injection line. De water temperature for this test was at ambient conditions. 4.2.6.3 Results Each IRWST injection flow path consists of two line sections. The first'line section runs from the bottom of the IRWST to the DVI tee junction De second line section is the DVI injection line described in the CMT test result section. This DVI line is common to the CMT, accumulator, and IRWST injection lines, and contains either the DVI venturi nozzle or orifice plates ORI-253 or b v= ORI 254. The test data are summarized in Table 4.2-9. Data were plotted in Figure 4.2-22 through 4.2 27 as the pressure drop versus the square of the flow rate and fitted with linear regression methodology. The plots show very good linearity, indicating that Darcy's equation can be used to predict pressure drops through the line. Line Resistance De same methodology used in the CMT injection flow test results section can be used to calculate the line resistance for the IRWST injection lines; however, the IRWST injection line inside diameter varied from section to secten. Thus, IRWST injection line resistance can not be calculated directly from the test results. Since pressure drop is more important from the scaling point of view, line resistances will not be calculated. However, the inside diameters of various sections can be found from the IRWST isometric drawings (Appendix H. Dwg. LKL 920200 and Dwg. LKL920201), and the line resistance can be calculated using Equation 3. Pressure Dron Comparison (without DVI Venturi Nozzles) Table 4.2-16 provides a comparison between the measured pressure drop, the design (predicted) pressure drop, and the specified acceptable pressure drop range (acceptance criteria), ne following
. observations were made for both IRWST-1 and IRWST-2 injection lines:
own536w-27.non:ite1298 4.2-19 REVISION 1
FLNAL DATA REPORT
- 1. All measured pressure drops were in agreement with Darcy's equation.
- 2. Both IRWST-1 and IRWST-2 pressure drop measurements were within the acceptable range.
This applies to the tests with only the orifice plates installed as well as those with DVI venturi installed.
- 3. The measured pressure drop was in agreement with the design (predicted) value, implying that the analytic model is valid. The measured total pressure drop for the entire IRWST-1 injection flow path was 21 in. H 2O versus the design value of 17.8 in. H2 O at the specified flow rat;: of 5.64 gpm - within 18.1 percent of each other.
For the IRWST-2 injection flow path, the measured pressure drop was 19.6 in. H 2O versus the design value of 17.8 in. H2 O at the flow rate of 5.64 gpm. - within 10.4 percent of each other. Note: The acceptance criteria range for both IRWST-1 and IRWST-2 is 13.4 to 26.8 in. H2 O for cases with and without DVI venturi nozzles installed. This observation applies only to tests with orifice plates ORI-253 and ORI-254 and without DVI venturi nozzles installed. No design (predicted) pressure drop information was available for tests with DVI venturi nozzles installed. The acceptance criteria, however, is the same.
- 4. The measun:d pressure drops in both IRWST injection lines for the case with orifice plates ORI-253 and ORI-254 but without DVI venturi nozzles installed were very close: 21 in. H 2O for IRWST-1 and 19.6 in. H2 O for IRWST within 7.3 percent of each other.
Comparison of Loss Coefficients from DVI Flanee (Nozzle) to Reactor Vessel Downcomer The exit loss coefficients from the DVI flange (nozzle) to the reactor vessel downcomer for both the DVI-l and DVI-2 lines were calculated using test data from IRWST-1 and IRWST-2 flow tests (Table 4.2-16). This line did not include orifice plates ORI-253 and ORI-254, or DVI venturi nozzles, but was common to CMT-1, CMT-2, ACC-1, ACC-2, IRWST-1, and IRWST-2 injection lines. Each test should measure the same K. Since the DVI line, flange, and exit geometries were identical for the tests of CMT, accumulator, and IRWST injections, the exit loss coefficient (K) calculated from these three tests should be comparable. The exit loss coefficient from the DVI flange to the reactor downcomer for these three tests in the first series of flow tests are compan:d below: O o^aP600\l536w.27.non:lb-081298 4.2-20 REVISION 1
.m_. ..m_- - . _ _ _ _ _ _ _ _ - _ _ _ _ . . _ . _ _ . . ~ . _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ . _ . _ _
FINA1. DATA Rzroar I b V K (dimensionless) K (dimensionless) i l Source DVI l DVI-2 CMT Flow Test 0.704 0.911 l Accumulator Flow Test 0.754 0.86 IRWST Flow Test 0.607 0.87 Average 0.688 0.880 l l The average of the DVI l exit coefficients is about 20 percent lower than the average of the DVI-2 l coefficients. This difference could be caused by either misalignment of the welded flange on the . l reactor vessel (which would cause a different trajectory for flow entering the downcomer and, therefore, different flow patterns) or protrusion of the flange gasket into the flow area (which would increase pressure loss). I l Since the coefficients for DVI-l are lower and exhibit greater scatter, it is likely that the difference l was caused by some variable factor, such as gasket protrusion. This difference probably had a very small effect on the balance of flows thtuugh the DVIs because the line pressure losses were much larger than the entrance losses. Pressure Dron Comparison (with DVI Venturi Nozzles) l IRWST injection lines with DVI venturi nozzles were not tested; therefore, pressure drop measurements are not available. The following methodology was used to estimate the pressure drop ! through the IRWST injection lines with the DVI venturi nozzles installed. i l
- 1. The pressure drop through orifice plates ORI-253 and ORI 254 was calculated using the analytic K values and the IRWST injection flow rate. This step is valid because the analytic K values for orifice plates ORI-253 and ORI 254 were judged to be acceptable (as discussed in the CMT injection flow test results section.)
- 2. The pressere drop from the bottom of IRWST to the DVI-l or DVI-2 flange at the IRWST injection flaw rate was calculated. This step includes pressure drops across orifice plates ORI-253 and ORI-254,
- 3. 'Ihe pressure drop for the orifice plates was subtracted from the corresponding pressure drops )
obtained in step 2, resulting in the pressure drop through the IRWST lines without the orifice plates or DVI venturi nozzles. O O c:Wish-27.noa:Ib-ost29s 4.2 21 REVISION 1
1 FNAL DATA REM)RT
- 4. The pressure drop through the venturi nozzles at the IRWST injection How rate was calculated using the K_ venturi calculated in the CMT injection flow test results section.
- 5. The results of steps 3 and 4 were added to arrive at the pressure drop from the bottom of the IRWST to the DVI flanges. Since the pressure drop downstream of the DVI flanges (with the venturi nozzles installed cases) was negligible, the sum of steps 3 and 4 is the total IRWST injection flow path pressure drop. CMT and accumulator injection flow tests both showed that the pressure drop from the DVI flanges to the reactor vessel downcomer was negligible for those cases with DVI venturi nozzles installed (Tables 4.2-7 and 4.2 8).
The following calculations were performed using the previous steps: 2 Q = 19.65*d_1 *C*h_Lo.5 (6) Crane 410 where: Q = rate of flow, gpm h_L = head loss, f t. d_1 = throat dia, in. 2 4 C = flow coefficient = (1 beta )/(beta
- K)o.5 (7) beta = d_1/d_2 d_2 = upstream pipe diameter, in.
For ORI-253: K = 1.27 d_1 = 0.899 in. Q = 5.54 gpm d_2 = 1.12 in. Using Equations 4 and 5: l h_L (ORI-253) = 0.18 ft. = 2.16 in. H2 O For ORI-254: l K = 2.023 , d_1 = 1 in. l Q = 5.54 gpm d_2 = 1.12 in. 1 1 o: sap 6am:53ew-27.non:itmsi298 4.2-22 REVISION 1
_ _ _ . -- _ _ _ _ _ . _ _ _ _ _ . _ _ . _ _ _ _ . _ _ . . __. _ m._ _ l FINA1. DrrA REromT l l 1 O Using Equations 7 and 8: I h_L(ORI 254) = 0.5 ft. = 6.0 in. H2O
)
De measured pressure drops from the bottom of the IRWST to the DVI l and DVI-2 flanges were measured by DP-701 and DP-702. These values may be calculated from data listed in Table 4.2-16 : 1 For IRWST-1: DP-701 = 20.7 in. H O 2 (includes ORI-253) For IRWST-2: DP-702 = 19.1 in. H O 2 (includes ORI-254)
'Iherefore, the pressure drop through the lines without orifice plates ORI 253 and ORI 254 are:
IRWST-1 to DVI-l flange pressure drop = 18.5 in. H2O I IRWST-2 to DVI-2 flange pressure drop = 13.1 in. H2 O The pressure drop through the DVI venturi nozzles can then be estimated as follows: For DVI l venturi nozzle: K = 5.17 d_1 = 0.642 in. Q = 5.54 gpm d_2 = 1.12 in. Using Equations 4 and 5: h_L (DVI venturi nozzle) = 0.39 ft. = 4.68 in. H2 O Finally, the pressure drop from the bottom of the IRWST to the DVI flanges can be calculated as follows. The DVI-l and DVI 2 venturi nozzles were identical. IRWST-1 bottom to DVI 1 flange pressure drop = 23.2 in. H2 O (includes DVI-l venturi nozzle) IRWST-2 bottom to DVI l flange pressure drop = 17.7 in. H2O (includes DVI-l venturi nozzle)
- These pressure drops also represent the pressure drops through the entire IRWST-1 and IRWST-2 injection lines, from the bottom of the IRWST to the reactor vessel downcomer.
i' l- The acceptance criteria for the IRWST injection lines is between 13.4 in, and 26.8 in. H2 0. Therefore, I. the as-built IRWST-1 and IRWST-2 injection lines were acceptable. L r% opish27.non:Ibal29s 4.2 REVISION I
FmAL DATA REroRT 4.2.7 Primary Sump Tank Injection Flow Test liydraulic measurements were made for the primary sump-l and primary sump-2 in this section. These lines originate at the bottom of the primary sump tank and run to the DVI tees and the reactor vessel downcomer via the DVI-l and DVI-2 nozzles. These measurements were performed in the first series of flow test only with only orifice plates ORI-253 and ORI 254 installed. The flow test with DVI venturi nozzles was not performed. Since all matrix tests were performed with the DVI venturi nozzles and without orifice plates ORI-253 and ORI-254 installed, the line resistance obtained from the flow test was adjusted to account for the addition of the DVI venturi nozzles and the removal of the orifice plates. De same calculation methodology used in the IRWST injection flow test section was used here to estimate pressure drops with the venturi nozzles installed. 4.2.7.I Procedures Figure 4.2-28 provides a schematic of the test arrangement for each primary sump tank injection line. Initially, both primary sump tank and the IRWST were drained. A blank flange was installed in the IRWST-1 injection line between the IRWST and the tee to the primary sump tank injection line. Another blank flange was installed in a similar manner for the IRWST-2 injection line. These two blank flanges prevented the IRWST from injecting. Isolation valves downstream of the primary sump tank injection lines (RCS-711 and RCS-712) were initially closed to allow filling and venting of the primary sump tank. Primary sump tank isolation valves were also closed (CSS-909 and CSS-910). He reactor vessel and the primary loops were drained to a level slightly above the DVI line penetrations to the reactor vessel. Since the DVI lines were filled prior to the start of the test, continuous primary sump tank injection flow at the beginning of the draining process was assured. The primary sump tank was then filled to the simulated curb level where it would overflow to the secondary sump tank. RCS-711 and RCS-712 were opened to gravity-drain the water from the bottom of the primary sump tank to the reactor vessel through DVI-l and DVI 2 nozzles, respectively. This draining process simulated the AP600 primary sump tank draining process. The draining flow rate and the desired pressure drops were measured and n-cordad. This test was performed with the primary sump tank injection isolation valves (CSS-909 and CSS-910) closed. 4.2.7.2 Instrumentation Figure 4.2-28 provides a schematic of the test arrangement and locations of the instrumentation. I A few modifications to the existing instrumentation were made. For the line from the DVI-l flange l (nozzle) to the downcomer of the reactor vessel, the high side of DP-128 was connected to a l thermocouple penetration at the bottom of the primary sump tank, allowing DP-128 to measure total pressure drop from the bottom of the primary sump tank to the reactor vessel downcomer, including the exit loss. This modification provided the total pressure drop from bottom of the primary tank to oAap600\l536w.27.non:lb-o81298 4.2-24 REVISION 1
, -. . . -_- . . . . . . . _ _ - - ~ _ , .. - - - - .
1 FINAL DATA REPORT eG the downcomer of the reactor vessel. For the line from the DVI-2 Dange (nozzle) to the downcomer b- of the reactor vessel, the high side of DP-129 was connected to a thermocouple penetration at the bottom of the primary sump tank, allowing DP-129 to measure the total pressure drop from the bottom of the primary sump tank to the reactor vessel, including the exit loss. FMM-901 and FMM-902 measured the Dow rate through DVI l and DVI-2, respectively. The primary sump tank level was also measured by LDP 901, 4.2,7.3 Results Raw test data are summarized in Table 4.2-10. Data were in Figures 4.2-29 and 4.2-30 plotted as the pressure drop versus the square of the now rate and fitted by linear regression methodology. The plots show very good linearity, indicating that Darcy's equation can be used to predict pressure drops through the line. Line Resistance The total line resistance for each primary sump tank injection can not be calculated directly from the
' test results because of the varying inside diameters from section to section. Since the pressure drop is more important from the scaling point of view, line resistance was not calculated; however, the inside diameters of various sections can be found in Appendix H, Dwg. LKL 920200 and Dwg. LKL920201.
The line resistance can be calculated using Darcy's equation. Pressure Dron Comparison (without DVI Venturi Nozzles but with ORI 253 and ORI 254 ) Table 4.216 compares the measured pressure drop with the design (predicted) pressure drop and the specified acceptable pressure drop range (acceptance criteria). The following observations were made for both the IRWST-1 and IRWST-2 injection lines:
- 1. The measured pressure drops exceeded the acceptance criteria for the primary sump injection lines. Primary sump-l measured 3.58 in. H 2O and primary sump-2 measured 5.37 in. H2 0.
These pressure drops were compared with'the acceptable range between 1.44 and 2.88 in. H2 O at an injection How rate of 1.16 0.3 gpm. However, numerous data points were obtained as the sump gravity-drained to the reactor vessel. These data provide a good characterization of the line, and no adjustments to the lines were necessary as long as the computer code models the same line resistance.
- 2. The two injection lines were not quite balanced. IRWST-1 dropped 3.58 in. H20; IRWST-2 l dropped 5.37 in. H2O - 50 percent more pressure than IRWST-1.
t t
- v owish-27.non:ltest298 4.2-25 REVISION 1 i
i
FmAL DATA REroRT l Pressure Drop ComDarison (with DVI Venturi Nozzles but without Orifice Plates ORI 253 and ORI 254 } Direct measurement of pressure drops in the primary sump tank injection lines were not made. However, these pressure drops can be estimated using the same methodology as in the IRWST injection flow test section. The following results were obtained using the primary sump tank injection flow rate of 1.16 gpm.
= Pressure drop through ORI-253 at 1.16 gpm = 0.095 in. H2 O e Pressure drop through ORI-254 at 1.16 gpm = 0.265 in. H2O
- Measured IRWST-1 pressure drop at 1.16 gpm = 3.58 in. H2O (including ORI-253)
- Measured IRWST-2 pressure drop at 1.16 gpm = 5.37 in. H2O (including ORI-254)
- Calculated IRWST-1 pressure drop at 1.16 gpm = 3.485 in. H2 O (without ORI-253)
- Calculated IRWST-2 pressure drop at 1.16 gpm = 5.105 in. H2 O (without ORI-254)
- Calculated pressure drop through DVI venturi nozzle at 1.16 gpm = 0.205 in. H2 O
= Calculated pressun: drop through IRWST-1 with DVI venturi nozzle at 1.16 gpm = 3.69 in. H2O e Calculated pressure drop through IRWST-2 with DVI venturi nozzle at 1.16 gpm = 5.31 in. H2O Pressure drops through the injection lines with venturi n.;zzles were estimated to be 3.69 in. H2 O for primary sump-l and 5.31 in. H 2O for primary sump-2. Both exceeded the maximum allowable pressure drop of 2.88 in. H2 O at the flow rate of 1.16 gpm.
4.2.8 Cold Leg Balance Line Injection Flow Test Hydraulic measurements of the cold-leg balance lines from the top of CMT-1 to CL-3 and from the top of CMT-2 to CL-1 were made in this section. These two lines are independent of the DVI lines. Therefore, the test data are applicable with or without DVI orifice plates ORI-253 or ORI-254 or the DVI venturi nozzles installed. These tests were only performed for the first series of flow tests. 4.2.8.1 Procedures Figures 4.2 31 and 4.2-32 provide schematics of the test arrangement for each cold leg balance line. In the cold leg balance line injection flow test, a flow path was set up using the RNS pump to pump water from the bottom of the primary sump tank to the DVI line, down to the DVI tee, and into the reactor vessel downcomer via the DVI nozzle. Flow was then directed to the cold leg and to the cold-leg balance line. Flow exited the cold-leg balance line and flowed to the top of the CMT. Since the CMT discharge was isolated, flow exited the top of the CMT via a temporary line leading to the top of the primary sump tank. The manway cover at the top of the primary sump tank was removed to receive the flow. O oAap60mish-2tnon:ib 08:298 4.2-26 REVISION 1
, FNAI, DATA REPCOT 1
/" He test was initiated by opening the temporary valve to the sump (TV-1 for CMT-1 and TV-2 for i
\
CMT-2), opening the cold leg balance line isolation valve, and starting the pump. De desired flow l rate was obtained by throttling the RNS pump discharge to the DVI isolation valve (RCS-801 for l CMT 1 and RCS-802 for CMT-2). The injection flow rate and desired pressure drops were measured and recorded. , This test was run with orifice plates ORI-253 and ORI 254 installed. No DVI venturi flow nozzles l were used. Test data are applicable to either case. 4.2.8.2 Instrumentation 1
\
. Figures 4.2-31 and 4.2-32 provide schematics of the test arrangement and locations of the '
instrumentation. 1 2 Two differential pressure indications were taken. For CMT-1, LDP-509 was calibrated as a differential pressure cell to provide the pressure drop from the balance line near CL-3 to the top of l CMT-1, he sense lines of LDP-509 were temporarily reversed for this test. The high side of l LDP 509 was tubed to the balance line near CL-3; the low side of LCP-509 was connected to the I inside of CMT-1. FMM 503 measured the flow rate through the CL-1 balance line.
'q .
Lf . For CMT-2, LDP-510 was calibrated as a differential pressure cell to provide the pressure drop from l CL-1 to the top of CMT-2. The sense lines of LDP-510 were temporarily reversed for this test. The i high side of LDP-510 was tubed to the balance line near CL-1; the low side of LDP-510 was connected to the inside of CMT-2. FMM-502 measured the flow rate through the CL-2 balance line, i The low sides of LDP-509 and LDP-510 were located inside the CMTs for this test, implying that the i measured pressure drops included entrance losses from the top of the CMTs. These measurements - also included the pressure drop through the steam distributors (steam diffusers) at the top of the l
' CMTs.
The special sensing line arrangement for LDP-509 and LDP-510 was unique to this test and did not apply to matrix tests. 4.2.8.3 Results Raw test data are summarized in Table 4.2-11. Data were plotted in Figures 4.2-33 and 4.2-34 as Jpressure drop versus square of flow rate and titted by linear regression methodology. The plots show very good linearity, indicating that Darcy's equation can be used to predict pressure drops through the 4- line. osap60mish27.non:ltws129s 4.2-27 REVISION l
FINAL DATA REPORT Line Resistance Line resistances (fIlD+K) for the CMT to cold-leg balance lines were calculated using the same methodology as in the CMT injection flow test section. Each balance line had an inside diameter 2 1.12 in. (flow area = 0.006842 ft ). The line resistances (including the CMT steam diffusers) are l summarized in Table 4.2-16. Pressure Dron Comparison ! Table 4.2-16 compares the measured pressure drop with the design (predicted) pressure drop and the specified acceptance criteria. The measured pressure drop includes the pressure drop through the CMT steam diffusers but the designed (predicted) pressure drop does not. This discrepancy stems from the ; fact that the steam diffusers were added to the test facility later in the facility construction phase after l the balance lines were already built. ne addition of the diffusers was due to the relatively late design changes made to AP600. Acceptance criteria remain unchanged with the addition of the steam diffusers. He following observations were maie for both the CMT-1 and CMT-2 cold-leg balance !ines: 1
- 1. The measured pressure drops in each line were in agreement with Darcy's equation.
- 2. The measured pressure drop was in agreement with the design (predicted) value in the CMT-1 O
balance line. They would agree even better if the pressure drop through the CMT diffuser was included in the design value. The measured total pressure drop for the entire CMT-1 balance line was 5.29 in. H 2O versus the design value of 4.46 in. H2O at the specified flow rate of 4.58 gpm. The acceptance criteria was between 4.0 and 5.4 in. H2O at the same flow mte. Therefore, the CMT-1 balance line met the acceptance criteria. For CMT-2 to the CL-1 balance line, the measured pressure drop was 4.24 in. 2H 0. No design value was available. However, the measured pressure drop met the acceptance criteria.
- 3. The resistances of these two lines were not quite balanced with each other. The CMT-1 balance line measured a pressure drop of 5.29 in. H2 O while CMT-2 balance line measured 4.235 in. H 2O - a 24.9 percent difference. The orifice plates in these lines could have been adjusted to improve the balance of the lines; however, this was not done. The lines were accepted because the results met the acceptance criteria.
4.2.9 ADS 1-3 Flow Test This test measured the pressure drop through the ADS 1-3 lines using the RNS pump to inject water from the primary sump tank and the IRWST into DVI-l and DVI-2, through the pressurizer, and back to the IRWST via the ADS valves. The orifice plates simulating the flow area of the ADS 1-3 valves oAap60m1536w-27.non:ltW)81298 4.2-28 REVISION 1
FINAt. DATA REPORT i t( were not installed in this test. A spacer was installed between the pressurizer outlet and the ADS 1-3 branches (Figure 4.2-35).- A discrepancy was discovered between Dwg. OSU 600203 (Appendix G) and the physical ADS 1-3 separator liquid line configuration. Specifically, the drawing shows orifice plate ORI-659 at the bottom of the ADS 1-3 separator. In reality, this orifice plate was never installed. i 4.2.9.1 Procedures Figures 4.2 35 and 4.2-36 provide schematics of the test arrangement. The primary sump tank and the IRWST were filled to the IRWST sump tank overflow level, and the sump recirculation bypass valves (CSS-909 and CSS-910) were closed. This arrangement ensured proper IRWST level and, consequently, proper backpressure on the sparger. The RNS pump was used as the motive force. Initially, the procedures called for the RNS pump to take suction from the bottom of the primary sump tank and discharge into two parallel lines leading to DVI-l and DVI-2. The field test engineer discovered that the total flow of the two DVI lines was not much different from the flow of only one DVI line. Therefore, the procedure was changed to discharge RNS flow into DVI l only. (RCS-801 was throttled to obtain various flow rates; RCS-802 remained closed.) Flow was directed , from DVI-l to the reactor vessel and the pressurizer, then back to the IRWST via the ADS valves. Only one ADS valve was opened at a time. Since the IRWST was initially filled to the overflow level, any additional flow into the IRWST would flow back to the primary sump tank and creating a closed-loop operation. ADS orifices ORI-655, ORI-656 and ORI-657) were not installed in the lines for this test in an attempt to characterize the ADS 1-3 lines. The RNS pump was started and one of the ADS valves was opened. 'Ihrottling RCS-801 provided the desired flow rate. Test data were recorded once the
. desired flow rate was obtained.
4.2.9.2 Instrumentation L Figures 4.2-35 and 4.2-36 provide a schematic of the test arrangement and locations of the instrumentation. The following instrumentation was used in the test:
- PT-604 Pressure upstream of ADS branch lines (Figure 4.2-35) t
= PT-603 Pressure at the top of the pressurizer
- L i
e FT-605 Pressure at the top of the ADS 1-3 separator (steam side but before the steam separating element) l . oAsp60mlsh-27 nortib-ost29s 4.2-29 REVISION 1 l j __ . __ - _ _ _ .
RNAL DATA REFORT
= FMM-205 Magnetic flow meter at DVI-1
= FMM-206 Magnetic flow meter at DVI-2
= FMM-601 Magnetic flow meter at ADS 1-3 separator liquid line
= FDP-605 Pressure drop across the flanges at the normal ORI-655 location (representing ADS-1) (Note: ORI-655 was not installed in this test)
= FDP-604 Pressure drop across the flanges at the normal ORI-656 location (representing ADS-2)
= FDP-606 Pressum drop across the flanges at the normal ORI-657 location (repmsenting ADS-3)
= LDP-701 IRWST water level measurement In addition to this instrumentation, a measurement tape was installed at the top of the reference leg for LDP-701 (IRWST level) with "zero" resting at the top of the reference leg. A temporary tygon hose was routed from root valve RCS-624 and upward. RCS-624 was the isolation valve of the bottom static pressure tap of LDP-601 (ADS l-3 separator level). This temporary level indication gave the separator level with its zero datum point at the top of the IRWST level reference leg. LDP-701 and the tygon hose level indications were recorded as water was pumped through the tect flow path. The relative distances between the tygon hose, RCS-624, LDP-701, and the ADS 1-3 separator inlet and outlet are shown in Figure 4.2-36.
4.2.9.3 Results Raw test data are summarized in Table 4.2-12. The nature of the piping and separator arrangement makes the test data rather difficult to understand. During the course of testing, the field engineer opened RCS-620, located at the top of the separator, and came to the conclusion that, because there was no water issuing from the vent line, the lines leading to the separator were not full. The Quick Look Report (LTCT-T2R-002) reported this conclusion also. However, as the test data were analyzed more carefully, there was evidence that the ADS separator inlet lines were completely filled with wat and that pressure in the separator forced the water through the liquid drain line. This evidence is discussed below. Comparison of Analytic Results with Measured Results l An analytic model (Appendix A) was set up to calculate the pressure drop from the pressurizer to the inlet of the separator using the measured flow rates for both the ADS-I and ADS-2 lines. The ADS-3 l line was identical to the ADS-2 line. Table 4.2-13 summarizes the results of the calculation. 1 I l o:\ap600\l536w.27.non:lb 081298 4.2-30 REVISION 1
i FINAt. DATA REPORT l i The test recorded pressure at the pressurizer (PT-603) and in the separator (171 %5). Since the test ( was conducted with only one ADS line open at a time, the difference between PY 603 and PT-605 represented the pressure drop through the line from the pressurizer to the separator,nlet. The measured pressure drops were compared with the predicted pressure drops in Table 4 2-13. The l measured pn:ssure drops agree with the predicted pressure drops, indicating that the imes were full. Figures 4.2-37,4.2-38, and 4.2-39 show the pressure drop versus the flow rate square for all three ADS lines using raw test data. A pressure offset exists when there is no flow. This pressure offset is consistent from ADS-1 to ADS-3, and was most likely due to the initial zero shift of the pressum cells. The plots also show that ADS-1 test results agree with the Darcy's equation but ADS-2 and ADS-3 do not. Therefore, only ADS-1 results were analyzed in more detail. The pressure offset for the ADS 1 test was corrected, and the results are summarized in Table 4.2-13 and plotted in Figure 4.2 40. The adjusted pressun: drops for the ADS-1 test agree with the predicted values. The analytic pressure drop is plotted as a function of the flow rate square in Figure 4.2-41. The derived equations in these two plots are virtually the same. Therefore, the adjusted ADS-1 test results are valid. The ADS-2 and ADS .1 test results were not acceptat& because of scattered data. Herefore, the
- pressure drop through the ADS-2 and ADS-3 parallel lines mg he calculated. The following steps were used:
- 1. Pressure drop through ADS-1 line was estimated.
Note: The ADSsl overall line consists of three sections. The first sectiori begins at the pressurizer and ends at the teejunction of the ADS 1-3 parallel lines. he second line section is the ADS-1 itself (parallel to ADS-2 and ADS-3). The third line section begins downstream of the ADS 1-3 parallel lines and ends at the separator inlet. The first and third lines are common to all three ADS parallel lines.
- 2. The calculated pressure drop for the ADS-1 parallel line was subtracted from the measured overall line pressure drop. The ADS-1 parallel line is relatively short and simple, and the analytic result should be valid.
- 3. Pressure drops for the ADS-2 or ADS-3 lines were calculated and added to the result of step I to arrive at the pressure drrap thre'irh the ADS-2 or ADS-3 lines.
Pressure Dron throuah Loon Seal Line The pressure drop from the bottom of the ADS 1-3 separator to the IRWST free surface consists of ! two sections: from the separator to the tygon hose connection (RCS-624) and from the tygon hose o:Wl536w-27.non:!b-081298 4,2,31 REVISION 1
FINAL, DATA REronT , z connection to the free surface of the IRWST. The reason for this breakdown is that the level measurements by the tygon hose and LDP-701 are very accurate and should be utilized fully. DP_ loss (from point 0 to point 2) = RHO *(hl - h3) (8) 2 Po = RHO *(h1 - h0) - RHO *Vo /(2*g) (9) P4= RHO *(hl - h4) + DP_ loss (point 4 to point 0) (10) Point 0 = Tygon hose connection level Point 1 = Tygon hose meastred water level Point 2 = downstream of sparger where velocity is negligible Point 3 = IRWST water level measured by LDP-701 Point 4 = Water level in ADS 1-3 separator Figure 4.2-36 provides the physical locations of assigned node points. Raw test data are summarized in Table 4.2-14. The line loss derived from the raw data is also included in the table. Equation 3 was used to estimate the pressure loss from the tygon hose connection to the sparger using the tygon hose and LDP-701 measurements. Figures 4.2-42,4.2-43, and 4.2-44 are correlations of the pressure drop as a function of the flow rate for ADS 1, ADS-2, and ADS-3, respectively. The plots show that the data for ADS-1 and ADS-3 agree with Darcy's equation. The ADS-2 data are scattered at low flow. However, all three plots show similar slopes and pressure drop offsets at no-flow conditions. His was to be expected because the liquid line is common to ADS-1, ADS-2, and ADS-
- 3. Figures 4.2-42 (ADS-1) Oi 4.2-44 (ADS-3) should be used to characterize the line.
Equation 10 can be used to estimate the pressure loss from the ADS 1-3 separator to the tygon hose connection. However, there are two unkowns in the equation: water level inside the separator and pressure loss through the line. Since the test did not measure the water level in the separator, there is not enough information to characterize tw short line section. However, since this line is a straight line with an entrance loss, the line loss should be negligible. These losses can be accurately calculated using Darcy's equation. Since there are insufficient test data, no conclusion can be drawn for this line. 4.2.10 Normal Residual Heat Removal Flow Balance his test was performed to assure balance flow at the two parallel lines at the discharge of the RNS pump. It was essential to have relatively equal flow delivered to both DVI-l and DVI-2. 4.2.10.1 Procedures Figure 4.2-45 yovides a schematic of the test arrangement. O o:\np600(1536w-27.non:lb-081298 4.2-32 ICVISION 1
FINAL DATA REPORT RNS flow balance was performed using the RNS pump to inject water from the IRWST and primary b sump to the DVI-l and DVI-2, through the pressurizer and ADS-1 valve, then back to the IRWST. The ADS 1 isolation valve (RCS-601) was used to throttle the flow rate to a pre-determined value of about 30 gpm. Total flow and flow to DVI-2 were measured. Flow to DVI-l was calculated. DVI-l ! and DVI-2 flow rates were compared. If they were not reasonably equal, orifice plate ORI-853 in the l ! line leading to DVI-l was adjusted, and the process was repeated until balanced flows were achieved. l 4.2.10.2 Instrumentation i Figure 4.2-45 provides a schematic of the test arrangement and locations of the instrumentation. FMM-805 measured total flow injected. FMM-803 measured RNS flow to DVI-2 only. 4.2.10.3 Results Because of changes in orifice locations in these lines, the original test data are no longer valid. About 5 minutes of RNS injection data (Table 4.2-15) was taken from Matrix Ten SB04 and was used to evaluate RNS flow balance. Average RNS injection flow in DVI 2 (FMM-803) was 9.96 gpm. Average total injection flow (FMM-805) was 20.12 gpm. Flow in DVI-I was calculated by subtracting the average of FMM-803 from the average of FMM-805. The calculated flow imbalance
-O between DVI-1 and DVI 2 was less than 2 percent, within the i10 percent required.
i' I
\.
. ow1536w.27.non:ib-os1298 4.2-33 REVISION 1
FINAL DATA REPORT TABLE 4.21
SUMMARY
OF REACTOR VESSEL AND PRIMARY LOOP INSTRUMENTATION USED IN FLOW TESTS Flow Test Series ID Description (First) (Third) LDP-102 Top of downcomer at 180 degrees to HL-2 Yes (U No DP-I l l Bottom of upper grid in core to top of upper core plate Yes Yes DP-il4 Bottom of upper support plate to top of upper support plate Yes Yes DP-121 CL-l nozzle to downcomer at 22.5 degrees Yes Yes DP-122 CL-2 nozzle to downcomer at 292.5 degrees Yes Yes DP.123 CL-3 nozzle to downcomer at 112.5 degrees Yes Yes DP-124 CL-4 nozzle to downcomer at 202.5 degrees Yes Yes DP-125 About 4.91 in. above upper core plate to bottom of HL-1 flange Yes Yes DP-126 About 4.9) in. above upper core plate to bottom of HL-2 flange Yes Yes DP-128 Bottom of DVI l flange next to reactor vessel to downcomer at top No Yes of core elevation and 270 degrees DP-129 Bottom of DVI-2 flange next to reactor vessel to downcomer at top No Yes of core elevation and 270 degrees DP-130 Top of upper head to top of downcomer (bypass hole DP) Yes Yes DP-201 Bottom of RCP-1 discharge flange to bottom of CL-1 nozzle Yes Yes DP-202 Bottom of RCP-2 suction flange to bottom of RCP-2 discharge flange Yes Yes DP-203 Bottom of RCP-1 suctior flange to bottom of RCP-1 discharge flange Yes Yes DP-204 Bottom of RCP-2 discharge flange to bottom of CL-2 nozzle Yes Yes DP-205 Bottom of RCP-3 suction flange to bottom of RCP-3 discharge flange Yes Yes DP-206 Bottom of RCP-4 suction flange to bottom of RCP-4 discharge flange Yes Yes DP-207 . Bottom of RCP-3 discharge flange to bottom of CL-3 nozzle Yes Yes DP-208 Bottom of RCP-4 discharge flange to bottom of CL-4 nozzle Yes Yes DP-209 Bottom of HL-1 flange to inlet of SG-1 Yes Yes DP-210 Bottom of HL-2 flange to inlet of SG-2 Yes Yes DP-211 SG-1 U-tube entrance loss (about 6.4 in. below tube sheet to short- Yes(D No tube entrance) o:\ap600(1536w 27.non:lt481298 4.2-34 REVISION 1
. - . - - - . -- - . - . . _ _ - . - - --_ -- - - - = - . _ .
FtNAt. DATA REPORT TABLE 4.21 (Continued)
SUMMARY
OF REACTOR VESSEL AND PRIMARY LOOP INSTRUMENTATION USED IN FLOW TESTS Flow Test Series ID Description (First) (Third) DP-211 SG-1 U-tube inlet to U-tube outlet pressure drop (both taps at 6.4 in. No Yes below tube sheet) DP-212 SG-2 U-tube entrance loss (about 6.4 in. below tube sheet to short- Yes(D No tube entrance) DP-212 SG-2 U-tube inlet to U-tube outlet pressure drop (both taps at 6.4 in. No Yes below tube sheet) DP-213 SG 1 U-tube entrance to exit (special hookup - both taps at 6.4 in. i Yes(l) No ! below tube sheet) DP-213 SG-1 fong tube exit loss - from tube sheet to 6.4 in. below tube No Yes sheet DP-214 SG-2 U-tube entrance to exit (special hookup - both taps at 6.4 in. YesO ) No l below tube sheet) (m ( ) DP-214 SG-2 long tube exit loss - from tube sheet to 6.4 in. below tube No Yes(D sheet DP-220 Special hookup from CL-4 nozzle to HL-2 nozzle - overall reactor vessel pressure drop DP-221 Special hookup frc.m bottom of CL-1 nozzle to bottom of No Yes(D downcomer at 0 degrees DP-801 Special hookup from bottom of HL-2 to PRHR HX inlet No Yes(D DP-802 Special hookup from PRHR HX inlet to outlet No Yes(U DP-803 Special hookup from PRHR HX outlet to SG-2 hot-leg channel head No Yes(U LDP-105(2) Downcomer pressure drop from DVI level to top of core level No Yes LDP-107(2) Bottom of lower plenum at 247.5 degrees to downcomer at 270 No Yes degrees and lower core plate 1: vel LDP-108(2) Bottom of lower plenum at 0 degree to inside core and just above No Yes lower core plate at 0 degree LDP-109(2) Differential pressure inside core from top of lower core plate to No Yes bottom of middle grid plate LDP-Il0(2) Differential pressure inside core form bottom cf middle grid plate to No Yes y/ bottom of top grid plate oiap600(1536w-27.non:lb-082598 4.2-35 REVISION 1
FINAL DATA REPORT TABLE 4.21 (Continued)
SUMMARY
OF REACTOR VESSEL AND PRIMARY LOOP INSTRUMENTATION USED IN FLOW TESTS Flow Test Series ID Description (First) (Third) LDP-ll2(2) Differential pressure inside upper plenum from top of upper core No Yes plate to about 10 in below center line of hot its LDP-116(2) Downcomer overall differential pressure from bottom of lower No Yes plenum at 247.5 degrees to top of downcomer at 270 degrees LDP-ll8(2) Downcomer differential pressure from lower core plate level at No Yes 270 degrees to top of core level at 270 degrees LDP-209(2) SG-1 inlet DP from inlet to 6.4 in. below tube sheet No Yes LDP-210(2) RCP-2 suction flange to SG-2 cold-leg side tube sheet (tap at 6.4 in. No Yes below tube sheet) LDP-211(2) RCP-3 suction flange to SG-1 cold-leg side tube sheet (tap at 6.4 in. No Yes below tube sheet) LDP-212(2) RCP-4 suction flange to SG-2 cold-leg side tube sheet (tap at 6.4 in. No Yes below tube sheet) LDP-213(2) RCP-1 suction flange to SG-1 cold-leg side tube sheet (tap at 6.4 in. No Yes below tube sheet) LDP-214(2) SG-2 inlet differential pressure from inlet to 6.4 in. below tube sheet No Yes TF-107 Temperature at bottom of CL-1 nozzle No Yes TF-108 Temperature at bottom of CL-2 nozzle No Yes TF-142 Temperature at bottom of HL-2 flange next to reactor vessel No Yes TF-143 Temperature at bottom of HL-1 flange next to reactor vessel No Yes TF-201 Temperature at bottom of RCP-1 suction flange No Yes TF-202 Temperature at bottom of RCP-2 suction flange No Yes TF-203 Temperature at bottom of RCP-3 suction flange No Yes TF-204 Temperature at bottom of RCP-4 suction flange No Yes FT 101 Pressure at bottom of CL-1 nozzle No Yes l'T-102 Pressure at bottom of CL-2 nozzle No Yes FT-103 Pressure at bottom of CL-3 nozzle No Yes 9 oAap60(A1536w-27.non lb-082798 4.2-36 REVISION 1
l l FINAt, DATA REPORT l i
4
!( / TABLE 4.21 (Continued)
SUMMARY
OF REACTOR VESSEL AND PRIMARY LOOP INSTRUMENTATION USED IN FLOW TESTS l l Flow Test Series l ID Description (First) (Third) FT-104 Pressure at bottom of CL-4 nozzle No Yes l FT-107 Pressure at top of reactor vessel upper head No Yes I'T-108 Pressure at bottom of reactor vessel No Yes . 1 FT-109 Pressure at bottom of DVI l flange (nozzle) No Yes FT 110 Pressure at bottom of DVI-2 flange (nozzle) No Yes FT-i l l Pressure at top of reactor vessel downcomer, just below bypass holes No Yes PT-112 Pressure at bottom of reactor vessel downcomer,just above lower No Yes plenum PT-113 Pressure below mid-core spacer grid No Yes l'T-201 Pressure at top of SG-1 fong tube No Yes FT-202 Pressure at HL-2 before inclined line leading to SG-2 No Yes FT-204 Pressure at top of SG-2 long tube No Yes PT-602 Pressure at steam outlet of pressurizer (narrow range) No Yes I'T-604 Pressure at steam outlet of pressurizer (wide range) No Yes Note: (1) Special hookup for the test (2) All LDPs were calibrated and used as differential pressure transmitters l l o:\ap600(15%w-27.non:lt>481298 4.2-37 *dEVISION 1
FINAL, DATA REPORT Tables 4.2 2 through 4.216 on pages 4.2 38 through 4.2-63 are not included in this nonproprietary document. t i 1 I l l l l 4 i 1 l 1 l l o:bp60(A1536w 27.non:Ib-082598 I 4.2-38 REVISION 1 ( 1
l FmAL DATA RzroRT l l l ' ,D l
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1 l l 1 FINAL DATA REront i
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FINAL DATA REPORT Figures 4.2-3 through 4.210 on pages 4.2-66 through 4.2-69 are not included in this nonproprietary document. O i i I I I l O' o:Wl536w-27a.non:Ib4)83198 4.2-66 REVISION 1
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O 4 N, Y c:\l 536w Rev l\l 536w-27b.non:l b-081798 4,2-70 REVISION 1 9
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l FINrt. DATA REPORT 1 1 Figures 4.213 through 4.2 20 on pages 4.2 72 through 4.2 75 are not included in this nonproprietary document. j
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FINAL DATA REronT Figures 4 2 22 through 4.2 27 on pages 4.2-77 through 4.2 79 are not included in this O nonproprietary document. O O c:\l 5hRev i tl 5 h-27c.non: I b-083198 4.2-77 REVISION 1
i l F5NAI. DATA REPORT I _ l O
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c:\l536wRevl\l536w 27c.non:Ib-081798 4,2 80 REVISION 1
_m._._..._.. _ . _ _ _ . . _ . . . _ _ _ . . _ . _ . . _ _ . _ _ _ _ _ FINAL DATA REM)RT A 4 4 Figures 4.2 29 and 4.2 30 on page 4.2 81 are not included in this nonproprietary document. l1 i 1 1 4 O o:\l536wRevi\l536w-27d.non:lb-082598 4.2 81 REVISION 1
1 l I i RNAL DATA REPORT l l l i
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FINAL DATA REPORT f~%\ t.g
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_ _ - - . - . _ - - - _ .- .. . _ . . . _ - . . ~ . _ _ . - FINAL DATA REPORT Figures 4.2-33 and 4.2-34 on page 4.2 84 are not included in this nonproprietary document. O O oA1536* Rev i\l 536w-27d.non: I b-082598 4,2 84 REVISION l
1 l 1 FINAL DATA REPC2T i l l 1 f%. G i narrr.nse
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WESTINGHOUSE PROPELIETA1Y Ct. ASS 2 FINA1. DATA REPORT l O l l[ A
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l l WESTINCHOUSE Paorm ETrav class 2 FINAL DATA REPORT j (
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Figures 4.2 37 through 4.2-44 on pages 4.2-87 through 4.2 90 are not included in this nonproprietary document. 4 i l t l t I 1 r i O i
, cA1536wRevi\l536w-27e.non lb-082598 REVISION 1 j 4.2-87 t
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FINA1. DATA Rzroar 4.3 HS01 Ambient Heat Losses The hot functional test, OSU-HS-01, was performed on May 22 through 24,1994. De objective of the test, to obtain data under a variety of conditions to determine OSU facility characteristics, was achieved through a series of steps summarized in the following six categories: Ambient heat losses of the facility at 100*,200*,300*, and 400*F. The pressurizer heater maintained facility pressure at [ ]* and the (SGs) were isolated to prevent loss due to steam flow. Steady-state characteristics of the facility at power levels of 100 kW,300 kW,500 kW, and 600 kW (or 100 percent power). The passive residual heat removal heat exchanger (PRHR HX) was isolated. SG pressure and level controls were set at automatic. Performance of the PRHR HX and characteristics of facility operation during forced convection (RCPs running) at power levels of 300 kW,500 kW, and 600 kW. Performance of SGs and characteristics of facility operation during natural circulation (RCPs tripped) at power levels of 300 kW,500 kW, and 600 kW. Performance of CMTs and characteristics of the facility operation duriig natural circulation j (RCPs tripped) at 500 kW. The PRHR HX was isolated. SG pressure and level controls were ! set at automatic. This results in a total of 18 distinct facility configurations and conditions for which 16 data files were created from the DAS at OSU. The raw data file identifscation, the corresponding procedure step number, the step description, and the date and time of the recording appear in Table 4.31. For the purposes of data validation, three of these test data files were closely reviewed. They were the first data file (U0000) with the facility at full pressure and 100*F; the last data file (U1636) with the facility at 600 kW power, RCPs tripped, and the PRHR HX in service; and one data file from the middle of the test (U1451) with the facility at 600 kW power, RCPs running, and the PRHR HX not in service. This review approach yielded acceptable results since the entire test was performed continuously without intermption. Therefore, if a channel did not fail at the start of the test (U0000) but failed somewhere in the middle or towards the end, it would show up as failed in the review of the last step / data file (U1636). As an extra safeguard, one of the steps / data files performed during the middle of test (U1451) was reviewed to determine if any channels acted as an exception. The review confirmed the accuracy of all data recorded throughout the test, except for the failed instmments noted in Table 4.3-2. There are, however, instances where data recorded are accurate but irrelevant for the purpose of determining facility characteristics. These are instances when steady-state conditions were a:usw.24.non:Isosi?9s 4.3-1 REVISION 1
FINAL DATA REPORT unachievable or when ponions of the facility were being used for component cooldown in between and during steps. The details of these circumstances are discussed in the following subsections. 43.1 Ambient IIeat Loss Data at 100*F (Test Procedure Step 4.13) ne recorded data in file U0000 are unsuitable for ambient heat loss calculations at 100"F. It was virtually impossible to maintain stable conditions at the low steam and feedwater rates necessary to maintain 100*F. All possible combinations of AUTO / MANUAL between reactor, feedwater, and steam controllers were attempted with no success. The recorded data are accurate, but since the acceptance criteria (i.e., steady-state conditions at 100 F) were not met, the data should not be used for ambient heat loss calculations. 4.3.2 CMT Cooldown On completion of step 4.43 (CMT Natural Convection Perfonnance Test), test engineers initiated a cooldown of the CMTs that continued through the first [ ]'6* of step 4.5.1. Using the RNS pump, the test engineers injected cool water into the CMTs and discharged the warmer water to other areas of the system. This action is acceptable, since the CMTs and the equipment used to cool them down are outside the test boundary for step 4.5.1. Since data recorded by the DAS include all instrument channels, the effects of the CMT cooldown on a representative sampling of instruments outside the test boundary are presented in Table 4.3-3. 43.3 IRWST Cooldown On completion of a step involving forced or natural cir:ulation flow through the PRHR HX, test engineers initiated a cooldown of the IRWST tnat contiaued through the next step of the procedure. The test engineers injected cool water into the IRWST and discharged the warmer water via the primary sump overflow. This action is acceptable, since the IRWST and the equipment used to cool it down are outside the test boundary for that particular-step. Since data recorded by the DAS include all instrument channels, the effects of the IRWST cooldown on a representative sampling of instruments outside the test boundary are presented in Table 43-4. This applies to step / data file 4.4.1/Ul 441. 43.4 Conclusion The purpose of the hot functional test was to obtain data under a variety of steady-state conditions to determine OSU facility characteristics. A thorough review of three data files, one at the beginning. I i one at the end, and one in the middle of the test, proved the data in those files to be accurate and 1 1 l l complete. Therefore, it is deduced that all of the data in the other 13 data files for this test are also l accurate and may be used by the test analysts for determination of OSU facility characteristics. l l o:\l536w 24.non:lb-081798 4.3-2 REVISION 1 l
FINAt DATA REPO." (~' TABLE 4.31 RAW DATA FILE IDENTIFICATION AND DESCRIPTION l Data File Procedure Recording l ID Step No. Step Description Date Time Span U0000 4.1.3 Ambient heat loss determination at 100*F 5/22/94 19:49 to 20:23 Ul415 4.1.5 Ambient heat loss determination at 200 F 5/22/94 21:06 to 21:38 Ambient heat loss determination at 300*F 5/22/94 23:17 to 23:52 Ambient heat loss determination at 400*F 5/23/94 01:20 to 01:55 U1424 4.2.4 100 kW power level testing 5/23/94 04:04 to 04:39 Ul431 4.3.1 300 kW power level testing 5/23/94 05:52 to 06:42 U1432 4.3.2 PRHR forced flow at 300 kW 5/23/94 07:19 to 08:06 U1441 4.4.1 500 kW power level testing 5/23/94 09:03 to 09:49 U1442 4.4.2 PRHR forced flow at 500 kW 5/23/94 10:14 to 10:47 U1443 4.4.3 CMT natural convection performance at 500 kW 5/23/94 13.43 to 14:25 without PRHR HX in service U1451 4.5.1 600 kW power level testing 5/23/94 16:46 to 17:22 U1452 4.5.2 PRHR forced flow at 600 kW 5/23/94 18:48 to 19:35 U1610 4.6.10 300 kW primary system natural convection 5/23/94 21:25 to 21:49 without PRHR HX in service U1616 4.6.16 300 kW primary system natural convection with 5/23/94 22:10 to 22:27
~
PRHR HX ia service U1623 4.6.23 500 kW primary system natural convection 5/23/94 22:44 to 23:18 without PRHR HX in service U1626 4.6.26 500 kW primary system natural convection with 5/23/94 23:33 to 23:48 PRHR HX in service U1633 4.6.33 600 kW primary system natural convection 5/24/94 00.05 to 00:39 without PRHR HX in service U1636 4.6.36 600 kW primary system naturrJ wnvection with 5/24/95 00:55 to 01:07 PRHR HX in service J O oAI536w.24.non:ll>081798 4.3-3 REVISION l
FINAL DATA REPORT TABLE 4.3 2 FAILED INSTRUMENTATION Instrument No. Description DP-212 SG-2 long tube exit loss HFM-112 Heat losses from CL-3 flange at reactor HFM-205 Heat losses from CL-3 flange at reactor HFM-505 Heat loss CMT-1 top HFM-510 Heat loss CMT-2/CL-1 balance line HFM-603 Heat loss pressurizer surge line HPS-509-1 CMT-1 cold-leg balance line heat transfer coel'icient HPS-509-2 CMT-1 cold-leg balance line heater dT above fluid temperature HPS-509-3 CMT-1 cold leg balance line fluid temperature SC-TH-102-4 TH-102-4/ group I heater rod signal conditioner TF-150 Upper reactor vessel layer H H at 186.2 degrees TFM-l l2 Thermocouple, type T, for HFM ll2 { TFM-205 Thermocouple, type T, for HFM-205 TFM-505 Thermocouple CMT-1, type T. for HFM-505 TFM-510 Thermocouple CMT-2/ cold-leg balance line, type T, for HFM-510 t i TFM-603 Thermocouple pressurizer surge line, type T, for HFM-603 l l 9 oA1536w-24.non:lb-081798 4.3-4 REVISION 1
i FINAL DATA REroar 1 i TABLE 4.3 3 INSTRUMENTATION OUTSIDE TEST BOUNDARY BUT AFFECTFS BY CMT COOLDOWN l Instrument No. Description l FMM-803 R'M to L'i.-2 flow i FMM-805 RNS discharge flow LDP-401 ACC-1 level LDP-402 ACC-2 level l LDP-502 CMT-2 wide-range le% LDP-505 CMT-1 wide-range leves (top) LDP-507 CMT 1 wide-range level PT-501 CMT-1 pressure PT-502 CMT 2 pressure Irr-802 RNS discharge pressure TF-549 CMT-1 discharge line
-_/ TF-550 CMT-2 discharge line 3
0:\l5h-24.non:Ib 081798 4.3-5 REVISION 1
FINAL, DATA REPORT TABLE 4.3-4 INSTRUMENTATION OUTSIDE OF TEST BOUNDARY BUT AFFECTED BY IRWST COOLDOWN Instnament No. Description FMM-701 IRWST-1 inject flow FMM-703 IRWST overflow FMM-803 RNS to DVI-2 flow FMM-805 RNS discharge fiow LCT-701 IRWST load cell LCT-901 Primary sump load cell LDP-502 CMT-2 wide-range level LDP-505 CMT-1 narrow-range level (top) LDP-507 CMT-1 wide-range level LDP-701 IRWST level LDP-901 Primary sump level LDP-902 Secondary sump level PT-802 RNS pump discharge pressure O c:\l5Mw-24.non:lt>481798 4.3-6 REVISION i
(, ,- _ . . . - - - - - - - - - - - - - ~ _ - - - . - - - - - - - - . - . . - FINAI. DATA REPC:fr 5.0 MATRIX TESTS RESULTS J l This section contains the results of the formal matrix test program. He objective of the test program, as defined iri Section L was to evaluate the passive safety system response to a range of small-break loss-of-coolar.t accidents (SBLOCAs) during transition to long-term cooling. The reference transient is a 2-in. cold-leg break with failure in the automatic depressurization system (ADS) in one valve in one ADS-4 line. His test was duplicated and repeated with variations in break size, backpressure, and operation of nonsafety systems. Other SBLOCAs were performed to evaluate the effect of moving the ; break k> cation to the hot leg, core makeup tank (CMT) balance line, and direct vessel injection (DVI) I line. Some tests did not involve LOCAs, but evaluated the inadvertent actuation of the ADS and an 1 inadvertent S signal. The matrix test report is organized by reference tests and comparison tests. The reference tests include: the cold-leg break with a single failure (Matrix Tests SB01 and SB18), a 2-in cold-leg break { with nonsafety systems in operation (Matrix Test SB04), a double-ended guillotine (DEG) break of the l CMT/ cold-leg balance line (Matrix Test SB10), a DEG break of the DVI line (Matrix Test SB12), j inadvertent ectuation of the ADS (Matrix Test SB14), an inadvertent S signal (Matrix Test SB31), and ! a 2-in break in the hot leg (Mati;x Test SB15). Comparison tests are similar to reference tests with f.#9mphasis on system modifications and compcaent responses in relation to the reference test. For example, the effect of a larger size break on a cold-leg LOCA response is measured in Matrix Test SB21, a double top and bottom 4-in. break, and is compared with the results of a 2-in. break evaluated J in Matrix Test SB01. Four other cold-leg break comparisone i re made in Subsection 5.1. There were l no comparison tests for the inadvertent S signal or the hot-leg t>reak. Each test is organized in a similar format. The system configuration and initial conditions are reviewed, the inoperable instruments are identified, and the sequence of events is listed. The test results and evaluation are reviewed on both a system and a component level. The test period is divided into the initial depressurization phase, which occurs from simulated break initiation to ADS-1 l actuation; the ADS phase, which is the period from ADS-1 actuation to the start of IRWST injection; l and the IRWST injection phase, which occurs from the start of IRWST injection to the end of the test. Facility responses are documented by data plots, referenced as figures in the text. i l The discussion of each test perfonnance is based primarily on instrument indications. When the components are losing inventory and coolant is flashing to steam, a steam percent is calculated based on compensated levels. These numbers should be considered preliminary. The computer program j used to calculate the compensated levels is being developed as part of the AP600 Low Pressure Integral Systems Test at Oregon State University. The program has been through initial configuratio i, and final validation will be completed as part of the AP600 Low-Pressure Integral Systems Test at
- Oregon State University, Test Analysis Report, WCAP-14292,C' which will review the steam percent i estimations from this report as part of the mass and energy balance calculation and update these values as required.
i c:\l 536wRev l\l 536w\l 536w.33.non:I b-081798 $.1 REVISION 1
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FINAL, DATA REPORT l , 5.1 ' Cold Leg Breaks with a Single Failure ! C There were six tests performed to evaluate the cold-leg small-break loss-of-coolant accident (SBLOCA) without operation of the nonsafety systems and a failure in one of the two automatic l depressurization (ADS) valves in one ADS-4 line. Test SB01 simulated a 2-in break in the bottom of the cold leg. This test was the first matrix test performed; it was repeated at the end of the test l program SB18 to ensure that the plant characteristics did not change. Reference test SB18 included a l vacuum breaker in the ADS 1-3 line. Another 2-in.-break was repeated to determine the impact of a l higher containment backpressure on the passive injection system (Matrix Test SB19). Three additional cold-leg break tests were performed to simulate smaller (0.5 in and 1.0 in.) and larger break sizes ' (double 4.0 in.). l 1 b 4 L) l I i k l l oil 5%w Rev l \ l 5Mw.34.non: l b-081798 5.1 1 REVISION 1
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FINAL DATA Rr.ronT I 5.1.1 2 In. Cold-Leg Break (Matrix Test SB01) 1
- This section provides the test results for Matrix Test SB01 (OSU Test U0001). Matrix Test SB01 simulated a 2-in. cold-leg (CL) break loss-of-coolant accident (LOCA) with long-term cooling end without operatiori 01 nonsafety systemw The break was located at the bottom of CL-3 with a i simulated failure of one of the automatic depressurization system-4 (ADS-4) lines and no vacuum
.I breaker in the ADS line. CL-3 is on the core makeup tank (CMT) side of the facility (Figures 5.1.1-1 and 5.1.1-2).
Subsection 5.1.1.1 provides details related to the test procedure, system configuration, and initial conditions. A description of inoperable instruments is provided in Subsection 5.1.1.2, and Subsection 5.1.1.3 references the sequence of events. Subsection 5.1.1.4 describes the test results and evaluation. Component responses are given in Subsecrh 5.1.1.5, and a summary of mass balance results is provided in Subsection 5.1.1.6. The conclusions as they apply to Matrix Test SB01 are given in Subsection 5.1.1.7. The facility responses to the break are documented by the data plots, referenced as figures in the text, at the end of this section. A data plot with the suffix x indicates extended time. Matrix Test SB01 was performed on June 1,1994. The performance of this test was successful because the reactor vessel heater rod cooling was maintained.
.O 5.1.1.1 System Configuration and initial Conditions The test was conducted per an approved written procedure. All actions were automatic after the test started with no operator response required.
A flow nozzle simulating one line of flow was installed in the ADS 4-1 line-hot leg-1 (HL-1) to the ADS 4-1 separator-to provide the single failure simulation, and a flow nozzle simulating two lities of flow was installed in the ADS 4 2 line-HL-2 to the ADS 4-2 separator. Additionalij, flow nozzles simulating two lines of flow each were installed in the ADS 13 inlet lines. The reactor heater control decay algorithm maintained maximum reactor heater power output for 140 seconds, and then power began to decay to simulate the total post-trip energy input of the AP600 nuclear fuel (Appendix F). This test was performed with reactor heater rod HTR-C2-317 removed and replaced with a dummy rod. Refer to Subsection 2.7 for pre-test operations. Table 5.1.1-1 shows initial facility conditions for Matrix Test SB01. There were th.m initial condition i parameters out of specification, none of which should invalidate this test.
- HL-2 temperature, indicated by SC-140, was [ P** or less than 0.1 percent above the
, h required temperature band. This was within the accuracy requirements of the instrumentation system.
c '.15hRevnl5h-8.non:lb 081798 5.1.1-1 REVISION I
l FINAL DATA REPG
- CMT-2 temperature, indicated by TF-532, was [ ]'6' or [ ]*** above the required
]' 6' TF-532 is located on the CMT-2 long thermocouple rod, temperature of less than [
about 0.5 in. down from the inside top of the CMT. The next three thermocouples down on the long thermocouple rod, their dimensions from the inside top of the CMT, and their average temperature for about [ ]'6* prior to break valve opening are listed in the following: Thermocouple Ilimension Averane Temperature n.b2 a.bs l TF-548 TF-530 i TF-526 Less than [ ]'6* percent of the CMT volume was at a temperature greater than [ ]' 6' l ! Test analysis using the CMT recorded temperature data, taking into account the temperature vanations at the top of the CMT, should still be possible.
= Accumulator-1 (ACC-1) pressure, indicated by PT-401, was [ ]'6' or [ ]**
below the required pressure band. The accumulator was pressurized to the required pressure, as indicated on local pressure indicator PI-401, prior to test actuation. The loss of pressure between tank pressurization and test actuation was possibly due to the nitrogen gas cooling in the accumulator. Test analysis staning with the recorded lower accumulator overpressure should still be possible. l The test ran for about 10.5 hours. 5.1.1.2 Inope:able Instruments l Table 5.1.1-2 is a list of instniments considered inoperable or invalid during all or portions of this test. Some of the instruments listed are on the Critical Instrument List (Subsection 3.2, Table 3.2-2) and, therefore, are addressed here. FDP-604, FDP-605, and FDP-606 measured the differential pressure (in. H;O) across the ADS 1-3 flow nozzles. The transmitters over-ranged momentarily when their respective ADS velve opened. Total flow through the ADS 1-3 valve complex can be determined by measuring ADS 1-3 separator liquid and steam flows from FMM-601 and FVM-601. O o:\l 536wRev i\l 536w-8.non: l b-081798 5.1.1-2 REVISION 1
FINAt. DATA REPORT l FMM-201, FMM-202, FMM 203, and FMM-204 measured flow (gpm) in each of the four cold legs.
, he data from these meters were invalid after [ ]"' due to possible steam in the cold legs.
He meters could not measure steam flow. FMM-402 measured ACC-2 injection flow into direct vessel injection-2 (DVI-2). FMM-402 data are approximately double DVI-2 flow data, as indicated by FMM-2%, when accumulator injection was i providing the only flow to DVI-2. FMM-206 was confirmed to be accurate by comparison to
)
FMM-504, which measured CMT-2 injection flow, and FMM-702, which measured IRWST-2 injection l flow, when each flow was the only injection flow in progress. ACC-2 injection flow rates can be ascertained by a comparison of the other injection flows through the DVI-2 nozzle. CMT-2 level data channel LDP-402 can also be used as a backup to this measurement. l l CMT-1 and CMT-2 injection flow meters FMM 501 and FMM-504 and passive residual heat removal (PRHR) inlet and outlet flow meters FMM-802 and FMM-804 provided accurate data when sensing liquid, but became inaccurate when sensing two-phase or steam flow. FMM-701 measured in-containment irfueling water storage tank l (IRWST-1) injection flow when the l
. primary sump valves were opened. De flow meter indicated a negative flow as water flowed from the primary sump to the IRWST. The meter was not designed to measure reverse flow, so this measurement was invalid. However, total IRWST flow was measured by FMM-702.
(%V) FMM-905 measured break separator loop seal flow to the primary sump. As the transient proceeded, the primary sump and break separator levels exceeded the elevation of the break at the bottom of CL-3. When this occurred, break flow initially stopped and then reversed. Flow reversal through the break occurred at about [ ]"' rendering subsequent data invalid. Steam generator (SG) tube level data (LDP-215, LDP 218, LDP-219, and LDP-222) were biased by vaporization of the water in the transmitter reference leg after the SG tubes started draining. However, the data provide accurate indication of the time when the tubes are empty. LDP-401 and LDP-402 measured ACC-1 and ACC-2 levels, respectively. Due to air trapped in the sense lines for the transmitters, the data from these transmitters were invalid. However, the initial level of the tank was established by a standpipe, so it was constant from test to test. The drain rate can be calculated using the ACC-1 and ACC-2 flow meters (FMM-401 and FMM-402, respectively). Altemately, a pressure correction may be applied directly to the level indications of LDP-401 and LDP-402. PT-201 measured reactor coolant system (RCS) pressure at the top of the SG-1 long tube. On August 15,1994, it was discovered that the transmitter had an incorrect zero compensation, which resulted in a negative erTor and negative data at low pressures. The transmitter zero was corrected at that time. PT-201 data obtained during Matrix Test SB01 had the zero correction performed, and the corrected data appear as PT 201. Negative data and corrected negative data can be used to determine oA1536w Rev I\1536w-8.non:l tro81798 $,],j-3 REVISION 1
FINAL DATA REPORT ' trends, but are considered inaccurate. PT_201 was not considered reliable for values less than 1.1 psig, but a sufficient amount of other pressure data are available. TF-501 and TF-504 measured CMT fluid temperature from the long thermocouple rod location near the bottom of each CMT. The thermocouples appear to have measured ambient conditions throughout the test, which would indicate a short somewhere in the thermocouple wiring. With these thermocouples inoperable, the required long thermocouple rod thermocouple availability of "seven out of ten and no more than one in succession failed" was met. Data provided by ADS-4 separator instrumentation prior to the ADS 4-1 and ADS 4-2 valves opening at 978 seconds were invalid dur; to the closed position of the ADS-4 valves and the ADS-4 separator loop seal valves. The instrumeuts affected are: FMM-602, FMM-603, FVM-602, FVM-603, LDP-611, and LDP-612. Test analysis will not be affected, since ADS-4 flow did not begin until the valves opened. Considering these critical instrument failures, sufficient instrumentation was available to allow the performance of mass balances as demoastrated in Subsection 5.1.1.6 and Appendix E. An energy balance will be performed and reported in the AP600 Low-Pressure Integral Systems Test at Oregon State University Test Analysis Report, WCAP-14292.* 5.1.1.3 Sequence of Events Table 5.1.1-3 contains the sequence of events for Matrix Test SB01. The first pages of the table provide the event times of selected events in the test. The subsequent pages of the table use bar charts of the data to provide a visual representation of the sequence of events. Both the numeric table and the bar chart sort the events in chronological order. The first pages of Table 5.1.1-3 indicate the source of the actual time data. A D in the Data Source column indicates the recorded time was obtained from a software program that monitored digital events in the facility. These events included purnp stans and stops, valve limit switch actuations, and alarms. The term valve opening means the valve has actuated and the closed limit switch is being opened (valve coming off the seat). An A in the Data Source column indicates the time data were obtained by reviewing test data obtained from the data acquisition system (DAS). Although the test data from the DAS were in digital format, the DAS monitored analog events such as pressure, flow, and temperature from the data. Because Matrix Test SB01 is the first test described in : Sis repcit, the following is a general discussion of the events and responses from test initiation through the end of the test. The test sequence was initiated by actuating the TEST pushbutton, which triggered the DAS to begin acquiring data. Approximately the first 2 minutes of data were used to establish the initial conditions for the test. Two minutes after TEST pushbutton actuation, a signal was sent to open the break valve located at the bottoai of CL-3. In the first [ ]* following the ::ignal to the break valve, the SG o A1536w Rev l\l 536w-8.non: I b-081798 5.1.1 -4 REVISION I
FnNAL DATA REronT d ( pressure setpoint was raised, the reactor shifted to power (kW) centrol mode with a programmed power demand, the main feedwater pump tripped and feedwater was isolated, the passive residual heat removal heat exchanger (PRHR HX) outlet valve and CMT discharge valves opened, and the reactor coolant pumps (RCPs) tripped. Forced flow was initiated through the DRHR HX and the CMTs until the RCPs stopped, at which time the flow changed to natural circulation flow. As the RCS depressurized and lost inventory through the break, pressurizer level decreased rapidly and steam formation began in the reactor vessel upper head. At about [ ]'6' the CMTs began to drain. De pressurizer and pressurizer surge line became completely empty of liquid, and at de , same time, steam formed in the SG tubes and pushed the remaining water out of the bottom of the tubes. Shortly thereafter, the SG channel heads began to empty. When CMT level decreased to the l low level setpoint, the ADS valve sequence initiated. l The ADS-1 valve opened, causing an increase in the rate of RCS depressurization. The ADS-2 and ADS-3 valves then sequenced open, further increasing the rate of depressurization. When RCS pressure decreased to below that of the accumulators, the accumulaters began to inject into the DVI l lines, which stopped CMT jnjection by closing the CMT discharge line check valves. l The reactor pressure low-low setpoint was reached, sending an automatic opt ting signal to the IRWST injection valves. IRWST injection did not start at this time because RCS pressure was still too high to f be oveicome by the IRWST static head. The accumulators emptied and depressurized, and CMT injection recommenced. When the CMT low-low level was reached, an automatic opening signal was sent to the ADS 4-1 and ADS 4-2 valves. These additional vent paths for the RCS helped Scrt,e RCS pressure low enough to initiate IRWST injection. I The IRWST injection continued, and about [ ]'6' later, the pressurizer surge line and then the pressurizer began to reflood The pressurizer subcooled and, with the ADS 1-3 sparger still I submer6ed in the IRWST, a partial vacuum was created in the pressurizer and ADS 1-3 separator, further increasing the fluid level in the pressurizer. About this same time, the CMT levels started to
-1 increase. After about [ ]'6' the ADS 1-3 sparger became uncovered, breaking the partial I vacuum and allowing the pressurizer to drain. A fluid level was maintained in the surge line for the remainder of the test.
Primary sump injection started several hours later with flow through the injection line check valves. The primary sump injection valves opened automatically when the IRWST reached its low-low level setpoint. The RCS remained on natural circulation flow with inventory leaving through ADS 4-1 and ADS 4 2 to the primary sump and returning from the sump through the DVI lines. The driving force was the heat input to the reactor. v OA1536wRev l\l 536w-8.non:l b-Od l 798 5.1.1 -5 REVISION 1
FINAI, DATA REPCOT 5.1.1.4 Test Results and Evaluation The LOCA event simulated in the OSU test facility resulted in interactions between the systems and components in the facility. For this mason, it is convenient to subdivide the event into different phases, each characterized by the systems' behavi6r and thermal-hydraulic phenomena occurring within the systems. The different event phases selected for the purpose of detailed evaluation of the I.OCA event are as follows:
- Initial Depressurization Phase: simulated break initiation to ADS-1 actuation e ADS Phase: ADS-1 actuation to start of IRWST injection
- IRWST Injection Phase: start of IRWST injection to end of test Initial Depressurization Phase The test began with the cetuation of the TEST pushbutton. After a 120-second delay, break valve TS-205, located at the bottom of CL-3, rueived an open signal (time zero). After an additional 0.5 second, an S signal was generated by the programmable logic controller (PLC), which time-sequences signals to initiate various events such as resetting controllers, stopping pumps, and repositioning valves.
The SG steam pressure controller reset to control pressure at [ ]'** to minimize heat remcval by the SGs. The value of [ ]'** was high enough that the valve would not re-open, yet low enough not to challenge the SG safety valves Simultaneously, the reactor controller transferred from the temperatun- (hot-leg average temperature in "F) control mode tc the power (kW) control mode, with demand programmed for 600 kW total power. The main feedwater pump tripped, and the SG feedwater control valves closed at 4 seconds, isolating feedwater to the SG. At[ ]'b' with the RCPs still running, the PRHR HX outlet valve opened, allowing forced flow through the PRHR HX to begin and, therefore, cooled fluid to enter the system at the SG-2 outlet channel head from the HX. At [ ]'6' the RCPs tripped and the flow through the PRHR HX shifted from forced to natural circulation flow. Also at [ ]'6' the two CMT injection valves opened, allowing forced flow from CL-1 and CL-3 through the CMTs and into the tractor vessel downcomer area through the two DVI nozzles. This also was cool ambient temperature fluid replacing inventory lost through the break. When the RCPs tripped, flow through the CMTs shifted from forced to natural circulation flow. At about [ ]'6' steam percent, as calculated from LDP-127 data, indicated that the reactor vessel began to lose inventory and flash to steam (Figure 5.1.1-3). Steam percent, as calculated from LDP-il5 data, indicated that the fluid in the reactor vessel upper head began to flash to steam at [ ]'6' and was essentially all steam at about [ ]'b' (Figure 5.1.1-4). The upper o:U536wRevnl536w-8mn:Ib-081798 5.1.1 -6 REVISION 1
FINAL DATA Rzroat I pieclim area of the reactor vessel steam percent, as calculated from LDP-139 data, began to show steam collection at approximately [ ]'6" Using data from the level channels and the calibrated range of the instruments, a core steam percent for each channel was calculated. The equation used to calculate steam percent was: Steam percent = 1 c mPensated level ' 100
, instrument range ,
he break flow caused a rapid decrease in pressurizer leve' (Figure 5.1.I-5) and emptied the pressurizer at approximately [ ]'6" At about [ h f the pressurizer surge line completely emptied, as indicated by LDP-602. CMT levels began to de. crease, making the transition from recirculation to draindown, as indicated by LDP-509 and LDP-510, at about [
]'6" respectively (Figure 5.1.1-6). After transition to draindown, the CMTs provided makeup to the RCS to compensate for the loss of inventory through the break.
After 140 seconds at 600 kW, reactor power began to follow the decay heat decay curve. (The OSU test facility power is held at a maximum of 600 kW for 140 seconds to simulate the total heat input of the AP600 nuclear reactor following a reactor trip.) At about [ J'6" a condensation /depressurization event took place in CMT-1. Data indicate s a rapid refill of the CMT-1 balance line with an increase of about [ ]'6' in tank level (tank full) once the balance line filled (Figure 5.1.1-77). An indication of the event was an [ ]'6" spike decrease in CMT-1 pressurt., whic'; resulted in about a [ ]'6' decrease in reactor upper head and CL-3 pressures (Figure 5.1.1-78). The CMT-1 inlet line and upper tank temperatures provided additional confirmation that a condensation /depressurization event occu Ted (Figures 5.1.1-79 and 5.1.1-80). Condensation /depressurization events are described more thoroughly in Subsection 7.1. The U-tubes of both SGs were completely empty by approximately [ ]'6' (Figures 5.1.1 7 and 5.1.18). De SG U-tube level instruments should be considered inoperab'2e once the U-tubes start to empty because their reference legs were routed internal through the SGs ar.d, therefore, steamed dry during the test. Loss of the reference legs gave a false full-level inuication. *.the SG cold-leg channel heads were all empty by about [ ]'6' which was prior to ADS-1 actuation, and the hot-leg channel heads emptied at about [ ]'6" or shortly after ADS-1 actuation (Figures 5.1.1-9 and 5.1.1-10). Although it was ranged improperly and considered inoperable, LDP-209 would still have responded to trends. The HL-1 level began to decrease at about [
]'6' and HL-2 level at about [ ]'6' as indicated by hot-leg elbow level instruments LDP-207 and LDP-208, respectively (Figure 5.1.1-11).
O Although they were ranged improperly and considered inoperable, LDP 207 and LDP-208 would still have responded to trends. The horizontal sections of the hot legs started to drain at about [ oA15hRevlM5h-8.non:ll> 081798 $,],].7 REVISION 1
FINAL DATA REPORT l ]'6' as indicated by LDP-205 and LDP-2%, respectively. The draining of the hot legs g was confirmed by LDP-139 (Figure 5.1.1-12), which indicated a level of about [ ]'6" at this time. W t-l LDP-139 measures level inside the core barrel between the bottom of the upper core support plate and the bottom of the upper support plate ([ ]'6* above the bottom of the reactor vessel), and l spans the hot legs. The hot leg temperatures during this time remained at saturation temperature (Figures 5.1.1-13 and 5.1.1-14). Even though they were partially or completely empty, the hot legs remained at l saturation temperature and never superheated due to a small flow of saturated steam from the reactor heater bundle to the SGs. The fluid level inside the reactor core barrel reached its minimum collapsed level of [ ]'** (the Indicated level at the top of the heated section of the heater rods is [ ]'6') during this test at [ ]'6' as it.dicated by LDP-127 (Figure 5.1.1-15). At t iis level, heater rod cooling was still maintained. At [ ]'6* CMT-1 reached its low level setpoint, thereby activating the ADS valve timing sequence portion of the PLC, and the ADS-1 valve opened at [ ]' ** l ADS Phase l l The opening of the ADS-1 valve resulted in two-phase flow through the pressurizer surge line and O pressurizer to the ADS 1-3 separator and then to the IRWST through a sparger. This flow caused the differential pressure instruments used to measure surge line and pressurizer level to initially produce an indicat. ion of the pressurizer and surge line reflooding, which may have been inf'uenced by line losses (Figure 5.1.1-5). With the accumulators at their maximum injection rate, the surge line and pressurizer began to reflood at about [ ]'** and the pressurizer attained its maximum level at [ ]'6* The pressurizer and surge line then drained back down and were completely empty at [ ]'6' This additional flow path, in conjunction with the break, caused RCS pressure to decrease at a more rapid rate. The opening of the ADS-1 valve, followed by the ADS-2 valve approximately [ ]'6* la caused an increase in the rate of RCS depressurization (Figure 5.1.1-45). When the RCS depressurized to approximately [ ]'6' at about [ ]'6' accumulator injection began. This reduced CMT-1 injection flow to 0 over the following [ ]*** and CMT-2 injection flow to less than [ ]'6' over the following [ ]'6' (Figure 5.1.1-16) by closing off the CMT discharge line check valves until the accumulators were almost empty and depressurizec. ACC-2 injection flow instrument FMM-402 appeared to have indications double the actual flov when compared to DVI-2 flow instrument FMM-206 and was considered inoperable. The AI1S-3 valve opened at approximately [ ]'6' but with RCS pressure having decreased to rbout [ ]'6" by this time, the valve opening had little effect on the rate of depressurizati >n. t o:\l536wRevl\l536w-8.non lb.081798 5.1.1-8 REVISION 1
FtNAL DATA Rt. roar When RCS pressure decreased to [ ]'6* at approximately [ ]'*' the two IRWST C l injection valves automatically opened (by a signal from the PLC), but IRWST injection could not l occur until RCS pressure decreased to near atmospheric, since the IRWST is a static system that depends on gravity flow and operates at break and ADS measurement system (BAMS) header pressure. CMT-2 injection flow started to increase at [ ' ]'6' and CMT-1 injection flow l started to increase at [ ]'6" The accumulators completed injection at [ ]'** for l ACC-1 and [ ] for ACC-2, at which time CMT injection resumed. At the end of accumulator injection, some of the nitrogen gas was injected into the DVI lines, momentarily cooling the injection lines. The nitrogen caused a momentary decrease in DVI l flow of about [ ]'6"and a decrease in ACC-1 outlet temperature of about [ ]'6' at [ ]'6' (Figures 5.1.1-17 and 5.1.1-18). On the ACC-2 side, there was no indicated change in DVI-2 flow, although the ACC-2 outlet temperature decreased [ ]'** at [ ]' 6
- At approximately [ ]'b' a sharp " bang" was heard, apparently emanating from the upper portion of the reactor vessel. This has been determined to be a steam condensation and resulting depressurization event in the reactor vessel upper downcomer area and has been further investigated (see Subsection 7.1). The sharp increase of DP-114 to [ ]'6
- at [ ]'6' DP-130 to
[ ]' 6 ' at [ ]*6* and the over-ranging of LDP-116 and LDP-140 at l [ ]'b' were major indications of the steam condensation event (Figures 5.1.1-19 and 5.1.1-15). Test data reveal that the collapse of the superheated steam bubble in the upper portion O of the reactor vessel downcomer annulus resulted in the downcomer fluid accelerating upward and impacting the bottom of the core barrel flange where the core bypass holes are located. The impact of i the downcomer liquid on the solid surface of the core barrel flange produced the " bang" heard during the test, he low pressure created in the upper downcomer annulus by the collapse of the steam bubble also resulted in a rapid increase in steam flow from the core barrel, through the upper head, and into the downcomer. DP-ll4 and DP-130 were both calibrated to indicate either positive or negative differential pressure, depending on the direction of flow. DP-130 was calibrated as if the system were full; therefore, when the fluid level is below the lower tap, the instrument is in error by the distance between its high and low-pressure taps,.i.e., [ ]'6' must be added to the indicated value to obtain the correct differential pressure. From [ ]'b' both DVI nozzle temperatures increased from essentially ambient conditions to as high as [ ]'b' and then returned to ambient conditions (Figure 5.1.1-20). The temperature increase was caused by two factors. First, there was rapid heating of the remaining water to be injected from the CMTs when they were at low levels. Second, the reactor vessel downcomer level was at the DVI nozzle level during this period, possibly partially uncovering the nozzles. He temperature transient was terminated when IRWST injection began refilling the reactor vessel at about [ ]'6# and tempemtures returned to arabient when the CMTs were empty, terminating the hot liquid injection. His temperature transient does not appear to have affected any other facility parameters. O o A1536wRev nl 536w-8.nosc i b-081798 5.1.1 -9 REVISION 1
FINAL DATA REPCOT At[ ]'** the ADS 4-1 and ADS 4-2 valves opened automatically on a signal from the PLC when CMT-1 level reached its low-low level setpoint. ADS-4 actuation started a decline in RCS inventory that would not be overcome until IRWST injection began (Figure 5.1.1-15). CMT-1 and CMT-2 were completely empty at [ ]'6' respectively (Figure 5.1.1-6). Pressurizer heater thermocouples TH-601, TH-602, and TH-603 (Figure 5.1.1-21) indicated that the pressurizer was slightly subcooled at about [ ]*** remained subcooled until primary sump injection began at about [ ]*** and then increased to saturation temperature. System heatup can be attributed to the wanner recycled injection water provided from the sump and lower liquid levels maintained in the RCS (Figures 5.1.1-22 and 5.1.1-23). The pressurizer remained at saturation temperature until about [ ]'** when the temperature began to rise into the superheated range (Figure 5.1.1-24). The PLC did not lower the pressurizer heater demand to [
]'6' as required by the logic, and the heaters stayed at 1.5 kW until about
[ ]'*' when they increased to about 3 kW and remained there until the end of the test. In subsequent tests, the heaters were de-energized by procedure, and later in the test program, the logic was changed to drop out the heater silicon-controlled rectifier (SCR) contactor. The IRWST appeared to overflow into the primary sump during the period of [ ]'** as indicated by FMM-703 (Figure 5.1.1-25). However, there was no indicated level change on LDP-701 (Figure 5.1.1-26) prior to or during this period. This early indication of overflow from the IRWST to the sump is believed to have been caused by the differential pressure between the IRWST and primary sump, which forced liquid from the overflow loop seal into the sump. The pressure in the two tanks was caused by a backpressure in the BAMS header as steam was released from the break separator following the break. A possible explanation is that the differential pressure was a m2 of the tank free volumes and vent pipe size (i.e., the sumps were empty and the IRWST was e filled to [ ]'** the IRWST has a [ j'** vent and the sump has a [ ]'*' va resultig in a faster pressure increase in the IRWST). From [ ]'6' an overflow and level increase occurred due to the loss-of-coolant inventory through ADS 1-3, as indicated by FMM-703 and LDP-701. At about [ ]'6* RCS pressure had decreased to about [ ]'** which was sufficiently low that the IRWST static head was greater than RCS pressure, and IRWST injection began. IRWST injection proceeded at a continually diminishing rate as the differential head between the IRWST and the RCS decreased. IRWST Inlection Phase The pressurizer and pressurizer surge line emptied for the second time at approximately [
]'6' respectively (Figure 5.1.1-5). The surge line then began to reflood almost immediately at [ ]'** and the pressurizer at about [ ]' 6
- This th reflooding was confirmed by HL-2, pressurizer surge line, and pressurizer subcooled temperatures (Figures 5.1.1-14 and 5.1.1-27). In addition, the surge line completely filled first, followed by the o:u536wRevN536w.8.non:lb-081798 5.1.1-10 REVISION 1
. . _ _ . _ __ _ ._ _ _ _ _ _ . _ _ _ _ _ ~ . . _ . . _ . _ _ _ _ _ . . _ . _ _ FINAL. DATA Rzront p level rise in the pressurizer. The reflood was caused by RCS levels increasing above the reactor vessel Q nozzles because IRWST injection exceeded the inventory losses (Figures 5.1.1-15, 5.1.1-17, and 5.1.1-28). He maximum pressurizer level attained was about [ ]'6" at [ ]'6' but immediately began to decrease and was empty at [ ]'6' The pressurizer remained empty for the remainder of the test. De surge line stayed full until [ ]'6' when the level decreased to about [ ] and remained there until [ ]'6' he level again decreased to about [ ] at [ ]'6" and oscillated between about [ ]'6' for the remainder of the test. During the time that the pressurizer had a liquid level, ADS 13 separator pressure (PT-605) and I ADS 1-3 sparger pressure (PT-606; Figure 5.1.1-29) both went below atmospheric pressure by as I much as [ ]'6" They were below atmospheric pressure from [ ]'b' he I negative pressure differential was broken when the level in the IRWST decreased below the sparger i nozzles. This negative pressure differential would explain much of the pressurizer level in that it would draw the level higher and hold it there. A vacuum breaker was installed on the sparger line inside the IRWST shortly after this test to prevent a recurrence of this problem. A vacuum breaker is included in the AP600 design. 1 1 Both CMT balance lines began to refill at about [ ]'6* when reactor vessel level, as indicated by LDP-140, was sufficiently high at [ ]'6" to cover and refill the cold legs g (Figures 5.1.130 and 5.1.1-31). At about [ ]'6' when the CMT-2 balance line had completely refilled, CMT-2 began to rapidly refill and reached the [ ]*6' level (about two-thirds full) in about [ ]'b' The refill of subcooled fluid caused a condensation /depressurization event to occur in CMT-2. This was reflected in reactor vessel pressure and levels and the CMT balance line level taking a sharp dip and then recovering. CMT-2 refilling also caused its internal temperatures to quickly decrease from the superheated region to subcooled at less than [ ]'6'and to remain subcooled until the CMT began to drain again (Figure 5.1.1-32). When the CMT-1 balance line had refilled completely at approximately [ ]'6' CMT-1 refilled to the [ ]'6' level (or about two-thirds full) in less than [ ] Again, the rapid refill with subcooled fluid caused another condensation /depressurization event to occur, this time in CMT-1. The CMT-1 internal temperatures responded to the subcooled fluid almost identically to those of CMT-2 (Figure 5.1.1-33). The CMTs reached a level and temperature equilibrium in which steam condensation and a resultant level increase were no longer occurring. During this period of time, there was no injection flow from the CMTs because the higher static head of the IRWST held the CMT discharge line check valves closed. When the CMT balance lines began to refill, CMT internal pressures began to decrease with respect to the reactor vessel upper head pressure. He cause of the pressure decrease is believed to be that once the cold legs filled with subcooled fluid, displacing what had been a blanket of superheated steam, the superheated steam in the CMTs began to slowly condense and the balance lines filled over a period of about [ ]'6# Once the balance lines filled and sprayed into the CMTs through the inlet diffusers, the subcooled fluid collapsed the superheated steam bubble in the CMTs, creating more of a oA15hRevi\l5h-8.non:Ib-081798 $,1,].)) REVISION 1
FINAt DATA REPORT decrease in pressure. The subsequent in-rush of fluid from the RCS was reflected in the decrease in reactor vessel levels and pressure. CMT-1 and CMT-2 remained at essentially constant levels for several [ ]'6' and then began slow draindowns at about [ ]'6" respectively. The draindown for both CMTs was slow and did not occur until IRWST relative level was [ ]'6* below that of the CMTs (Figure 5.1.1-34). Data indicate that the CMTs drained for a while, and then the differential head between the IRWST and the CMTs again closed the CMT discharge check valves, terminating draining until the time that the differential shifted the other way and draining recommenced. Both CMTs were completely empty at about [ ]'** which coincides closely with the primary sump injection valve opening at [ ]' 6 ' A possible correlation is that when the primary sump valve opened, the IRWST had just reached its minimum level of about [ ]'6" which is I about [ ]*** below the instrumented level for the CMTs, and that there was still a slight partial I vacuum remaining in the CMTs. Also, when the primary sump injection valves opened, there was a short time period in which the IRWST and primary sump levels equalized, causing a decrease in RCS fluid levels and resulting in a rather fast drop in CMT levels from about [ ]'** LDP-905 revealed that break separator level (Figure 5.1.1-35) began to increase at the same rate as the primary sump at about [ ]* 6# This occurred when sump level reached the height of the break separator loop seal. As a result of this level increase, break separator level reached the height of the break in CL-3, causing break flow to reverse and flow from the break separator into the RCS through the break at about [ ]*** (Figure 5.1.1-36). The break flow then remained essentially 0 or slightly negative throughout the rest of the test. Primary sump injection began through the check valves around the sump injection valves at approximately [ ]*** (Figure 5.1.1-37) when primary sump and IRWST levels were essentially equal (Figure 5.1.1-35). At [ ]'** the primary sump injection valves automatically opened when the IRWST reached its low-low level setpoint of [ ]'** When sump injection began, the reactor downcomer temperatures rapidly increased to match the sump fluid temperature (Figures 5.1.1-23 and 5.1.1-38). When the primary sump injection valves opened, the DVI flows decreased and the sump and IRWST levels equalized (Figure 5.1.1-39). During the l reduced DVI flow period, there was an upward spike in reactor downcomer temperatures. The downcomer thermocouples located above the DVI nozzle elevation increased to saturation at this time and remained at saturation for the rest of the test because the reactor vessel collapsed level was at the DVI nozzles (Figure 5.1.1-22). PRHR HX outlet temperature, indicated by TF-804 (Figure 5.1.1-40), remained subcooled in the range ! of[ ]'** during most of the test, but after sump injection, began to rise and was just reaching saturation temperature at the end of the test. The inlet temperature, indicated by TF-803, was subcooled between [ ]'** After reaching saturation at about 12,000 seconds, it remained at saturation temperature for the rest of the test. l o:\l536* Revl u S 36w-8.non: l tso81798 5.1.1-12 REVISION 1 1
FINAt, DATA REPORT In the early stages of the transient, the PRHR HX instrumented long- and shon-tube temperatures
's generally remained below [ ]'b' with the shon tube predominantly hotter,and maintaining a wider delta temperature between the tube inlet and outlet. This would indicate that most of the flow was through the short tubes. This difference between long- and short-tube performance was evident until about [ ]'b' when the shon-tube temperatures rapidly dropped to align themselves with the long-tube temperatures (Figure 5.1.1-41). He drop in temperature did not coincide with any other event and was an indication of a flow cessation through the PRHR HX, with the indicated delta temperature across the HX being the result of the temperature gradient in IRWST fluid. Flow cessation is observed from CL-2 and CL-4 data, which show that temperatures at the bottom of the reactor flange for both loops began to increase at about this same time, followed saturation from about
[ ]'b# went into the superheated region at about [ ]'b' and then became subcooled when the cold legs staned to refill at about [ ]'6' (Figures 5.1.1-42 and 5.1.1-43). The PRHR HX inlet temperature became subcooled coincident with the ADS-4 valves opening at [ ]'6' and over the next [ ]'6# dropped to and paralleled the outlet temperature. Again, this is an indication that there was no flow through the HX during this time frame i l (Figure 5.1.1-40). At [ ] the PRHR HX inlet temperature instantly jumped from [ ]'** to saturation temperature. This happened about [ ]'6' after pressure, level, and l flow oscillations began in the facility and was possibly caused by the inlet line " burping" and once again allowing the line to fill with saturated steam. Following the " burp," all of the PRHR l V temperatures began to slowly approach saturation. When primary sump injection started through the check valves around the sump injection valves, IRWST injection was approximately [ ]'b' per side (Figure 5.1.1-37). The flow then split, with approximately [ ]*b' per side from the IRWST and [ l'** per side from the primary sump. When the primary sump injection valves opened, the IRWST-2 side flow increased to approximately [ ]'b' the IRWST-1 side flow indication went negative, the primary sump-l side flow increased to approximately [ ]'b' and the primary sump-2 side decreased to approximately [ ]'** This flow split is discussed more thoroughly in the IRWST Response ponio i of this I section. Overflow from the primary sump to the secondary sump staned at about [ ]'6' (Figure 5.1.1-35). Starting at about [ ] there was a series of pressure, level, and flow oscillations that f occurred throughout the components of the facility lasting until about [ ]*** Since they l occur in other tests, these oscillations are addressed in the Test Analysis Repon.* In the long-term cooling mode of operation, system inventory was lost through the ADS 4-1 and ADS 4-2 valves to the primary sump. System inventory was made up through primary sump and 4 IRWST injection through the DVI lines and some small flow from the primary sump through the
< O break separator and into the break. He driving force for this flow was the decay heat simulation in the reactor heater rods. The hotter fluid produced in the reactor flowed out through the hot legs and 1
o:\l5hRevl\1536w-8.non:Ib 081798 5 ],1-13 REVISION 1
FINAt. DATA RF. PORT l I ADS-4 to the primary sump, and the cooler fluid in the sump returned from the bottom of the sump to the reactor vessel downcomer via the sump injection lines. The test continued for an additional 4 hours in the sump recirculation mode. The final total core power at the end of the test was [ ]'6' (Appendix F). 5.1.1.5 Component Responses Reactor The reactor and its associated connections to the other systems in the facility are depicted in Dwg. OSU 600203, and its heater locations are depicted in Dwg. OSU 600007. The reactor instrumentation is shown in Dwg. OSU 600101, Sh. I and 2. These drawings are in Appendix G. When the TEST pushbutton was pressed, the reactor controller was in auto-local, controlling hot-leg average temperature at 420 F (the reactor controller automatically controls that temperature by varying the demand signal to the heaters). At time zero, the PLC sent a signal to open the break valve and then 0.5 second later signaled the reactor controller to shift control to auto-remote with total power demand initially at 600 kW (the setpoint is generated by an algorithm programmed into the controller, and the controller automatically controls the demand to the heaters to control the setpoint kW). The power algorithm programs full power (600 kW) for the first 140 seconds and then iets power decay at an exponential rate that simulates the decay heat input of the AP600 nuclear reactor following a trip from full power (Appendix F). The DVI flow into the reactor vessel began immediately when the CMT discharge isolation valves opened at 6 seconds (Figure 5.1.1-17). This flow continued at a rate averaging about [ ]'b' p r side until about [ - ] when RCS pressure decreased sufficiently for accumulator injection to start and DVI flow increased to [ ]'b' per side. When accumulator injection ended, the l DVI flow again dropped to about [ ]'6' per side while the remainder of the CMT fluid was l injected. Steam percent, as calculated from LDP-127 data (Figure 5.1.1-3), showed that the reactor vessel began to lose inventory, i.e., water began to flash to steam, at approximately [ ]'6' Steam percent, l as calculated from LDP-115 data (Figure 5.1.1-4), showed that the fluid in the reactor vessel upper head began to drain and flash to steam at about [ ] and was essentially all steam at about ! [ ]*6' Steam percent in the upper plenum area of the reactor vessel, as calculated from LDP-139, began to show steam collection at approximately [ ]'b' The initial increase in reactor vessel downcomer level indication on LDP-il6 and LDP-140 was because the RCPs tripped at [ ]'b' and the instruments were no longer affected by the dynamic flow differential pressure and indicated a true full-range level (Figure 5.1.1-15). LDP-116 and LDP-140 showed that the level in the reactor vessel downcomer began to decrease at o:\l 536wRev i\l 536w-8.non: I b-081798 5.1.1-14 REVISION 1
FmAL DATA RzromT l l [- J'*' The reference leg taps for these level differential pressure transmitters are located at (, [ ]'** respectively, and approximately [ ]*** below the core barn 1 support flange. l This means that the level in the downcomer started decreasing some time prior to indication. The two level differential pressure transmitters share a common lower tap located at [ ]'b' i Fluid levels in both the reactor vessel downcomer and inside the core barrel continued to decrease due l to inventory loss through the break, with LDP-il6 and LDP-140 showing a level of [ ]*6*at [ ]'6' and LDP-127, which measured collapsed level inside the core barrel, showing a minimum level of [ ]'** at [ ]'** The level in the downcomer was near the top of the cold legs and remained there until ADS-1 actuation. The level in the core barrel completely uncovered the hot-leg nozzles, but maintained cooling in the heated section of the heater rods. A verification that the heater rods remained cooled is that none of the upper heater rod thermocouples showed a significant increase in temperature as would be expected if they had not been sufficiently cooled (Figure 5.1.1-44). ! l After reaching the minimum, the reactor core barrel level increased slowly as injection flow became slightly greater than the break flow due to the decreasing RCS pressure (Figure 5.1.1-15). De level in the core barrel increased to [ ]' 6
- at [ ]'*' which was still about [ ]'** below the ;
hot-leg nozzles. The ADS-1 valve opened at [ ]'** causing an increased rate of RCS depressurization (Figure 5.1.1-45) and an additional flow path out of the RCS. His, in tum, caused l the following to occur: the level in the reactor vessel downcomer dropped rapidly to [ l'**at [ ]*** and then began another slow recovery, and the level in the core barrel rose rapidly to [ . ]'6
- at [ ]'b' and then began another slow increase. These rapid level changes are attributed to the decreased pressure in the hot-leg area after the ADS 1 valve opened, causing an increased flow out of the reactor vessel to the pressurizer and also increased steam percent and, l therefore, swell in the core barrel.
From [ ]'** there was reverse flow from the upper head, through the core bypass ! holes, and into the upper downcomer region (Figure 5.1.1-19). His flow then rapidly returned to 0, l which again was an effect of the ADS-1 valve opening. The upper head area then essentially became stagnated until about [ ]'** During this stagnant period, upper head and upper downcomer thermocouples indicated that superheating occurred in those areas (Figures 5.1.1-46 and 5.1.1-47). The superheating could have been the result of both the reduction in pressure and radiant heating from the hotter vessel wall metal. At[ ]'6' test personnel heard a loud " bang" in the upper portion of the reactor vessel. The test personnel were located on the grating at the top of the reactor vessel and, therefore, were able to pinpoint the location of the noise. Later investigation determined that the " bang" was the result of a condensation /depressurization event in the upper downcomer area, causing the downcomer water level to rise very rapidly and hit the core barrel flange, resulting in the noise. The sharp increase of DP-114 l' to [ ]'6' at [ ]'** DP-130 to [ ]'** at [ ]'** and the
\ over-ranging of LDP-116 and LDP-140 at [ ]'** are major indications of the steam oA15hRevlu5b-8.non:Ib-081798 5,] ,] .15 REvtsloN l
FINAL DATA REPORT condensation event (Figures 5.1.1-19 and 5.1.1-15). This and other similar events have been investigated (see Subsection 7.1). The conclusion is that the forces resulting from these condensation /depressurization events were not sufficient to cause any facility damage. At[ ]'6* after accumulator injection had been completed and only CMT injection was available to replace lost RCS inventory, the core barrel level again began to slowly decrease. At [ ]'6' ADS-4 actuated, providing two additional depressurization paths and decreasing RCS pressure to near atmospheric. The decreasing reactor core barrel level continued until about [ ]'6' when the CMTs were essentially empty and IRWST flow began to increase RCS makeup to a rate slightly greater than the losses, thereby causing a slow increase in reactor vessel levels. Minimum indicated levels reached during this period were approximately [ ]'6' in the core barrel and [ ]'6' in the downcomer (Figures 5.1.1 12, 5.1.1-15, and 5.1.1-17). As RCS pressure decreased, IRWST injection rate increased to about [ ]'6" per side (Figure 5.1.1-48), which in turn raised reactor vessel levels to the highest they would attain during this test. The downcomer level instruments over-ranged at about [ ]'6* At this same time, core barrel level eq iled downcomer level, and the two levels remained together at [ ]'6' for about the next ]'6' and then increased slightly and returned to about [ ]'6'at about[ ,'6* (Figures 5.1.1-15 and 5.1.1-22). While t.he reactor vessel levels increased, the upper head differential pressures again indicated steam flowing from the core barrel, through the head, and into the downcomer region. Flow stopped at about the same time that the downcomer level instruments, LDP-116 and LDP-140, over-ranged. The flow staned again when reactor vessel levels dropped back down to about [ ]*6#at [ ]'6* and contirued throughout the remainder of the test (Figures 5.1.1-22 and 5.1.1-49). Based on reactor vessel level effects on the flow through the core bypass holes, it can be observed that whenever the cold legs are partially to fully uncovered, there is flow; whenever the downcomer level covers the cold legs, the flow is stopped and the steam in the upper head becomes stagnant. Reactor performance for the remainder of the test remained essentially unchanged until the primary su;np injection valves opened at [ ]'6* when the reactor vessel levels dropped to about [ ]'6' and remained there until test termination. Core Makeup Tanks 1 Thermal-hydraulic responses of the two CMTs were very similar throughout the performance of l Matrix Test SB01. Although their responses with respect to time were different, data representing l CMT-l's response are presented in this subsection, with CMT-2 times in parentheses. Where the difference in actual response is noteworthy, CMT-2 data are also addressed. The CMTs were filled and vented with ambient temperature water prior to the initiation of facility stanup and heatup for the performance of the test. The CMT balance line isolation valves were in the o:\l 5 %w Rev i \l 5 %w-8.non: l t>-081798 5.1.1-16 REVISION 1
=. - . .. -. -_ --- _-- - .-.~.---.- - ---_--~
FINA1. DATA REPORT j AUTO and OPEN positions, and the CMT discharge isolation valves were in the AUTO and CLOSED positions. His alignment prevented flow through the CMTs, thus allowing them to remain at ambient temperature while being pressurized with the RCS during the facility startup. The CMT piping is shown in Dwg. OSU 600206, and CMT instrumentation is shown in Dwg. OSU 600501 and Dwg. l OSU 600502. These drawings can be found in Appendix G. l RCP-forced flow through the CMTs to the DVI nozzles began immediately when the CMT discharge valves opened at [ ]'6' (Figure 5.1.1-16). Then, [ ]'6* later, they continued to inject due to natural circulation flow from the cold legs through the CMTs and into the DVI nozzles when the RCPs were tripped by the PLC signal. The natural circulation continued until the CMTs began to drain at [ l'6' ([ ]'6') as determined by a decreasing balance line level. At [ ]'6' ([ ]'6'), CMT wide-range level began to decrease (Figure 5.1.1-6). From that point on, all CMT injection was due to gravity and the manometer effect of colder, more dense water at a higher relative level in the CMTs, since the RCS and the CMTs were l affected by the same overpressere from the steam in the facility. At about [ ]'6' a condensation /depressurization event took place in CMT-1. De data indicate a rapid refill of the CMT-1 balance line with an increase of about [ ]'6# in tank level once the balance line filled (Figure 5.1.1-77). Other indications of the event were an [ ]'6' spike decrease in CMT-1 pressure, which resulted in about a [ ]'6' decrease in reactor upper head and g CL-3 pressures (Figure 5.1.1-78). The CMT-1 inlet line and upper tank temperatures provided O4 edditional confirmation that a condensation /depressurization event occurred (Figures 5.1.1-79 and 5.1.1-80). This refill of CMT-1 probably caused sona delay in the ADS 1-3 valve opening sequence. Condensation /depressurization events are described more thoroughly in Subsection 7.1. ' At[ ]'6#([ ]'6'), RCS pressure was sufficiently low, about [ ]' 6# that accumulator injection began. The ADS-1 valve opened at [ ]*6' and caused a rapid decrease in RCS pressure. RCS depressurization, coupled with the pressurized nitrogen gas-assisted
. accumulator injection, resulted in backpressure against the CMT discharge check valves, caused
, CMT-1 injection flow to decrease, then drop to essentially 0 at about [ ]'6' and remain there until ACC-1 injection was completed at [ J'6' During this same time frame, CMT-2 injection flow also decreased, then dropped to about [ ]'6' and remained at that rate until ACC-2 injection was completed at [ ]'6' (Figure 5.1.1-16). l At the same time that the CMT injection flows were tapering off with the CMTs at about the
.[ ]'** level (or two-thirds full), with RCS pressure and, therefore, saturation temperature decreasing fairly rapidly, the steam space at the top of the CMTs became superheated primarily due to the pressure decrease and radiant heating from the hot CMT wall metal (Figures 5.1.1-50 and 5.1.1-51). The CMT steam space remained superheated, but at a slowly diminishing level of superheat until the CMT reflooded at [ ]'6'([ ]'**), 'when subcooled fluid O entered the CMT and cooled it (Figures 5.1.1-32 and 5.1.1-33). CMT temperature response is discussed more thoroubly in Subsection 7.2.
oAl$MwRevl\l5%w 8.non:IM)81798 3,],}.17 REVISION 1
FINAL DATA REPORT When accumulator injection was complete for each side, CMT injection was restored at a rate of [ ]'6' per side. Injection continued at a slowly diminishing rate down to about [ ]'*" when the CMTs emptied at [ ]'*' ([ ]'6' Figure 5.1.1-16). As the levels came down, the fluid was displaced by superheated steam, caused by the decreasing pressure and hot CMT wall metal. The balance line levels oscillated around zero during the period when the accumulators were injecting, but then began to indicate a level of [ ]'6' and stayed constant until they started to refill at about [ ]'6" This same level indication would be repeated again later in the test after the second draining of the CMTs and continue for the duration of the test. The effects of a level differential pressure transmitter acting as a flow differential pressure transmitter probably was not the cause, since the amount of steam flow required to initiate that effect would be sufficient to cool the CMTs, resulting in condensation of the steam and a rising fluid level. Also, based on reactor vessel downcomer and cold-leg levels (Figums 5.1.1-15 and 5.1.1-52) and cold-leg temperature indications in the superheated region (Figures 5.1.1-53 and 5.1.1-54), the cold legs were not liquid solid and therefore could not support a column of fluid in the balance lines. A logical explanation for the continuous balance line level indication is that the level transmitter reference legs were not completely filled, causing an erroneous level indication. Both CMT balance lines began to refill at about [ ]'** when reactor vessel level, indicated by LDP-140, was sufficiently high at [ ]'** to cover and refill the cold legs (Figures 5.1.1-30 and 5.1.1-31). At about [ ]'6' when the CMT-2 balance line had refilled completely, CMT-2 began to rapidly refill and in about [ ]'** was at the [ ]'6' level, or about two-thirds full. The refill with subcooled fluid caused a condensation /depressurization event to occur in CMT-2. This was reflected in the reactor vessel pressure and levels and CMT balance line level taking a sharp dip and then recovering. CMT-2's refilling also caused its internal temperatures to l decrease quickly from the superheated to subcooled region at less than [ ]'6" and to remain subcooled until the CMT drained again (Figure 5.1.1-32). When the CMT-1 balance line refilled completely at approximately [ ]'6' CMT-1 refilled to the [ ]'** level, or about two-thirds full in less than [ ]' 6
- Again, the rapid refill with subcooled fluid caused another condensation /depressurization event to occur, this time in CMT-1. The CMT-1 internal temperatures responded to the subcooled fluid almost identically to those of CMT-2 (Figure 5.1.1-33). The CMTs appear to have reached a level and temperature equilibrium where steam condensation and a resultant level increase were no longer taking place. Another possibility for the CMTs not refilling completely is that some of the nitrogen from the accumulators may have collected in the CMTs. During this time, there was no injection flow from the CMTs due to the higher static head of the IRWST holding the l CMT discharge line check valves closed.
When the CMT balance lines began to refill, CMT internal pressures began to decrease with respect to the reactor vessel upper head pressure. It is possible that pressure decreased because once the cold legs filled with subcooled fluid, displacing what had been a blanket of superheated steam, the superheated steam in the CMTs began to slowly condense in the balance lines, thus filling them over a 0:U536wRevi\l536w-8.non:Ib-o81798 5.1.1-18 REVISION 1
FINAs. DATA REPORT period of about [ ]'6" Once the balance lines filled and overflowed into the CMTs, the subcooled fluid sprayed through the inlet diffuser and collapsed the superheated steam bubble in the CMTs, causing more depressurization and the subsequent in-rush of fluid from the RCS that was reflected in the decrease in reactor ressel levels and pressure. Both CMTs remained at essentially constant levels for several thousand seconds, and then they began slow draindowns at about [ ]'6' respectively. ne draindown for both CMTs
. was slow and did not occur until IRWST relative level was [ ]'6* below that of the CMTs (Figure 5.1.1-34). Data indicate that the CMTs drained for some time, and then the differential head I:
between the IRWST and the CMTs closed the CMT discharge check valves, tenninating the draining until such time as the differential shifted the other way and draining recommenced. Both CMTs were completely empty at about [ ]*6' which coincides closely with the primary sump injection valve opening at [ l'6' A possible correlation for the coincidence is that when the primary sump valve opened, the IRWST htd just reached its minimum of about [ ]'6* which is about [ ]'6* below the instrumented level for the CMTs, and that there was still a slight negative pressure remaining in the CMTs. Also, when the primary sump injection valves opened, there was a short time period in which the IRWST and primary sump levels equalized, causing a decrease in RCS fluid levels and resulting in a rather rapid drop in CMT levels from about [ ]*6* Accumulators O Q Prior to test initiation, the accumulators were filled with ambient temperature fluid to an internal standpipe level, which ensures that they are filled to the same level for each test. They.were then pressurized to [ ]'6" with nitrogen and placed in service by the accumulator discharge valves being opened once the RCS was at normal operating temperature ar,d pressure. The accumulators represent a pure passive system that injects fluid through the DVI nozzles to reflood the reactor vessel once RCS pressure has dropped below accumulator pressure. The piping is shown in Dwg. OSU 600206 (Appendix G). ACC-2 injection started at [ ]'6' and ACC-1 at [ ]'6" (Figure 5.1.1-16). Since RCS pressure and, therefore, CMT overpressure were less than accumulator pressure, the accumulators' injection virtually shut off CMT injection by creating a backpressure against the CMT discharge check valves. De accumulators injected at approximately [ ]'6' per tank, resulting in about a [ ]'6# increase in the reactor vessel levels. Injection continued at that rate until [ J'6" for ACC-1-and [ J'6" for ACC-2. The completion of accumulator injection and subsequent resumption of CMT injection resulted in a reactor vessel core barrel level decrease of about [ ]'6' and a virtual stabilization of downcomer levels. l l l- Later in the OSU test program, it was discovered that when the accumulators were pressurized, the indicated level on the DAS increased several inches. The problem was the bourdon tube local pressure instruments, PI-401 and PI-402, were tubed to the bottom of the accumulator level differential
, pressure instrument reference leg tubing. As pressure increased during the final preparations for the test, the reference legs were compressed into the pressure indicators, resulting in the indicated level o:ushRevi\l5h-8.non:lb-081798 5.1.1 19 REVISION l 1
i
FINAL DATA REPORT increase. By re-routing the pressure indicator tubing to the top of the level differential pressure instrument reference legs, accumulator pressurization no longer had an effect on indicated level. The true starting level for the accumulators should be [ ]'6# for ACC-1 and [ ]'** for AC-2. For this test, the starting levels on the DAS were [ ]'6' for ACC-1 (LDP-401) and [ ]'6* for ACC-2 (LDP-402). However, due to the accumulators being filled to their standpipe levels, the correct critical levels were obtained. Pressurizer The pressurizer level at test initiation was [ ]*6' (uncompensated), and the pressure was [ ]'** with pressure control in automatic. The PLC was programmed to adjust the pressurizer heater demand to 0 at 5.6 seconds. When the break valve opened, the initial depressurization caused heater output to go to maximum, and at 5.6 seconds, heater power decreased to about [ ]'6'as opposed to 0 kW per design (Figure 5.1.1-24). At [ ]'*# heater output increased to about [ ]'6" and remained there for the duration of the test. It can be posruiated that the heater SCR control circuitry was not properly tuned. In subsequent tests, the heaur breaker was manually opened several seconds after the S signal. Later in the test program, P1'! logic was changed to open the heater contractor for positive de-energization at 5.6 seconds. The pressurizer connections to the other components in the facility are shown in Dwg. OSU 600203 (Appendix G). When the break valve opened, pressurizer level immediately started to decrease, and the pressurizer was empty at [ ]'6" (Figure 5.1.1-5). The pressurizer surge line was empty at [ ]'** It is believed that the steam environment and the fairly low heater power were sufficient to maintain the pressurizer heater sheath temperatures at saturation temperature when the pressurizer was empty (Figure 5.1.1-21). The pressurizer and surge line remained empty until the ADS-1 valve opened at [ ]'**at which time test data indicate that they began to reflood at the same time. The ADS-2 valve opened at [ ]'6* causing flow rate through the surge line and pressurizer to increase even more (Figure 5.1.1-55). When the ADS-3 valve opened at [ ]'** it had little or no effect on the rate of depressurization. The data indicate two-phase flow through the surge line and pressurizer to the IRWST. This created significant pressure drops through each component that may have influenced the level instruments (Figure 5.1.1-5). At about [ ]'6' HL-2 elbow level and calculated steam percent data indicate that it was full (Figures 5.1.1-56 and 5.1.1-57). The surge line data indicate that within several seconds, it also had filled and that the pressurizer had started to reflood at about [ ]'6" (Figures 5.1.1-58 and 5.1.1-59). During the period of accumulator injection between [ ]*** makeup to the RCS was approximately double the losses through ADS 1-3 and the break (Figures 5.1.1-60,5.1.1-61, and 5.1.1-62). The increasing inventony raised system fluid levels and when the hot legs filled caused oA1536w Rev i\l 536w-8.non: I b-081798 5.1.1-20 REVISION 1
,-. . - , .- _- _ _ . . _. . , . - = - . -. _ . _ FNAt. DATA RuonT l (q) reactor and pressurizer pressures to diverge. The reactor pressure was greater (Figure 5.1.1-63). This pressure differential was the force that pushed low steam percent fluid into the surge line and pressurizer. ADS 1-3 separator steam flow went to 0 at [ ]*6' which was a good indication of low steam percent fluid entering the separator (Figure 5.1.1-61). ACC-1 and ACC-2 were essentially empty at [ ]'6" and [ ]'6' respectively (Figure 5.1.1-64), leaving only CMT injection in progress. RCS inventory began to decrease again. ADS 1-3 separator level went above normal at [ ]"6' (Figure 5.1.1-65), and its loop seal flow began to increase (Figure 5.1.1-61), confirming that low steam percent fluid was being pushed through ADS 1-3. Another good indication of low steam percent fluid being lost through ADS 1-3 was that ADS 1-3 flow differential pressure instruments (Figure 5.1.1-55) oscillated around zero while the separator level increased. The pressurizer attained its maximum fluid level / minimum steam percent at [ ]'6"and almost immediately the level started to decrease rapidly (Figures 5.1.1-58 and 5.1.1-59). This decrease { rate is believed to be the result of two occurrences. First, the pressurizer was gravity-draining to the l RCS to make up for a decreasing inventory in the RCS. Second, the constantly decreasing pressurizer pressure and the heaters still being powered at [ ]'6' was causing the percent steam to increase, ! thereby pushing more fluid into the ADS 1-3 separator and on to the IRWST. At[ ] the ADS 1-3 separator reached its maximum level and maximum loop seal flow of ( . ]'6' (Figures 5.1.1-65 and 5.1.1-61). By [ ]'6# it retumed to normal level, and flow was significantly lower. The rate of pressurizer level decrease / steam percent increase decreased at about that same time, since the method of inventory movement out was by gravity to the RCS. What appears to be noise on the pressurizer level data during this period is believed to be caused by the flashing of liquid to steam to keep the pressurizer at saturation pressure. When the ADS-4 valves opened, the ADS 1-3 separator liquid flow decreased to 0 at [ ]'6' The pressurizer level continued to decrease by gravity until it was empty at [ ]'6' The reactor and pressurizer pressures converged to be essentially equal at [ ]'6' and then started to diverge when the surge line again began to refill. At[ ]'6' with RCS levels approaching their maximum, the surge line began to refill. When it was full at [ ]'6' the pressurizer began to reflood. Coincident with the pressurizer reflooding, the ADS 1-3 separator and ADS 1-3 sparger pressures went negative by as much as [ ]*6* further enhancing the level rise in the pressurizer (Figure 5.1.1-29). The negative pressure was caused by the level in the pressurizer, the ADS 1-3 sparger nozzles being submerged, and condensation due to cooling of the components in between. The pressurizer reflood
, this time can be confirmed by HL-2, pressurizer, and surge line temperatures being subcooled \ (Figures 5.1.1-14 and 5.1.127). In addition, the surge line completely filled first, followed by a level oA1536wRev nl 536w-8.non:I b-081798 5.1.1-21 REVISION 1
FINAL DATA REPORT l rise in the pressurizer. The negative pressure problem was corrected for subsequent tests by installation of a vacuum breaker on the sparger line inside the IRWST. The pressurizer filled to a maximum of [ ]'** at [ ]'6" and immediately began a rapid decrease to empty at [ ]'6* (Figure 51.1-5). The rapid decrease in level was coincident with IRWST level dropping below the sparger nozzles ac,d breding the vacuum. The rapid decrease in pressurizer level also caused an increase in reactor vessel levels until they regained equilibrium with the vest of the system (Figure 5.1.1-22). For the remainder of the test, the pressurizer remained empty. l Surge line level started to decrease at [ ]'** to about [ ]'** where it remained until [ ]' 6* The level again decreased to about [ ]'** at [ ]*** and then oscillated between [ ]'6* for the remainder of the test. 1 l Passive Residual Heat Removal Heat Exchanger The PRHR HX was filled arad vented with the rest of the system and then isolated by closing its outlet valve prior to the facility startup and heatup. Prior to test initiation, the outlet valve was placed in auto so that it would open as programmed by the PLC after the S signal. This maintained the fluid temperature in the HX equal to the IRWST temperature of [ ]'6* at test initiation. The PRHR HX inlet is piped from the ADS 4-2 line, which is attached to HL-2, and the outlet ties into the SG-2 outlet channel head, which is connected to CL-2 and CL-4 (Dwg. OSU 600203 and OSU 600206). The thermocouple arrangement is shown in Dwg. OSU 600701. These drawings can be found in Appendix G. Two minutes after the TEST pushbutton was pressed, the break valve opened (time zero). At 6 seconds, the PRHR HX outlet valve received its signal from the PLC to open. The RCPs were still running, so the initial flow through the PRHR HX was forced flow. Flow became solely natural circulation flow [ ]'6* later, when the RCPs were stopped by an automatic signal from the PLC. The inlet flow rose from 0 to about [ ]'6' and the outlet flow rose from 0 to about [ ]'6* at [ ]'6* (Figure 5.1.1-66). At the same time, the temperature difference between the inlet and outlet of the HX began to increase and was [ ]'6* at [ J'6* l (Figure 5.1.1-40). The difference between inlet and outlet flows could be attributed to the density , difference of the fluid and inlet flow, FMM-802, response to two-phase flow. l l In the early stages of the trander.<, the HX instrumented long- and short-tube fluid temperatures l generally remained below [ ]'** with the short tube predominantly hotter and maintaining a l wider delta temperature. This difference between long- and short-tube performance is evident until about [ ]'** when the short-tube temperatures rapidly dropped and aligned with the long-j tube temperatures (Figure 5.1.1-41). The drop in temperature did not coincide with any other event and is believed to be a flow stoppage through the PRHR HX; the indicated delta temperature was the l result of the temperature gradient in the IRWST fluid. Stoppage of flow is confirmed by CL-2 and l CL-4 data which show that the temperatures at the bottom of the reactor flange for both loops began ) to incr ase around this same time, followed saturation temperature for about [ ]'6' went l l o:\l 536w Rev lu S36w-8.non: l t>.081795 5, } , } .22 REVISION I
l' FmAr. DATA REPORT l 1 l into the superheated region for about [ ]'6' and then became subcooled when the cold legs ;
'( started refilling at about [ ]'6' (Figures 5.1.1-42 and 5.1.1-43). '
The PRHR HX inlet tempereture became subcooled coincident with the ADS-4 valves opening at [ ]'6' and over the next [ ]'6" dropped to and followed the outlet temperature. This was again an indication that there was no flow through the HX during this time frame (Figures 5.1.1-40). At [ ]'6' the PRHR HX inlet temperature instantly jumped from ( . ]'6" to saturation temperature. This happened about [ ]'6' after pressure, level, and flow oscillations began in the facility and is believed to be caused by the inlet line " burping," allowing the line to fill with saturated steam. Following this occurrence, all of the PRHR temperatures began to l slowly approach saturation. The PRHR HX remained full until [ ]'6' and then began to decrease to about [ ]'6' ( indicated level at [ ]'6' which was the lowest level that would be encountered during this ! test. A level of [ ]'6' was maintained during most of the test (Figire 5.1.1-67). The PRHR HX wide-raage level data indicate that the HX was partially filled between about [
]'6"(Figure 5.1.1-68). An explanation for this occurrence is that during the long period without flow through the HX and with the RCS loops full, the cooldown of the HX and associated piping caused it to go into a negative pressure, thereby drawing fluid up into it. Dere was insufficient irorumentation installed on the HX to verify this. He effects of condensation events in s the CMTs were reflected in the level changes at about [ ]'6' When RCS levels began to decrease at about [ ]'6' the PRHR HX level also began to i
decrease and was back to its minimum indicated level of about [ ]'6" by [ ]'6' I The level oscillations from about [ ]*6' until the PRHR HX reached its minimum are believed to be a result of the overall facibiy oscillations as described in the Test Analysis Reporr/h Review of the data does not reveal that flow m ever re-established through the HX during the l remainder of the ' test. 1
\
Steam Generators l The SG primary sides were filled and vented with the RCS. He secondary sides were also filled prior ! to facility startup and heatup. Prior to test initiation, the SG levels were established at [ ]'6' on I the narrow range instruments with the feedwater pump and feedwater control valves in AUTO. The steam pressure controller was also placed in AUTO to control secondary-side pressure at [ ]'6' Coincident with the S signal being generated at 0.5 second, the steam pressure controller's setpoint was j reset to [ ]'6" which was high enough so that the control valve would remain closed yet low enough that it would not challenge the SG safety valves. At 4 seconds, the PLC signaled the feedwater pump to trip and the feedwater control valves to close, essentially bottling up the SGs to minimize any cooldown of the RCS. The SGs are shown in Dwg. OSU 600002 and OSU 600203, and SG instrumentation is shown in Dwg. OSU 600301. (These drawings can be found in Appendix G.) t . oA15hRevl\l5h-8.non:lt>081798 5,1,].23 REVISION 1 1
FmAL DATA REPORT l The SGs initially absorbed some heat from the RCS, and their secondary pressure increased to a maximum of [ ]'6' at [ ]'6' (Figure 5.1.1-45). The pressure then stayed fairly constant until [ ]'6' when secondary pressure and primary pressure were essentially equal. Secondary pressure then followed primary pressure until about [ ]'6# at which time the SGs started transferring heat to the RCS. Following this time, the SG secondary-side pressure decay was dictated only by SG heat losses to the RCS and to ambient conditions. At the end of the test, they were at about [ ]'6' On the primary side, the SG-2 tubes began to drain at about [ ]'6" and SG-1 tubes began to drain at about [ ]'6' (Figures 5.1.1-8 and 5.1.1-7). Between [ ]'6' the SG 1 U-tubes were completely empty; between [ ]'6' the SG-2 U-tubes were completely empty. The delay in SG-2 could be caused by the effects of PRHR HX flow entering the outlet channel head, reducing the steam expansion rate in the U-tubes. The SG level data then indicate that the tubes began to refill after about I minute and return to their original level. It is possible that this apparent refill indication is the result of the vaporization of level instrument reference legs, which are routed intemal to the SG tube bundle. This phenomenon is treated separately in Section 2.4. Cold Lees and Hot Lees Due to the totally different responses between the cold legs and the hot legs, they are addressed separately here. The loop components, piping, and instrumentation are depicted in Dwg. OSU 600203 (Appendix G). The cold legs began to drain at about [ ]'6' when the level in the reactor vessel downcomer dropped below the tops of their nozzles (Figure 5.1.1-15). When the nozzles started to uncover, the nozzle thermocouple temperatures located at the top of the nozzle flanges quickly rose into the superheated region (Figures 5.1.1-42, 5.1.1-43, 5.1.1-53, and 5.1.1-54). At [ ]'6'a condensation /depressurization event took place which rapidly refilled and then partially redrained the cold legs. This can be seen on the cold-leg temperature plots as a sharp drop into the subcooled region with a quick return to superheat. 'Ihe temperature data from the CL-3 reactor vessel nozzle flange bottom thermocouple show that temperature oscillated between subcooled and superheated conditions following the ADS 1 valve opening at [ ]'6' Then, at about [ ]'6' data indicate superheat, remaining in the superheat region until [ ]'b' when the RCS regained enough inventory from IRWST injection to start refilling the cold legs. It is possible that the break location, at the bottom of CL-3, placed it at a slightly lower pressure than the other cold legs and allowed the small amount of liquid in the bottom of the pipe to flash to steam, resulting in this phenomenon. During the period of about [ ]'6* which coincides with CL-3 being empty, there was no indicated flow through the break (Figure 5.1.1-28). O c:\1536wRev i\l 536w-8.non: I b-081798 5.1.1-24 REVISION l
FINAL DATA REPORT Temperature plots indicate that the other three ccid-leg reactor vessel nozzle flange bottom
) thermocoupla did not superheat until between [ ]'6' indicating that they were not totally empty until then. All four cold-leg bottom thermocouples became subcooled at the same time, indicating that the legs were refilling. The four cold legs were completely filled with liquid between about [ ]'6' when the data show that their top thermocouples became subcooled and stayed filled until approximately the time that primary sump injection began and all the cold-leg thermocouple temperatures went to saturation temperature. The temperatures remained at saturation for the duration of the test, and the legs were possibly filled with a two-phase mixture.
It is believed that the method by which cold legs become superheated is that a saturated two-phase mixture moves from the heater bundle area of the reactor througn the hot legs and into the SGs. where it is superheated. The superheated steam then flows back to the reactor vessel downcomer through the cold legs. During the period that the cold legs were superheated, there was a negative differential pressum arross the reactor vessel upper support plate and across the core barrel flange bypass holes. This means that the reactor vessel downcomer/ cold legs were at a lower pressure than the core barrel / hot legs, which created the differential required to cause steam to flow through the SG U-tubes and become superheated (Figure 5.1.1-19). The hot legs followed the saturation curve thrcughout the test, except for a few instances that will be discussed here (Figures 5.1.1-13, 5.1.1-14, and 5.1.1-15). 'Ihe HL-2 thermocouple, located in the elbow at the SG, data show that the elbow became at least partially empty between [ O' ]'6# and the steam volume was superheated by the massive heat source of the SG. Both
. hot leg reactor vessel nozzle flange temperatures indicated subcooling between about [
l
]'6# coincident with the period that reactor vessel full range level (LDP-127) showed that i the hot legs were submerged (Figure 5.1.1-15). The reactor vessel level decreased at about
)
[ ]'6# resulting in at least partial uncovering of the hot legs and the subsequent return to ! saturation temperatures. Temperatures remained at saturation until about [ ]'6" when reactor vessel levels again increased and refilled the hot legs with subcooled liquid. The hot legs remained subcooled until about [ ]'6" when the reactor vessel level again decreased. The hot legs partially voided and thus allowed saturated steam to enter from the reactor upper plenum area. Hot-leg temperatures stayed at saturation for the remainder of the test. It is believed that the hot legs, even though partially or completely empty during the test, remained at saturation temperature and never superheated due to a small flow of saturated steam from the reactor core to the SGs, which kept the piping at saturation temperature. This steam could take several paths l
- on the HL-2 side
- it could flow to the PRHR HX; it could flow to the pressurizer through the surge line; it could flow to the SG, where it would be superheated by the secondary side; or it could flow to the ADS 4-2 system once ADS-4 actuation had taken place. On the HL-1 side of the facility, it could only flow to the ADS 4-1 system or to the SG.
oA15%wRevluSMw-8.non:lb 081798 5.1.1-25 REVISION l
FINAL DATA REPCOT l In Containment Refaeline Water Storace Tank I Prior to test initiation, the IRWST was emptied, load cell transmitter LCT-701 zeroed, and the tank l filled to the overflow with ambient temperature water, it was then drained to the level of the fill I nozzle, which was the level set during pre-operational testing. The initial tank conditions were [ ] of liquid, [ ]'6' liquid level, and [ ]'6' liquid temperature. The IRWST injection valves were closed and in AUTO. Dwg. OSU 600206 shows how the IRWST is connected in the facility. Dwg. OSU 600701 depicts relative locations of the instrumentation. These drawings can be found in Appendix G. The break valve opened at 0 second, and the resultant break flow into the break separator caused a spike pressure increase in the break separator to [ ]'6' and a subsequent backpressure spike in the IRWST of [ ]'6' It did not result in a pressure spike in the primary sump, thereby creating a differential pressure between the IRWST and the primary sump that lasted until about [ ]' 6' This differential pressure could been a result of the tank free volumes and size of the vent pipes connecting to the BAMS header. is, the IRWST was full of liquid to the fill nozzle, and the sumps were empty; the IRWST has a [ ]'6' vent, and the two sumps share one [ ]'6' vent piped to the BAMS header. The differential pressure caused a small overflow indication (FMM-703) from the IRWST to the primary sump for approximately [ ]'6' This was not a real overflow but, rather, fluid from the IRWST overflow loop seal being pushed into the sump (Figure 5.1.1-69). This can be confirmed by the fact that there was no change in IRWST level during this period (Figure 5.1.1-26). The IRWST thermocouples located at or above the initial fluid level almost immediately following the opening of the break valve began to indicate an increasing temperature (Figures 5.1.1-70, 5.1.1-71, 5.1.1-72 and Dwg. OSU 600701, Appendix G). The increase is due to a small backflow of steam from the BAMS header into the IRWST vapor space. The ADS 1-3 separator, associated automatic valves, and piping were discussed extensively in the Pressurizer Response description in this section; therefore, only those aspects thnt directly affect IRWST response will be discussed here. l The piping from the ADS 1-3 valves, including the separator and where the separator steam and liquid I out piping join, is heat-traced and therefore elevated in temperature prior to test start. The ADS-1 valve opened at [ ]'6" initiating two-phase flow from the RCS through the pressurizer, ADS 1-3 separator, and the ADS 1-3 sparger into the fluid volume of the IRWST. There was an immediate rise in sparger tip temperature, indicated by TF-719, followed by a rapid rise in long-rod terr.prature 'IF-706 located on the long rod about [ ]'6' below the sparger nozzles, and a slower rise in TF-705, also located on the long rod but an additional [ ]'6' down, when the ADS-2 valve opei.ed at [ ]'6* (Figures 5.1.1-70 and 5.1.1-72). The initial effects on TF-706 and TF-705 were due to their close proximity to the sparger nozzles. o:\l5X.<Revl\l536w-8.non:lb o81798 3,] ,] .26 REVISION 1
FmAt. DATA REroar
} When the ADS 1-3 valves opened and added inventory to the IRWST prior to IRWST injection, the
) level increased to tank overflow (Figure 5.1.1-26), causing overflow to the primary sump from about j [ ]'6' (Figure 5.1.1-25). The cessation of overflow came within [ ]'6' of the beginning of IRWST injection to the DVI nozzles.
ADS 1-3 flow rate to the IRWST peaked at about [ ]'** (Figure 5.1.1-61), which is coincident with the peak temperature seen at TF-719, located at the tip of the sparger nozzle ! (Figure 5.1.1-70). De nozzle temperature then decreased from about [ ]'6' over the next [ ]'6' coincident with the ADS 1-3 separator flow decrease from its peak to 0 at [ ]'6' For the remainder of this test, there was no discernable flow from ADS 1-3 into I the IRWST, and temperatures remained essentially constant until IRWST injection began to drop the fhiid level in the tank. Reactor vessel pressure (Figure 5.1.1-45) decreased to [ ]'6' its low-low pressure setpoint, at l [ ]'6' and sent a signal to the PLC to automatically open the IRWST injection valves. l Injection from the IRWST did not begin at this time, since RCS pressure was significantly higher than the IRWST static head. When the ADS 4-1 and ADS 4-2 valves opened at [ ]'6# providing two additional depressurization path' or the RCS, RCS pressure dropped to just below [ ]'6' allowing IRWST injection to start ' the no. 2 side at [ ]'6' and on the no. I side at [ ]'6' (Figure 5.1.1-37). IRW ' injection increased slowly to its maximum for O i this test of[ ]*** per side at about [ ]'6* and then the rate slowly decreased over time as the tank static head decreased. i As IRWST injection cominued and the tank level decreased, the long-rod thermocouple temperatures i rose very slowly. At the time of switchover to primary sump injection / recirculation mode, none were above [ ]'6' When they uncovered, their temperatures rose but still maintained a gradient with the warmest at the top and the coolest at the bottom. This is believed to be the result of no mixing, since there was no discernable ADS 1-3 flow into the tank after about [ ]'6', and there
- would also be a temperature gradient in the tank wall with the warmest at the top due to the rather slow gravity draindown.
At the time primary sump injection started through the check valves, the IRWST injection flow rate was approximately [ ]'6' per side. De flow then split at approximately [ ]'** per side f rom the IRWST and [ ]'b' per side from the primary sump. When the primary sump injection valves opened, the IRWST-2 side flow increased to approximately [ ]'6' the IRWST-1 side flow indication went negative, the primary sump-l flow increased to approximately [ J'6' and primary sump-2 flow decreased to approximately [ ]'6' hese are true indications of what l was happening. A portion of the primary sump-l flow was diverted to the IRWST, and only about i
. [ ]'6* flowed to the DVI nozzle. On the no. 2 side, the total flow to the DVI nozzle was the sum of the IRWST and primary sump injection flows on that side (Figure 5.1.1-37). De IRWST and
( pnmary sump flows remained essentially constant for the remainder of the test. This flow pattern is 1 oAl5WRevN$h-8.non:Ib-081798 3,1,]-27 REVISION 1
FINAL hATA RuoRT most probably due to larger diameter piping from the IRWST to the primary sump connection on the DVI manifold on the no. I side ([ ]*"" 1.D. versus [ ]'** 1.D.). Around [ ]'** all of the IRWST tempemtures increased, and some of the upper thermocouples reached saturation. This increase is believed to be due to the hotter primary sump fluid that was being injected into it though its no. I side injection line. Break and ADS Measurem.nt System For the test initial conditions, the BAMS steam header was lined up so that the break separator steam flow would be directed through its 6-in. vent line and all BAMS header steam would be released through the 6-in. line containing valve CSS-901 and steam flow vortex meter FVM-901. The valves in the header were all positioned with their controls in manual. The primary sump [ ]'** vent line was not installed for this test. The break separator, tue ADS 4-1 and ADS 4-2 separators, and the ADS 1-3 separator loop seal liquid line were all heat-traced with their temperature controls set to maintain [ ]'"* The loop seal liquid lines for the break separator and the ADS 4-1 and ADS 4-2 separators were not heat-traced. All of the steam pping and the ADS 1-3 reparator were lieat-traced with their temperature controls set to maintaiD 't ]*6* The loop seals for all four separators were filled with [ ]'** fluid from the con densate retum system just prior to test initiation. The BAMS is shown in Dwg. OSU 600901 (Appendix G). Almost instantly when the break valve opened, break separator pressure started te increase with a peak O pressure of about [ ]'** at [ ]"* All of the other components in the BAMS exhibited delayed and similar, but smaller peaks, as the result of backpressures created from the header with the exception of the primary sump (Figure 5.1.1-73). This is believed to be due to the relative sizes of components vent piping with the primary sump having the smallest vent pipe at [ ]'** Even with the RCS at steady-state conditions between [ ]*** the BAMS component pressures all agreed within about 1 psig, which can be considered well within allowable instrument errors (Figures 5.1.1-74 and 5.1.1-74x). When the break valve opened, data indicate that the break separator level dropped to a low of [ ]'** at [ ] and then recovered right back to normal level at [ ]'** (Figure 5.1.1.26). The reference leg for break separator level instrument LDP-905 is installed on the side of the tank and near the top, and the variable leg is connected to the vertical section of the loop seal leaving the separator. This type of installation, combined with the fact that the level transient was completed within [ ]'** makes it very likely that the data reflect the dynamic effects of flow rather than a real level transient. All of the ADS separators experience this same type of initial response when their associated depressurization valves open. The break separator and BAMS header steam flows are shown on Figure 5.1.1-75 and the break separator loop seal flow on Figure 5.1.1-28. It can be seen that both steam and liquid flow began immediately following the break valve opening with steam flow out of the system essentially stopped cA1536w Rev i\ l 536w-8.non: I b-081798 $,},].28 REVISION 1
Fsw.t. DATA REroar 1 at [ ]'6' He loop seal liquid How peaked at [ ]'6' at [ ]'6' and decreased (A to 0 at [ ]'6" when the ADS 4-1 and ADS 4-2 valves opened and the cold legs began to approach saturation temperature and were emptying. The data indicates that loop seal flow resumed at [ ]'6# which is coincident with the cold legs refilling, and continued at [ ]'6' until it went negative at about the same time the primary sump began to inject through its check valves. It remained negative or 0 for the duration of the test. He reason for the flow reversal is that the sump level became higher than the break level. It is also believed that the reason that the break
- flow went to 0 for about [ l'6' is that the cold legs were filled with steam.
ADS 4-1 and ADS 4-2 opened at [ ]'b' and resulted in less than [ ]'6' of steam flow each for a period of less than [ ]'6' (Figure 5.1.1-76). The liquid How from the < separators had initial peaks of about [ ]'** for ADS 4-1 and ADS 4-2, respectively (Figure 5.1.1-28). The liquid flow continued throughout the remainder of the test ranging from about ! [ ]'6 per side, j The ADS 1-3 separator does not tie directly into the BAMS and was extensively covered in the IRWST Response and Pressurizer Response portions of this section, so it is not covered further here. During this test, the BAMS controller calculated the total steam loss from the facility to be [ ]'6" Of that amount, [ ]'6" would have to be made up to the IRWST and n [. ]'6# to the primary sump. The equivalent amount of liquid from the condensate return system v at [ ] would have been [ ]'6' His was not made up since the system controls were not fine enough. Rese amounts were recorded in the test log. ! 5.1.1.6 Mass Balance The mass balance results for Matrix Test SB01 test data were calculated based on water inventory before and after the test and are provided in Appendix E. De mass at the end of the test was within [ ]'6' of the mass at the beginning of test. 5.1.1.7 Conclusions he test was performed with minimal problems and is considered acceptable. Although not all of the facility initial conditions met the specified acceptance criteria, the deviations did not impact the quality of the data. De instrumentation problems encountered were not critical to the performance of the facility mass and energy balances. Facility response to the test was as anticipated for the conditions that were established. The data clearly demonstrate that cooling of the reactor heater rods was maintained throughout the duration of the test.
\
o:\l536wRevl\l536w-8. ann:ltr081798 5.1.1-29 REVISION 1
l Fv4AL DATA REPORT i TABLE 5.1.1 1 MATRIX TEST SB01 INITIAL CONDITIONS Instrument Specified Initial Actual Initial Parameter No. Condition Condition Comments _ as Pressurizer pressure'" FT-604 370 2 2 psiF HL-1 temperature'" SC-141 42012'F HL-2 temperature") SC-140 420 t 2*F [ ]"" above required temperatere SG-1 pressure'" FT-301 285 2 5 psig SG-2 pressure'" FT-302 285 t 5 psig Pressurizer level'" LDP-601 65 2 5 in. Uncompensated level (corrected for specific volume change) SG-1 narrow-range LDP-303 26 $ 3 in. Uncompensated level (corrected level'" for specific volume change) SG 2 narrow-range LDP-304 26 2 3 in. Uncompensated levci (corrected level'" for specific volume change) IRWST temperature"' TF-709 < 80 F CMT-1 temperature"' TF-529 < 80'F CMT-2 temperature"' TF-512 < 80 F [ ]"' above required temperature ACC-1 temperature"' TF-403 < 80 F ACC-2 temperature"' TF-404 < 80 F IRWST level") LDP-701 Level established by fill line elevation ACC 1 level"J' LDP-401 Level established by standpipe at 37 in. ACC-2 level"J' LDP-402 Level established by standpipe at 37 in. ACC-1 pressure"' PT-401 232 2 2 psig [ ]"# below required level ACC-2 pressure"' Irr-402 232 2 2 psig _ _ o:\l 536wRev l \l 536w-8.non:l tr081798 5.1.1-30 REVISION 1
I FINAL DATA REPORT TABLE 5.I.I I (Continued) U MATRIX TEST SB01 INITIAL CONDITIONS Instrument Specified Initial Actual Initial j Parameter No. Condition Condition Comments l CMT-1 level <2> LDP-507 Full
- CMT-2 level <23 LDP-502 Full _ _
Iielt: (1) The test procedure for Matrit Test SB01 did not require recording the control board indications prior to this test, but the specified conditions were verified on the control board prior to test actuation. The initial ; conditions from DAS, averaged over 2 minutes prior to the break valve opening, are recorded here. (2) Data were r.ot recorded in the procedure, but the test engineer verified that specified conditions were achieved while establishing initial conditions. The value of the parameter was determined post-test by calculating the average DAS indication for a time of about 2 minutes before the break valve opened.
'3) The bourdon pressure tube local indicator (PI-401 or PI-402) was tub d to the lower portion of the reference leg for differential pressure transmitter (LDP-401 or LDp-402). As pressure in the accumulator increased, the air inside the bourdon tube was compressed, thereby lowering the reference leg liquid level. This resulted ie a false indication of measured level. .
i I i l l l ( oAl5%wRevl\l5Mw-8.non:Ib481798 5.1.1-31 REVISION l
FINAt. DATA REPORT TABLE 5.1.12 MATRIX TEST SB01 INOPERABLE INSTRUMENTS / INVALID DATA CHANNELS Instrument No. Instrument Type Description of Problem FDP-6M* Differential pressure transmitter Over-ranged momentarily when associated ADS FDP-605* flow velve opened FDP-606* FMM-20l* Magnetic flow meter Data invalid after 16 seconds due to possible FMM-202* steam in cold leg FMM-203* FMM-204* FMM-402* Magnetic flow meter Inoperable - reads twice DVI flow during accumulator injection FMM-50l* Magnetic flow meter Data invalid between [ ] and after [ ] when CMT is empty FMM 502 Magnetic flow meter Data invalid after [ ]'6' due to possible steam in balance line FMM 503 Magnetic flow meter Data invalid after 104 seconds doe to possible steam in balance line FMM-5N* Magnetic flow meter Data invalid between [ ]'6' and after [ ] when CMT is - empty FMM-70l
- Magnetic fiow meter Negative values after primary sump valves open at [ ]'6' are invalid (Subsection 5.1.1.2)
FMM-802* Magnetic flow meter Data invalid after steam forms in PRHR HX inlet line, which appears to be at about [ ] FMM-804* Magnetic flow meter Data valid until PRHR HX initially drained at [ ] after this time, the possibility of steam in the outlet line invalidatcs the data FMM-905
- Magnetic flow meter Negative values after break separator level exceeds the break elevation at about
[ ] are invalid HFM-ll2 Heat flux meter Out of service at test start HFM-505 Heat flux meter Data appear erratic HFM-703 Heat flux meter Inoperable throughout test HPS-509-1 through 3 Heated phase switch Inoperable throughout test 9 c:\l 536w Rev i\l 536w-8.non: I b-081798 5,1,] .32 REVISION 1
FINAn. DATA Raro1T O TABLE 5.I.I 2 (Continued) MATRIX TEST SB01 INOPERABLE INSTRUMENTS / INVALID DATA CHANNELS l Instrunnent No. Instrument Type Desemiption of Problem I l' ' LDP-201 Differential pressure transmitter - Data invalid due to effect of vertical portion of I LDP-202 level sense line attached to top of pipe; data can show
-l LDP-203 level trends, when pipe is empty or starts to l LDP-204 l drain, but absolute level indication cannot be l l LDP-205 used I LDP-206 LDP-207 Differential pressure transmitter - Inoperable - ranged improperly; data can show LDP-208 level level trends, but absolute level indication cannot LDP-209 be used l
LDP-215* Differential pressure transmitter - Inoperable - when tube voids, reference leg ! LDP-216 level steams off (Subsection 2.4) LDP-217 LDP-218* LDP-219* LDP-220 LDP-221 LDP-222* LDP-40l* Differential pressure transmitter - Data invalid (Subsection 5.1.1.2) p LDP-402* level LDP-509 Differential pressure transmitter - Reference leg appears to not have been LDP-510 level completely filled LDP-801 Differential pressure transmitter - Inoperable -level never changed during test level IrT_101 Pressure transmitter Data less than 6.1 psig invalid PT_102 Pressure transmitter Data less than 6.2 psig invalid IrT_103 Pressure transmitter Data less than 6.2 psig invalid FT_104 Pressure transmitter Data less than 6.4 psig invalid IrT_108 Pressure transmitter Data less than 8.4 psig invalid IrT_109 Pressure transmitter Data less than 6.3 psig invalid IrT_Ill Pressure transmitter Data less than 6.0 psig invalid IFT_Il2 Pressure transmitter Data less than 8.8 psig invalid IrT_ll3 Pressure transmitter Data less than 6.4 psig invalid PT 20l* Pressure transmitter Data less than 1.1 psig invalid PT_202 Pressure transmitter Data less than 5.9 psig invalid IrT 205 Pressure transmitter Data less than 6.1 psig invalid
- s 1
c:\l5hRevi\l5h 8.non:lb-061798 5.1.1-33 REVISION 1
FINAL DATA REroRT l l l l l TABLE 5.I.12 (Continued) MATRIX TEST SB01 INOPERABLE INSTRUMENTS / INVALID DATA CIIANNELS Instrument No. Instrument Type Description of Problem TF-169 Thermocouple fluid temperature Data invalid due to leaking O-ring in core barrel l TF-50l
- Thermocouple fluid temperature Inoperable - indicates ambient throughout test TF-5(M
- Thermocouple fluid temperature Inoperable - indicates ambient throughout test l
TF-536 Thermocouple fluid temperature Inoperable - in<!icates ambient throughout test l l TF-542 Thermocouple fluid temperature Inoperable - leads swapped t TF-619 Thermocouple fluid temperature Inoperable - indicates ambient throughout test l TFM-703 Thermocouple for HFM 703 Inoperable - indicates ambient throughout test TH-317-1 through 4 Thermocouple heater rod Inoperable - heater rod C2-317 removed prior to test i TW-210 Thermocouple wall temperature Data appear erratic TW-503 Thermocouple wall temperature Inoperable throughout test DV-526 l l TW-530 l TW-542 TW-534 Thermocouple wall temperature Inoperable -indicates ambient temperature TW-552 throughout test Note:
- Instruments marked with an asterisk are critical instruments. See Subsection 5.1.1.2 for discussion.
1 l l l l 9 oA15hRevl\1536w-8 non:1L 081798 5.1.1-34 REVISION 1
l FINAI DATA REP 0t.T 4 TABLE 5.1.13 MATRIX TEST SB01 SEQUENCE OF EVENTS Time after Data Break Event
- Description in Bar Chart
- Source * (sec.)
TEST Pushbutton Depressed TEST PB Pressed D - Break Valve Open Signal Break Viv Open Sig D Break Valve Starts to Open - Break Viv Open D Feed Pump Trips Feed Pump Trips D CMT-1 Outlet Valve Starts to Open CMT-1 Inj Viv Open D CMT-2 Outlet Valve Starts to Open CMT-2 Inj Viv Open D PRHR HX Outlet Valve Starts to Open PRHR HX Viv Open D Reactor Coolant Pumps Trip RCPs Trip D CMT-1 Recirculation Flow Stops (LDP-509) CMT 1 Recire Flow Stops A CMT-2 Recirculation Flow Stops (LDP-510) CMT 2 Recire Flow Stops A-Pressurizer Empty (LDP-601) Pressurizer Empty A Pressurizer Surge Line Empty (LDP-602) Surge Line Empty A SG-1 Hot-Leg Short Tube Empty (LDP-217) SG-1 HL Sbit Tube Empty A SG-1 Hot Leg Long Tube Empty (LDP-215) SG-1 HL Lng Tube Empty A CL-1 Channel Head Empty (LDP-21!) CL l Chan Head Empty A CL-3 Channel Head Empty (LDP-213) CL-3 Chan Head Empty A SG-1 Cold-Leg Short Tube Empty (LDP-221) SG-1 CL Shrt Tube Empty A SG-1 Cold Leg Long Tube Empty (LDP-219) SG-1 CL Lng Tube Empty A SG-2 Cold-Leg Short Tube Empty (LDP-220) SG-2 CL Shrt Tube Empty A SG-2 Cold-Leg Long Tube Empty (LDP-222) SG-2 CL Lng Tube Empty A SG-1 Hot Leg Short Tube Empty (LDP-216) SG-2 CL Shrt Tube Empty A Time of Minimum Reactor Level Observed Time of Min Rx Level A During Test (LDP-127) SG-2 Hot-Leg Long Tube Empty (LDP-218) SG-2 HL Lng Tube Empty A HL 1 Pipe Starts to Drain (LDP-205) HL-1 Pipe Starts to Drain A CL-4 Channel Head Empty (LDP 212) CL-4 Chan Head Empty A CL-2 Channel Head Empty (LDP-210) CL-2 Chan Head Empty A _ _ oAID6wRevi\l5%w-8.non:lb.081798 3,],].35 REVISION I
FINAt. DATA Rt.roat TABLE 5.1.13 (Continued) MATRIX TEST SB01 SEQUENCE OF EVENTS Time After Data Break Event
- Description in Bar Chart
- Source * (sec.)
CMT-1 Low Level Signal CMT-1 Level ' o D ACC-2 Injection Starts (FMM-402) ACC-2 Inj Starts A ACC-1 Injection Starts (FMM-401) ACC 1 Inj Starts A ADS 1 Valve Starts to Open ADS-1 Viv Open D CMT-2 Low Level Signal CMT-2 Level Lo D SG-2 Hot-Leg Elbow Starts Draining (LDP-208) HL-2 Elbow Starts to Drain A SG-2 Hot-Leg Channel Head Empty (LDP-214) HL-2 Chan Head Empty A HL-2 Pipe Starts to Drain (LDP-206) HL-2 Pipe Starts to Drain A HL-2 Pipe Empty (LDP-206) HL-2 Pipe Empty A ADS-2 Valve Starts to Open ADS-2 Viv Open D HL-1 Pipe Empty (LDP-205) HL-1 Pipe Empty A ADS-3 Valve Starts to Open ADS-3 Viv Open D Reactor Pressure Low Reactor Pressure Lo D IRWST-2 Injection Valve Starts to Open IRWST-2 Inj Viv Open D SG-2 Hot Leg Elbow Minimum (LDP-208) HL-2 Elbow Level Min A IRWST 1 Injection Valve Starts to Open IRWST-1 Inj Viv Open D Pressurizer Renoods (LDP-601) Pressurizer Renoods A ACC-! Empty (LDP-401) ACC-1 Empty A ACC-2 Empty (LDP-402) ACC-2 Empty A CMT-1 Level Low-Low CMT-1 Level Lo-Lo D ADS 4-1 Valve Starts to Open ADS 4-1 Viv Open D ADS 4-2 Valve Starts to Open ADS 4-2 Viv Open D CMT-2 Level Low-Low CMT-2 Level Lo-Lo D CMT-2 Empty (LDP-502) CMT-2 Empty A IRWST-2 Injection Starts (FMM-702) IRWST-2 Inj Starts A IRWST-1 Injection Starts (FMM 701) IRWST-1 Inj Starts A CMT-1 Empty (LDP-507) CMT-1 Empty A CMT-2 Starts to Renood (LDP-502) CMT-2 Renoods A _ _ g oA1536w Rev ill 536w-8.non:l t481798 5,1,1-36 REVISION 1
_ . - . _ _ . _ . . _ _ _ _ . _ . - _ . . _ _ . _ _ . - _ - . _ . _ _ _ .__ _ _. _ _ _ .~.. _ __. FINAL DATA REPORT s TABLE 5.1.13 (Continued) MATRIX TEST SB01 SEQUENCE OF EVENTS Time After Data Break Event
- Description in Bar Chart
- Source * (sec.)
- ~
CMT-l Starts to Reflood (LDP-507) CMT-1 Refloods A Primary Sump Starts to Overflow to Secondary Pri Sump Overflows A Sump (LDP-901) Primary Sump-2 Injection Starts (FMM-902) Pri Sump-2 Inj Starts A Primary Sump-l Injection Starts (FMM-901) Pri Sump-l Inj Starts A Primary Sump-l Injection Valve Starts to Open Pri Sump-l Inj Viv Open D Primary Sump-2 Injection Valve Starts to Open Pri Sump-2 Inj Viv Open D SG-1 Hot-Leg Channel Head Empty (LDP-209) HL-1 Chan Head Empty A SG-1 Hot-Leg Elbow Starts Draining (LDP-207) HL-1 Elbow Starts to Drain A SG-1 Hot-Leg Elbow Minimum (LDP-207) HL-1 Elbow Level Min A _ __ E9ls: (1) Data from the instrument channel in parenthesis were used to determine level, flow, or pressure conditions. (2) The attached bar chart provides a graphic representation of the timing of events. (3) D = time data obtained from a software program that monitored the input and output of the facility's PLC. A = time data obtained by reviewing data from the instrument channel listed in the Event Description column. (4) 0.0.S. = out of service b oA15%wRevi\tSMw 8.non:Ib-081798 5.1.1-37 REVISION 1
FINAL DATA REPORT The bar charts for Table 5.1.13 on pages 5.1.138 through 5.1.1-44 are not included in this nonproprietary document. 8 O O c:\l536wRevl\l536w-Ba.non:Ib481798 5.1.1-38 REVISION 1
..._,..____._____..____.._.____._.,_...._...._._.._._._,__.-._.___..__.._._.m_.. . . _ _ _ . . _ . _ _ _ . . _ _ _ _ _ _ _ . . _ _ _ _ .
FINAL DATA REPORT E i i O l l 1 l i 1 l
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. rd 4 si -
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h 5 'l 3- E A
-F i --nar 4 E-m- i i-F D F 1- 1F 1 Figure 5.1.1-1 Primary Loop and Break Piping Layout o:\l536wRevl\l536w-8a.non:lb4)81798 5.1,] .45 REVISION 1
1 FINAL DATA REPORT j s i 9 i JWL i l (DOC 4 l _J L M_. n TS-201 T : {03 .
)
PLUG m. dit x _ (f RV227 N ' CL-3 DP - 215 ,, , _f '
-QJ1 !
I s,e < T ____ RV237 - 4
! TS-205 1 !
Figure 5.1.1-2 Primary Loop and Break Pipe Arrangement O o:\l 536m Rev l\l 536w-8tnon: I b-081798 5, } , } .46 REVISION 1 1
4 1 FmAL DATA REPORT i i { Figures 5.1.13 through 5.1.180 on pages 5.1.1-47 through 5.1.1-125 are not included in this l 5 nonproprietary document. l d .i A k t 1 i 4 ( c:\l 5%*Rev i\l 5%w-8a.non: I b-082598 5.1.1-47 REVISION 1
i FINAL DATA REroar i 5.1.2 Test Repeatability (Matrix Test SB18 Comparison with Matrix Test SB01) Dis section identifies and describes the response of the facility based on a comparison between Matrix Test SB18 (OSU Test U0018) and Matrix Test SB01 (OSU Test U0001). The simulated break for Matrix Test SB18 was located at the bottom of CL-3 with a simulated failure of one of the ADS-4 l lines. CL-3 is on the CMT side of the facility (Figures 5.1.2-1 and 5.1.2 2). Matrix Test SB18, the l final Westinghouse-performed matrix test, duplicated Matrix Test SB01, the first test performed. The purpose of performing Matrix Test SBl8 was to confirm the ability of the facility to replicate its response to a small-break loss-of-coolant accident (SBLOCA) with the same configuration from the beginning to the end of the test program. Matrix Test SBl8 was performed on September 15,1994. This test was considered successful because the reactor heater bundle cooling was maintained throughout the test. The transient was initiated when i break valve TS-205 opened and continued through ADS actuation; CMT, accumulator, and IRWST injection; and primary sump recirculation injection. Subsection 5.1.2.1 provides details related to the systems' configuration and initial conditions. A description of inoperable instruments is provided in Subsection 5.1.2.2, and Subsection 5.1.2.3 references the sequence of events. A discussion of the test results and evaluation can be found in Subsection 5.1.2.4, and Subsection 5.1.6.5 is a comparison of component responses. A summary of the mass balance results is provided in Subsection 5.1.2.6. Conclusions, as they apply to test Matrix Test SB18, are in Subsection 5.1.2.7.
~
The facility responses to the break are documented by the data plots, referenced as figures in the text, at the end of this section. The numbering and content of the data plots for Matrix Test SB18 are identical to the data plots provided in Subsection 5.1.1 for Matrix Test SB01. For example, the data plot for instrument channel LDP-601 for Matrix Test SB01 is shown in Figure 5.1.15; the data plot for the same channel for Matrix Test SB18 is shown in Figure 5.1.2 5. Not all of the figures in the data plot package at the end of the section are referenced. Only those figures required to explain a different response from that of Matrix Test SB01 are referred to in the text. The additional figures are provided for the benefit of readers who wish to compare additional Matrix Test SB18 parameters to similar parameters for Matrix Test SB01. A data plot with the suffix x indicates extended time. 5.1.2.1 System Configuration and Initial Conditions Matrix Test SB18 was performed per an approved written procedure, met all but two of the specified initial conditions, and all actions were automatic with no operator responses required. The two conditions that werent met are discussed below. The test was performed without the operation of nonsafety-related systems. The normal residual heat removal system (RNS) and chemical and volume control system (CVS) pumps did not operate during the test. h LJ o:\l5hRevl\l5h.13.non:Ib-081798 5.1.2-1 REVISION 1
FINAL DATA REPORT A flow nozzle simulating one line of flow was installed in the ADS 41 line-HL-1 to ADS 4-1 separator-to provide the single failure simulation, and a flow nozzle simulating two lines of flow was installed in the ADS 4 2 line-HL-2 to the ADS 4 2 separator. Additionally, flow nozzles simulating two lines of flow each were installed in the ADS 13 inlet lines. He reactor heater control decay algorithm maintained the maximum reactor heater power output for [ ]* and then power began to decay to simulate the total decay energy input of the AP600 nuclear fuel (Appendix F). This test was performed with reactor heater rod HTR-C2-317 electrically disconnected to simulate the heater conditions during the performance of Matrix Test SB01. He differences in facility configuration for the two tests were:
- After the performance of Matrix Test SB01, a vacuum breaker was installed on the ADS 1-3 sparger line inside the IRWST to eliminate negative pressures in the pressurizer and ADS 1-3 separator.
- FMM-201, FMM-202, FMM-203, and FMM-204, which measured cold-leg flow, were removed from the system and replaced with pipe spools because of continuous failures. RCP differential pressure instrument data (DP-202, DP-203, DP-205, and DP-206) could be used to monitor pump degradation.
- Component and piping insulation had deteriorated significantly because of the amount of maintenance (such as gasket replacement) that was required during the test program.
- Pressurizer heater logic was changed so that the PLC initiated a signal to open the pressurizer heater SCR contactor at [ ]'6* after S signal actuation, thereby ensuring de-energization of the heaters.
l l l
- CMT balance line isolation valves RCS-529 and RCS-530 were in the AUTO and OPEN l
positions, as part of the initial conditions for Matrix Test SB01, but were closed and opened by the operator 1 minute after the TEST pushbutton was pressed in Matrix Test SB18 to ! prevent heatup at the top of the CMTs prior to break valve opening. The facility fill and vent, startup, and heatup were performed per the same approved operating procedures used for Matrix Test SB01. A zero check was performed for all differential pressure instmments as was done for the reference test. Initial conditions for the test were established and recorded in the procedure. Refer to Subsection 2.7 for pre test operations. The test ran for about 6 hours. I 1 l Table 5.1.2-1 shows the initial conditions recorded from the operator's panel and the average of the same parameters for about 2 minutes prior to the break valve opening from the DAS. o:\l 536w Rev i\l 536w- 13.non: l tro81798 5.1.2-2 REVISION 1
- . . - - - _ - . _ _ . - - - _ - . _ ~ - _ . - - . , . . . . . _ - - _~ -
1 FMAL DATA REPORT q There were two initial condition parameters out of specification, neither of which should invalidate this V test.
- ACC-1 pressure, indicated by PT-401, was [ ]'6' or [ ]'6' percent below the required pressure band. The accumulator was pressurized to the required pressure, as indicated on local pressure indicator PI-401, prior to test actuation. The loss of pressure between tank pressurization and test actuation was possibly due to nitrogen gas cooling in the accumulator.
Test analysis starting with the recorded lower accumulator overpressure should still be possible.
- ACC-2 pressure, indicated by N-402, was [ ]'6' or [ ]'6' percent below the required )
pressure band. The accumulator was pressurized to the required pressure, as indicated on local
)
pressure indicator PI-402, prior to test actuation. The loss of pressure between tank pressurization and test actuation was possibly due to nitrogen gas cooling in the accumulator. Test analysis starting with the recorded lower accumulator overpressure should still be possible. 5.1.2.2 Inoperable Instruments Table 5.1.2 2 is a list of instruments considered inoperable or invalid during all or portions of this test. Some of the instmments listed are on the Critical Instmment List (Subsection 3.2, Table 3.2 2) and, therefore, are addressed here: 1 FMM-201, FMM-202, FMM-203, and FMM 204 measured flow (gpm) in each of the four cold legs. A decision was made to continue testing without the availability of these instruments. Replacement flow meters repeatedly failed; their continued use was precluded due to cracking of the ceramic liners from thermal stratification in the loop piping. The necessary boundary conditions for loop flow could be determined from cold-leg differential pressure transmitters DP-202, DP-203, DP-205, and DP-206. CMT-1 and CMT-2 injection flow meters FMM-501 and FMM-504 and PRHR inlet and outlet flow l - meters FMM-802 and FMM 804 provided accurate data when sensing liquid, but became inaccurate when sensing two-phase or steam flow. L FMM-905 measured break separator loop seal flow to the primary sump. As the transient proceeded, the primary sump and break separator levels exceeded the elevation of the break at the bottom of CL-3. When this occurred, break flow initially stopped and then reversed. Flow reversal through the break occurred at about [ ]'6' rendering subsequent data invalid. SG tube level data (LDP-215, LDP-218, LDP-219, and LDP-222) were biased by vaporization of the water in the transmitter reference leg after the SG tubes started draining. However, the data provide ( accurate indication of the time when the tubes are empty, c:\l 5 Mw Rev i\l 5%w.13.non: l b.081798 5.1.2-3 REVISION 1
FINA1, DATA REPORT TF-103 and TF-104 measured CL-3 and CL-4 bottom of-pipe fluid temperatures entering the reactor vessel. Both thermocouples were n: moved to accommodate installation of thermal stratification measurement instrumentation it was permissible for both thermocouples to be inoperable because TF-101 and TF-102, which measured the CL-3 reactor flange top and CL-4 reactor flange top, were operable during the performance of Matrix Test SB18. TF-501 and TF-504 measured CMT fluid temperature from the long thermocouple rod location near the bottom of each CMT. The thermocouples appear to have measured ambient conditions throughout the test, which would indicate a short somewhere in the thermocouple wiring. With these thermocouples inoperable, the required long thermocouple rod thermocouple availability of "seven out of ten and no more than one in succession failed" was met. Data provided by ADS-4 separator instmmentation prior to the ADS 4-1 and ADS 4-2 valves opening at [ ] were invalid due to the closed position of the ADS-4 valves and the ADS-4 separator loop seal valves. The instruments affected are: FMM-602, FMM-603, FVM-602, FVM-603, LDP-611, and LDP-612. Test analysis will not be affected, since ADS-4 flow did not begin until the valves opened. Considering these critical instmment failures, sufficient instrumentation was available to allow the performance of mass balances as demonstrated in Subsection 5.1.2.6 and Appendix E. An energy balance will be perfonned and reported in the AP600 Low-Pressure Integral Systems Test at Oregon State University Test Analysis Report, WCAP-14292.* 5.1.2.3 Sequence of Events Table 5.1.2-3 contains the sequence of events for Matrix Tests SB18 and SB01. The first pages of the l table provide selected event times from both tests and the difference between event times for the l reference SB18 to SB01. The subsequent pages of the table provide a visual representation of the time comparison using bar charts. On both the numeric table and the bar charts, the events are sorted in the chronoiogical order in which they occurred in Matrix Test SB18. He table defines the source of the actual time values. A D in the Data Source column indicates the recorded time was obtained from a software program that monitored digital events in the facility. Rese events included pump starts and stops, valve limit switch actuations, and alarms. The term valve opening means the valve has actuated and the closed limit switch is being opened (valve coming off the seat). An A in the Data Source column indicates the time data were obtained by reviewing test data recorded by the DAS. Although the test data from the DAS were in digital format, the DAS monitored analog events such as pressure, flow, and temperature. O oA1536w Re v i\l 536w .13.non: l t>.081798 5.1.2-4 REVISION 1
FINAL, DATA Raroar 3 5.1.2.4 Test Resuhs and Evaluation l His section compares the results of reference Matrix Test SB18 with the results of test SB01. In doing so, the overall system response to the LOCA event in Matrix Test SB18 is evaluated. The section is divided into three different phases, each characterized by the systems' behavior and thermal-hydraulic phenomena occurring in the systems. He phases are as follows:
- Initial Depressurization Phase: simulated break initiation to ADS-1 actuation ADS Phase: ADS-1 actuation to start of IRWST injection IRWST Injection Phase: start of IRWST injection to end of test initial Depressurization Phase As with Matrix Test SB01, this test began with the actuation of the TEST pushbutton. Break valve TS-205 received an open signal from the PLC [ ]'6" later (time zero). After an additional 0.5 second, an S signal was generated by the PLC, which time-sequences signals to initiate various events such as resetting controllers, stopping pumps, and repositioning valves.
The initial depressurization phase for Matrix Test SB18 began similarly to Matrix Test SB01. PLC timing for various event initiations was within I second of the same event for Matrix Test SB01. At Od about [ ] steam percent, as calculated from LDP-127 data, indicated that the reactor vessel began to lose inventory, i.e., steam formation began (Figure 5.1.2-3), which was the same as in Matrix Test SB01. CMT transition from recirculation to draindown occurred within [ ] of the transition in Matrix Test SB01, as indicated by decreasing levels on LDP-509 and LDP-510 (Figure 5.1.2-6). In Matrix Test SB18, CMT-1 transitioned [ ]'b' earlier and CMT-2 transitioned [ ]'6* later than the same events in Matrix Test SB01. De pressurizer and pressurizer surge line emptied within seconds for the two tests (Figure 5.1.2-5). De core decay heat simulation of reactor heater power followed the programmed algorithm just as it did in Matrix Test SB01 (Appendix F). In Matrix Test SB18, it took slightly longer for the SG-2 U tubes to completely empty, but the SG-1 U-tubes and SG 1 and SG-2 channel heads and hot legs drained slightly faster than in reference test SB01 (Figures 5.1.2-7 through 5.1.2.11). There was essentially no difference in the PRHR HX response during this phase of the test (Figures 5.1.2-41, 5.1.2-66, and 5.1.2-67). In Matrix Test SB18, the fluid level inside the reactor vessel core barrel reached its minimum collapsed level of [ ] at about [ ]'6# as indicated by LDP-127 (Figure 5.1.2-15). o:\l5hRevi\l5h-13.non:Ib-081798 5.1.2-5 REVISION 1
FINAL DATA REPORT De minimum level occurred about [ ]'6' later and about 1 in, higher than in Matrix Test SB01. Matrix Test SB18 also had a low core barrel level of [ ]'6* during the depressurization phase prior to the ADS-1 valve opening. Heater rod cooling was maintained (Figure 5.1.2-44). Also in Matrix Test SB18, during a condensation /depressurization event that occurred at about [
]'6' the indicated core barrel level transitioned from [ ]'6' and back to [ ]'6' over a [ ]'6' period. Heater rod cooling was mainta ined.
The initial RCS depressurization was similar for the two tests from break valve opening until [
]'6' when the ADS-1 valve opened in Matrix Test SB18 and caused pressure to decrease at a greater rate than for Matrix Test SB01. In Matrix Test SB01, the ADS-1 valve did not open until
[ ]'b' The period of quasi-equilibrium pressure between the RCS and the secondary side of the SGs was about the same for Matrix Test SB18 and Matrix Test SB01, about [ ]' 6' (Figure 5.1.2-45). In the first few seconds after the break valve opened, the SGs were isolated on both the feedwater and steam sides in order to minimize RCS heat losses. He only cooling available to the SGs was heat losses to the ambient. With RCS pressure decreasing due to the break and SG pressure increasing due to heat absorption from the RCS, the pressures converged at about [ ]'6' At about [ ] when inventory losses through the break caused k.CS pressure to drop below the pressure on the SG secondary side, the SGs became a heat source for the RCS. ADS Phase CMT-1 reached its low level setpoint at [ ]'6' thus triggering the ADS 1 valve opening at about [ ]'6' in Matrix Test SB18. This can be compared with the CMT-1 low level setpoint being reached at [ ]'6' and the ADS-1 valve opening at [ ]'6' for Matrix Test SB01 (Figure 5.1.2-6). The RCS pressure response following the ADS-1 valve opening was the same for both tests. Facility response to the opening of the ADS-2, ADS-3, and ADS-4 valves was quite similar to that of Matrix Test SB01. CMT injection flow data appear to be consistent for the two tests (Figure 5.1.2-16). In Matrix Test SBl8, the CMT-2 low level setpoint was not reached until about [ ]'6' after h CMT-1 low level setpoint was reached. In Matrix Test SB01, the two CMTs attained their low level setpoints within [ ]*6' of each other. A possible explanation for the delay for CMT-2 in Matrix Test SB18 is the timing of initiation of accumulator injection, which began for both accumulators about [ ]'6' after the ADS-1 valve opened. Accumulator injection created a backpressure in the DVI header that partially or completely shut off CMT injection by closing the CMT injection line check valves. In Matrix Test SB18, accumulator injection began about [ ]'6' earlier than in Matrix Test SB01 (Figure 5.1.2-16). The earlier accumulator injection in Matrix Test SB18 can be directly attributed to the ADS-1 valve opening earlier. The accumulator flow profiles and interaction wie the - CMTs were consistent for both tests (Figure 5.1.2-16). o A15 %w Rev i\l 5 %w.13.non: l b-081798 5.1.2-6 PEVISION 1
FmAL DATA RuonT Early in Matrix Test SBl8, a condensation /depressurization event occurred at about [ ]'6' V as compared with [ ]'b' in Matrix Test SB01. DP-114 and DP-130, which measured reactor vessel upper head differential pressure, exhibited large negative spikes, indicative of instantaneous high steam flow from the reactor vessel upper plenum through the upper head and into the downcomer area (Figure 5.1.2-19). At the same time, LDP-127 showed a sharp spike decrease in ) collapsed level indication, and LDP-116 and LDP-140 indicated sharp spike increases (Figure 5.1.215).
'Ihe parameters of RCS pressure, reactor vessel levels, and DVI rates were almost identical at the time l of the event in both tests (Figures 5.1.2-45, 5.1.2-15, and 5.1.2 17).
Test data show that the collapse of the superheated steam bubble in the upper portion of the reactor vessel downcomer annulus resulted in the downcomer fluid accelerating upward and impacting the bottom of the core barrel flange where the core bypass holes are located. The impact of the downcomer liquid on the solid surface of the core barrel flange produced the " bang" heard during the test. The low pressure created in the upper downcomer annulus by the collapse of the steam bubble also resulted in a rapid increase in steam flow from the core barrel, through the upper head, and into the downcomer. Refer to Subsection 7.1 for a more detailed treatment of condensation / depressurization events at the OSU test facility. Pressurizer and surge line response during the ADS phase of the test was similar for Matrix Test SB18 l l and SB01, with the exception of a slight offset in timing due to the earlier opening of the ADS-1 valve in Matrix Test SB18 (Figure 5.1.2-5). IRWST injetion started within [ ]'6* for the two tests at about [ ]'6' (Figure 5.1.2 37). IRWST Inlection Phase During the performance of Matrix Test SB01, the pressurizer reflooded to about [ ]'6' from about [ ]'6' This phenomenon was attributed to a negative pressure being formed in the pressurizer and ADS l-3 separator when HL-2 filled with subcooled fluid to a level that started to fill the surge line while the ADS 1-3 sparger was still submerged in the IRWST. A vacuum breaker was installed on the ADS 1-3 sparger line inside the IRWST following the performance of Matrix Test SB01 to prevent a recurrence of a reflood caused by negative pressure. The vacuum breaker performed its function, and under similar conditions, a pressurizer reflood did not occur during Matrix Test SBl8 (Figure 5.1.2-5). Both CMTs reflooded during the second hour of Matrix Test SB18 as the CMTs did in Matrix Test SB01 (Figures 5.1.2-30 through 5.1.2-34). The reflood occurred about [ ]'6' earlier, and the draindown was completed about [ ]'6' earlier in Matrix Test SB18 than in the reference test. In Matrix Test SB18, the CMTs were completely empty about the time that sump injection started through the primary sump injection line check valves, but about [ ]'6' prior to the sump injection valves opening. In Matrix Test SB01, the CMTs were completely empty at o A15%w Rev i\l 5%w-13.non:l b-081798 5.1.2 7 REVISION 1
FmAL DATA REPORT about the time that the primary sump injection valves opened. The drain rate for the CMTs was faster in Matrix Test SB18, even though the IRWST drain rate for both tests was consistent. Condensation /depressurization events were observed during the CMT redood during Matrix Test SB18 similar to those during Matrix Test SB01. Primary sump injection started at about [ ]'6' earlier in Matrix Test SB18 than in Matrix Test SB01. DAS rack 1 failed and stopped acquiring data at [
]'6' which eliminated all pressure, differential, and level transmitter data from that time to the end of the test. The test engineer recorded data in the Test Log from the local transmitters at 15 minute intervals for the remainder of the test for injection flows, IRWST level, and auctioneered reactor heater sheath temperature. The recorded data appears in Table 5.1.2-4.
Primary sump injection valves opened at [ ]'6' and the test was stopped 30 minutes later. The facility response during sump injection recirculation during Matrix Test SB18 appears to be consistent with the data obtained during Matrix Test SB01. 5.1.2.5 Comparison of Component Responses Reactor Reactor response during the performance of Matrix Test SB18 was similar to reactor response during Matrix Test SB01, with the exception of event times. The lowest collapsed level recorded in the LDP-127 test data was [ ]'6' at [ ]'6' which was about [ ]'6' highet and [ ]'6' later than Matrix Test SB01 (Figure 5.1.2-15). During a condensation / depressurization event that occurred at about [ ]'6' the collapsed level transitioned from [ ]'6' and back to [ ]'6' over a [ *6' period. Reactor heater rod cooling was maintained during both level excursions (Figure 5.1.2-44). Core Makeup Tanks CMT response during the performance of Matrix Test SB18 was similar to CMT response in Matrix Test SB01, with the exception of event times. In Matrix Test SB18, both CMTs emptied about [ ]'6' into the test, which was very consistent with CMT performance in the reference test (Figure 5.1.2-6). CMT reflood during the second hour of both matrix tests was consistent because CMT-2 was the first to reflood in each test (Figures 5.1.2-30 and 5.1.2-31). The levels attained in the CMTs during reflood were within [ ]'6' from one test to the other. Also in both tests, the rapid reflood of the CMTs was the result of a condensation /depressurization event occurring in the CMTs when subcooled fluid filled the balance lines and spilled into the tanks filled with superheated steam. In both tests, the final draindown of the CMTs was controlled by IRWST draindown due to interaction of the CMT and IRWST levels on the CMT discharge line check valves. o:\l 536wRev nl 536w.13.non: l t481798 5.1.2-8 REVISION l
l FINAt. DATA Rzec::t Accumulators Accumulator response during the performance of Matrix Test SB18 was similar to accumulator response in Matrix Test SB01, with the exception of event timing. Pressurizer De pressurizer response during Matrix Test SB18 produced the only major differences between the tests. During the performance of Matrix Test SB01, the pressurizer exhibited a second reflood at about [ l'b' but a similar event did not occur during Matrix Test SB18 (Figure 5.1.2-5). As was described in Subsection 5.1.1.4 in the Pressurizer Response, the reflood was caused by negative pressure developing in the pressurizer and ADS 1-3 separator with respect to the pressure in the reactor vessel. Negative pressure was precipitated by several occurrences: 1) the ADS 1-3 sparger located in the IRWST was submerged,2) the hot legs had refilled with subcooled fluid into the surge line, and 3) the combination of I and 2 resulted in condensation of steam in the pressurizer and ADS 1-3 separator with subsequent cooling. The problem of negative pressure developing in the pressurizer and ADS 1-3 separator was corrected prior to the performance of Matrix Test SB18 by a facility modification that consisted of a vacuum breaker being installed on the ADS 1-3 sparger line inside the IRWST. There was one operational difference between the two tests. During Matrix Test SB01, the pressurizei heaters remained energized at about 1.5 kW Bis was corrected prior to the performance of Matrix Test SB18 by a change in heater control logic. During Matrix Test SB01, the logic was supposed to drive the heater power demand to zero, resulting in zero power from the pressurizer heater SCRs to the heaters. According to the data for Matrix Test SB01, it can be postulated that the SCR control circuitry had not been properly tuned. The pressurizer heater control problem was corrected prior to the performance of Matrix Test SB18 by a change in the logic which opened the heater SCR contractor to de-energize the heaters at the start of the test (Figure 5.1.2-24). All other pressurizer responses during Matrix Test SB18 were similar to the responses during Matrix Test SB01, with the exception of event timing. Passive Residual Heat Removal Heat Exchanner De PRHR HX response during Matrix Test SB18 was similar to Matrix Test SB01, with the exception of event timing and the response of the HX wide-range level data after about [ l'6" During Matrix Test SB01, the HX partially refilled when the RCS loops were filled with subcooled fluid. Following the refill, when small pressum and level oscillations began to occur, it is possible that the oscillations caused the PRHR HX inlet line to " burp," allowing a negative pressure in the HX to equalize v,ish the RCS. Equalization of pressures allowed equalization of PRHR levels with those of O the RCS. Daring Matrix Test SB01, once the levels equalized, the PRHR level remained essentially constant. o A15 h Rev l\l 536w.13.non: l b-081798 5.1.29 REVISION 1
FINA1 DATA REFORT Data obtained during the performance of Matrix Test SB18 indicate that PRHR HX performance was consistent with the performance during Matrix Test SB01 until the final draindown of the HX. During Matrix Test SB18, the HX began to drain, but the data then indicated the drain stopped and the HX began a slow refill for as long as data were obtained (Figure 5.1.2-68). It is possible that the HX was refilling, but a more logical possibility is that wide-range level transmitter LDP-802 was slowly losing its reference leg due to a low saturated pressure in the HX tubes. Steam Generators SG response during the performance of Matrix Test SB18 was similar to SG response in Matrix Test SB01, with the exception of event timing. Cold Lens and Hot Lees The RCS cold-leg and hot-leg response during the performance of Matrix Test SB18 was similar to the response in Matrix Test SB01, with the exception of event timing. In-Containment Refueline Water Storare Tank IRWST response during the performance of Matrix Test SB18 was similar to IRWST response during the reference test, except that during Matrix Test SB18, it took about [ ] longer to reach the low-low level setpoint, which opened the primary sump injection valves. The tank draindown rates were identical until about [ ]'** when the available DAS data ended for Matrix Test SB18 (Figure 5.1.2-35). At about [ ]'** in Matrix Test SBl8 the time that primary sump injection began through the check valves in the sump injection lines in both tests, it is possible that sump injection flow caused a reduction in IRWST injection rate and, therefore, the extended draindown time (Table 5.1.2-4). l Break and ADS System Response The BAMS response during the performance of Matrix Test SB18 was similar to the response in Matrix Test SB01, with the exception of event timing. 5.1.2.6 Mass Balance l The mass balance results for Matrix Test SB18 test data were calculated based on water inventory before and after the test and are provided in Appendix E. The mass at the end of the test was within [ ]'6' percent of the mass at the beginning of test as compared to [ ]'b' percent for Matrix Test SB01. l 9 l oa t 536w Rev i\l 536w.13.non: l b.081798 5.1.2-10 REVISION 1 l
FINAL DATA REPORT l i N 5.1.2.7 Conclusions i The test was performed with minimal problems and is considered acceptable. Although not all of the l l facility initial conditions met the specified acceptance criteria. the deviations did not impact the quality l of the data. The instmmentation problems encountered were not critical to the performance of the ; l facility mass and energy balances. i, Facility response to the test was as anticipated for the conditions that were established. The data l clearly demonstrate that cooling of the reactor heater rods was maintained throughout the duration of the test. i l l 1
.p I
l 1 i l I i e
- O c:\l5hRevl\l5h-13.non:Ib-081798 5.1.2-11 REVISION 1
FINAL DATA REP (Gr TABLE 5.1.21 MATRIX TEST SB18 INITIAL CONDITIONS Instrument Specified Initial Actual Initial Parameter No. Condition Condition Comments Pressurizer pressure'" IrT-604 370 2 2 psig HI, teny - + ' re'" SC-141 420 2'F HL-2 temperature'" SC-140 420 2*i-SG-1 pressure'" PT-301 285 2 5 psig SG-2 pressure'" Irr-302 285 5 psig Pressurizer level'" LDP-601 65 2 5 in. Level signal temperature-compensated by TF-605 SG-1 narrow-range LDP-303 2613 in. Level signal temperature-level'" compensated by TF-301 i SG-2 narrow-range LDP-3M 26 x 3 in. Level signal temperature-level (" compensated by TF-310 IRWST temperature") TF-709 < 80 F CMT- temperature") TF-529 < 80"F i CMT-2 temperature"> TF-532 < 80*F ACC-1 temperature"' TF-403 < 80*F ACC-2 temperature"' TF-4M < 80*F IRWST level"' LDP-701 Level established by fill-line elevation ACC-1 levelr2> LDP-401 Level established by standpipe at 37 in. ACC-2 level ' LDP-402 Level established by standpipe at 37 in. ACC-1 pressure"' IrT401 232 2 2 psig [ ]'6' below required pressure ACC-2 pressure") PT-402 232 2 psig [ J below required pressure O o A15c,ow Rev i\l 536w- 13.rion: l b481798 5.1.2-12 REVISION 1
FINAL DATA REPORT s TABLE 5.I.21 (Continued)
's MATRIX TEST SBI8 INITIAL CONDITIONS Instrument Specified Initial Actual Initial Parameter No. Condition Condition Comments
'.DP-307 Full de fMT-1 level
- l CMT-2 level
- DP-502 Full Ho_tsi (1) Data for the indicated parameter were recorded in the test procedure as an initial condition for the test. The value was determined by the test engineer from the appropriate control board indicator.
(2) 1. sta were not recorded in the procedure, but the test engineer ve>ified that the specified conditions were acefeved while establishing initial conditions. The value of the parameter was determined post-test by j caledating the average DAS indication for a time of about 2 minutes before the break valve opened. I l b O l O V oA15h Rev l \ l 536w.13.nort: l l>081798 $ ].2-13 REVISION l
FINA1. DATA REPORT TABLE 5.1.2 2 MATRIX TEST SB18 INOPERABLE INSTRU'.fENTS/ INVALID DATA CHANNELS 4 Instrument No. Instn ment Type Description of Problem DP-905 Differential pressure transmitter Data questionable throughout test FDP-605 Differential pressure transmitter flow Over-ranged early in test FMM-20l* M ;netic flow meter Removed from system FMM-202' FMM-203* : FMM-2M
- FMM 50l* Magnetic flow meter Data invalid between [
]"' and after [
]'b' when the CMT is empty FMM-502 Magnetic flow meter Data invalid after [ ]'6' due to possible steam in balance line FMM-503 Magnetic flow meter Data invalid after [ ]'6" due to possible steam in balance line FMM-SM* Magnetic flow meter Data invalid between [
] and after
[ ] when the CMT is empty FMM-802* Magnetic flow meter Data invalid after steam forms in PRHR HX inlet line, which appears to be at about [ ] FMM-804* Magnetic flow meter Data valid until PRHR HX initially drained at [ ]'"' after this time the possibility of steam in the outlet line invalidates the data FMM-905
- Magnetic flow meter Negative values after break separator level exceeds the break clevation at about [ l'"' are invalid HFM-103 Heat flux meter Inoperable throughout test HFM-105 HFM-111 HFM-201 HFM-505 HFM-510 HFM-703 HFM-801 HPS-509-1 through 3 Heated phase switch 9
OA15hRevlu5h-13satib-081798 5.1.2 14 REVISION 1
.- _ .~ . .. .- - . . . . . - . . . . ._
FINA1. DATA REM.tT l r
-( , TABLE 5.1.2-2 (Continued)
V MATRIX TEST SBI8 INOPERABLE INSTRUMENTS / INVALID DATA CHA.NNELS l l Instrument No. Instrument Type Description of Problem l LDP-201 Differential pressure transmitter - level Data invalid due to effect of vertical LDP-202 portion of sense line attached to top of LDP-203 pipe; data can show level trends and LDP-204 when pip empties or starts to draia. LDP-205 but absolute level indication cannot be LDP-206 used l , LDP-215* Differential pressure transmitter - level Inoperable - when tube voids reference LDP-216 leg steams off(Subsection 2.4) LDP-217 LDP-218* LDP-?l9* LDP-220 LDP-221 LDP-222* l LDP-801 Differential pressure cell - level Inoperable -indicates full when HX not full TF-103* Thermocouple fluid temperature Replaced with thermal stratification thermocouples ("%i TF- 104* .'t Thermocouple fluid temperature Replaced with thermal stratification d thermocouples TF-170 Thermoccuple fluid temperature Affected by leakage from reactor vessel downcomer TF-50l
- Thermocouple fluid temperature Inoperable - indicates ambient t throughout test TF-504
- Thermocouple fluid temperature Inoperable - indicates ambk.at l throughout test l TF-702 . Thermocouple fluid temperature Inoperable - indicates ambient throrghout test
.l TFM-103 Thermocouple wall temperature Inoperable throughout test l TFM-105 l TFM-Ill
! TFM-703 f- TW-503 Thermocouple wall temperature Inoperable throughout test i i N_gg:
- Instruments marked with an asterisk are critical instruments. See Subsection 5.1.2.2 for discus ion.
L V j oA15hRevl\l5h.13mn:lt481798 5.1.2-15 REVISION 1
FINAL DATA REPORT Table 5.1.2-3 on pages 5.1.216,17, and 18 is not included in this nonproprietary document. O 1 I 1 I O I oAINRevlu5h-13.non:b482798 5.1.2-16 REVISION 1
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FINAL DATA REPORT The bar charts for Table 5.1.2-3 on pages 5.1.2-19 through 5.1.2 24 are not included in this
- j. nonptoprietary document.
i t l l l I' l l i l l l t i E-5 E 5 1 t I T i. t ..
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FINAL DATA Rzroar Table 5.1.2-4 on page 5.1.2 25 is not included in this nonproprietary document. { [ l l I l l~ l l l t ( .' I r I i i i f 1 4 l i~ l i j oA15hRevi\f5h.13.non:lWm 5.1.2-25 REVISION I
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Figure 5.1.21 Primary Loop and Break Piping Layout s O oA15hRevi\l5h-13.non:Ib 081798 5.1.2-25 REVISION 1
FINAI, DATA REPORT O J'5L (DAV O d b
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/-- TS-201 03 !
PLUG min b1W ( [~ RV227 s CL-3 15 dhi I?'"b,____ s (
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O~ Figure 5.1.2-2 Primary Loop and Break Pipe Arrangement oA1536wRev i\1536w- 13.non: I b481798 5.1.2-27 REVISION l
FINAt. DATA REPORT Figures 5.1.2 3 through 5.1.2-80 on pages 5.1.2 28 through 5.1.2-106 (except for Figure 5.1.2-61 on page 5.1.2-86) are not included in titis nonproprietary document. l O l l l 1 REVISION 1 oA1536wRevi\lSMw-13.non:ltM1898 5.1.2-28
em C, U pd l OSU TEST: U0018 h ADS 1-3 LIQUID AND STEAM FLOYS h Volumetric Flow Rate (gpm) l A - FMM-601 123 0 0 ADS 1-3 Loop Seal Flw 3 Volumetric Flow Eate (cfm) l B FVM-601 126 0 0 ADSt-3 Sep Stm Flow _5 _ 50 q .. . . . . . 1400 g -
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FINAt. DATA REPORT l l 5.1.3 Effect of Backpressure (Matrix Test SB19 Comparison with Matrix Test SB01 and SB18) This section identifies and describes the response of the facility based on a comparison between Matrix l l Test SB01 (OSU Test U0001), SB18 (OSU Test U0018) and Matrix Test SB19 (OSU Test U0019). l The simulated break for Matrix Test SB19 was located on the bottom of CL-3 with a simulated failure of one of the ADS-4 lines. CL-3 is on the CMT side of the facility (Figures 5.1.3-1 and 5.1.3-2). Matrix Test SB19 was identical to Matrix Test SB01, except that in Matrix Test SB19, containment backpressure was simulated by automatic control of BAMS header pressure. The purpose of performing Matrix Test SB19 was to gain a comparison of how the facility responded to a small-break LOCA, with the same physical configuration, with and without containment backpressure simulation. The transient began when break valve TS-205 opened and continued through ADS actuation; CMT, ; accumulator, and IRWST injection; and primary sump recirculation injection. Matrix Test SB19 was performed on July 14,1994. Test duration was about 6 hours. The performance of this test was considered successful because reactor vessel heater rod cooling was maintained. Subsection 5.1.3.1 provides details re!ated to facility configuration and initial conditions for the test. A description of inoperable instruments is provided in Subsection 5.1.3.2, and Subsection 5.1.3.3 references the sequence of events. A summary of the overall system response and any component responses that differed from the reference test SB01 are described in Subsections 5.1.3.4 and 5.1.3.5, respectively. A summary of the mass balance results is provided in Subsection 5.1.3.6. Conclusions, as they apply to Matrix Test SB19, are in Subsection 5.1.3.7. Facility responses to the break are documented by the data-plot package at the end of this section. The numbering and content of the data plots for Matrix Test SB19 are identical to the data plots provided in Subsection 5.1.1 for Matrix Test SB01, with the exception that some ranges and time l samples have been changed to allow for differences between tests. For example, the data plot of instmment channel LDP-601 for Matrix Test SB01 appears in Figure 5.1.1-5. The data plot for the same channel in Matrix Test SB19 appears in Figure 5.1.3-5. The discussion of Matrix Test SB19 does not refer to all of the figures in the dat. glot package at the end of the section. Only figures required to explain a different response from that of Matrix Test SB01 are referred to in the text. A data plot with the suffix x indicates extended time. 5.1.3.1 System Configuration and Initial Conditions The test was performed per an approved written procedure, met the specified initial conditions with three exceptions described in this subsection, and required no operator action. Marix Test SB19 simulated a 2-in. cold-leg break LOCA with long-term cooling and no nonsafety systems operating. The RNS and CVS pumps did not operate during this test. Containment o:U5hRevi\l5h-15.non:Ib-082598 5.1.3-1 REVISION 1
FINAL DATA REPORT backpressure was simulated by automatic control of BAMS header prenure. A tlow nozzle simulating one line of flow was installed in the ADS 4-1 line (HL-1 to the ADS 4-1 separator) to provide the single-failure simuladon, and a flow nozzle simulating two lines of flow was installed in the ADS 4-2 line (HL-2 to the ADS 4-2 separator). Flow nozzles simulating two lines of flow each were installed in the ADS 1-3 inlet lines. The reactor heater control decay algorithm maintained the maximum reactor heater power output for 140 seconds. Reactor heater power then began to decay to simulate the total decay energy input of the AP600 nuclear fuel (Appendix F). Differences between the facility configuration for the two tests were:
- After the performance of Matrix Test SB01, a vacuum breaker was installed on the ADS 1-3 sparger line inside the IRWST to eliminate negative pressures in the pressurizer and ADS 1-3 separator.
- Cold-leg flow instruments FMM-201, FMM-202, FMM-203, and FMM-204 were removed from the system and replaced with pipe spools due to continuous failures (RCP differential pressure instruments DP-203, DP-202, DP-205, and DP-206 were used to monitor for punip degradation).
- CMT balance line isolation valves RCS-529 and RCS-530 were in the AUTO and OPEN positions as initial conditions for Matrix Test SB01. For Matrix Test SB19, the valves were closed and opened by the operator i minute after pressing the TEST pushbutton to prevent heatup at the top of the CMTs prior to the break valve opening.
A[ ]" vent line with an in-line check valve was installed from the top of the primary sump to the BAMS header downstream of valve CSS-906. The [ ]'6'vm h a b setvice for this test which allowed primary sump steam flow to be monitored by break separator [ ]'6" steam flow transmitter FVM-906. He facility fill and v:nt, startup, and heatup were performed per the same approved operating procedures as were used for Matrix Test SB01. A zero check was performed for all differential pressure instruments. Initial conditions for the test were established. Refer to Subsection 2.7 for pre-test operations. He test engineer failed to record the control board indications prior to this test, but all controllers had been in the AUTO position and controlling within the specified band for about 20 minutes prior to test actuation. The initial conditions from DAS, averaged over about 2 minutes prior to the break valve opening, are presented in Table 51.3-1. l The following three initial condition parameters, none of which invalidate this test, were out of specification:
- HL-1 temperature (SC-141) was [ ]'b' percent below the required temperature band. This was within the instrumentation system accuracy requirements.
oA15hRevi\l5W 15.non:lt>O81798 5.1.3-2 REVISION I
FINAt. DATA REPORT q = ACC-1 pressure was [ ]'6" percent below the required pressure band. The V accumulator was pressurized to the required pressure, as indicated on local pressure indicator PI-401, prior to test actuation. The loss of pressure between tank pressurization and test actuation was possibly due to cooling of the nitrogen gas in the accumulator. Test analysis staning with the recorded lower accumulator overpressure is still possible.
- ACC-2 pressure was [ ]'6# percent below the required pressure band. The accumulator was pressurized to the required pressure, as indicated on local pressure indicator PJ-402, prior to test actuation. The loss of pressure between tank pressurization and test actuation was possibly due to cooling of the nitrogen gas in the accumulator. Test analysis starting with the recorded lower accumulator overpressure is still possible.
5.1.3.2 Inoperable Instruments Table 5.1.3-2 is a list of instruments considered inoperable during all or portions of this test. Some of the instmments listed in the table are on the Critical Instmmentation List (Subsection 3.2, Table 3.2-2) and, therefore, are addressed here. FMM-201, FMM-202, FMM-203, and FMM 204 measured flow (gpm) in each of the four cold legs. A decision was made to continue testing without the availability of these instruments. Replacement flow meters repeatedly failed; their continued use was precluded due to cracking of the ceramic liners from thermal stratification in the loop piping. The necessary boundary conditions for loop flow could be determined from cold-leg differential pressure transmitters DP-202, DP-203, DP-205, and DP-206. CMT-1 and CMT-2 injection flow meters FMM-501 and FMM-504 and passive residual heat removal (PRHR) inlet and outlet flow meters FMM-802 and FMM-804 provided accurate data when sensing liquid, but became inaccurate when sensing two-phase or steam flow. FMM-701 measured IRWST-1 injection flow. When the primary sump valves were opened, the flow meter indicated a negative flow as water flowed from the primary sump to the IRWST. The meter was not designed to measure reverse flow, so this measurement was invalid: however, total IRWST flow was measured by FMM 702. FMM-905 measured break separator loop seal flow to the primary sump. As the transient proceeded, the primary sump and break separator levels exceeded the elevation of the break at the bottom of CL-3. When this occurred, break flow initially stopped and then reversed. Flow reversal through the break occurred at about [ ]'6# rendering subsequent data invalid. SG tube level data (LDP-215, LDP-218, LDP-219, and LDP-222) were biased by vaporization of the water in the transmitter reference leg after the SG tubes started draining. However, the data provide O V accurate indication of the time when the tubes were empty. oA15hRevlu5h 15.non:Ib o81798 1.1.3 3 REVISION 1
FINA1. DATA REPORT N-201 measured RCS pressure at the top of the SG-1 long tube. On August 15,1994, it was discovered that the transmitter had an incorrect zero compensation which resulted in a negative error and negative data at low pressures. He transmitter zero was corrected at that time. N-201 data obtained during Matrix Test SBly had the zero correction performed on the transmitter. and the corrected data appears as N_.201. Negative data and corrected negative data can be used to determine trends but is considered inaccurate. PT_201 data is considered unreliable for values less than 1.1 psig, but a sufficient amount of other pressure data are available. Data provided by ADS-4 separator instrumentation prior to the ADS 4-1 and ADS 4-2 valves opening at 1133 seconds were invalid due to the closed position of the ADS-4 valves and the ADS-4 separator loop seal valves. The instruments affected are: FMM-602, FMM-603, FVM-602, FVM-603, LDP-611, and LDP-612. Test analysis will not be affected, since ADS-4 flow did not begin until the valves opened. TF-501 and TF-504 measured CMT fluid temperature from the long thermocouple rod location near the bottom of each CMT. The thermocouples appeared to have measured ambient conditions throughout the test, indicating a short in the thermocouple wiring. With these thermocouples inoperable, the required long thermocouple rod availability "seven out of ten and no more than one in succession failed" was met. Considering these critical instrument failures, sufficient instrumentation was available to allow the performance of mass balances as demonstrated in Subsection 5.1.3.6 and Appendix E. An energy balance will be performed and reported in the AP600 Low-Pressure Integral Systems Test at Oregon State University Test Analysis Report, WCAP-14292.* 5.1.3.3 Sequence of Events l Table 5.1.3-3 provides the sequence of events for Matrix Test SB19 and compares them with the l reference tests SB01 and SB18. The following discussion compares the performance of SB19 with l SB01; differences between SB01 and SB18 are discussed in Subsection 5.1.2. The performance of l SB01 is affected during ADS 1-3 since it was conducted without a vacuum breaker in the sparger. l SB18 was conducted with the vacuum break installed but is missing pressure, differential pressure and l level test data after 14,312 seconds. The first pages of Table 5.1.3-3 provide the time of occurrence l for selected events in the tests and the difference in time of event occurrence with SB01 and SB18. l De subsequent pages of the table provide a visual representation of the time comparison by use of a l bar chart. On both the numeric table and the bar chart, events are sorted in chronological order for l Matrix Test SB19. The first pages of Table 5.1.3-3 indicate the source of the actual time data. A D in the Data Source column indicates that the recorded time was obtained from a software program that monitored digital events in the facility, including pump starts and stops. valve limit switch actuations, and alarms. An A in the Data Source column indicates that the time data were obtained by reviewing test data recorded oAl5%wRevluS%w 15 non:ltso81798 5.1.3-4 REVISION 1
- , n . . - .. . . - - - - .----_----- . . - . - - .
FINAL DATA REPORT by the DAS. Although test data from the DAS were in a digital format, the DAS monitored analog V events such as pressure, flow, and temperature. l 5.1.3.4 Test Results and Evaluation his section will evaluate the (sverall system response to the LOCA event in Matrix Test SB19 and will be subdivided into three detferent phases. De different event phases, each of which are characterized by the system s behavior and thermal-hydraulic phenomena occurring in the systems are as follows:
- Initial Depressurization Phase: simulated break initiation to ADS-1 actuation I ADS Phase: ADS-1 actuation to start of IRWST injection
- IRWST Injection Phase: stant of IRWST injection to end of test Initial Depressurization Phase l
l As with the reference test, Matrix Test SB19 began when the TEST pushbutton was pressed. Break ! valve TS-205 received an open signal from the programmable logic controller (PLC) 120 seconds later l (time zero on the data plots). All time references in this section are with respect to time zero. At 0.5 second, an S signal was generated by the PLC. ! PLC timing for various similar event initiations were within a second for Matrix Tests SB01 and SB19. CLDP-127 indicated that the reactor vessel began to lose inventory, i.e. steam formation began, I at about [ ]'** (Figure 5.1.3-3), about [ ]'6' later than when steam formation occurred in reference test SB01. l In Matrix Test SB19, the pressurizer and pressurizer surge line emptied at [ ]'b'
. respectively (Figure 5.1.3-5), about [ ]'b' earlier for the pressurizer and [ ]
[ earlier for the pressurizer surge line than in Matrix Test SB01. ne difference in time can be L attributed to higher break flow rates early in Matrix Test SB19 (Figure 5.1.3-62). L i The early emptying of the pn:ssurizer and pressurizer surge line in Matrix Test SB19 did not result in earlier CMT transition from recirculation to draindown (Figure 5.1.3-6). CMT-1 trans;%n occurred at about [ ]'6' in Matrix Test SB19, compared with [ ] in Matrix Test SB01. It j appears CMT-2 started to transition at about 100 seconds in Matrix Test SB19 but the balance line l . refilled. CMT-2 transition actually occurred at about 270 seconds in Matrix Test SB19, compared
. with [ ]'6# in Matrix Test SB01.
i ! ne algorithm for core decay heat simulation of reactor power used during Matrix Test SB19 was the same algorithm used during reference test SB01. ne core decay heat simulation of reactor heater
- power followed the programmed algorithm just as it did in Matrix Test SB01 (Appendix F).
.A15maalush.15.aon:lb481798 5.1.3 5 REVISION 1 4
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FmAL DATA REPO2r l During Matrix Test SB19, the SG U-tubes emptied from between [ ]'6' earlier than the U-tubes during Matrix Test SB01 (Figures 5.1.3-7 and 5.1.3-8). The SG channel heads and hot legs drained earlier during Matrix Test SB19 than during Matrix Test SB01, with the exception of HL-2 which drained about [ ] later during Matrix Test SB19 (Figures 5.1.3-9, 5.1.3-10, and 5.1.3-11). The difference in drain time for the SG U-tubes, channel heads, and hot legs can be attributed to the higher break flow rates early in Matrix Test SB19 (Figure 5.1.3-62). There was essentially no difference in the PRHR HX response between Matrix Tests SB01 and SB19 during this phase of the test (Figures 5.1.3-41, 5.1.3-66, and 5.1.3-67). During Matrix Test SB19, the fluid level inside the reactor vessel core barrel reached its minimum collapsed level of [ ]'6' at about [ ]'6' per CLDP-127 (Figure 5.1.3-15), about [ ]'6' later and about [ j'6* lower than during Matrix Test SB01. A minimum core barrel level of 60 in, was reached during the depressurization phase. Also in Matrix Test SB19, during a condensation /depressurization event that occurred at about [ ]*6' the collapsed level transitioned from [ ]'6' then back to [ ]'6' Nod. Heater-rod cooling was maintained during both instances (Figure 5.1.3-44). Initial RCS depressurization was similar for both tests from break valve opening until [ ]' 6" when the ADS-1 valve opened in Matrix Test SB19 causing pressure to decrease at a greater rate than for Matrix Test SB01 where the ADS-1 valve did not open until [ ]'6' The period of quasi-equilibrium pressure between the RCS and the secondary side of the SGs was about [ ]'b' (Figure 5.1.3-45) for both tests. In the first few seconds after the break valve opened the SGs were isolated on both the feedwater and steam sides to minimize RCS heat losses. The only cooling available to the SGs was heat losses to ambient conditions. With RCS pressure decreasing due to the break and SG pressure increasing due to heat absorption from the RCS, the pressures converged at about [ ]'6' At about [ l'6' when inventory losses through the break caused RCS pressure to drop below the pressure on the SG secondary side, the SGs became a heat source for the RCS. ADS Phase CMT-1 reached its low level setpoint at [ ]'6' triggering the ADS-1 valve opening at about [ ]'** in Matrix Test SB19, as compared with [ l*6' for the CMT-1 low level setpoint to be reached and [ ]*6# for the ADS-1 valve to open in reference test SB01 (Figure 5.1.3-6). The RCS pressure response following ADS-1 opening was similar for both tests (Figure 5.1.3-45); facility response to the opening of ADS-2, ADS-3, and ADS-4 valves was also similar. CMT injection flow data appeared consistent between the two tests (Figure 5.1.3-16). In Matrix Test SB19, the CMT-2 low level setpoint was not reached until about [ ]*6' after NT-1 reded its setpoint. In Matrix Test SB01, the two CMTs attained their low level setpoints whhin o:uSMwRevNN 15.non:Ib-o81798 5.1.3-6 REVISION 1 % w t _ _ ___ . .
_ . - - . . _ - __ .-~ .. . . . - - -. . . . - - . FINAL DATA REPCOT l 1 l i' [ ]'6' of each other and prior to accumulator injection beginning. A possible explanation for ! y' the delay of setpoint achievement in CMT-2 during Matrix Test SB19 is the timing of initiation of l l accumulator injection which started for both accumulators about [ ] before ADS-1 ) opened. Accumulator injection created backpressure in the DVI header that partially or completely shut off CMT injection by closing CMT injection line check valves; the accumulators must necessarily have a higher overpressure than the RCS and the CMTs in order to inject. Accumulator injection during Matrix Test SB19 began about [ ]'6' before the ADS 1 valve opened, about [ ]'6' earlier than in Matrix Test SB01 (Figure 5.1.3-16). He earlier accumulator injection in Matrix Test SB19 was attributed to the faster depressurization rate following the quasi-equilibrium pressure plateau between primary and secondary pressures (Figure 5.1.3-45). The faster depressurization rate in Matrix Test SB19 can be attributed to higher break flow rates over ! the first [ ]'6' (Figure 5.1.3-28). The accumulator flow profiles and interaction with the CMTs was consistent for both tests (Figure 5.1.3-16). Early in Matrix Test SB19, condensation /depressurization events occurred at about 545 and 615 seconds, as compared with [ ]*6# during reference test SB01. DP-114 and DP-130 exhibited large negative spikes indicative of instantaneous high steam flow from the reactor vessel upper plenum through the upper head and into the downcomer area (Figure 5.1.3-19). At the same time, CLDP-127 showed a sharp-spike decrease in level indication, and CLDP-ll6 and CLDP-140 T indicated sharp-spike increases (Figure 5.1.3-15). The parameters of RCS pressure, reactor vessel L levels, and DVI rates were almost identical at the time of the event in both tests (Figures 5.1.3-45, 5.1.3-15, and 5.1.3-17). A loud " bang" was heard in the BAMS steam discharge header and recorded in the test log at [ ]'6# into the test. A condensation /depressurization eveat in the BAMS steam discharge header may have been the precursor for the event at [ ]'6' Test data revealed that the collapse of the superheated steam bubble in the upper portion of the reactor l vessel downcomer annulus resulted in the downcomer fluid accelerating upward and impacting the bottom of the core barrel flange where the core bypass holes are located. The low pressure created in the upper downcomer annulus by the collapse of the steam bubble also resulted in a rapid increase of l steam flow from the core barrel, through the upper head, and into the downcomer. Subsection 7.1 provides a more detailed treatment of condensation /depressurization events. The pressurizer and pressurizer surge line reflooded during blowdown through the ADS 1-3 valves
- (Figure 5.1.3-5) during both tests. During Matrix Test SB19, the pressurizer re-emptied about j [ ] earlier than during Matrix Test SB01, probably the result of faster RCS depressurization and the vacuum breaker installed on the ADS 1-3 sparger line inside the IRWST (Figure 5.1.3-45).
The PRHR HX inlet head level transmitter recorded HX level changes during Matrix Test SB19 but
, did not respond during Matrix Test SB01 (Figures 5.1.3-66 and 5.1.3-67). During Matrix Test SB01, lb the HX appeared to reach a minimum level of [ ]'*' at about [ ]'6' refill to [ ]'6' a
o.ushRevN5b-15 non:lt4Pl798 5.1.3-7 REVISION 1
FINAL DATA REFoRT at about [ ] and then return to a minimum again at [ ]'** In Matrix Test SB19, the HX level was predominantly greater than [ ]'6' during the ADS phase and appeared to maintain more flow. One possible cause for what appeared to be a more efficient PRHR HX response during Matrix Test SB19 is that RCS levels remained higher than the levels during Matrix Test SB01 (Figure 5.1.315). IRWST injection began at [ ]'6" about [ ]'6" earlier during Matrix Test SB19 than during Matrix Test SB01 (Figure 5.1.3-37). Since the start of IRWST injection is dependent on the differential pressure between the reactor vessel and the IRWST, and the force for injection is gravity, the early injection during Matrix Test SB19 was probably the result of the faster RCS depressurization and the increased pressure in the IRWST due to containment pressure simulation. IRWST Inlection Phase The early start of IRWST injection at 1030 seconds during Matrix Test SB19 had two major effects on the facility's overall response. First, CMT injection was stopped by the IRWST injection head prior to the CMTs emptying completely, while in Matrix Test SB01 the CMTs emptied before IRWST injection started (Figures 5.1.3-6 and 5.1.3-16). Second, with no delay between CMT and IRWST injection, the cold legs refilled and subcooled about [ ]'** earlier than during Matrix Test SB01 (Figures 5.1.3-42, 5.1.3-43, 5.1.3-53, and 5.1.3-54). Because the cold legs maintained some level in the early part of Matrix Test SB19, flow was maintained through the break; in Matrix Test SB01, when the cold legs emptied, break flow stopped between about [ ]'** (Figure 5.1.3 28). Both CMTs reflooded during both tests (Figures 5.1.3-6 and 5.1.3-30 through 5.1.3-34); however, the timing of CMT reflood and CMT response during reflood were different. CMT-1 reflood occurred e' about [ ]'6' in Matrix Test SB19, [ ]'** earlier than in Matrix Test SB01. CMT-2 reflood occurred at about [ ]*** earlier than in Matrix Test SB01. The earlier reflood of the CMTs in Matrix Test SB19 was directly attributed to the earlier reflood of the cold legs. Reflood of the CMTs occurred for the same reasons described in reference test SB01 (Subsection 5.1.1). In Matrix Test SB19, the CMT was at higher temperatures than in Matrix Test SB01 when the balance line refilled. When the cooler fluid in the balance line reached a long horizontal section of pipe at about [ ]'** the fluid flashed causing a spike increase in CMT pressure. The pressure spike resulted in a rapid reduction of level in the balance line and a spike discharge from the CMT into the DVI. The CMT pressure then began to decrease again but from a higher pressure. The pressure spike phenomena occurred once in CMT-1 and four times in CMT-2. This increasing pressure spike phenomena did not occur in the reference test SB01 data. Condensation /depressurization events were observed during the CMT reflood during SB19 like occurred during Matrix Test SB01. , I 5.1.3-8 REVISION I oA15WRevN5h 15.non:lb-081798
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l l l~ FINAL DATA Rt.romT l f During reference test SB01, the pressurizer reflooded to about [ ]*6" over the period of about l j O- [ ]** This phenomena was attributed to negative pressure formed in the pressurizer and ADS 1-3 separator when HL-2 filled to the surge-line level with subcooled fluid while l the ADS 1-3 sparger was still submerged in the IRWST. A vacuum breaker was installed on the ADS l l-3 sparger line inside the IRWST following the performance of Matrix Test SB01 to prevent a l recurrence of a reflood caused by negative pressure. A pressurizer reflood did not occur during Matrix l Test SB19 (Figure 5.1.3-5). l Primary sump injection started at about [ ]"' in Matrix Test SB19, about [
]'6" later than in Matrix Test SB01 (Figure 5.1.3 37). During Matrix Test SB01, injection flow started through the primary sump injection line check valves when the fluid levels in the sump and the IRWST equalized, about [ ]"' before the sump injection valves opened. During Matrix Test SB19, injection flow did not start until almost [ J'6' after the primary sump injection valves opened. A possible explanation for the delay in sump injection during Matrix Test SB19 is that BAMS pressure increased from about [ ]'6' and caused a slight differential l pressure between the IRWST and the primary sump (Figure 5.1.3-74x).
l Sump injection recirculation continued for 2 hours. The facility response while in the sump injection recirculation mode during Matrix Test SB19 appeared consistent with the data obtained during Matrix Test SB01. f ! 5.1.3.5 Comparison of Component Responses i Reactor Reactor response was almost identical for Matrix Tests SB01 and SB19, with the exception of times. The lowest collapsed level recorded in the CLDP-127 test data for Matrix Test SB19 was [ ]"' at [ ]'6' later than in Matrix Test SB0)(Figure 5.1.3-15). During a condensation /depressurization event that occurred at about [ ]"' the indicated level transitioned from [ ]"' Reactor heater rod cooling was maintained during both level excursions (Figure 5.1.3-44). Core MakeuD Tanks The early emptying of the pressurizer and pressurizer surge line in Matrix Test did not result in earlier CMT transition from recirculation to draindown (Figure 5.1.3-6). CMT-1 transition occurred at about
]"' in Matrix Test SB19, compared with [
[ ]'6' in Matrix Test SB01. CMT-2 l' transition occurred at about [ ]*' in Matrix Test SB19, compared with [ ]"' in Matrix Test SB01. To initiate the CMT transition from recirculation to draindown, the fluid coupling
] between the cold legs and the balance lines must be broken. In both tests, CL-1 and CL-3 fluid temperatures (measured [ ]'6' from the top inside diameter of the reactor nozzle flange) were subcooled when the CMTs transitioned from recirculation to draindown (Figures 5.1.3-53 and oA15hneviusw-15.non: b-os 79s 5.1.3-9 REVISION I
FINA1. DATA REPORT 5.1.3-54). One indication that the cold legs have partially drained is that their top thermocouples will indicate superheat. A possible explanation for the transition delay in Matrix Test SB19 is that containment backpressure simulation delayed the start of cold leg draindown; i.e., the transition from single to two-phase now. Due to the location of SC-105 and SC-101 in the cold leg now stream, the broken fluid coupling between the cold legs and the balance lines would occur before the thennocouples would sense cold-leg draindown. Both CMTs reflooded during both tests (Figures 5.1.3-6 and 5.1.3-30 through 5.1.3-34); however, the timing of CMT reHood and CMT response during reflood were different. CMT-1 reHood occurred at about [ ]# in Matrix Test SB19, [ ]'6" earlier than in Matrix Test SB01. CMT-2 reflood occurred at about [ ]'6* earlier than in SB01. Reflood of the CMTs occurred for the same reasons described in reference test SB01 (Subsection 5.1.1). The reason for the earlier CMT renood in Matrix Test SB19 was that the cold legs refilled with subcooled fluid earlier than in Matrix Test SB01. The balance lines refilled due to the condensation of steam in 2 the CMT or the balance line and the subsequent reduction in CMT pressure with respect to RCS pressure. In Matrix Test SB19, the CMT was at higher temperatures than in Matrix Test SB01 when the balance line refilled. When the cooler Guid in the balance line reached a long horizontal section of pipe at about [ ]'** (representing a larger hot surface area than the vertical section of pipe), the fluid flashed causing a spike increase in CMT pressure. The pressure spike resulted in a rapid reduction of level in the balance line and a spike discharge from the CMT into the DVI. The CMT pressure would then began to decrease again but from a higher pressure. The pressure spike phenomena occurred once in CMT-1 and four times in CMT-2. In CMT-2 the first two events occurred at the [ ]'** level and the last two occurred at the next horizontal pipe run located at about the [ ]'6# level. This increasing pressure spike phenomena did not occur in the reference test SB01 data. In addition, in Matrix Test SB19, CMT-1 was the first to reflood and the first to drain completely; in Matrix Test SB01, CMT-2 was the first to renood and the first to drain completely. Condensation /depressurization events were observed during the CMT reflood during Matrix Test SB19 like occurred during Matrix Test SB01. CMT injection was suf6cient to maintain reactor heater rod cooling until IRWST injection began. Accumulators I Accumulator response was identical for Matrix Tests SB01 and SB19, with the exception of event timing. The accumulators increased or stabilized reactor vessel indicated fluid levels. I oA15hRevnl5W 15.non:ltro81798 5.1.3-10 REVISION 1 1
. . - _ _ _ - _ - - _ . _ - - - - - . - - - . _ . - _ - . - . - . . . . _ . ~ _ - ~
FINAL DATA RErorr Pressuriser in Matrix Test SB19, the pressurizer and pressurizer surge line emptied at [ ]"' respectively (Figure 5.1.3 5), about [ ]"' earlier for the pressurizer and [ ]' 6 ' earlier for the pressurizer surge line than in Matrix Test SB01. The difference in time can be attributed to higher break flow rates early in Matrix Test SB19 (Figure 5.1.3-62). The pressurizer refloodeu during Matrix Test SB01, but a similar event did not occur during Matrix Test SB19 (Figure 5.1.3 5). As described in Subsection 5.1.1.4, the reflood was caused by a negative pressure developing in the pressurizer and ADS 1-3 separator with respect to the pressure in the reactor vessel. The negative pressure was precipitated by several things: the ADS 1-3 sparger located in the IRWST was submerged; the hot legs had refilled with subcooled fluid into the surge line; and the combination of the previous items resulted in condensation of steam in the pressurizer and ADS 1-3 separator with subsequent cooling. The problem of a negative pressure being developed in the pressurizer and ADS 1-3 separator had been corrected prior to the performance of Matrix Test SB19 by a facility modification that installed a vacuum breaker on the ADS 1-3 sparger line inside the IRWST. The other difference between the two tests was that durirw ,latrix Test SB01, the pressurizer heaters
- remained energized at about 1.5 kW. This was correcta aor to performance of Matrix Test SB19 by a procedural change that required the test crew to open
- pressurizer heater breaker when the S signal was verified (Figure 5.1.3 24). During Matrix Test SB01 the logic was supposed to drive the heater power demand to zero resulting in zero power from the heater SCRs to the heaters. According to the data for Matrix Test SB01 it can be postulated that the SCR control circuitry had not been properly tuned.
All other pressurizer responses during Matrix Test SB19 were similar to the responses during Matrix Test SB01 with the exception of timing. Passive Residual Heat Removal Heat Exchanner There was a difference in the PRHR HX performance during the ADS phase of SB19 than during SB01 (Figures 5.1.3-66 and 5.1.3-67). The PRHR HX inlet head level transmitter recorded heat exchanger level changes during Matrix Test SB19 but did not record a level change during Matrix Test SB01. During Matrix Test SB01 the heat exchanger appeared to reach a minimum level of [ ]'** at about [ l and refill to [ ]'6' at about [ ]'*' and return to a minimum again at [ ]"' In Matrix Test SB19 the heat exchanger level was predominantly greater than [ ]** during this phase of the test and appeared to maintain more flow. One possible cause for what appears to be more efficient PRHR HX response during Matrix Test SB19 is that RCS levels
, [ remained higher than the levels did during Matrix Test SB01 (Figure 5.1.3-15).
4 oA15hRevhl5h.15.non:1t481798 5.1.3-11 REVISION 1
FmAL DATA RzrcQT The PRHR HX response during Matrix Test SB19 was similar to reference test SB01 except for the timing of events and the response of the HX wide-range level data after about [ ]'6" During Matrir Test SB01 the HX partially refilled during the period of time that the RCS loops were filled with subcooled fluid. Following the refill, when small pressure and level oscillations began to occur, it is possible that the oscillations caused the PRHR HX inlet line to " burp" allowing a negative pressure in the HX to equalize with the RCS. Equalization of pressures allowed equalization of PRHR levels with those of the RCS. During Matrix Test SB01, once the levels equalized the PRHR level remained essentially constant. Data obtained during the performance of Matrix Test SB19 indicate that PRHR performance was consistent with the performance during Matrix Test SB01 until the final draindown of the HX. During Matrix Test SB19 the HX began to drain but, the data then suggests that the drain stopped and the HX began a slow refill for as long as data was obtained (Figure 5.1.3-68). It is possible that the HX was l refilling, but another possibility is that the wide-range level instrument LDP-802 was slowly losing its reference leg due to vaporization from the lower saturated pressure in the HX. r j Steam Generators l I l SG response during the performance of Matrix Test SB19 was similar to their response during Matrix l Test SB01 with the exception of event timing. 1 I Cold Leos and Hot Lees O The RCS cold-leg response during Matrix Test SB19 was different than during Matrix Test SB01 in that the cold legs refilled earlier in the test (based on reactor vessel downcomer levels Figure 5.1.3-15). During Matrix Test SB01 the cold legs began to empty at about [ ]'** i and did not refill until about [ j'b' During Matrix Test SB19 the cold legs began to empty about [ J'** earlier but refilled at about [ ]'6' A possible explanation , for the early refill during Matrix Test SB19 is that IRWST injection started earlier and DVI flow l was greater than break flow (Figures 5.1.317 and 5.1.3-28). l The RCS hot legs response during the performance of Matrix Test SB19 was similar to their response during Matrix Test SB01 with the exception of event timing. In Containment Refueline Water Storane Tank IRWST response during the performance of SB19 was similar to IkWST response during the reference test (SB01) except that during Matrix Test SB19 the low-low level setpoint and opening the primary sump injection valves occurred about [ ]'b' earlier. The earlier opening of the sump injection valves during Matrix Test SB19 may have been the result of sump injection not occurring through the sump injection line check valves allowing IRWST injection to remain at a higher rate for longer, c:\l5hRevi\l5h.15.non:Ib-081798 5.1.3-12 REVISION I
FINAt. DATA REronT The IRWST injection was sufficient to recover reactor vessel levels and maintain the reactor heater Q rods cooled. Break and ADS M;;.=;...;r.t System BAMS response during the performance of Matrix Test SB19 was significantly different from the response during Matrix Test SB01 due to the containment back-pressure simulation used during Matrix Test SB19 (Figures 5.1.3 73, 5.1.3-74, 5.1.3-74x, r.ad 5.1.3-75). Following break valve opening, the BAMS header pressure tracked the projected algorithm curve thr containment backpressure simulation and reached a maximum pressure of about [ ]'6' The projected algorithm curve maximum pressure for the facility was about [ ]'6' The BAMS header pressure stayed at [ ]'6# until [ J'6" and then began to decrease. Actual pressure remained below projected pressure early in the test except for a brief period betiveen [ ]'6# The BAMS pressure control valve (CSS-901) was in the AUTO and CLOSED positions at the start of the test. De control valve closed when header pressure dropped benow the programmed value, then re-opened to control pressure at about 16,000 seconds. BAMS header pressure attained the projected pressure again at about [ ]'6' and maintained projected pressure for the remainder of the test. A possible explanation as to why the actual pressure did not niach the projected pressure and maintain it throughout the test is that the BAMS heat losses were greater than anticipated. 5.1.3.6 Mass Balance ( i ne mass balance results for Matrix Test SB19 test data were calculated based on water inventory before and after the test and are provided in Appendix E. De mass at the end of the test was within 0.5 percent of the mass at the beginning of the test r.s compared to 5.8 percent for Matrix Test SB01. 5.1.3.7 Conchasions De test was performed with minimal problems and is considered acceptable. Although not all of the facility initial conditions met the specified acceptance criteria, the deviations did not impact the quality of the data. De instrumentation problems encountered were not critical to the performance of the facility mass and energy balances. Facility response to the test was as anticipated for the conditions that were established. The data clearly demonstrate that cooling of the reactor heater rods was maintained throughout the duration of the test. I% oA15manmsw.is.non: bas 79s 5.1.3 13 REVISION 1
i i FtNAL DATA REPORT l l l TABLE 5.1.31 MATRIX TEST SB19 INITIAL CONDITIONS l l Specified Instrument Initial Actual Initial Parameter No. Condition Condition Comments ! Pressurizer pressure"' Irr-604 370 2 psig HL-l temperature"' SC-141 420 *F [ l l below required temperature HL-2 temperature"' SC 140 420 *F SG-1 pressure"' Irr 301 28515 psig l SG-2 pressure"' Irr-302 285 5 psig l Pressurizer level"> LDP-601 65
- 5 in. Uncompensated level (corrected for specific volume change)
SG-1 narrow- LDP-303 26 2 3 in. Uncompensated level range level"' (corrected for specific I volume change) SG-2 narrow-range LDP-304 26 3 in. Uncompensated level level"' (corrected for specific volume change) I IRWST temperature <23 TF-709 < 80*F l CMT-1 temperature'2> .IF-529 < 80 F j CMT-2 temperature"' TF-532 < 80*F l ACC-1 temperature <23 TF-403 < 80*F l ACC-2 temperature'2> TF-4G4 < 80*F IRWST level (2' LDP-701 Level l established by ; fill line I rievation ACC-1 level:2> LDP-401 Level established by standpipe at 37 in. i ACC-2 level <2) LDP-402 Level established by - - l I standpipe at 37 in. , 1 1 l O' oA15hRevi\l5h-15.non:Ib-081798 5.1.3-14 REVISION 1
FINAt, DATA REPORT l O TABLE 5.1.3-1 (Continued) b MATRIX TEST SB19 INITIAL CONDITIONS Specified Instrument Initial Actual Initial Parameter No. Condition Condition Comments ACC-1 pressure
- FT-401 232 2 2 psig
**e
[ ]'** below required pressure ACC-2 pressure
- FT-402 232 2 2 psig [ ]'$e FT-402 below required pressure CMT-1 level
- Full LDP-507 CMT-2 levelm Full LDP-502 _ _
N.9.iti (1) The test engineer failed to record the control board indications prior to this test, but all controllers had been in the AUTO position and controlling within the specified band for about 20 minutes prior to test actuation.
'Ihe initial conditions from DAS, averaged over about 2 minutes prior to the break valve opening, are recorded here.
(2) Data were not recorded in procedure, but the test engineer verified that specified conditions were achieved while establishing initial conditions. The value of the parameter was determined post-test by calculating the average DAS indication for a time of about 2 minutes before the break valve opened. () O I O o:\l536wRevi\l536w 15.non:!b-081798 5.1.3-15 REVISION l
FINA1, DATA REPORT TABLE 5.1.3 2 MATRIX TEST SB19 INOPERABLE INSTRUMENTS / INVALID DATA CHANNELS Instrument Instrument Type Description of Problem No. FMM 20l* Magnetic flow meter Removed from system (Subsection 5.1.3.2) l I FMM-202* FMM-203* FMM-204* FMM-50l
- Magnetic flow meter Data invalid after 13,053 seconds when CMT was empty FMM-502 Magnetic flow meter Data invalid after 270 seconds due to possible steam in balance line FMM 503 Magnetic flow meter Data invalid after 246 seconds due to possible steam in balance line FMM-504* Magnetic flow meter Data invalid after 13,249 seconds when CMT was empty FMM-70l
- Magnetic flow meter Negative values after break separator level exceeded break ,
elevation at about 12,450 seconds invalid 1 FMM-802* Magnetic flow meter Data invalid after steam formed in PRHR HX inlet line f l (about 130 seconds) , FMM-804* Magnetic flow meter Data valid until PRHR HX initially dreined at 1750 l seconds; after this time, the possibility of steam in the j l outlet line invalidated the data FMM-905 Magnetic flow meter Negative values invalid after break separator level exceeded the break elevation at about 13,400 seconds HFM-103 Heat flux meter inoperable throughout test HFM-105 HFM-505 HFM-510 HFM-703 HFM-801 HPS-203-1 Heated phase switch Inoperable throughout test through HPS-203-3 HPS-509-1 through HPS-509-3 LDP-201 Differential pressure transmitter - Data invalid due to efhet of vertical portion of sense line LDP-202 level attached to top of pipe; data can show level trends and LPD-203 when pipe empties or stans to drain, but absolute level LDP-2(M indication can not be used LDP-205 LDP-206 LDP-207 Differential pressure transmitter - Inoperable; ranged improperly; data can show level trends l LDP-208 level but absolute level indication can not be used LDP-209 LDP-215* Differential pressure transmitter - Inoperable; when tube voids, reference leg steams off; see LDP-216 level Subsection 2.4 LDP-217 LDP-218* LDP-219* LDP-220 LDP-221 LDP-222' o:\l536wRevi\1536w-15.non:lb-081798 5.1.3-16 REVISION !
FINAL DATA REPORT [] TABLE 5.1.3 2 (Continued) 'Q MATRIX TEST SB19 INOPERABLE INSTRUMENTS / INVALID DATA CHANNELS Instrument Description of Problem Instrument Type LDP-801 Differential pressure cell - level Inoperable; indicated full when HX not full IT_101 Pressure transmitter Data less than 6.1 psig invalid Fr_102 Pressure transmitter Data less than 6.2 psig invalid FT_.103 Pressure transmitter Data less than 6.2 psig invalid I'T_104 Pressure transmitter Data less than 6.4 psig invalid PT_108 Pressure transmitter Data less than 8.4 psig invalid PT_109 Pressure transmitter Data less than 6.3 psig invalid 3 PT_111 Pressure transmitter Data less than 6.0 psig invalid ' PT_112 Presrure transmitter Data less than 8.8 psig invalid IT_113 Pressure transmitter Data less than 6.4 psig invalid IT_20l
- Pressure transmitter Data less than 1.1 psig invalid FT_202 Pressure transmitter Data less than 5.9 psig invalid IT_203 Pressure transmitter Data questionable throughout test Fr_205 Pressure transmitter Data less than 6.1 psig invalid I IT-610* Pressure transmitter Over-ranged from 358 to 1330 seconds FT-611* Pressure transmitter Over-ranged from 340 to 1466 seconds ,
IT-701 Pressure transmitter Over-ranged from 344 to 1420 seconds I TF-170 Thermocouple fluid temperature Affected by leakage from reactor downcomer I (A1 TF-203 TF-50l
- Thermocouple fluid temperature Inoperable; indicated ambient throughout test TF-504
- TF-615 TF-619 TF-702 TFM-103 Thermocouple wall temperature Inoperable throughout test TFM-105 TFM-703 TH-317-1 Thermocouple heater rod Heater rod replaced with non instrumented rod TH-317-4 TW-503 Thermocouple wall temperature Inoperable throughout test TW-534 Thermocouple wall temperature Inoperable; indicated ambient throughout test TW-547 TW-552 Note:
- Instruments marked with an asterisk are critical instruments. See Subsection 5.1.3.2 for discussion.
/ b' o A1536w Rev i\l 536w- 15.non: I b-081798 5.1.3-17 REVISION 1
FINAL, DATA Rr.ront l l ) Table 5.1.3-3 on pages 5.1.318 through 5.13-25 is not included in this non-proprietary document. l l J i i 4 i i
)
i gI I i l 9 o:\l536wRevl\l536w 15.non:lb-082598 5.1.3-18 REVISION l
. . - , ~ . . . . ~ . - . - .. . . . . . . . . . - ~ . - . . . -
1 l-FINAL DATA REroRT g i a j J 1 em.mme a i t i 1 4 J 1 Il l l ::_ .. . g::p i s
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-F i --omrw F. 'w -- i 8-r + r + r i O Figure 5.1.3-1 Primary Loop and Break Piping Layout c:\l 536w Rev l\l 536w- 15.non:1 b-081798 5.1.3-26 REVISION 1
. - . _ - - _ . _ . _ _ . _ . . _ . - . - - . _ . =
l FINAL DATA REPORT l l l O ~ J'R
- 4 i (lef 1
i i
~
_) L o V em n TS-201 PT 203 p PLUG
... y se
.RV227 cs ,
CL-3 DP - 215 ,,, _f '
-QJ(
s,c ) ( RV237 e 4 -
! TS-205 l
O Figure 5.1.3 2 Primary Loop and Break Pipe Arrangement oA15?6w Rev lil 536w- 15.non: l W81798 5.1.3-27 REVISION 1
FINAL DATA Rzroar Figures 5.1.3 3 through 5.13-80 on pages 5.13 28 through 5.13106 are not included in this nonproprietary document. I l l i l l l l O O c:\l536wRevlil536w.15.non:Img 5.1 3-28 REVISION ]
FmAL DATA Rzroar n l 5.1.4 Effect of a Larger Break Size (Matrix Test SB21 Comparison with Matrix Test SB01 and Vl SB18) In this section, the test results from Matrix Test SB21 (OSU Test U0021) are compared with those of l Matrix Test SB01 (OSU Test U0001 and SB18 (OSU Test U0018). The simulated large-break LOCA (LBLOCA) was located on the top and the bottom of CL-3 with a simulated single failure being one of the ADS-4 lines. CL-3 is on the CMT side of the facility (Figures 5.1.4-1 and 5.1.4-2). The purpose of performing Matrix Test SB21 was to compare facility response to a LBLOCA, using the same physical configuration as Matrix Test SB01, with the exception of break size. Matrix Test SB21 was performed on August 3,1994. He transient began when break valves TS-0Al and TS 205 opened and continued through ADS actuation; CMT/aecumulator/IRWST injection, and primary sump recirculation injection. The performance of this test was considered successful because reactor heater rod cooling was maintained. Subsection 5.1.4.1 provides details related to the systems' configuration and initial conditions for the test. A description of inoperable instruments is provided in Subsection 5.1.4.2, and Subsection 5.1.4.3 references the sequence of events. A summary of the overall system response and any component responses that differed from the reference test are described in Subsections 5.1.4.4 and 5.1.4.5, respectively. A summary of the mass balance results is provided in Subsection 5.1.4.6. Conclusions, O as they apply to Matrix Test SB21, are in Subsection 5.1.4.7. V The facility responses to the break are documented by the data plots, referenced as figures in the text, at the end of this section. De numbering and content of the data plots for Matrix Test SB21 are identical to the data plots provided in Subsection 5.1.1 for Matrix Test SB01, with the exception that some ranges and time samples have been changed to allow for differences in the two tests. For example, the data plot for instmment channel LDP-601 for Matrix Test SB01 is shown in Figure 5.1.1-5; the data plot for the same channel in Matrix Test SB21 is shown in Figure 5.1.4-5. Not all of the figures in the data plot package at the end of the section are referenced. Only those figures required to explain a different response from that of Matrix Test SB01 are referred to in the text. He additional figures are provided for the benefit of readers who wish to compare additional Matnx Text SB21 parameters to the comparable parameters for Matrix Test SB01. A data plot with the sufD. x indicates extended time. 5.1.4.1 System Configuration and Initial Conditions ne test was performed per an approved written procedure and met the specified initial conditions with three exceptions that are discussed in this subsection. All actions were automatic, with the exception that operator response was required for closing CSS-902 and CSS-906 in the BAMS header when total steam flow decreasr.d to approximately [ ]* O Matrix Test SB21 simulated a double 4-in. cold-leg break LOCA with long-term cooling, without the operation of nonsafety systems. The normal residual heat removal system (RNS) and chemical and o:\l 5%wRev i\l 5 %w-21.non: I b-081798 5,},4.] REVISION 1
FINA1, DATA REPORT , volume control system (CVS) pumps did not operate during this test. Inserts simulating the 4-in. top and bottom cold-leg breaks were installed in the break spools located at the top and bottom of CL-3.
'Ihe top 4-in. break flow was initially directed vertically (up) and then horizontally to the break separator upper nozzle. The bottom 4-in. break flow we initially directed vertically (down) aad then horizontally to the break separator cold-leg nozzle.
A flow nozzle simulating one line of flow was installed in the ADS 4-1 line-HL-1 to ADS 4-1 separator-to provide the single failure simulation, and a flow nozzle simulating two lines of flow was installed in the ADS 4-2 line-HL-2 to the ADS 4-2 separator. Additionally, flow nozzles simulating two lines of flow each were installed in the ADS 1-3 inlet lines. The differences in facility configuration for the two tests were:
- After the performance of Matrix Test SB01, a vacuum breaker was installed on the ADS 1-3 sparger line inside the IRWST to eliminate negative pressures in the pressurizer and ADS 1-3 separator.
= FMM-201, FMM-202, FMM-203, and FM-204, which measured cold-leg flow, were removed from the system and replaced with pipe spools because of continuous failures. RCP differential pressure instrument data (DP-202, DP-203, DP-205, and DP-206) could be used to monitor pump degradation.
. CMT balance line isolation valves RCS-529 and RCS-530 were in the AUTO and OPEN positions, as part of the initial conditions for Matrix Test SB01, but were closed and opened by the operator 1 minute after the TEST pushbutton was pressed in Matrix Test SB21 to prevent heatup at the top of the CMTs prior to break valve opening.
- A 6-in. vent line with an in-line check valve was installed from the top of the primary sump to j the BAMS header downstream of CSS-906. The 6-in. vent line had a blank installed and was
! not in service for this test.
. TF-103 and TF-104 were removed from CL-3 and CL-4, respectively, and were replaced with thermocouple rods connected to an independent recording system to provide data for a cold-leg thermal stratification study.
- A new set of decay heat curve algorithms was programmed into the reactor controller to maintain reactor heater power at 600 kW for 300 seconds prior to initiating the power decay during Matrix Test SB21 (Appendix F). During Matrix Test SB01, power was held at 600 kW for 140 seconds prior to initiating the power decay.
= BAMS steam header valves CSS-901, CSS-902, CSS-905, and CSS-9% were all open at the beginning of Matrix Test SB21, and then CSS-902 and CSS-906 were closed by the operator oA1536w Rev lu S36w.21.non: I b-o81798 5.1.4-2 REVISION 1
. _ _ - . . _ .-___ _ _ - ... ~ . . _ . _ _ . .___. - ~ _ _ . . _ _ _ _ _ _ _ . _ _ . _ _ _ _
FINAL DATA REPORT
]'6' as measured by a combined indication from FVM-901 when steam flow was at [
y and FVM-902 (Appendix G, Dwg. OSU 600901). CSS-902 and CSS-906 were closed at the start of reference test SB01, ne facility fill and vent, startup, and heatup were performed per the same approved operating procedures used for Matrix Test SB01. A zero check was performed for all differential pressure instruments as was done for the reference test. Initial conditions for the test were established and recorded in the procedure. Refer to Subsection 2.7 for pre-test operation. De test ran for about 5-1/2 hours. Table 5.1.4-1 shows the initial conditions recorded from the operator's panel and the average of other initial condition parameters for approximately 2 minutes prior to the break valves opening from the DAS. There were three initial condition parameters out of specification, none of which should invalidate this
. test.
. CMT-2 temperature, indicated by TF-532, was [ ]'6' above the limiting maximum temperature of 80*F. TF-532 is located on the CMT-2 long thermocouple rod, about [ ]'6' down from the inside top of the CMT.
The next three thermocouples down on the long thermocouple rod, their dimensions from the inside top of the CMT, and their average temperature for about 2 minutes prior to break valve opening are listed in the following: Thermocouple Dimension Average temperature TF-548 [ ]'6' [ ]' 6' TF-530 [ ]'6' [ ]'6' TF-526 [ ]'6' [ ]'6' nese data suggest that a small percentage of CMT volume was at a temperature greater than [ ]'6' Test analysis using the CMT recorded temperature data, taking into account the temperature variations at the top of the CMT, should still be possible.
- ACC-1 pressure, indicated by PT-401, was [ l'"' or [ ]'6" below the required pressure band. De accumulator was pressuried to the required pressure, as indicated on local pressure indicator PI-401, prior to test actuation. De less of pressure between tank
- pressurization and test actuation was possibly due to nitrogen gas cooling in the accumulator.
Test analysis starting with the recorded lower accumulator overpressure should still be possible.
. o:uswaevntsw.2tmn:ibasi?98 5.1.4-3 REVISION 1
FINAI. DATA REronT
= ACC-2 pressure, indicated by PT-402, was [ ]'6" or [ ]'6' below the required pressure band. The accumulator was pressurized to the required pressure, as indicated on local pressure indicator PI-402, prior to test actuation. The loss of pressure between tank pressurization and test actuation was possibly due to nit ogen gas cooling in the accumulator.
Test analysis starting with the recorded lower accumulator overpressure should still be possible. 5.1,4.2 Inoperable Instruments Table 5.1.4-2 contains a list of instruments considered inoperable or invalid during all or portions of this test. Some of the instruments listed are on the Critical Instrument List (Subsection 3.2, Table 3.2-2) and, therefore, are addressed here. FMM-201, FMM-202, FMM-203, and FMM-204 measured flow (gpm) in each of the four cold legs. A decision was made to continue testing without the availability of these instruments. Replacement flow meters repeatedly failed; their continued use was precluded due to cracking of the ceramic liners from thermal stratification in the loop piping. The necessary boundary conditions for loop flow could l be determined from cold-leg differential pressure transmitters DP-202, DP-203, DP-205, and DP-206. FMM-401 measured ACC-1 injection flow and provided accurate data when sensing liquid, but became erratic when filled with nitrogen following accumulator injection. ACC-1 level channel LDP-401 may be used as a backup for FMM-401. CMT-1 and CMT-2 injection flow meters FMM-501 and FMM-504 and PRHR inlet and outlet flow meters FMM-802 and FMM-804 provided accurate data when sensing liquid, but became inaccurate when sensing two-phase or steam flow. 1 FMM-701 measured IRWST-1 injection flow when the primary sump valves were opened. The flow
. meter indicated a negative flow as water flowed from the primary sump to the IRWST. The meter i was not designed to measure reverse flow, so this measurement was invalid. However, total IRWST l flow was measured by FMM-702.
1 1 FMM-905 measured break separator loop seal flow to the primary sump. As the transient proceeded, the primary sump and break separator levels exceeded the elevation of the break at the bottom of CL-3. When this occurred, break flow initially stopped and then reversed. Flow reversal through the break occurred at about [ ]'6# rendering subsequent data invalid. SG tube level data (LDP-215, LDP-218, LDP-219, and LDP-222) were biased by vaporization of the water in the transmitter reference leg after the SG tubes started dnining. However, the data provide accurate indication of the time when the tubes are empty. 4 o:\l5hRevl\l536w-21.non:Ib o81798 5.1.4 4 REVISION 1
. - - . - - .. - _ ~ _ _ . . - - _ - . _ - . _ - - . - - . - --
FINAL DATA Raroar PT-201 measured RCS pressure at the top of the SG-1 long tube. On August l$,1994, it was ( discovered that the transmitter had an incorrect zero compensation, which resulted in a negative error , and negative data at low pressures. The transmitter zero was corrected at that time. PT-201 data obtained during Matrix Test SB21 had the zero correction performed, and the corrected data appear as PT 201. Negative data and coTected negative data can be used to determine trends, but are considered inaccurate. PT_201 was not considered reliable for values less than 1.1 psig, but a sufficient amount of other pressure data are available. TF-103 and TF-104 measured CL-3 and CL-4 bottom of-pipe fluid temperatures entering the reactor , vessel. Both thermocouples were removed to accommodate installation of thermal stratification measurement instrumentation. It was permissible for both thermocouples to be inoperable because TF-101 and TF-102, which measured the CL-3 reactor flange top and CL-4 reactor flange top, were operable during the performance of Matrix Test SB21. l TF-501 and TF-504 measured CMT fluid temperature from the long thermocouple rod location near the bottom of each CMT. The thermocouples appear to have measured ambient conditions throughout l the test, which would indicate a short somewhere in the thermocouple wiring. With these thermocouples inoperable, the required long thermocouple rod thermocouple availability of "seven out of ten and no more than one in succession failed" was met. Data provided by ADS-4 separator instrumentation prior to the ADS 41 and ADS 4-2 valves opening i at[ ]* were invalid due to the closed position of the ADS-4 valves and the ADS-4 separator loop seal valves. The instruments affected are: FMM-602, FMM-603, FVM-602, FVM-603, LDP-611, and LDP-612. Test analysis will not be affected, since ADS-4 flow did not begin until the valves opened. Considering the critical instmment failures listed, sufficient instrumentation was available to allow the performance of mass balances as demonstrated in Subsection 5.1.4.6 and Appendix E. An energy , balance will be performed and reported in the AP600 Low-Pressure Integral Systems Test at Oregon State University, Test Analysis Report, WCAP-14292.* 5.1.4.3 Sequence of Events l Table 5.1.4-3 provides the sequence of events for Matrix Test SB21 and compares them with the l reference tests SB01 and SB18. The following discussion compares the performance of SB21 with l SB01; differences between SB01 and SB18 are discussed in Subsection 5.1.2. The performance of l l SB01 is affected during ADS 1-3 since it was conducted without a vacuum breaker in the sparger. l l SB18 was conducted with the vacuum break installed but is missing pressure, differential pressure and l level test data after 14,312 seconds. The first pages of Table 5.1.4-3 provide the time of occurrence , l for selected events in the tests and the difference in time of event occurrence with SB01 and SB18. [ l 'Ihe subsequent pages of the table provide a visual representation of the time comparison by use of a i oatshnevitism.21.non:ib-osi79s 5.1.4-5 REVISION 1
FINAI, DATA REPORT l bar chart. On both the numeric table and the ban chart, events are sorted in chronological order for l Matrix Test SB21. The table defines the source of actual time values. A D in the Data Source column indicates the recorded time was obtained from a software program tlat monitored digital events in the facility. These included pump starts and stops, valve limit switch actuations, and alarms. The term valve opening means the valve has actuated and the closed limit switch is being opened (valve coming off the seat). A 4 A in the Data Source column indicates the time data were obtained by reviewing test data recorded by the DAS. Although the test data from the DAS were in digital format, the DAS monitored analog events such as pressure, flow, and temperature from the data. 5.1.4.4 Test Results and Evaluation This section cc:npares the results of Matrix Test SB21 with the results of reference test SB01. In doing so, the overall system response to the LBLOCA event in Matrix Test SB21 is evaluated. The section is divided into three different phases, each characterized by the systems' behavior and thermal-hydraulic phenomena occurring in the systems. The phases are as follows:
. Initial Depressurization Phase: simulated break initiation to ADS-1 actuation
- ADS Phase: ADS-1 actuation to start of IRWST injection
- IRWST Injection Phase: start of IRWST injection to end of test Initial Depressurization Phase As with Matrix Test SB01, this test began with the actuation of the TEST pushbutton. Break valves TS-201 and TS-205 received an open signal from the PLC [ ]'6' later (time zero). After an additional 0.5 second, an S signal was generated by the PLC, which time-sequences signals to initiate various events such as resetting controllers, stopping pumps, and repositioning valves.
! The initial depressurization phase for Matrix Test SB21 began similarly to Matrix Test SB01. PLC timing for various event initiations was within I second of the same event for Matrix Test SB01, with the exception of the main feed pump and RCP trips. The main feed pump tripped [ ]'6'later I and the RCPs tripped [ ]'6" earlier in Matrix Test SB21 than in Matrix Test SB01. At about [ ]'6" steam percent, as calculated from LDP-127 data, indicated that the reactor vessel began to lose inventory, i.e., steam formation began (Figure 5.1.4-3). This was about [ ]'6' earlier than when steam formation occurred in Matrix Test SB01. . l In Matrix Test SB21, the pressurizer and pressurizer surge line became empty at [
]'6" respectively (Figure 5.1.4-5). This occurred about [ ]*6' earlier for the pressurizer and [ ]'** earlier for the surge line than in Matrix Test SB01. The time oA1536wRevluS36w-21.non:lb 081798 REVISION 1 l 5.1.4 6
FINAt. DATA REPORT p difference can be attributed to the fact that the larger break resulted in break flow rates being about
\ double early in Matrix Test SB21 compared with Matrix Test SB01 (Figurt 5.1.4-62).
De early emptying of the pressurizer and surge line resulted in earlier CMT transition from recirculation mode to draindown, i.e., the time that the CMT balance line level began to decrease (Figure 5.1.4-6). CMT-1 transition occuned at about [ ]'b' ir Matrh Test SB21, as compared with [ ]'6' in Matrix Test SB01. CMT-2 transition eccurred at about [
'] in Matrix Test SB21, as compared with [ ]'6' in Mat:5 Tet 5B01.
The core decay heat simulation of reactor heater power followed the programmed algorithm just as it did in reference test SB01 (Appendix F). I The SG U-tubes were completely empty anywhere between [ ]'6" in Matrix Test SB21, which was about [ ]'6# earlier than the U-tubes emptied in Matrix Test SB01 (Figures 5.1.4-7 and 5.1.4-8). The SG cold-leg channel heads drained earlier in Matrix Test SB21 by
. as much as [ ]'6' for SG-1 and [ ]'6" for SG-2 than in Matrix Test SB01 (Figures 5.1.4-9 and 5.1.4-10). De CL-3 channel head was the first to empty at [ ]'b' and the CL-2 channel head the last at [ ]'6' in Matrix Test SB21. De difference in drain time for the SG U-tubes and channel heads can be attributed to the larger break, resulting in higher break flow rates early in Matrix Test SB21 (Figure 5.1.4-62).
Both accumulators began to inject into the DVI nozzles at [ ]'6' as compared with about [ - ]'6' for Matrix Test SB01 (Figure 5.1.4-16). The combination of accumulator and CMT injection was sufficient to compensate for the loss of RCS inventory through the break and maintain reactor vessel levels essentially constant after about [ ]'S' (Figure 5.1.4-15). LDP-127 indicated a core barrel level of about [ ]'6" and the downcomer level transmitters, LDP-ll6 and LDP-140, indicated a level of about [ ]'6# There was a marked increase in both steam and liquid flows from the break separator at about [' ]'6' during Matrix Test SB21 (Figure 5.1.4-62). The flow increase coincided with the operator's closing CSS-902 and CSS-906 when total steam flow leaving the BAMS header decreased to approximately [ ]'6' The reason for the steam flow increase recorded by FVM-905 is that closing CSS-906 isolated the 8-in. vent line from the break separator, allowing flow only through the 6-in. vent line. Closing CSS-902 and CSS-906 also caused about a [ ]'6" increase in the break
' separator pressure, with essentially no change in primary sump pressure, resulting in a pressure differential that increased liquid flow from the break separator (Figure 5.1.4-74).
The PRHR HX response during this phase was similar for both tests (Figures 5.1.4-41,5.1.4-66, and 5.1.4-67). For Matrix Test SB21, fluid level inside the reactor vessel core barrel reached its minimum collapsed level of [ ]'6' at about [ ]'b' as indicated by LDP-127 data (Figure 5.1.4-15). De l oAlswneviush-21mno.081798 5.1,4-7 REVISION 1
FtNAL DATA REPORT minimum level occurred about [ ]'6" earlier and about [ ]'6' lower than during Matrix Test SB01. Reactor vessel heater rod cooling was maintained at that level (Figure 5.1.4-44). Due to the break size and resultant rapid loss of RCS fluid inventory during Matrix Test SB21, the initial depressurization, when cortpared with Matrix Test SB01, was very different (Figure 5.1.4-45). A period of quasi-equilibnum pressum existed between the RCS and the secondary side of the SGs for about [ ]'6' in Matrix Test SB01. In Matrix Test SB21, RCS pmssure decreased below that of the SG secondary side at about [ ]'6# after being at a quasi-equilibrium pressure for only about [ ]'** In the first few seconds after break valve opening, the SGs were isolated on both the feedwater and steam sides ;n order to minimize RCS heat losses. The only cooling available to the SGs was heat losses to ambient. With RCS pressure decreasing due to the break and SG pressure increasing due to heat absorption from the RCS, the pressures converged at about [ ]' 6' At about [ ]*' inventory losses through the break caused RCS pressure to drop below the pressure on the SG secondary side, and the SGs became a heat source for the RCS. The rapid RCS depressurization and early start of CMT draindown in Matnx Test SB21 resulted in CMT-2 reaching its low level setpoint at [ ]'6' and the ADS-1 valve opening [
]'6" later at [ ]' 6' This was [ ]'6' earlier than the ADS-1 valve was observed to open during Matrix Test SB01.
ADS Phase When the ADS-1 valve opened, injection flow from each accumulator increased by about [ ]'6# and injection flow from each CMT decreased by about [ ]*6# (Figure 5.1.4-16). The reduction in CMT flow was possfaly due to induced backpressure from the accumulator flow increase reacting against the CMT injection line check valves. When the ADS-2 valve opened at [ l'6' injection flow from each accumulator increased to about [ ]'6' and the CMT injection flows decreased to zero. Similar injection results were seen in reference test SB01, but about [ ]'6* later in the test and with lower flow rates. The ADS-3 vaive opening had little effect on any other parameters. The initial flow from ADS l-3 during Matrix Test SB21 were about double the flow during Matrix Test SB01, but the flow durations were shorter due to the rapid depressurization from the larger break (Figure 5.1.4-61). Pressurizer reflood began within [ ]'6# of the ADS-2 valve opening (Figure 5.1.4-5). Pressurizer reflood during Matrix Test SB21 was possibly due in part to the divergence of pressurizer and reactor vessel upper head pressures, with reactor vessel pressure being the higher of the two, as it was in Matrix Test SB01 (Figure 5.1.4-63). Also, about the same time that the ADS-1 valve opened in Matrix Test SB21, irregular oscillations in temperatures, pressures, and flow began (Figures 5.1.4-15,5.1.4-47, and 5.1.4-63). The oscillations were not observed in Matrix Test SB01. The oscillations were possibly produced by another o:U536wRevl\l536w 21.non:lb 081798 5.1.4-8 REVISION 1
FrNAL DATA REPORT phenomenon that was not observed in the reference test, detectable flow through the SGs. Flow through the primary side of the SGs, from the hot legs to the cold legs, was sufficient to cause differential pressure indications on DP-211 and DP-212 (Figure 5.1.4 81). Detectable flow through the SGs and the oscillations stopped about the time the CMT-1 balance line was filling, around [ ]'6" Accumulator injection flow began to decrease as the accumulators were nearly empty, allowing CMT injection flow to recommence (Figures 5.1.4-16 and 5.1.4-64). When the accumulators were empty, CMT injection flow stabilized at about [ ]'6" per tank. ACC-1 and ACC-2 were empty at [
]'6" respectively. These times were [ J'6" earlier than for the respective accumulator in Matrix Test SB01.
PRHR HX cooling in Matrix Test SB21 appears to have decreased to I to 2 gpm at about [ ]'b' versus about [ ]'6* in Matrix Test SB01 (Figures 5.1.4-40, 5.1.4-41, and 5.1.4-66). Matrix Test SB21 data then indicate an increase of flow through the PRHR HX short tubes at about [ ]'6' which would continue until about [ ]' 6' When RCS pressure decreased to [ ]'b' at approximately [ ]'6' (about [
]'6' earlier than in Matrix Test SB01), the two IRWST injection valves were automatically opened by a signal from the PLC. IRWST injection could not occur until RCS pressure decreased to near atmospheric, since the IRWST is a static system depending on gravity flow and operating at O- BAMS steam header pressure (Figure 5.1.4-45).
At[ ]'6' CMT-2 reached its low-low setpoint, causing the ADS 4-1 and ADS 4-2 valves to open. These two additional RCS vent paths decreased RCS pressure sufficiently so that IRWST injection to the DVI nozzles was initiated at about [ ]'6' earlier than during Matrix Test SB01 (Figure 5.1.4-48). The actuation of IRWST injection is dependent on the differential pressure between the reactor vessel and the IRWST, and the force for injection is gravity. Therefore, the early injection during Matrix Test SB21 was probably the result of the faster RCS depressurization.
. IRWST Inlection Phase The early start of IRWST injection during Matrix Test SB21 had two major effects on the facility overall response. First, CMT injection was stopped by the IRWST injection head prior to the CMTs being completely empty as in Matrix Test SB01 (Figure 5.1.4-6). Secondly, the cold legs refilled and subcooled at about [ ]'6" earlier than in Matrix Test SB01 (Figures 5.1.4-42,5.1.4-43,5.1.4-53, and 5.1.4-54). An effect of the cold legs maintaining some level in the early part of the test was that flow was maintained through the break in Matrix Test SB21, whereas in Matrix Test SB01, when the cold legs emptied, break flow stopped between about [
]'6' (Figure 5.1.4-28).
I a:ushnevnish.21.non:1w!M 5.1.4-9 REVISION 1
FmAL DATA REPORT Both CMTs reflooded during Matrix Test SB21 as they did during Matrix Test SB01 (Figures 5.1.4-6, 5.1.4-30, and 5.1.4-31). Major differences between Matrix Tests SB21 and SB01 were the timing of the CMT reflood and CMT response during the reflood. CMT-1 reflood occurred at about l [ ]'6" in Matrix Test SB21, or [ ]'6' earlier than in Matrix Test SB01. CMT- ) 2 reflood occurred at about [ ]'6" earlier than in Matrix Test SB01. The reflood of the CMTs occurred for the same masons as described for referer.ce test SB01 in Subsection 5.1.1. The reason for the earlier CMT reflood in Matrix Test SB21 is that the cold legs refilled with subcooled fluid earlier than in Matrix Test SB01. The balance lines refilled due to the condensation of steam from the CMT on the cooler fluid in the balance line and the subsequent reduction in CMT pressure. Condensation /depmssurization events were observed during the CMT reflood during Matrix Test SB21 like those that occurred during Matrix Test SB01. In Matrix Test SB01, the pressurizer reflooded to about [ ]'6# over the period of about [
]'6# This phenomenon was attributed to a lower pressure being formed in the pressurizer and ADS 1-3 separator when HL-2 filled with subcooled fluid to a level that started to fill the surge line while the ADS 1-3 sparger was still submerged in the IRWST. A vacuum bmaker was installed on the ADS 1-3 sparger line inside the IRWST following the performance of Matrix Test SB01 to prevent a recurrence of a reflood caused by negative pressure. The vacuum breaker performed its function and, under similar conditions, a pressurizer reflood did not occur during Matrix Test SB21 (Figure 5.1.4-5).
Primary sump injection started at about [ ]'6* earlier in Matrix Test SB21 than in Matrix Test SB01 (Figure 5.1.4-37). The start of primary sump injection in Matrix Test SB21 was somewhat different than in Matrix Test SB01. In Matrix Test SB01, injection flow started through the primary sump injection line check valves when the fluid levels in the sump and the IRWST equalized, about [ ]'6" before the sump injection valves opened. In Matrix Test SB21, injection flow did not start until about [ ]'6" after the primary sump injection valves opened. A possible explanation for the delay in sump injection during Matrix Test SB21 is that the IRWST and primary sump levels were not yet equalized; the IRWST was at the higher level, shutting off flow from the primary sump (Figures 5.1.4-35 and 5.1.4-37). The test continued in the sump injection recirculation mode for 2 hours. For Matrix Test SB21, the facility response while in this mode appears to be consistent with the data obtained in Matrix Test SB01. O o$1536wRev nl 536w.21.non:I b-081'198 5,],4 10 REVISION 1
FMAL DATA REPORT p 5.1.4.5 Comparison of Component Responses V Reactor The algorithm for core decay heat simulation of reactor power used during Matrix Test SB21 was a different algorithm than the one used during reference test SB01. A new set of decay heat curve algorithms was programmed into the reactor controller to maintain reactor heater power at 600 kW for 300 seconds prior to initiating the power decay in Matrix Test SB21. During Matdx Test SB01, the power was held at 600 kW for 140 seconds prior to initiating the power decay. 'Ihe core decay heat simulation of reactor heater power followed the programmed algorithm just as it did in Matrix Test SB01 (Appendix F). Reactor response during the performance of Matrix Test SB21 was consistent with rector response during Matrix Test SB01, with the exception of event timing. Core Makeup Tanks The early emptying of the pressurizer and surge line resulted in earlier CMT transition from recirculation mode to draindown, i.e., the time that the CMT balance line level began to decrease (Figure 5.1.4-6). The CMT-1 transition occurred at about [ ]'6# in Matrix Test SB21 compared with [ ]'** in Matrix Test SB01. The CMT-2 transition occurred at about [ V ] in Matrix Test SB21 compared with [ ]'** in Matrix Test SB01. To initiate the CMT transition from recirculation mode to draindov'n, the fluid coupling between the cold legs and the balance lines must be broken. In both tests, CL-1 and CL-3 fluid temperatures, measured [
]'6" from the top inside diameter of the reactor nozzle flange (as indicated by SC-105 and SC-101),
were subcooled at the time the CMTs transitioned from recirculation mode to draindown (Figures 5.1.4-53 and 5.1.4-54). One method of determining that the cold legs have partially drained is that their top thermocouples indicate superheat. Due to the location of SC-105 and SC-101 in the cold-leg flow stream, the fluid uncoupling between the cold legs and the balance lines would occur before the thermocouples could sense cold-leg draindown. Both CMTs reflooded in Matrix Test SB21 as they did in Matrix Test SB01 (Figures 5.1.4-30 and 5.1.4-31). In Matrix Test SB21, the CMT draindown was terminated prior to the tanks injecting their full volume due to the early initiation of IRWST injection. Another major difference between Matrix Tests SB21 and SB01 was the timing of the CMT reflood. CMT-1 reflood occurred at about [ ]'** in Matrix Test SB21, or [ ]'6" carlier than in Matrix Test SB01. CMT-2 reflood occurred at about [ ]'6" earlier than in Matrix Test SB01. The reflood of the CMTs occurred for the same reasons as described for reference test SB01 in Subsection 5.1.1. The reason for the earlier CMT reflood in Matrix Test SB21 is that the cold legs refilled with subcooled fluid earlier than in Matrix Test SB01. The balance lines refilled due to the condensation of steam from the CMT on the cooler fluid in the balance line and the subsequent reduction in CMT pressure with respect to RCS pressure. oA15%wRevlM5%w-21.non Ib481798 $,1,4-1 ] REVISION 1
FINA1. DATA REPORT Condensation /depressurization events were observed during the CMT reflood in Matrix Test SB21 like those that occurred in Matrix Test SB01. Accurnulators Accumulator response during the performance of Matrix Test SB21 was similar to the response of Matrix Test SB01, with the exception of event timing. Pressurizer In Matrix Test SB21, the pressurizer and pressurizer surge line were empty at [ ]' b" respectively (Figure 5.1.4-5). This occurred earlier by about [ ]'** for the pressurizer and [ ]'b' for the surge line than in reference test SB01. The time difference can be attributed to higher break flow rates early in Matrix Test SB21 (Figure 5.1.4-62). During the performance of Matrix Test SB01, the pressurizer exhibited a second reflood at about [ ]'b' but a common event did not occur during Matrix Test SB21 (Figure 5.1.4-5). As was described in Subsection 5.1.1.4 in the Pressurizer Response, the reflood was caused by a lower pressure developing in the pressurizer and ADS 1-3 separator with respect to pressure in the reactor vessel. The negative pressare was precipitated by several things: 1) the ADS 1-3 sparger located in the IRWST was submerged,2) the hot legs had refilled with subcooled fluid into the surge line, and
- 3) the combination of I and 2 resulted in condensation of steam in the pressurizer and ADS 1-3 separator with subsequent cooling.
The problem of a lower pressure being developed in the pressurizer and ADS 1-3 separator than in the RCS had been :orrected prior to the performance of Matrix Test SB21 by a facility modification, the installation of a vacuum breaker on the ADS 1-3 sparger line inside the IRWST. Another difference between the two tests was that during Matrix Test SB01, the pressurizer heaters remained energized at about [ ]'b' This was corrected prior to the performance of Matrix Test SB21 by a procedural change that required the operators to open the pressurizer heater breaker when the S signal was verified (Figure 5.1.4-24). During Matrix Test SB01, the logic was supposed to drive the heater power demand to zero, resulting in zero power from the pressurizer heater SCRs to the heaters. The data for Matrix Test SB01 suggest that the SCR control circuitry had not been properly tuned at that time. All other pressurizer responses during Matrix Test SB21 were similar to the responses during Matrix Test SB01, with the exception of event timing. c:\l536wRevi\l536w.21 non:Ib o81798 3,],4 12 REVISION 1 l ?
1 FINAL DATA Rzront
., Passive Residual Heat Removal Heat Excha==er
'Ihere was a difference in PRHR HX performance during the ADS phase of Matrix Test SB21 (Figures 5.1.4-66 and 5.1.4-67). The PRHR HX inlet head level transmitter recorded HX level changes during Matrix Test SB21 but did not record a level change during Matrix Test SB01. In Matrix Test SB01, the HX appeared to reach a minimum level of [ ]'6' at about [ ]'6' refill to [ ]'6" at about [ ]'*' and return to a minimum again at [ ]'6' In Matrix Test SB21, the HX level was predominantly higher than [ ]'6' during the first [ ]'6'and higher than [ ]'** until about [ J'*' It also appeared to maintain more flow. A
. possible cause for what appears to be more efficient PRHR HX response during Matrix Test SE21 is tnat RCS levels remained higher than Matrix Test SB01 levels (Figure 5.1.4-15).
The PRHR HX response during Matrix Test SB21 was similar to reference test SB01, except for event timing and the response of the HX wide-range level data after about [ ]*** In Matrix Test SB01, the HX partially refilled when the RCS loops were filled with subcooled fluid. Following the refill, when small pressure and level oscillations began to occur, it is possible that the oscillations caused the PRHR HX inlet line to " burp," allowing negative pressure in the HX to equalize with the RCS. Equalization of pressures allowed equalization of PRHR levels with those of the RCS. In Matrix Test SB01, once the levels equalized, the PRHR level remained essentially constant. O Data obtained during the performance of Matrix Test SB21 indicate that PRHR performance was consistent with the performance during Matrix Test SB01 until the final draindown of the HX. In Matrix Test SB21, the HX drained to [ ]'** at about [ ]'6' but then appeared to begin a slow refill for as long as data were obtained (Figure 5.1.4-68). It is possible that the HX was , refilling, but a more logical possibility is that LDP-802 was slowly losing its reference leg due to vaporization at a lower saturation pressure. Steam Generators At about the same time that the ADS-1 valve opened in Matrix Test SB21 ([ ]'6'h irrquiar oscillations in temperatures, pressures, and flow began (Figures 5.1.4-15, 5.1.4-47, and 5.1.4-63). The oscillations were not observed in Matrix Test SB01. The oscillations were probably produced by another phenomenon not observed in Matrix Test SB01, detectable flow through the SGs. Flow through the primary side of the SGs, from the hot legs to the cold legs, was sufficient to cause differential pressure indications on DP-211 and DP-212 (Figure 5.1.4-81). A possible cause for the oscillations is fluid at saturation temperature in the hot legs being superheated in the SGs and then condensing in the cold legs, resulting in multiple depressurization events in localized areas. Detectable flow through the SGs and the oscillations stopped about the time that the CMT-1 balance line was filling, around [ ] For the balance line to fill, the cold legs had to be full. O During the period of flow through the primary side of the SGs, there were oscillations of about a [ ]'6' magnitude in the SG wide-range level (Figure 5.1.4-82). The flow also subcooled much of cA15hRevl\l5h-21.non:lt481798 5.1.4-13 REVISION 1
FrNAL DATA REronT the fluid in the secondary side of the Sris. The SG fluid temperatures leveled off when primary-side flow stopped. Den as saturation temperature dropped, the temperatures followed saturation (Figures 5.1.4-83 and 5.1.4-84). SG downcomer fluid temperatures, located [ ]'b' above the tube sheet, subcooled almost immediately during Matrix Test SB21 and were still subcooled at the end of the test (Figures 5.1.4-85 and 5.1.4-86). Cold Legs and Hot Lers ne RCS cold-leg response during Matrix Test SB21 was different than during Matrix Test SB01 in that the cold legs refilled earlier in the test (based on reactor vessel downcomer levels; Figure 5.1.4-15). A possible explanation for the early refill in Matrix Test SB21 is that IRWST injection started earlier and DVI flow was greater than break flow (Figures 5.1.4-17 and 5.1.4-28). Here was a difference in detectable flow through the SG primary sides in Matrix Tests SB21 and SB01. The primary side flow is discussed in the Steam Generator Response. Another difference was that the cold legs never superheated during Matrix Test SB21 (Figures 5.1.4-81, 5.1.4-42, 5.1.4-43, 5.1.4-53, and 5.1.4-54). The cold legs probably did not superheat because of their early refill. In Containment Refueline Water Storaee Tanis The IRWST revonse during the performance of Matrix Test SB21 was similar to the IRWST response during reference test SB01, except that during Matrix 'le:4 SB21, the low-!ow level setpoint and the primary sump injection valves opened about 4800 seconds earlier. The earlier opening of the sump injection valves during Matrix Test SB21 may have been the result of sump injection not occurring through the sump injection line check valves, allowing IRWST injection to remain at a higher rate for longer. Break and ADS Meascrement System i The BAMS response during the performance of Matrix Test SB21 was different from the response during Matrix Test SB01 because of the initial header valve alignment and the larger break (Figures 5.1.4-73,5.1.4-74, and 5.1.4-75). Following break valve opening, BAMS header pressure was about [ ]'6# higher than in Matrix Test SB01. Peak BAMS header steam Ibws were about a factor of [ ]'** greater than the steam flows in Matrix Test SB01, but with the more rapid depressurization rate in Matrix Test SB21, the steam flow duration was shorter. l l Here was a marked increase in both steam and liquid flows from the break separator at about l [ ]'b' during Matrix Test SB21 (Figure 5.1.4-62). The flow increase coincided with the operator's closing CSS-902 and CSS-906 when total steam flow leaving the BAMS header decreased i to approximately [ ]'6" The reason for the steam flow increase, as recorded by FVM-903, i= that closing CSS-906 isolated the [ ]'6" vent line from the break separator, allowing flow only l through the [ ]'b' vent line. Closing CSS-902 and CSS-906 also caused about a [ ]'b' 1 oA1536wRevi\l 536w-21.non: l b-081798 5.1.4-14 REVISION 1 l
. . . - - . . - . _ _ . - . . . - . - . - - . . ~ ~ . . - . - . - . - - _ - . _ . - . - . - . _ . - . . - - . -
d FINAL DATA Rzroar s
,s increase in the break separator pressure, with essentially no change in primary sump pressure, and
- causing a pressure differential that resulted in the increased liquid flow from the beak separator (Figure 5.1.4-74).
' 5.1.4.6 ' Mass Balance ne mass balance results for . Matrix Test SB21 were calculated based on waLJ Sventory before and after the test and are provided in Appendix E. The mass at the end of the test was within
[ ]d' pement of the mass at the beginning of the test as compared to [ ]** percent for reference test SB01. 5.1.4.7 Conclusions i De test was performed with minimal problems and is considered acceptable. Alnough not all of the facility initial conditions met the specified acceptance criteria, the deviations did not i:npact the quality of the data. He instrumentation problems encountered were not critical to the performarce of the
- facility mass and energy balances.
1 Facility respon e to the test was as anticipated for the conditions that were established. The data ! clearly demonstrate that cooling of the reactor heater rods was maintained throughout the duration of
' the test.
i 1 i I s i J
. o:uswaeviusw.21mn:it>osi?98 5.1.4-15 REVISION 1 v w ,,
v---3r
l l FINA1. DATA REPORT l l l TABLE 5.I.41 l MATRIX TEST SB21 INITIAL CONDITIONS i Specified Instrument Initial Actual Initial Parameter No. Condition Condition Comments
~ ~
Pressurizer pressure"' PT-604 ?/0 2 2 psig 0C HL-1 tcmperature(" SC-141 420 t 2'F HL-2 temperature") SC-140 420 2'F SG-1 pressure'" IT-301 285 t 5 psig _ SG-2 pressure"' 14-302 285 5 psig Pressurizer level") LDP-601 65 5 in. Level signal te.wperature-compensated by TF-605 SG-1 narrow-range LDP-303 2613 in. Level signal temperature-level'" compensated by TF-301 SG-2 narrow-range LDP-304 26 2 3 in. Level signal temperature-level"' compensated by TF-310 IRWST temperature
- TF-709 < 80'F CMT-1 temperature
- TF-529 < 80 F CMT-2 temperature <2' TF-532 < 80 F [ ]'*" above required temperature ACC-1 temperature
- TF-403 < 80 F ACC-2 temperature
- TF-4(M < 80'F IRWST level
- LDP-701 Level established by fill line elevation ACC-1 level <2> LDP-401 Level established by standpipe at 37 in.
ACC-2 level
- LDP-402 Level established by standpipe at 37 in.
l ACC-1 pressure
- IT-401 232 2 2 psig [ ]'6* below i
required pressure ACC-2 pressure
- iT-402 23212 psig [ ]'"' below required pressure I
l t v:\l536wRevl\l536w-21.non:lt>.081798 5,],4-] 6 REVISION 1 i
l l FINA1. DATA REPORT 1 ( TABLE 5.1.41 (Continued) MATRIX TEST SB2I INITIAL CONDITIONS Instrument Specified Initial Actual Initial Parameter No. Condition Condition Comments CMT-1 level * #* LDP-507 Full CMT-2 level
- LDP-502 Full _. _
Nois: (1) Data for the indicated parameter were recorded in the test procedure as an initial condition for the test. The value was determined by the test engineer from the appropriate control board indicator. (2) Data were not recorded in the procedure, but the test engineer verified that specified conditions were achieved while establishing initial conditions. The value of the parameter was determined post-test by
.alculating the average DAS indication for a time of about 2 minutes before the break valve opened.
l l l l l' i t l l l l l l Nj l l i o:\l536wRevl\l536w-21.non Ib-081798 5.1.4-17 REVISION 1
1 l FINAL DATA REPO2T l TABLE 5.1.4 2 MATRIX TEST SB21 INOPERABLE INSTRUMENTS / INVALID DATA CHANNELS Instrument No. Instrument Type Description of Problem DP-216 Differential pressure transmitter Inoperable throughout test FMM-20l* Magnetic flow meter Removed from system (Subsection 5.1.4.2) FMM-202* FMM-203* FMM-204* FMM-40l
- Magnetic flow meter Inoperable after ACC-1 emptied at [ P6' when it appeared to have filled with nitrogen FMM-50l
- tiagnetic flow meter Data invalid after [ ]'6" seconds when the CMT was empty FMM-502 Magnetic flow meter Data invalid after [ ]'6# due to possible steam in balance line FMM-503 Magnetic flow meter Data invalid after [ ]'6' due to possible steam in balance line l FMM-504* Magnetic flow meter Data invalid after [ ]'6' when the CMT l was empty FMM-70l
- Magnetic flow meter Negative values after primary sump valves opened at
[ ]'6' invalid (Subsection 5.1.3.2) FMM-703 Magnetic flow meter Inoperable throughout test 1 . FMM-802* Magnetic flow meter Data invalid after steam formed in PRHR HX inlet line at about [ ]'6" FMM-8(M* Magnetic flow meter Data valid until PRHR HX initially drained at [ J'6' after this time, the possibility of steam in the outlet line invalidated the data FMM-905* Magnetic flow meter Negative values after break separator level exceeded the break elevation at about [ ]'6' invalid i HFM-103 Heat flux meter Inoperable throughout test l HFM-105 l HFM-505 HFM-703 l HPS-203-1 Heated phase switch Inoperable throughout test through -3 and HPS-509-1 through -3 i LDP-139 Differential pressure transmitter - luoperable throughout test l LDP-il3 level l 1 e 1 c:\l 536w Rev !\l 536w.21.non: 11481798 5.1.4-18 REVISION 1
I FINAL DATA Rzroar O\ (Vj TABLE 5.1.4-2 (Continued) MATRIX TEST SB21 INOPERABLE INSTRUMENTS / INVALID DATA CHANNEL Instrument ( No. Instrument Type Description of Problem 1 LDP-201 Differential pressure transmitter Data invalid due to effect of vertical portion of sense line LDP-202 - level attached to top of pipe; data can show level trends and LDP-203 when pipe empties or starts to drain, but absolute level LDP-204 indication can not be used LDP-205 LPD-206 LDP-215* Differential pressure transmitter inoperable - when tube voided, reference leg steamed off LDP-216 - level (Subsection 2.4) LDP-217 LDP-218* LDP-219* 3 LDP-220 LDP-221 LDP-222* l LDP-611 Differential pressure transmitter Data prior to ADS 4-1 opening at [ ]"' should
- level be ignored LDP-612 Differential pressure transmitter Data prior to ADS 4-2 opening at [ ]"' should
( - level be ignored (' LDP-80! Differential pressure cell - level Inoperable -indicated full when HX not full IrT_101 Pressure transmitter Data less than 6.1 psig invalid FT_102 Pressure transmitter Data less than 6.2 psig invalid PT_103 Pressure transmitter Data less than 6.2 psig invalid IrT_lG4 Pressure transmitter Data less than 6.4 psig invalid PT_108 Pressure transmitter Data less than 8.4 psig invalid IrT_109 Pressure transmitter Data less than 6.3 psig invalid IrT_111 Pressure transmitter Data less than 6.0 psig invalid frT_Il2 Pressure transmitter Data less than 8.8 psig invalid PT_ll3 Pressure transmitter Data less than 6.4 psig invalid l FT_20l* Pressure transmitter Data less than 1.1 psig invalid FT_202 Pressure transmitter Data less than 5.9 psig invalid IrT_205 Pressure transmitter Data less than 6.1 psig invalid Irr-801 Pressure transmitter Not used for this test
,f g PT-802 Pressure transmitter Not used for this test i a V TF-103
- Thermocouple fluid temperature Replaced with thermal stratification thermocouples oil 5%wRev i \t 5%w-21.non: l t>481798 $,],4 19 REVISION 1
FINAL, DATA REFORT TABLE 5.1.4 2 (Continued) MATRIX TEST SB21 INOPERABLE INSTRUMENTS / INVALID DATA CHANNEL Instrument No. Instrument Type Description of Problem TF-104
- Thermocouple fluid temperature Replaced with thermal stratification thermocouples TF-170 Thermocouple fluid temperature Inoperable due to leaking 0-ring in core barrel TF-203 Thermocouple fluid temperature Inoperable -indicated ambient throughout test TF-50l
- TF-504
- TF-615 TF-619 TFM-103 Thermocouple wall temperature Inoperable throughout test TFM-105 TFM-703 TH-317-1 Thermocouple heater rod Heater rod replaced with non-instruinented rod through TH-317-4 TW-503 Thermocouple wall temperature Inoperable throughout test TW-534 Thermocouple wall temperature Inoperable - indicated ambient throughout test l TW-547 TW-552 Note:
- Instruments marked with an asterisk are critical instruments. See Subsection 5.1.4.2 for discussion.
l l l O c:\1536wRev l\1536w-21.non:I b-081798 5.1.4-20 REVISION 1 i
FINAL DATA RaromT Table 5.1.4 3 on pages 5.1.4-21 through 5.1.4-29 is not included in this non-proprietary O document. 4 O O c:ushRevid5h-21.non:Ib481798 5.1.4-21 REVISION 1 {
FINAL DATA REPORT O 1 I w.::. .; e p s
'N s's l
g p%' L ,,R.','A n
~
s gl6 w rg ,N in g r s, / ( w l[\ i r[ , n rj it# +
-cou--ig -
j + rN
-e . 8) = __p' % W i a__q_
m
^,
-., , M M ,' '
j
, > + - - .
-F i --omra 1-one-- i 4-F T r T T 1 O
Figure 5.1.41 Primary Loop and Break Piping Layout o:usMwRevluS%w-21.non:lb481798 3,},4 30 REVISION 1
FINAL DATA REPORT O nt Dar J L C I n - TS-201
@ PLUG s,c _
y s RV227 s CL-3 DP
\
215 ,, , _f 'QJL dID - RV237 - 4
!- TS-205 i
O Figure 5.1.4-2 Primary Loop and Break Pipe Arrangement 1 0:\l 535w Rev i\l 536w-21.non: I b-081798 5.1.4-31 REVISION 1
FWAL DATA REPORT Figures 5.1.4 3 through 5.1.4-86 on pages 5.1.4-32 through 5.1.4-116 are not included in this nonproprietary document. t l l O O oA1536wRevitI536w.2f Jion:Ib-082798 5.1.4 32 REVISION 1
FINAt. DATA REPORT 5.1.5 ( Effect of a Smaller Break Size (Matrix Test SB23 Comparison with Matrix Test SB01 l and Matrix Test SB18) In this section, the results of Matrix Test SB23 (OSU Test U0023) are compared with the results l of Matrix Test SB01 (OSU Test U0001) and Matrix Test SB18 (OSU Test U0018). Rese two tests were identical, except that Matrix Test SB23 simulated a 1/2-in. pipe break and reference test SB01 simulated a 2-in. pipe break. Both simulated breaks were located at the bottom of CL-3. In both , tests, failure of one leg of one ADS-4 line was simulated by orificing one ADS-4 line for 50-percent design flow area and the other for 100-percent design flow area. Matrix Test SB23 was perfonned on August 9,1994, with a test duration of about 7.8 hours. During the transient, the CMTs and accumulators injected their water inventories, the IRWST injected water, and circulation through the primary sump was established. The test was terminated at 28,000 seconds after long-term cooling had been demonstrated. De core remained cooled throughout the test, and excessive temperature increases were not observed; therefore, this test was considered successful. Subsection 5.1.5.1 provides details related to the system configuration and initial conditions. Subsection 5.1.5.2 provides the description of inoperable instruments, and Subsection 5.1.5.3 lists the sequence of events. A discussion of the test results and evaluation can be found in Subsection 5.1.5.4, and Subsection 5.1.5.5 provides a comparison of component responses. A summary of the mass balance results appears in Subsection 5.1.5.6. The conclusions, as they apply to the comparison of results of Matrix Tests SB23 and SB01, can be found in Subsection 5.1.5.7. System responses to the break are documented by data plots, referenced as figuies in text, at the end j of this section. The numbering and content of the data plots for Matrix Test SB23 are identical to the ! data plots provided in Section 5.1.1 for Matrix Test SB01, with the exception that some ranges and ; time samples have been changed to allow for differences in the two tests. For example, the data plot j for instrument channel LDP-601 for Matrix Test SB01 is shown in Figure 5.1.1-5; the data plot for the same channel for Matrix Test SB23 is shown in Figure 5.1.5-5. Not all figures in the data plot package are referenced; only those figures required to explain a different response from that of Matrix Test SB01 are referred to in text. 5.1.5.1 System Configuration and Initial Conditions Matrix Test SB23 was performed in accordance with an approved written procedure. The test facility was configured in the normal arrangement described in Section 2. The configuration was identical to Matrix Test SB01, except for the size of the simulated break. All actions were automatic after the test ' started with no operator response required. He appropriate prerequisites were completed, and the initial conditions were satisfied. The required break simulation piping and break instrumentation were installed per Dwg. OSU 600904 O- and break piping layouts (Figures 5.1.5-la and 5.1.5 lb). A break hole ([ ]*), c:\l5hRevi\l5h-17.non:lb-081798 5.1.5-1 REVISION 1
FINAt. DATA REPORT simulating a 1/2-in. cold-leg break in the AP600, was installed in the bottom of the pipe break spool in CL-3. A flow nozzle simulating one line of flow was installed in the ADS 4-1 line (HL-1 to the ADS 4-1 separator) to provide the single tailure simulation, and a flow nozzle simulating two lines of flow was installed in the ADS 4-2 line (HL-2 to the ADS 4-2 separator). Additionally, flow nozzles simulating two lines of flow each were installed in the ADS l-3 inlet lines. He nonsafety syetems were not in operation for Matrix Tests SB23 or SB01. Pretest operations such as fill and vent processes were performed and are defined in greater detail in Subsection 2.7. Instruments were checked for required calibrations. Table 5.1.5-1 is a comparison of the specified and actual pre-test conditions for Matrix Test SB23. Here was one initial condition parameter which did not meet the specification. Pressure in ACC-1 and ACC-2 (PT-401 and PT-402) was [ ]' 6 ' respectively, lower than the specification. The loss of pressure between tank pressurization and test actuation was possibly caused by nitrogen gas cooling in the accumulator. Test analysis with the recorded lower accumulator overpressure should still be possible. The heater rod bundle power was adjusted prior to break initiation to achieve the required hot-leg temperatures. At the initiation of the break, rod bundle power was set at 600 kW. He actual power decay curves are provided in Appendix F. The differences between the actual and specified power decay are acceptable. Pressurizer power was terminated at initiation of the break. 5.1.5.2 Inoperable Instruments Table 5.1.5-2 contains a listing of instruments considered inoperable or invalid during all or portions of this test. Some of the instruments listed are on the Critical Instrument List (Subsection 3.2, Table 3.2-2) and, therefore, are addressed here. FMM-201, FMM-202, FMM-203, and FMM-204 measured flow (gpm) in each of the four cold legs. A decision was made to continue testing without the availat>ility of these instruments. Replacement flow meters repeatedly failed; their continued use was precluded due to cracking of the ceramic liners from thermal stratification in the loop piping. De necessary boundary conditions for loop flow could be determined from DP-202, DP-203, DP-205, and DP-206. FMM-501, FMM-504, FMM-802, and FMM-804 provided accurate data when filled with water, but became inaccurate when sensing a two-phase mixture or steam. FMM-701 measured IRWST-1 injection flow when the primary sump valves were opened, the flow meter indicated a negative flow as water flowed from the primary sump to the IRWST. He meter oA15hRevl\l536w-17.non:lt@81798 5.1.5-2 REVISION 1
FrNAL DATA REPORT O w[ as not designed to measure reverse flow, so this measurement was invalid!
]'6' However, total IRWST flow was measured by FMM 702. J FMM-905 measured break separator loop seal flow to the primary sump. As the transient proceeded, the primary sump and break separator levels exceeded the elevation of the break at the bottom of CL-3. When this occurred, break flow initially stopped and then reversed. Flow reversal through the break occurred between about [ ]'6' SG tube level data (LDP-215, LDP-218, LDP-219, and LDP-222) were biased by vaporization of the water in the transmitter reference leg after the SG tubes started draining. However, the data provide accurate indication of the time when the tubes are empty.
N-201 measured RCS pressure at the top of the SG-1 long tube. On August 15,1994, it was discovered that the transmitter had an incorrect zero compensation, which resulted in a negative error and negative data at low pressures. The transmitter zero was corrected at that time. FT-201 data obtained during Matrix Test SB23 had the zero correction performed, and the corrected data appear as Fr_201. Negative data and corrected r:egative data can be used to determine trends, but are considered inaccurate. PT_201 is not considered reliable, but a sufficient amount of other pressure data is available. O Data provided by ADS-4 separator instrumentation prior to the ADS 4-1 and ADS 4-2 valves opening at 2575 seconds were invalid due to the closed position of the ADS-4 valves and the ADS-4 separator loop seal valves. The instruments affected are: FMM-602, FMM-603, FVM-602, FVM-603, LDP-611, and LDP-612. Test analysis will not be affected, since ADS-4 flow did not begin until the valves opened. Considering these critical instrument failures, sufficient instrumentation was available to allow the performance of mass balances as demonstrated in Subsection 5.1.5.6 and Appendix E. An energy balance will be performed and reported in the AP600 Low-Pressure Integral Systems Test at Oregon ' State University Test Analysis Report, WCAP-14292.m l 5.1.5.3 Sequence of Events l Table 5.1.5-3 provides the sequence of events for Matrix Test SB23 and compares them with the l reference Matrix Tests SB01 and SB18. The following discussion compares the performance of SB23 l with SB01; differences between SB01 and SB18 are discussed in Subsection 5.1.2. The performance 4 l of SB01 is affected during ADS 1-3 since it was conducted without a vacuum breaker in the sparger. l SB18 was conducted with the vacuum break installed but is missing pressure, differential pressure, and l level test data after 14,312 seconds. The first pages of Table 5.1.5 3 provide the time of occurrence l for selected events in the tests and the difference in time of event occurrence with SB01 and SB18. l 'Ihe subsequent pages of the table provide a visual representation of the time compsrison by use of a oA15hRevl\l5W-17.non:lb-081798 5,1,$.3 REVISION 1
FINAL DATA REPORT l bar chart. On both the numeric table and the bar chan, events are sorted in chronological order for s ) l Matrix Test SB23. Because the simulated break area in Matrix Test SB23 is about [ ]'b' of the area of the simulated break in Matdx Test SB01, the resultant lower primary coolant loss rates caused delays of [
]'6" in most of the events that occurred after the break. Exceptions to this were the smaller delays in reflooding of CMT-1 ([ ]'6*), primary sump overflow ([ ]' 6#h and primary sump injection ([ ]'6#). Because of the low rate of flow through the simulated break, the primary sump ir:jection valves opened about [ ]'6' later in Matrix Test SB23 than in Matrix Test SB01. Because of limited How through the smaller break, the time for the IRWST to drain to the low level trip signal, which initiated the opening of the primary sump injection valves, was delayed.
The test proceeded through CMT, accumulator, IRWST, and primary sump injection. Matrix Test SB23 was terminated at about [ ]'6" after recirculation from the primary sump maintained steady-state operation for more than 2 hours. 5.1.5.4 Tests Results and Evaluation Facility responses for Matrix Test SB23 are reviewed in the same three phases employed in reference test SB01:
- Initial Depressurization Phase: simulated break initiation to ADS-1 Actuation l
= ADS Phase: ADS-1 actuation to start of IRWST injection
= IRWST Injection Phase: start of IRWST injection to end of test l
Initial Depressurization Phase ! Equipment responses to the S signal after actuation of the TEST pushbutton for Matrix Test SB23 were within I second of those for reference test SB01. The slight difference probably resulted from interpolation errors introduced by the data acquisition scanning rates. l After the break valve opened, liquid flowed from the break at an average rate of about [ ]'6* (Figure 5.1.5-28). Because the break was so small, steam produced by flashing of the break flow was below the detection limit for the BAMS steam flow meters (Figure 5.1.5-75). l l Reactor pressure increased initially, since the small break flow was insufficient to remove the energy being added to the primary system from the core heater bundle. At about [ ]'6" the pressure relief valves opened for about [ ]'6' After the relief valves closed, reactor pressure fell slowly to about [ ]'6' (Figure 5.1.5-2) at about [ ]'6' and then dropped about [ ]'6' when the SG primary and secondary pressures reached quasi-equilibrium onl5%wRevnl5%w 17.non;ltW81798 REVISION 1 5.1.5-4
. . - - - - . --. - - - . - ~ ~ - . - - . . . - . . --
FmAr DATA Rzroar O (Figure 5.1.5-45). Reactor pressure continued to decrease slowly for the remainder of the initial depressurization phase, reaching [ ]'6" he pressurizer emptied more slowly, requiring [ ]'6' in Matrix Test SB23 compared with [ J'6" in Matrix Test SB01. De pressurizer surge line emptied at [ ]*6' which also reflected the effect of the smaller break flow on inventory loss in the RCS (Figure 5.1.5-5). De SG-1 tubes and channel heads in Matrix Test SB23 drained between [ ]'6' l about [ ]'6" later than in reference test SB01. He SG-1 tubes drained earlier than the l SG-2 tubes by [ ]'b' possibly because of the lower backpressure in SG-1 caused by : the simulated break in the cold leg connected to this SG (Figures 5.1.5-7 and 5.1.5-9).
)
1 Once the SG tubes started to drain, their level measurements were erroneous because of water vaporization in the reference legs. As a result, refilling of the SG tubes, if it occurs, must be determined from temperature measurements. In this test, the SG did not refill; steam in the SGs remained superheated after draining (Figures 5.1.5-83 and 5.1.5-84). He ADS-1 valve opened at [ ]'6' later in Matrix Test SB01 than in Matrix Test SB23, terminating the initial depressurization phase. Since opening of this valve was initiated by the low level signal from CMT-1, the difference in timing was a result of the delay in j CMT drainage because of the smaller break flow. ADS Phase i Both accumulators started to inject their water inventories just after the ADS-1 valve opened when direct vessel injection (DVI) pressure reached the accumulator pressures (Figure 5.1.5-16). When the ADS-1 valve opened and the pressurizer was vented, the pressurizer refilled, partially drained, refilled again, then drained, emptying at about 3600 seconds (Figure 5.1.5-5). The combined effect of cold water injection from the accumulators and the opening of the ADS valves resulted in a rapid decrease in reactor pressure from about 170 psig to atmospheric pressure between 2260 and 2650 seconds (Figure 5.1.5-45). Core heater temperatures near the top of the core decreased simultaneously from about 420 F to about 270*F (Figure 5.1,5-44). Injection of water from the IRWST to the DVI lines started at [ ]'6' respectively, when pressure in the IRWST exceeded pressure in the two DVI lines (Figure 5.1.5-48). IRWST Inlection Phase After IRWST flow began, the core remained cooled (Figure 5.1.5-44x). IRWST flow reached a maximum of [ ]'6' in each line at about [ ]'6' and then fell steadily to about [ ]'6'just before primary sump injection began at about [ ]'** (Figure 5.1.5-48). IRWST flow increased sharply by [ ]'6' at about [ ]'6' pbably as a cA15hRevl\l5W.17.non:Ib-081798 5,1.5 5 REVISION 1
i l FINAL DATA REPORT result of condensation events in the CMTs that caused sudden low pressures in the downcomer and DVI lines. After IRWST flow decreased at [ ]'6' the flow rate fluctuated by several gpm in both lines until about [ ]'6' After this time, IRWST flow steadied, then fell when primary sump injection began. 1 Flow from the primary sump was initially established through CSS-921, CSS-922, CSS-923, and l CSS-924 in the lines bypassing the primary sump injection valves when the level in the IRWST was reduced sufficiently to permit these check valves to open. When the IRWST low-low level setpoint I
]'6' the primary sump injection valves were opened by the PLC. The I was reached at [
test was terminated about [ ]'6' later. Steady-state flows and temperatures were achieved in the reactor during this period. Core coolant reached saturation in the upper portion, indicating that this part of the core the core was being cooled by boiling (Figure 5.1.5-81). 5.1.5.5 Component Responses The component responses in Matrix Test SB23 were similar to those in Matrix Test SB01, except that l they were delayed up to several thousand seconds because of the smaller break. l l 5.1.5.6 Mass Balance Mass balance results for Matrix Test SB23 were calculated from water inventories in the facility components at the start and conclusion of the test. The final water inventory agreed with the initial inventory within [ ]'6" percent. Details of the mass balance calculations are provided in Appendix E. 5.1.S.7 Conclusions ne test was performed with minimal problems and is considered acceptable. Although not all of the facility initial conditions met the specified acceptance criteria, the deviations did not impact the quality of the data. De instrumentation problems encountered were not critical to the performance of the facility mass and energy balances. Facility response to the test was as anticipated for the conditions that were established. The data clearly demonstrate that cooling of the reactor heater rods was maintained throughout the duration of the test. l O c:\ t 536w Rev i\l 536w- 17.non: l t>081798 5.1.5-6 REVISION 1
FINAL, DATA REPORT f\ TABLE 5.1.51 MATRIX TEST SB23 INITIAL CONDITIONS Instrument Specified Initial Actual Initial Parameter No. Conditica Condition Comments Pressurizer PT-604 370 2 2 ps;g
- 4 pressure"'
HL-1 SC-14) 420 2'F temperature") HL-2 SC-140 420 2'F Temperature"' SG-1 pressure"> PT-301 285 5 psig l 1 SG-2 pressure"> PT-302 285 5 psig Pressurizer LDP-601 65 2 5 in. Level signal was level"' temperature-compensated by TF-605 SG-1 LDP-303 26 2 3 in. Level signal was narrow range temperature-compensated level"' by TF-301 f- S-2 LDP-304 26 t 3 in. Level signal was (% narrow range temperature-compensated level") by TF-310 IRWST TF-709 < 80'F temperature
- CMT1 TF-529 < 80*F temperature
- CMT-2 TF-532 < 80'F temperature
- ACC-1 TF-403 < 80'F temperature
- ACC-2 TF-404 < 80'F temperature
- IRWST level
- LDP-701 Level established by fill-line elevation
\
o:\l5hRevi\l536w 17.non:Ib-081798 5.1.5-7 REVISION 1
r l l l l Fmrt. DATA REPORT l I TABLE 5.I.51 (Continued) MATRIX TEST SB23 INITIAL CONDITIONS Instrument Specified Initial Actual Initial Parameter No. Condition Condition Comments ( ACC 1 level"3' LDP-401 Level established by standpipe at l 37 in. Level established by ACC-2 level"J' LDP-402 standpipe at 37 in. l ACC-1 pressure
- I'T-401 232 2 2 ptig Pressure was [
]* low; condition acceptable ACC-2* Pressure IYT-402 232 2 psig Pressure was [
]'" low; condition I I
acceptable CMT-1 level"' Full LDP-507 I CMT-2 level
- Full LDP-502 _ _
I l Note: (1) Data for the indicated parameter were recorded in the test procedure as an initial condition for the test. The value was determined by the test engineer from the appropriate control board indicator. (2) Data were not recorded in procedure, but the test engineer verified that specified conditions were achieved I while establishing initial conditions. The value of the parameter was determined post-test by calculating the I average DAS indication for a time of about 2 minutes before the break valve opened. (3) The bourdon pressure tube local indicator (PI-401 or PI-402) was tubed to the lower portion of the reference leg of the accumulator level transmitter (LDP-401 or LDP-402). As pressure in the accumulator was increased, , l air inside the bourdon tube was compressed, thereby lowering the reference leg liquid level, resulting in a false indication of measured level. l I l l J oA1536w Rev l\l 536w- 17.rm l b-081798 5.1.5-8 REVISION 1
r FmAL DATA Rzroar
)
1 l l TABLE 5.1.5-2 MATRIX TEST SB23 INOPERABLE INSTRUMENTS / INVALID DATA CHANNELS I Instrument No. . instrument Type Description of Problem l FMM-20l* Magnetic flow meter Removed from system l l FMM-202* Magnetic flow meter Removed from system FMM-203* Magnetic flow meter Removdd from system FMM-204* Magnetic flow meter Removed from system FMM-401 Magnetic flow meter Feiled FMM-502* Magnetic flow meter Dita invalid after 2310 seconds because of stum in balance line FMM-503* Magnetic flow meter Failed FMM-703 Magnetic flow meter Over-ranged from about 2450 to 2610 second i FMM-802* Magnetic flow meter Data invalid after steam forms in PRHR HX l inlet line at about 2325 seconds FMM-804* Magnetic flow meter Data valid until PRHR HX initially drained a 2325 seconds after which the possibility of steam in the outlet line invalidated the data HFM-103 Heat flux meter Failed HFM-105 Heat flux meter Failed HFM-505 Heat flux meter Failed HFM-510 Heat flux meter Failed i HPS-2031 through Heated phase switch Inoperable throughout test l i HPS-203-3 l HPS-509-1 through Heated phase switch Removed HPS-509-: i LDP-102 Differential pressure transmitter - level Failed l-LDP-201 Differential pressure transmitter - level Data invalid due to effect of vertical portion j LDP-202 of sense line attached to top of pipe; data LDP-203 can show level trends, when the pipe is empt; LDP-204 or starts to drain, but absolute level LDP-205 indication can not be used LDP-206 1 i o:\l 536wRev i\l 536w.17.non:l t41798 $,1,$-9 REVISION 1
FINAL DATA REPORT TABLE 5.I.5 2 (Continued) MATRIX TEST SB23 INOPERABLE INSTRUMENTS / INVALID DATA CHANNELS Instrument No. Instrument Type Description of Problem LDP-207 Differential pressure transmitter - level Inoperable; ranged improperly; data can show through trends, but absolute level indication can not LDP-209 be used LDP-215* Differential pressure transmitter - level Data invalid when tube drained and reference LDP-215 leg started to vaporize LDP-216 LDP-217 LDP-218* LDP-219* LDP-220 LDP-221 LDP-222* LDP-802* Differential pressure transmitter - level Data valid until PRHR HX initially drained at LDP-804 2325 seconds; data suspect because of possibl< vaporization of their common reference line FT_101 Pressure transmitter Data less than 6.1 psig invalid PT_102 Pres,ure transmitter Data less than 6.2 psig invalid FT_103 Pressure transmitter Data less than 6.2 psig invalid FT_104 Pressure transmitter Data less than 6.4 psig invalid Fr_108 Pressure transmitter Data less than 8.4 psig invalid PT_109 Pressure transmitter Data less than 6.3 psig invalid Fr_Ill Pressure transmitter Data less than 6.0 psig invalid Fr_112 Pressure transmitter Data less than 8.8 psig invalid FT_Il3 Pressure transmitter Data less than 6.4 psig invalid PT_.20l* Pressure transmitter Data less than 1.1 psig invalid PT_202 Pressure transmitter Data less than 5.9 psig invalid Fr_205 Pressure transmitter Data less than 6.1 psig invalid TF-103
- Hermocouple fluid temperature Removed and replaced with thermocouple for thermal stratification test TF-1(M* Hermocouple fluid temperature Removed and replaced with thermocouple for thermal stratification test TF-170 hermocouple fluit Emperature Read low throughout test
'lTM-103 nermocouple for HFM-703 Inoperable; indicated ambient temperature TFM-105 throughout test oA1536wRevl\l536w-17.non lb-081798 $, } ,$-10 REVISION 1 l
FINAt DATA REPORT ; l r
\ TABLE 5.1.5 2 (Continued)
MATRIX TEST SB23 INOPERABLE INSTRUMENTS /INVALII) %TA CHANNELS Instrument No. Instrument Type Description of Problem TH-3171 Thermocouple heater rod Inoperable; heater rod removed prior to test through TH-317-4 TW-503 'Ihermocouple wall temperature Inoperable throughout test B.elv
- Instruments marked with an asterisk are critical instruments. See Subsection 5.1.5.2 for discussion.
1 1 i i l O b O oA1536w Rev l\l 536w.17.non:I b41798 5.] .5-11 REVISION 1
FINAL DATA REPORT Table 5.1.5 3 on pages 5.1.512 through 5.1.5-20 is not included in this nonproprietary document. O l 9 c:\l536wRevi\l536w-17.non:lb 081798 3,},$.12 REVISION 1
FINAL DATA REPORT O l 1 I l g I l 1 ( b:L e ., s 6 '%'s - g
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-1 > --oma.+ Foner- i 4-F D e l V f 4 O Figure 5.1.5-la Primary Loop and Break Pipe Arrangement o:\l536wRevi\l536w-17.non:ltW)82598 5.1.5-21 REVISION 1
- . - . ._ _ __ _ . . - . . . . - ._ . .~ . _ - . _ - _ . _ ~ . - . __ _ __ ~ . . - . - FINA1 DATA REPORT O JML (7 J k I _)
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- TS-201 PT 203
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Figure 5.11-Ib Primary Loop and Break Pipe Arrangement 0 oA1536wRevi\l536w 17.non:Ib-082598 5.1.5-22 REVISION 1
p... ..v. FINAL DATA REPORT O Figures 5.1.5 2 through 5.1.5-84 on pages 5.1.5-22 through 5.1.5-106 are not included in this nonproprietary document. O O
- c:\l536wRevl\l536w 17.non:lN98 5.1.5-23 REVISION 1
. . = .
_. . _ _ _ _ . . . _ . - _ _ . _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ . ~ . _ . _ . FINAL DATA RaromT 1 5.1.6 Effect of an Interinediate Break Size (Matrix Test SB05 Comparison with ; l Matrix Test SB01 and Matrix Test SB18) J In this section, the results of Matrix Test SB05 (OSU Test U0005) are compared with those of Matrix l l Test SB01 (OSU Test U0001) and Matrix Test SB18 (OSU Test U0018). Rese two tests were identical, except that Matrix Test SB05 simulated a 1-in. pipe break and reference test SB01 simulated I a 2-in. pipe break. Both simulated breaks were located at the bottom of CL-3. In both tests, failure of one leg of one ADS-4 line was simulated by orificing one ADS-4 line for 50-percent design flow area
- - and the other for 100-percent design flow area.
I Matrix Test SB05 was performed on June 21,1994. During the transient, the CMTs and accumulators l injected their water inventories, the IRWST injected water, and circulation through the primary sump was established. De test was terminated at 23,000 seconds after long-term cooling had been demonstrated. The core remained cooled throughout the test, and excessive temperature increases were not observed; therefore, the test was successful. Subsection 5.1.6.1 provides details related to the system configuration and initial conditions. Subsection 5.1.6.2 provides the description of inoperable instmments, and Subsection 5.1.6.3 lists the sequence of events. A discussion of the test results and evaluation can be found in Subsection 5.1.6.4, l and Subsection 5.1.6.5 is a comparison of component responses. A summary of the mass balance results appears in Subsection 5.1.6.6. The conclusions, as they apply to the comparison of the tests,
/]
D can be found in Subsection 5.1.6.7. System responses to the break are documented by data plots, referenced as figures in text, at the end of this section. The numbering and content of the data plots for Matrix Test SB05 are identical to the I data plots provided in Section 5.1.1 for Matrix Test SB01, with the exception that some ranges and j time samples have been : hanged to allow for differences in the two tests. For example, the data plot j for instrument channel LDP-601 for Matrix Test SB01 is shown in Figure 5.1.1-5; the data plot for the L same channel for Matrix Test SB05 is shown in Figure 5.1.6-5. Not all of the figures in the data plot I package are referenced; only those figures required to explain a different response from that of Matrix Test SB01 are referred to in text. An x suffix in a data plot figure number indicates extended time.
- 5.1.6.1 System Configuration and Initial Conditions Matrix Test SB05 was performed in accordance with an approved written procedure. The test facility was configured in the normal arrangement described in Section 2. The configuration was identical to Matrix Test SB01, except for the size of the simulated break. All actions were automatic after the test
( started with no operator response required. 4
- The required break simulation piping and its instrumentation were installed per Dwg. OSU 600904
{_ (Appendix G) and the break piping layouts in Figures 5.1.6-la and 5.1.6-lb. A break hole ! oA15hRevl\l5h.14.non:Ib-081798 $,},6 1 REVISION 1 i
.- - , , - c .
FINAL DATA REPORT 1 [ ]* simulating a 1-in. cold-leg break in the AP600, was installed in the bottom of the pipe break spool in CL-3. A flow nozzle simulating one line of flow was installed in the ADS 4-1 line (HL-1 to the ADS 4-1 separator) to provide the single failure simulation, and a flow nozzle simulating two lines of flow was installed in the ADS 4-2 line (HL-2 to the ADS 4-2 separator). Additionally, flow nozzles simulating two lines of flow each were installec in the ADS 1-3 inlet lines. He nonsafety systems were not in operation for Matrix Tests SB05 and SB01. Pre-test operations such as fill and vent processes were performed and are defined in greater detail in Subsection 2.7. Instmments were ches ked for required calibrations. Table 5.1.6-1 is a comparison of the spec:fied and actual pre-test conditions for Matrix Test SB05. There was one initial condition parameter which did not meet the specification. Pressure in ACC-2, indicated by FT-402, was [ ]'b' lower than the specification. The loss of pressure between tank pressurization and test actuation was possibly caused by nitrogen gas cooling in the accumulator. Test analysis with the recorded lower accumulator overpressure should still be possible. The heater rod bundle power was adjusted prict to break initiation to achieve the required hot-leg temperatures. At the initiation of the break, the rod bundle power was set at 600 kW. The actual power decay curves are provided in Appendix F. The differences between the actual and specified power decay are acceptable. Pressurizer power was terminated at initiation of the break. 5.1.6.2 Inoperable Instruments Table 5.1.6-2 is a list of instruments considered inoperable or invalid during all or portions of this test. Some of the instmments listed are on the Critical Instrument List (Subsection 3.2, Table 3.2-2) and, therefore, are addressed here. FMM-201, FMM-202, FMM-203, and FMM-204 measured flow (gpm) in each of the four cold legs. A decision was made to continue testing without the availability of these instruments. Replacement flow meters repeatedly failed; their continued use was precluded due to cracking of the ceramic liners from thermal stratification in the loop piping. The necessary boundary conditions for loop flow could be determined from DP-202, DP-203, DP-205, and DP-206. FMM-501, FMM-504, FMM-802, and FMM-804 provided accurate data when sensing liquid, but became inaccurate when sensing two-phase or steam flow. FMM-701 measured IRWST-1 injection flow. When the primary sump valves were opened, the flow meter indicated a negative flow as water flowed from the primary sump to the IRWST. The meter oA1536w Rev i\l 536w.14.non: I b-081798 5,1,6 2 REVISION 1
FINAL DATA Raroar was not designed to measure reverse flow, so this measurement was invalid. However, total IRWST V flow was measured by FMM-702. FMM-905 measured break separator loop seat flow to the primary sump. As the transient proceeded, the primasy sump and break separator levels exceeded the elevation of the break at the bottom of CL-3, When this occurred, break flow initially stopped and then reversed. Flow reversal through the break occurred at about [ ]* rendering subsequent data invalid. SG tube level data (LDP-215, LDP-218, LDP-219, and LDP-222) were biased by vaporization of the water in the transmitter reference leg after the SG tubes started draining; however, data provide I accurate indication of the time when the tubes are empty. i LDP-401 and LDP-402 measured ACC-1 and ACC-2 levels, respectively. Due to air trapped in the sense lines for the transmitters, data from these transmitters were invalid. However, the initial level of the tank was established by a standpipe, so it was constant from test to test. 3e drain rate can be calculated using FMM-401 and FMM-402. Altemately, a pressure correction may be applied directly i to the level indications of LDP-401 and LDP-402. PT-201 measured RCS pressure at the top of the SG-1 long tube. On August 15,1994, it was discovered that the transmitter had an incorrect zero compensation, which resulted in a negative error { O and negative data at low pressures. The transmitter zero was corrected at that time. PT-201 data obtained during Matrix Test SB05 had the zero correction performed, and the corrected data appear as PT_201. Negative data and corrected negative data can be used to determine trends, but are considered inaccurate. PT_201 was not considered reliable for values less than 1.1 psig, but a sufficient amount of other pressure data is available. TF-501 and TF-504 measured CMT fluid temperature from the long thermocouple rod location near the bottom of each CMT. 'Ihe thermocouples appear to have measured ambient conditions throughout the test, which would indicate a short somewhere in the thermocouple wiring. With these thermocouples inoperable, the required long thermocouple rod thermocouple availability of "seven out of ten und no more than one in succession failed" was met. Data provided by ADS-4 separator instrumentation prior to the ADS 4-1 and ADS 4-2 valves opening at 978 seconds were invalid due to the closed position of the ADS-4 valves and the ADS-4 separator loop seal valves. The instruments affected are: FMM-602, FMM-603, FVM-602, FVM-603, LDP-611, and LDP-612. Test analysis will not be affected, since ADS-4 flow did not begin until the valves opened. Considering these critical instrument failures, sufficient instrumentation was available to allow the performance of mass balances as demonstrated in Subsection 5.1.6.6 and Appendix E. An energy
- i. balance will be performed and reposted in the AP600 ImPressure Integral Systems Test at Oregon State University Test Analysis Report, WCAP-14292.*
c:\l5hRevl\l5h-14.non:Ib-081798 3,1,6-3 REVISION 1
FINAL DATA REPORT 1 5.1.6.3 Sequence of Events
)
l Table 5.1.6-3 provides the sequence of events for Matrix Test SB05 and compares them with the l reference Matrix Tests SB01 and SB18. The following discussion con 4 ares the performance of SB05 ; l with SB01; differences between SB01 and SB18 are discussed in Subsection 5.1.2. 'Ihe performance l of SB01 is affected during ADS 1-3 since it was conducted without a vacuum breaker in the sparger. ; l SB18 was conducted with the vacuum break installed but is missing pressure, differential pressum, and l level test data after 14,312 seconds. The first pages of Table 5.1.6-3 provide the time of occurrence ; l for selected events in the tests and the difference in time of event occurrence with SB01 and SB18. l l The subsequent pages of the table provide a visual representation of the time comparison by use of a I l bar chart. On both the numeric table and the bar chart, events are sorted in chronological order for l Matrix Test SB05. l l The first page of Table 5.1.6-3 indicates the source of the actual time data. A D in the Data Source column indicates the recorded time was obtained from a software program that monitored digital events in the facility, including pump starts and stops, valve limit switch actuations, and alarms. The term vane opening means the valve has actuated and the closed limit switch is being opened (valve coming off the seat). An A in the Data Source column indicates the time data were obtained by reviewing test data obtained from the data acquisition system (DAS). Althugh the test DAS were in digital format, the DAS monitored analog events of pressure, flow, and temperature. Because the simulated break area in Matrix Test SB05 is about 1/4 the area of the simulated break in Matrix Test SB01, the rates of depressurization and water inventory loss from the RCS were lower, resulting in delays of [ ]"' in most of the events that occurred after the break. The CMTs reflooded earlier in Matrix Test SB05. The CMTs reflooded earlier because the RCS refilled more rapidly when IRWST injection started because the rate of water loss through the simulated break was smaller than the reference test SB01 loss rate. The pressurizer in Matrix Test SB05 renooded earlier than in Matrix Test SB01 because the net water volume in the RCS increased faster when the accumulators injected and the break flow was smaller. 1 The test proceeded through CMT, accumulator, IRWST, and primary sump injection. Matrix Test SB05 was terminated at about 23,000 seconds, after recirculation from the primary sump ; maintained steady-state operation for several hours. 5.1.6.4 Test Results and Evaluation Facility responses for Matrix Test SB05 will be reviewed in the same three phases employed in reference test SB01:
. Initial Depressurization Phase: sirautated break initiation to ADS-1 actuation
. ADS Phase: ADS-1 actuation to start of IRWST injection
. IRWST Injection Phase: start of IRWST injection to end of test I
1 o A1536wRev i\l 536w- 14.non: l b-081798 5.1.6-4 REVISION 1
FINAL DATA ILEPORT l i 3 Initial DtDrtS8UrIEstIOR Phase For Manix Test SB05, equipment responses to the S signal after the TEST pushbutton was pressed were within I second of those of Matrix Test S801. The slight time difference was probably the l result of interpolation errors introduced by the data recording scanning rates, i When the break valve opened, liquid flow reached a maximum of [ ]'6' (Figure 5.1.6-62), compared with [ ]'6' in Matrix Test SB01. De Matrix Test SB05 maximum steam flow was [ ]'6' compared with about [ ]'6' for Matrix Test SB01. These lower flows are expected because of tne smaller break diameter in Ahtrix Test SB05 (Figure 5.1.6-75). For the first [ ] (Figure 5.1.6-45), RCS pressure decreased because of water loss through ! the break, cooling provided by the PRHR, and cold water circulation from the CMTs. From about l [ J'b' pressure remained nearly constant at about [ ]'6" his period j coincided with the period during which primary and secondary pressure in the SG was in quasi-equilibrium (Figure 5.1.6-45). Here were small RCS piessure rises of about [ ]'6' that occurred at [ ]'6# and were accompanied by CMT flow increases. Dese pressure rises may have been caused by the cold-leg side of the SG tubes draining, since the levels in these tubes were fluctuatir.g at this time (Figures 5.1.6-7 and 5.1.6-8), or the pressurizer surge line draining (Figure 5.1.6-5). These fluids were the only source of transient energy input that could have caused this pressure increase at this time.
\.
He SG-2 tubes in Matrix Test SB05 drained [ ]'6' after the same tubes drained in reference test SB01. Draining of the SG-1 tubes in Matrix Test SB05 was delayed [
- ]'6' compared with the same tubes in Matrix Test SB01 (Figures 5.1.6-7 and 5.1.6-8). Once l the SG tubes started to drain, their level measurements were erroneous because of water evaporation in
! the common referenc~e legs of the SG tube level transmitters; however level data can be used to indicate trends. t l The measured level in the upper plenum of the reactor (LDP-139) decreased from an initial level of - l [ ]'6' (Figure 5.1.6-12). 9ased on these data, the maximum l %m content in the upper plenum was about [ ]'6# put a about [ ]'6' (Figure 5.1.6-4). This level was higher than the equivalent levei in Matrix Test SB01 and occurred several hundred seconds later. This is consistent with the expected results from the smaller break. Note: This steam content is an average and is comprised of a two-phase mixture with a steam volume above the mixture. De distribution of steam between the two-phase mixture and the steam above the mixture cannot be determined from the data. j As indicated by the lack of large temperature increases in the core heater wall temperatures during the i entire test, the core remained cooled (Figure 5.1.6-44). He short-term temperature rises (duration of
- about [ J'6') were probably spurious signals, as indicated by their short duration. A oAlswacvmsw.14.non strost?98 5.1.6-5 REVISION 1
1 1 FINAL DATA REPORT l temperature increase of about [ ] at about [ ]*6' was probably real and coincided with the pressure rise discussed previously. The pressurizer emptied at [ ]'6' in Matrix Test SB05 compared with [ ]'6' in Matrix Test SB01. Similarly, the Matrix Test SB05 pressurizer surge line emptied at [ ] compared with [ ]'6' for reference test SB01 (Figure 5.1.6-5). These results are consistent with the relationship of break size in these tests. In Matrix Test SB05, the pressurizer started to reflood at [ ]'6' as a result of steam being vented when the ADS-1 valve opened (Figure 5.1.6-5). A peak level was attained while the accumulators were injecting. A second peak was reached when CMT flow increased after the accumulatots emptied. The decrease in pressurizer level between these peaks may have been the result of the compressibility of nitrogen from the accumulators, which would inject into the system when the accumulater water inventory was exhausted. The level fell as the CMT flows decreased, and the ADS-4 valves opened until the pressurizer emptied again at [ ]'6' CMT-1 transitioned from recirculation to dlaindown at [ ]'6' later than in Matrix Test SB01 (Figure 5.1.6-6). CMT-2 reached this transition at [ ]'6' later than in Matrix Test SB01. When the transition occurred, flow rates from each CMT approximately doubled (Figure 5.1.6-16). At [ ]'6' the CMT-1 low level setpoint was reached, and the signal to open the ADS-1 valve was initiated [ ]'** later. This CMT-1 low level was reached [ ] earlier.in Matrix Test SB01. Bef re the ADS-1 valve opened, however, system pressure fell below the accumulator pressure, and both accumulators started injection of their water inventories at [ ]'6' after the accumulators began to inject in Matrix Test SB01. In Matrix Test SB01, the accumulators injected [ ]'b' before the ADS-1 valve openea. Accumulator flow rates are provided in Figure 5.1.6-16. ADS Phase Per facility logic, the ADS-1 valve opened [ ]'6' after the CMT-1 low level setpoint was i reached at [ ]'6' about [ ]'6' later than in Matrix Test SB01. The rate of RCS depressurization increased when the ADS-1 valve opened; however, ADS-2 and ADS-3 valves opening did not affect the rate of pressure reduction (Figure 5.1.6-45). The ADS-2 and ADS-3 valves opened l about [ ]'6' later than the valves in Matrix Test SB01. This difference is consistent with I the delayed opening of the ADS-1 valve, since they were triggered by the same signal, which had a longer delay in Matrix Test SB05. O oA15%wRevnl5%w-14.non:Ib-081798 5.1.6-6 REVISION 1
. .. . - _ - - . - - . - . - - _. - - .~ - . . - -
FINAr. DATA Rarc 7
~
ne CMT-2 transition from recirculation to draindown was delayed relative to CMT-1 because the pressure in CL-3 was slightly lower than in the other cold legs. The same effect delayed the achievement of CMT-2's low level setpoint. Therefore, CMT-1 transitioned from recirculation earlier and drained slightly earlier. Dere was only one downcomer condensation event in Matrix Test SB05, whereas several such events occurred in Matrix Test SB01. De sudden condensation of steam in the downcomer at about [ ]'6" caused a sharp, short rise in downcomer levels (Figure 5.1.6-15), and a negative differential pressure spike across the upper support plate and through the upper core support plate , bypass holes (Figure 5.1.6.-19). The latter negative differential pressure indicated a high flow rate of I steam through the bypass holes from the upper head to the downcomer, correspondent to a similar event in Matrix Test SB01 at about [ ]'6d Injection in IRWST-1 and IRWST-2 started at [ ]'6' compared with [
]'6" respectively, in Matrix Test SB01 (Figure 5.1.6-48). This difference is consistent with the effect of the smaller flow in Matrix Test SB05. !
IRWST Inlection Phase Both CMTs in Matrix Test SB05 started to reflood at about [ ]'6" (Figure 5.1.6-6x), about O [ ]'6* earlier than the respective CMTs in Matrix Test SB01. However, CMT-2 stopped refilling for about [ ]*** then refilled completely. De CMTs in Matrix Test SB01 reflooded at widely different times, i.e., [ ]'6' for CMT-1 and [ ]'6' for CMT-
- 2. The difference between CMT behavior in Matrix Test SB05 versus Matrix Test SB01 is consistent with the lower break flow rate in Matrix Test SB05, which led to earlier refilling of the cold-leg /CMT balance line. When these balance lines filled and spilled over into the top of the CMT, the cold liquid
, caused a sudden condensation of steam in the CMT. The resulting pressure decrease caused rapid filling of the tank with liquid from the RCS through the DVI and CMT injection lines. This phenomenon is discussed more completely in Subsection 5.1.1.4. The CMTs reflooded within about [ ]'6" of each other in Matrix Test SB05l whereas reflooding in Matrix Test SB01 occurred over a much wider time range (about [ ]'6#). This difference in timing probably resulted from the effect of the larger break on the CL-1 backpressure in Matrix Test SB01.- Primary sump injection in line 1 and line 2 started at abom t ]'**
. respectively, more than [ ]'6" later than in Matrix Test SB01. Flow from the primary sump was initiated through CCS-922 and CCS-924 when pressure in the primary sump equaled pressure at the break (Figure 5.1.6-37). The primary sump injection valves opened at [ ]'6' when the low-low level setpoint was reached in the IRWST. Test operation continued for almost 2 more O hours with cooling provided by recirculation through the primary sump.
U o:\l 5%w Rev l\l 5Mw- 14.now l b-061798 3,],6 7 REVISION 1
FINAL DATA REPORT 5.1.6.5 Comparison of Component Responses Other than differences in timing and some rates of change because of the smaller break in Matrix Test SB05, the responses of the components in Matrix Tests SB01 and SB05 were comparable, except for the pressurizer. Pressurizer "Ihe pressurizer in Matrix Test SB05 only reflooded once (at [ ]'6#), whereas the pressurizer in Matrix Test SB01 reflooded a second time, late in the transient, at about [ ]'6" This diffemnt response was caused by the modifications made to the system to eliminate negative p:essures in the pressurizer (described in Subsections 5.1.1.5 and 5.1.2.5). In Matrix Test SB05, pressurizer power was de-energized at the stan of the test by the PLC, whereas power to 'he pressurizer heaters was maintained at about 1.5 kW in Matrix Test SB01. 5.1.6.6 Mass Balance Mass balance results for Matrix Test 5905 were calculated frorn water inventories in the facility components at the start and conclusion of the test. The final water inventery agreed with the initial inventory within [ ]'b* Details of the mass balance analysis are provided in Appendix E. 5.1.6.7 Conclusions l Facility response to Matrix Test SB05 was as anticipated for the established conditions. Although not all of the facility initial conditions met the specified acceptance criteria, the deviations did not impact the quality of the data. The instrumentation problems encountered were not critical to the performance of the facility mass and energy balances. The lower break flow rates in the simulated 1-in. break (Matdx Test SB05) produced similar system l l responses (including CMT recirculation, draining, and reflooding; accumulator injection; IRWST injection; and primary sump injection) compared with the simulated 2-in. break in reference test SB01. Actuation of these components was delayed from [ ]'6# compared with Matrix Test SB01 because of the lower break flow and resultant slower rates of depressurization and level changes in the CMTs, accumulators, and IRWST. O oA1536wRevi\1536w.14.non:lb o81798 5.1.6-8 REVISION I
FINAs. DATA Rzroar TABLE 5.1.61 MATRIX TEST SB05 INITIAL CONDITIONS Specifbd Instnament laitial Actual Initial Parameter No. Condition Condition Comments Pressurizer PT-604 370 2 2 psig pressure'" HL-1 SC-141 420 t 2*F temperature'" HL-2 SC-140 420 2'F temperature'" SG 1 pressure'" PT-301 285 2 5 psig SG-2 pressure'" FT-302 285 2 5 psig Pressurizer level
- LDP-601 65 5in. Level signal was temperature- !
compensated by SC-605 l SG-1 narrow-range LDP-303 26 2 3 in. Level s.gnal was temperature-level
- compensated by TF-305 and TF-307 l SG-2 narrow-range LDP-304 26 3in. Level signal was temperature-level
- compensated by TF-306 and TF-308 IRWST TF-709 < 80'F temperature
- p C M T-1 TF-529 < 80*F I temperature
- CMT-2 TF-532 < 80 F temperature
- ACC-1 TF-403 < 80'F temperature
- ACC 2 TF-404 < 80*F temperature
- IRWST level
- LDP-701 Level established by fill-line elevation ACC-1 level"* LDP-401 Level established by standpipe at 37 in.
ACC-2 level"* LDP-402 ' Level established by standpipe at 37 in. ACC-1 pressure
- PT-401 232 2 psig ACC-2 pressure
- FT-402 232 2 2 psig Pressure was [ ]'** low; condition acceptable O
c:\l5hRevi\l5h-14.non:Ib 081798 $,],6-9 REVISION 1
FINAt. DATA REront TABLE 5.1.61 (Continued) MATRIX TEST SB05 INITIAL CONDITIONS Specified Instrument Initial Actual Initial Parameter No. Condition Condition Comments CMT-1 level <2' LDP-507 Full CMT-2 levelc2> LDP-502 Full ._ Note: (1) Data for the indicated parameter were recorded in the test procedure as an initial condition for the test. He value was determined by the test engineer from the appropriate control board indicator. (2) Data were not recorded in procedure, but the test engineer verified t. at the specified conditions were achieved while establishing initial conditions. He value of the parameter was determined post test by calculating the average DAS indication for a time of about 2 minutes before the break valve opened. (3) The bourdon pressure tube local indicator (PI-401 or PI-402) was tubed to the lower portion of the reference
- leg of the accumulator level transmitter (LDP-401 or LDP-402). As pressure in the accumulator increased, air inside the bourdon tube was compressed, thereby lowering the reference leg liquid level, resulting in a false indication of measured level.
1 O O c:\l536wRevi\l536w-14non lb-081798 5.1.6 10 REVISION 1
FINAL DATA REPORT l TAllLE 5.1.6-2 d MATRIX TEST SB05 INOPERABLE INSTRUMENTS / INVALID DATA CHANNELS Instrument No. Instrument Type Description of Problem FMM-201
- Magnetic flow meter Removed from system FMM-202* Magnetic flow meter Removed from system FMM-203* Magnetic Saw meter Removed from system FMM-204* Magnetic flow meter Removed from system FMM-502* Magnetic flow meter Data invalid after 544 seconds because of steam in balance line FMM-503* Magnetic flow meter Data invalid after 456 seconds because of steam in balance line FMM-70l
- Magnetic flow meter Data invalid after 16,700 seconds because of reverse flow from primary sump to IRWST i FMM-802* Magnetic flow meter Data invalid after steam formed in PRHR HX inlet line at about 300 seconds FMM-804
- Magnetic flow meter Data valid until PRHR HX initially drained at 580 seconds after which the possibility of steam in the outlet line invalidated data FMM-902 Magnetic flow meter Negative values indicated because of reverse FMM-905
- flow from primary sump / break separator until levels in IRWST/ primary sump / break separator were equal
, HFM-103 Heat flux meter Failed HFM-105 Heat flux meter Failed HFM-201 Heat flux meter Failed HFM-601 Heat flux meter Failed HPS-203-1 Heated phase switch Inoperable throughout test through HPS-203-3 HPS-509-1 Heated phase switch inoperable throughout test through HPS-509-3 LDP-127 Differential pressure transmitter - level Data invalid before 8640 seconds; isolation valve inadvertently closed until this time LDP-201 Differential pressure transmitter - level Data invalid due to effect of vertical portion of LDP-202 sense line attached to top of pipe; data can show LDP-203 level trends, when pipe is empty or starts to LDP-204 drain, but absolute level indication can not be LDP-205 used LDP-206 LDP-207 Differential pressure transmitter - level Inoperable; ranged improperly; data can show through trends, but absolute level indication can not be LDP-209 used r
u oil 536wRevi\l536w 14.non:Ib 081798 3,],6.]l REVISION 1 l
FNAL DATA REPORT r==:r TABLE 5.I.6-2 (Continued) MATRIX TEST SB05 INOPERABLE INSTRUMENTS / INVALID DATA CHANNELS Instrument No. Instrument Type Description of Problem Differential pressure transmitter - level Data invalid when tube drains and reference leg LDP-215* LDP-215 starts to vaporize LDP-216 LDP-217 LDP-218* LDP-219* LDP-220 LDP-221 LDP-222* LDP-40l* Differential pressure transmitter - level Data invalid; see note 3 of Table 5.16-2 LDP 402* LDP-802* Differential pressure transmitter - level Data valid until PRHR HX initially drained at LDP-804 1000 seconds after which data suspect because of ssible vaporization of the common reference ne Pressure transmitter Data less than 6.1 psig invalid PT_101 Pressure transmitter Data less than 6.2 psig invalid IrT_102 Pressure transmitter i sata less than 6.2 psig invalid IrT_103 Pressure transmitter Data less than 6.4 psig invalid PT_104 Pressure transmitter Data less than 8.4 psig invalid IrT_108 Pressure transmitter Data less than 6.3 psig invalid IFT_109 Data less than 6.0 psig invalid IrT_lli Pressure transmitter Pressure transmitter Data less than 8.8 psig invalid frT_ll2 Pressure transmitter Data less than 6.4 psig invalid IYT_113 Pressure transmitter Data less than 1.1 psig invalid frT_20l* Pressure transmitter Data less than 5.9 psig invalid frT_202 Pressure transmitter Data less than 6.1 psig invalid IrT_205 Irr-801 Pressure transmitter Not reliable after 1000 seconds TF-170 Thermocouple fluid temperature Read low throughout test TF-501 Thermocouple fluid temperature Read low throughout test TF-504 Thermocouple fluid temperature Read low throughout test TFM-103 Thermocouple for HFM-703 Inoperable; indicated ambient temperature TFM-105 throughout test TFM-201 TFM-601 TH-317-1 Thermocouple heater rod Ino - J' : heater rod removed prior to test through TH-317-4 TW-503 Thermocouple wall temperature liioperable throughout test
'nV-534 Thermocouple wall temperature Inoperable; indicated ambient throughout test TW-552 Thermocouple wall temperature Inoperable: indicates ambient throughout test Note:
- Instruments marked with an asterisk are critical instruments. See .otoection 5.1.6.2 for discussion.
o:\l536wRevl\l536w-14.non:Ib-081798 5.1.6-12 REVISION I
FINAL DATA REPORT Table 5.1.6-3 on pages 5.1.6-13 through 5.1.6-21 are not included in this nonproprietary version. 1 l 4 l 1 i l l I l r oA15%wnevius*w-14.non::b-os2598 5.1.6-13 REVISION 1
FINAL DATA REPORT O u k I
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FINAL DATA REPORT l O y,-,t (m a l' l
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RV237 - 4 TS-205 i Figure 5.1.6-1b Primary Loop and Break Pipe Arrangement (Sh. 2 of 2) c:\l 536w Rev i\l 536w- 14.non: I b-082798 5.1.6-23 REVISION 1
FINAL DATA REPORT Figures 5.1.6 2 through 5.1.6-80 on pages 5.1.6-24 through 5.1.6-104 are not included in this J nonproprietary document. l O 9 c:\is36wnevisis36w-14.non:ius2798 5.1.6-24 REVISION I
FmAL DATA REPORT 5.2 Cold Leg Breaks with Operation of Nonsafety Systems Matrix Tests SB04 (OSU Test UO204) and SB24 (OSU Test U0024) were performed with the nonsafety-related systems, chemical and volume control system (CVS) and normal residual heat removal system (RNS), operating. Both systems functioned as planned during the simulated break transients. In this section, the results of Matrix Test SB04 (2-in. cold-leg break with operation of nonsafety systems) are discussed in detail, and the results of Matrix Test SB24 (1/2-in. cold-leg break l with operation of nonsafety systems) are compared with the data from this large simulated break. D V O oA1536RevlM5W4.non:lt41798 5.2-1 REVISION 1
FINAL DATA REPORT 5.2.1 Reference 2-in. Cold-Leg Break (Matrix Test SB04) This section provides the test results for Matrix Test SB04, including the initial conditions, data plots, evaluation of overall results, and detailed discussions of test facility components performance. The simulated 2-in. break was located at the bottom of cold leg-3 (CL-3), which returns primary coolant from steam generator-1 (SG-1) to the reactor. To simulate failure of one of the ADS-4 lines, a flow nozzle with a 50-percent design flow area was installed in ADS 4-1, and a flow nozzle with the full design flow area was installed in ADS 4-2. During this test, the RNS and CVS pumps remained in operation to provide system response data for a 2-in. cold-leg break with these conditions. Facility responses to the break are documented by data plots, referenced as figures in the text. De test was terminated at about 3400 seconds after the core makeup tanks (CMTs) refilled. He CMTs refilled with water supplied by the RNS pump to the reactor coolant system (RCS). During the entire test, heater rods in the reactor vessel simulating core fuel rods remained cooled by water or a two-phase water / steam mixture. Performance of this test was successful. Subsection 5.2.1.1 provides details related to the test procedure, system configuration, and initial conditions. A description of inoperable instruments appears in Subsection 5.2.1.2, and Subsection 5.2.1.3 references the sequence of events. Subsection 5.2.1.4 describes the test results and evaluation. Component responses appear in Subsection 5.2.1.5, and a summary of mass balance results appear in Subsection 5.2.1.6. The conclusions, as they apply to Matrix Test SB04, appear in Subsection 5.2.1.7. 5.2.1.1 System Configuration and Initial Conditions Matrix Test SB04 was performed on June 27,1994, in accordance with an approved written test procedure. The test facility was configured in the normal arrangement described in Section 2, with the exception of the following specific modifications: 2-in. simulated break located at the bottom of CL-3
- Simulated failure of one leg of one of the ADS-4 lines
=
Nonsafety-related systems (CVS and RNS pumps) operated as programmed throughout the test Valves to the RNS suction line were aligned to connect the in-containment refueling water storage tank (IRWST) instead of the feedwater storage tank The 2-in. break was simulated by a [ ]* diameter orifice in the downward branch line from a O spoolpiece installed in CL-3. A normally closed, pneumatically actuated ball valve isolated the orifice from the system prior to break initiation. The required break simulation piping and break a:us36aeviush-6.non:IM81798 5.2.1-1 REVISION 1
FINAL DATA REPORT instrumentation were installed per Drawing OSU 600904 (Appendix G) and the break piping layouts in Figures 5.2.1-la and 5.2.1-lb. Here, the liquid and vapor fractions of the leakage flow were separated and individually measured. Liquid flowed to the primary sump tank, and vapor was discharged from I l the facility. Prior to initiation of the test, steady-state operating data were recorded for about 120 seconds to record pre-test conditions. Table 5.2.1-1 provides a comparison of the pre-test conditions averaged over the 120-second period with the conditions required by the test specifications. All pre-test operating conditions were achieved prior to the test, with the exception of several hot-leg temperatures that were 1* to 2*F above the limit. Decay heat from the reactor fuel was simulated by electrically powered heater rods installed in the core barrel of the reactor vessel. Comparison of the actual power in this test with the specified decay heat power is provided in Appendix F. Prior to initiation of the test, a check list of required prerequisites was completed. Key prerequisites in this check list included:
. Installation of simulated break piping and instrumentation
. Installation of proper flow nozzles in the ADS l = Valve line-up in accordance with an approved operating procedure l = Fill-and-vent system in accordance with an approved operating procedure
= Data acquisition system (DAS) instrumentation verification
. Water conductivity in specification
= Verification of relief valve setpoints l
= Temporary break valve control lines and instrumentation lines correctly connected Following satisfaction of these prerequisites, the facility was brought to full power using approved operating procedures. Control valves, pump switches, fan switches, and controllers were set in accordance with the instructions in the written test procedure, and their positions were verified by a written check list. After initializing the DAS, the software program used to monitor digital events in the plant was made operational and synchronized with the DAS. When the initial conditions were satisfied, the test was initiated by pressing the TEST pushbutton.
5.2.1.2 Inoperable Instruments Table 5.2.1-2 is a ".st of instruments considered inoperable or invalid during all or portions of this test. Some of the inramena listed are on the Critical Instrument List (Subsection 3.2, Table 3.2-2) and, therefore, are addressed here. FMM-201, FMM-202, FMM-203, and FMM-204 measured flow (gpm) in each of the four cold legs. A decision was made to continue testing without the availability of these instruments. Replacement flow meters repeatedly failed; their continued use was precluded due to cracking of the ceramic liners oA1536Revl\l536w-6.non:lb 081798 5.2.1-2 REVISION 1
t FINAL DATA REPORT from thennal stratification in the loop piping. De necessary boundary conditions for loop flow could
\ be determined from DP-202, DP-203, DP-205, and DP-206.
FMM-501, FMM-504, FMM-802, and FMM-804 provided accurate data when sensing liquid, but ; became inaccurate when sensing two-phase or stea'n flow. Steam generator (SG) tube level data (LDP-215, LDP-218, LDP-219, and LDP-222) were biased by , vaporization of the water in the transmitter reference leg after the SG tubes started draining. However, I the data provide accurate indication of the time when the tubes are empty. LDP-401 and LDP-402 measured ACC-1 and ACC-2 levels, respectively. Due to air trapped in the sense lines for the transmitters, data from these transmitters were invalid. However, the initial level of the tank was established by a standpipe, so it was constant from test to test. De drain rate can be calculated using FMM-401 and FMM-402, respectively. Alternately, a pressure correction may be applied directly to the level indications of LDP-401 and LDP-402. l PT-201 measured reactor coolant system (RCS) pressure at the top of the SG-1 long tube. On August 15,1994, it was discovered that the transmitter had an incorrect zero compensation, which resulted in a negative error and negative data at low pressures. De transmitter zero was corrected at that time. PT-201 data obtained during Matrix Test SB04 had the zero correction performed, and the corrected data appear as PT_201. Negative data and coirected negative data can be used to determine , trends, but are considered inaccurate. PT_201 is not reliable, but a sufficient amount of other pressure l data are available. TF-501 and TF-504 measured CMT fluid temperature from the long thermocouple rod location near the bottom of each CMT. The thermocouples appear to have measured ambient conditions throughout the test, which would indicate a short somewhere in the thermocouple wiring. With these thermocouples inoperable, the required long thermocouple rod thennocouple availability of "seven out of ten and no more than one in succession failed" was met. Data provided by ADS-4 separator instrumentation were invalid because the ADS-4 valves and the ADS-4 separator loop seal valves did not open during this test. The instruments affected are: FMM-602, FMM-603, FVM-602, FVM-603, LDP-611, and LDP-612. Considering these critical instrument failures, sufficient instrumentation was available to allow the performance of mass balances as demonstrated in Subsection 5.2.1.6 and Appendix E. An energy balance will be performed and reported in the AP600 Low-Pressure integral Systems Test at Oregon State University Test Analysis Report, WCAP-14292.* N. ousmaevmsn.6.non:nros 79s 5.2.1-3 REVISION 1
FINAt. DATA REPORT 5.2.1.3 Sequence of Events ne response of the test facility to the simulated break is summarized both in tabular and bar-chart form in Table 5.2.1-3. This test configuration resulted in ADS 1-3 actuation and CMT, accumulator, CVS, and RNS injections. He RNS pump injected cold water from the IRWST into the SG-2 channel head. The test was terminated about 13 minutes after the CMT refilled. Total test time was about 3400 seconds (i.e., less than I hour). After the TEST pushbutton was pressed, steady state data were recorded for about 120 seconds. At time zero, the break valve opening signal was generated, and the indication that the break valve had opened (lower limit switch) was recorded at [ ]' 6' The pressurizer heater was manually tripped several seconds prior to the S signal. At [ ]'** the safety system's actuation signal S initiated the following actions:
- Feedwater to second.uy side of SGs shut off ([ ]***)
- Passive residual heat removal (PRHR) outlet isolation valve opened ([ ]'6')
= Valves on CMT injection lines opened ([ ]'6')
= Reactor coolant pumps (RCPs) tripped at [ ]'** and coasted to zero flow at
[ ]'*' As primary coolant flowed through the break, pressure in the primary system decreased rapidly, and the indicated level in the reactor dropped. At [ ]'6' the indicated level in the upper head had fallen, resulting in the formation of steam. CVS pump injection sta:ted at [ ]'*' The level in the reactor continued to decrease. The SG tube began to drain at about [ ]'** then emptied completely by [ ]'6' Eight to ten seconds after the CMT balance lines started to drain, the associated CMT started to drain ([ ]'*' for CMT-1 and [ ]'** for CMT-2). Both accumulators injected water at [ ]'6' when direct vessel injection (DV1) pressure fell below the total accumulator pressure (nitrogen pressure, plus static head). At [ j'** RNS flow started. At [ ]'** the ADS-1 valve opened ([ ]**# after the CMT low level was reached), causing the pressurizer to refill at [ ]'6" since water flow into the system was greater than flow out of the simulated break. Pressure continued to drop as the system was cooled by the injected water. The ADS-2 valve opened at [ ]'*'([ ]'6' after the CMT low level was reached), and the ADS-3 valve opened at [ ]'6*Q ]*6' after the CMT low level was reached). When reactor pressure fell to less than [ ]'6' both IRWST isolation valves opened. Since the RNS pump was operating, water from the IRWST was pumped into the system when DVI pressure fell below the discharge pressure of the pump. At [ ]'** CMT-1 refilled; CMT-2 refilled at [ ]'** Although the refilling of either CMT had tx en specified as the end of the test, an additional 13 minutes (780 seconds) of operation were recordeo to ensure that stable conditions had been achieved. oA1536Revhl536w-6.non:lb-081798 5.2.1 4 REVISION I l
FINAL DATA REPORT
.A 5.2.1.4 Test Results and Evaluation .% )'
Complex interactions among the subsystems that compose the plart model were observed during l Matrix Test SB04; therefore, events during this test were divided into the following three phases to facilitate discussion of overall system and subsystem performance: 1 2
= Initial Depressurization Phase: simulated break initiation to ADS-1 actuation 1
=
ADS Phase: ADS-1 actuation to start of IRWST injection l IRWST Injection Phase: start of IRWST injection to end of test ; h i laitial Deoressurization Phase , 1 As high-pressure, high-temperature water flowed through the simulated break nozzle, it flashed into a i , two-phase mixture of steam and water because of the sudden pressure reduction. The liquid and steam flow components were separated in the break separator of the break and ADS measurement system - (BAMS), and their flows were individually measured. The liquid flow rate (break sepasator loop seal flow) from the break reached a maximu-m of about [ ]'6' then fell to about [
]'6" (Figure 5.2.1-2). At [ ]'6* liquid flow decreased to about [ ]'** then rose slowly to about [ ]'6" and then decreased slowly to about [ ]'6# during the remainder of the test. The region of nearly constant flow (between [
--]'6' ) occurred during the period of quasi-equilibrium between the primary- and b
secondary-system pressure / temperature (Figure 5.2.1-45) 'Ihe decrease in break flow at [ ]'6" resulted frorn the effects of ptessure reduction and cold water from the CMTs, which injected at [ ]'6" Steam flow from the break (Figure 5.2.1-3) reached a maximum of [ ]'6' and fell to about [ ]'6" (the SG transitioned from recirculation to draindown and the pressurizer drained). Steam flow appeared to decrease sharply at [ ]'6' as a result of RCS pressure reduction and CMT injection. After the ADS-1 valve opened at about [ ]'6" steam flow through the simulated break became negligible for the remainder of the test. After the simulated break valve opened, the measured level in the reactor decreased (Figure 5.2.1-4), i and its pressure fell (Figure 5.2.1-5). At about [ ]'6# the CMT isolation valves opened, and cold water flowed from the CMTs through the DVI line into the reactor downcomer. At this time, the CMT balance lines remained full, and water flowed into the DVI from the CMTs by natural circulation, driven by the density difference between hot water in the balance line and cold water in the CMTs. During this period, the indicated reactor level continued to decline, since flow through the simulated break was greater than the makeup flows. In this period, makeup water flowed into the reactor vessel from the CMTs, pressurizer, and CVS. Pressurizer water was driven into HL-2 by the higher pressure in the pressurizer until it empt'ed at about [ ]'6' (Figures 5.2.1-6 and 5.2.1-7). When the pressurizer reached the low-low level oAss36aeviusw-6_nonsos179s 5.2.1-5 REVISION 1
FINAL DATA REro2r l l setpoint at about [ ]'6' the CVS pump injected cold water from the feed storage tank to the l SG-2 cold-leg channel head. Figure 5.2.1-8 is a graph correlating CVS and RNS flows with time. Boiling occurred in the upper section of the core heaters at about [ ]'6' as indicated by l I slightly higher than saturation temperatures of the heater surface thermocouples (Figure 5.2.1-10). As the system depressurized, water in the core flashed from the decrease in pressure and vaporized because of heat input from the core. He maximum steam percent in the upper core occurred between [ ]'6' and was estimated to be [ ]*6' percent. The steam percent was estimated using the upper core level data (LDP-110) provided in Figure 5.2.1-9. Steam percent was estimated from the equation below: Steam percent = level water) -Iml(wh steam) x 100 level (100% water) During this period, the lower core remained filled with liquid water (LDP-109). Core heater temperatures (Figure 5.2.1-10) were [ ]'6" higher than the saturation temperature during the earlier stages of the transient, suggesting that boiling occurred at all locations measured (at core heights of [ ]'6' ). At about [ ]'6' the temperature at the [ ]'b' level fell below saturation, inferring subcooled heat transfer below the level. The remainder of the heater temperatures remained substantially above the saturation temperature, indicating that hoiling heat transfer occurred in the upper portion of the core (reflected by thermocouples at [ ]'6'b The heaters remained cooled during the entire transient-by two-phase steam / water for the first [ ]'6' and then by nucleate boiling. Short temperature excursions of about [ ]'6' occurred during condensation events at [ ]'6' These excursions resulted from flow reversals in the core barrel as a result of low pressures in the downcomer caused by condensation events. Core coolant temperatures increased by about [ ]'6" between [
]'6* (Figure 5.2.1-10). This temperature rise followed the ef!ect of the pressure transient on the saturation temperature. The cause of this pressure transient is not certain, but it appears to coincide with the transition from recirculation to tube drainage in the SG which provided a flow path to equalize hot-leg and cold-leg pressures.
Level data for the [ ]'6' volume above the upper-core plate (LDP-112; Figure 5.2.1-9) indicated that it remained full of water during the entire transient. However, since the volume below the upper-core plate contained about [ ]'6' percent steam and the volume above the upper-core plate was nearly [ ]'6' percent steam during its minimum, the volume just above the upper-core plate could not have remained liquid. One possible explanation for this apparent anomaly is an instrumentation failure; however, all other similar instrumentation in the reactor behaved normally. Derefore, it is unlikely that there was an instrumentation error. A more probable explanation is that the instrumentation measured an annular layer of liquid water along the wall (these pressure taps are flush el536anius36w-6mlM81798 5.2.1-6 REVISION 1
r i FINAL DATA REPCTT
} /
with the core barrel wall). Since the two-phase steam / water mixture was flowing at a high velocity through the holes in the upper core plate, an annular layer of water could be maintained at the outer diameter of the core barrel. It is this continuous layer of water at the wall that may have been measured by LDP-il2. De indicated level in the reactor vessel (LDP-127) reached a minimum of about [ ]"" between [ ]"' (Figure 5.2.1-4). This level was slightly above to the upper core plate. When DVI pressure (Figure 5.2.1-5) reached accumulator pressure [ ]"' both accumulators injected water into the reactor annulus through the respective DVI lines (Figure l 5.2 l.19). De measured level in the reactor started to rise when the accumulators began injecting water. Water injected from the RNS ([ ]'6#), resulted in a continued increase in reactor ! level, even after the accumulators emptied at [ ]"' The RNS pumps started at about [
] when system pressure reached [ ]'6* however, the full flow did not develop until the CVS pump was shut off (Figure 5.2.1-8).
As the system depressurized due to cominued flow through the simulated break and cold-water injection, liquid in the hot legs reached saturation temperature at about [ ]'6# and began to flash (Figures 5.2.1-13 and 5.2.1-14). This steam caused apparent level fluctuations in the SG tubes as the steam bubbles passed through the tubes (Figures 5.2.1-15 and 5.2.1-16). Water in the SG tubes also flashed as pressure decreased The SG tubes started to drain between [ ]'6' l !h and the SG tubes emptied between [ [ ] and the SG-2 channel head emptied by [
]'** De'SG-1 channel head emptied
]"' (Figures 5.2.1-17 and by i 5.2.1-l 8).
l l CMT-1 reached the low-level setpoint at [ ]'6#.which caused the ADS-1 valve to open at )
-[ -]'6" Opening of the ADS-1 valve concluded this phase of the test.
ADS Phase The ADS-1 valve opened at about [ ]'b' about [ )*** after CMT-1 reached its l low level setpoint, resulting in the following system effects:
- The pressurizer refilled, reaching its maximum level at about [ ]'6" (Figure 5.2.1-7).
Flow rates from the accumulators rose from about [ ]"# to a maximum of [ ]'b' (Figure 5.2.1-19). L
- RNS flow increased from [ ] (Figure 5.2.1-8).
=
De rate of depressurization of the entire system (including reactor vessel, pressurizer, hot legs, cold legs, and CMTs) increased (Figures 5.2.1-5, 5.2.1-6, 5.2.1-20, and 5.2.1-21). l o:uswaevlu5h-6.non:lb481798 5.2.1-7 REVISION 1 l
FINAL, DATA REPORT l ADS-2 actuated at [ ]'6# after the CMT-1 low level setpoint was reached). System pressures continued to decrease at about the same rate measured after the ADS-1 valve opened, and then asymptotically approached atmospheric pressure at about [ ]'6# The reactor vessel continued to refill with negligible effects on the refilling rate from the ADS-2 valve opening (Figure 5.2.1-4). Liquid flow through ADS 1-3 ranged from about [ ]'6# with a short peak (about a [ ]'6* duration) at [ ]'6# when ADS-2 actuated (Figure 5.2.1-2). From [
]'6' to test termination, liquid flow averaged about [ ]*6# This flow was probably a two-phase mixture of steam and water that had refilled the pressurizer at [ ]'6# During this period, steam vaporized from the liquid in the pressurizer (which had refilled at about [
]'6#). Steam flow from the ADS 1-3 separator ranged from [ ]'6# with the maximum occurring when ADS-2 actuated. Separator steam flow was [ ]'6# seconds to the end of the test (Figure 5.2.1-53). ,
At[ ]'6* the CVS pump shut down (Figure 5.2.1-8). This shutdown was initiated when the liquid level in the pressurizer reached its normal level. 1 IRWST-1 and IRWST-2 isolation valves opened at [ ]'6# respectively, when the reactor fell below the low-low pressure setpoint ([ ]'6#); however, flow from this tank was blocked by higher pressure from the RNS pump, which is connected to the same piping that leads to the DVI lines. Check valves in the IRWST lines prevented RNS flow from entering the IRWST. IRWST Inlection Phase Both accumulators emptied by about [ ]'6" (Figure 5.2.1-22). By this time, system pressure had reached atmospheric pressure. The system water level slowly rose as the RNS injected water at a higher rate than the loss rate through the simulated break. During this period, system temperatures continued to drop. At[ ]'6* respectively, CMT-1 and CMT-2 began to refill as indicated by the rise in level indication (Figure 5.2.1-23x). The CMTs refilled as a result of condensation events. This phenomenon is discussed in detail in Subsection 7.1. Although refilling of the CMTs satisfied the test completion objective, the test was continued for about 13 minutes to provide steady-state data. 5.2.1.5 Component Responses Responses of the major test facility components are discussed in detail in this section. The components included in this review are:
- Reactor a CMTs oA1536Revi\l536w-6.non lM81798 5.2.1 -8 REVISION 1
.- .- - - - _ . - - . _ _ - . . . . . . - ~ . - - - -
I FINA1. DATA REPonT l [~] V
' Accumulators Pressurizer 1 1
- PRHR HX
- SGs Cold legs and hot legs
- IRWST
- BAMS l
- Nonsafety related systems Reactor l 'As soon as the break valve opened, the indicated level in the upper plenum of the reactor fell sharply while the level in the downcomer region remained nearly constant just below the upper bypass holes in the upper head. The upper-plenum levels are shown in Figures 5.2.1-9 and 5.2.1-9x, for the first
[ ]**' and from [ ]'6' to test termination, respectively. Downcomer and wide-range reactor levels are shown in Figures 5.2.1-4 and 5.2.1-4x. Indicated levels in the upper plenum and' core barrel fell because water flowing out of the break was greater than the water flowing into the reactor from the pressurizer. The CVS pump staned at [ ]'6' The indicated level in the core barrel increased several inches for shon periods of time as the CVS b injected and the SG tubes drained, allowing water to drain into the channel heads. and from there, into { the hot and cold legs. The overall level in the core barrel, however, decreased steadily and reached a minimum of about [ ]'6' The level started to rise at this time because both accumulators injected water into the DVI line. The rate of the rising level increased further at
.[ ']'b' when RNS injection began. Several condensation events that caused very short-lived, but sharp, level drops occurred at about [ ]'b' Sudden condensation of steam at the top of the downcomer caused low pressure in this volume, producing reverse flow from the lower plenum into the downcomer. Sudden drops in core barrel level and sudden increases in' I downcomer level resulted from these short, but intense, flow reversals during the condensation events. -
Subsection 7.1 provides a detailed discussion of these condensation events. ADS-2 actuation at [ !
]'6' also caused an apparent decrease in level. Since these indications were very rapid (less than [ ]'6' ), it is probable that they were dynamic pressure pulses rather than actual level j changes.
After [ ~ ]'6' the core barrel level became slightly greater than the downcomer level, and
. both levels remained relatively constant until the test was terminated.
Pressures in the reactor, DVI, and downcomer are shown during the initial [ ]'6# period in Figure 5.2.1-5. These pressures follow each other very closely throughout the test. After [ ]'6# these pressures had decreased to near atmospheric pressure. Actual data records revealed that the downcomer pressure was about [ ]'6' higher than the reactor upper-plenum region while the RCPs were operating. During the pump coastdown period, this pressure difference decreased e:usnaevmsh6.non:lM81798 5.2.1-9 REVISION 1
FINu, DATA REPORT until the upper plenum pressure became greater than the downcomer pressure at about [ ]' 6 ' l This difference, which was [ ]'6' continued during the depressurization period until both ; 1 pressures approached atmospheric pressure. This pressure differential resulted in the level difference between the downcomer and the upper-plenum / core-barrel region discussed previously. l Reactor and DVI line pressures followed each other closely during the entire test. The pressures dropped sharply for the first [ ]'*# until the RCPs shut off. The pumps maintained a high flow rate through the simulated break, resulting in a high rate of depressurization (Figure 5.2.1-5). At about [ ]'6# the pressures rose about [ ]'** because hot water from the SGs partially drained into the hot and cold legs. From [ ]*** the pressures remained nearly constant at about [ ]'6* This pressure plateau was caused by a quasi-equilibrium in between the primary-system temperature and the secondary-system temperature in the SGs (Figure 5.2.1-45). As the SG tubes drained completely at about [ ]'6# RCS pressure began to fall again. At [ ]'** the ADS-1 valve opened, increasing the rate of depressurization. Small inflections in pressures of several pounds per square inch occurred from the condensation events at about [ ]'6* The pressures reached atmospheric pressure at about [
]'*' (Figure 5.2.1-5).
Differential pressures across the upper support plate and the upper head are illustrated in Figures 5.2.1-24 and 5.2.1-24x for the first [ ]'** and from [ ]'** to test termination, respectively. Initially, differential pressure across the upper-support plate was about [ ]'6# of water, indicative of the dynamic pressure drop from forced flow. Differential pressure across the upper-support plate fell to 0 at about [ ]'** which coincided with the end of RCP coastdown. This differential pressure remained at 0, except for short-term negative spikes that coincided with condensation events. The negative pressure differential spikes that coincided with the condensation events indicated increased pressure differences across the upper head and higher steam flows through the bypass holes. Large negative differential pressures (less than negative 22 in of water) indicated significant flow through the bypass holes in the upper head frcm the upper plenum to the downcomer. Short, intense, negative differential pressure spikes that occurred coincident with condensation events confirmed high rates of steam flow from the upper plenum to the downcomer during these events. After [
]'** differential pressure across the upper head remained at about [ ]'** of water, with the exception of an excursion at about [ ]'** as a result of a condensation event. This negative differential pressure was caused by evaporation of water in the lines connecting the pressure l
transmitter and actually represents no or very little flow. Subsection 2.4 discusses this instrumentation
- effect in detail.
The following temperatures are graphically presented in the identified figures and are discussed in this section: l o:\ l 536Rev i\l 536w-6.non: l b-081798 5.2.1-10 REVbiON 1
FINA1, DATA REPORT I l . e' l
/" Figures 5.2.1-25 and 5.2.1-25x: Reactor fluid temperatures at the top of the reactor, reactor l
upper vessel at 90 degrees, upper reactor vessel at 270 degrees, and 1/2-in. above the upper 1 l support plate; and saturation temperature based on PT-107 1 Figures 5.2.1-26 and 5.2.1-26x: Downcomer fluid temperatures at the hot-leg centerline, hot-leg bottom, below DVI line, upper flange, and bottom flange; and saturation temperatures I based on PT-107 l 1 l
=
Figures 5.2.1-27 through 5.2.130x: Fluid temperatures in the reactor core at various ! elevations from 10.5 to 51.86 in., saturation temperatures based on PT-113 l I~ l Figures 5.2.1-10 and 5.2.1-10x: Heater rod temperatures, and saturation temperatuies based on PT-113 Fluid temperatures at the top of the reactor (TF-120; Figure 5.2.1-25) and the top of the downcomer (TF-168; Figure 5.2.1-26) are compared with saturation temperatures based on pressure at the reactor 1
- top. Dese temperatures were slightly subcooled prior to break initiation. As pressure fell, the top of the reactor reached saturation temperature in about [ ]*** and the downcomer steam volume at its top reached saturation temperature at about [ ]'6' Both temperatures followed the saturation temperature to about [ ]'** as the reactor depressurized. De downcomer temperature superheated relative to reactor pressure beyond this time because the SGs stopped draining and fumishing saturated liquid to the cold leg. As the level in the downcomer dropped, steam in the top portion remained hot (i.e., superheated relative to reactor pressure), and pressure continued to decrease. Steam at the reactor vessel top also became superheated relative to the local pressure as pressure declined, and heat was supplied to the vapor by the reactor vessel. Condensation events (at about [ ]d' ) resulted in temperature decreases from superheat to well below l saturation in the downcomer top, indicating that subcooled water from the bottom of the reactor was -
drawn into this region as steam condensed.- ' During the first two condensation events in the downcomer, superheated steam in the top of the reactor (TF-120) fell to saturation. His probably was caused by cool water ejected into the top of the reactor through the holes in the downcomer top by the level surge in the downcomer caused by the sudden pressure reduction during these condensation events. For the third condensation even ([
]**), the steam temperature at the reactor top was only slightly reduced and remained above the saturation temperature. In this case, the downcomer pressure surge may have been smaller and did not eject cold liquid into the reactor top. The small temperature decrease probably resulted from the expansion of the steam as the core barrel level decreased during this condensation event.
Downcomer temperatures at lower elevations are shown in Figures 5.2.1-26 and 5.2.1-26x. These downcomer temperatures remained subcooled for the entine test because of the continuous injection of cold water from the CVS and RNS, in addition to injections of cold water from the CMTs and accumulators. De highest temperatures in the downcomer occurred adjacent to HL-2, which reached oat 536anm5h4.non:1b481798 5.2.1-11 REVISION 1
FINA1. DATA REPORT l the saturation temperature between [ ]'6' Hot-leg fluid temperatures were near saturation during this period because saturated water was draining from the SG into the hot leg and flowing into the downcomer. Fluid temperatures in the downcomer and hot leg varied between [
]'6# and then became nearly steady at about [ ]'6"for the remainder of the test. He apparent discrepancy in temperatures measured by TF-169 and TF-170 was probably caused by leakage of cooler water from the downcomer contacting TF-170. Therefore, temperatures recorded by this thermccouple should be disregarded.
Fluid temperatures in the core at elevations from [ ]'6' are compared with the saturation temperature at midcore in Figures 5.2.1-27 through 5.2.1-30x. These plots have been divided into four sets for clarity, each with several core temperatures and the saturation temperature. Temperatures increase as elevation through the core increases. However, during the test, the saturation temperature was never exceeded, demonstrating that dryout did not occur in the core during this test. Fluid temperatures rose briefly for all locations at about [ ]'6# because of flow reversals caused by condensation events in the downcomer. Core heater rod temperatures measured by thermocouples installed near the surface of the heater rod sheaths were corppared with the saturation temperature at the upper core spacer grid (Figures 5.2.1-10 and 5.2.1-10x). Core heater rod temperatures were never significantly higher than the saturation temperature, demonstrating that the heaters remained cooled by liquid or a two-phase mixture during the entire test. Short-term temperature rises recorded during the test may have been caused either by brief periods of lack of coolant contact with the heater surface or an electric field effect from the heaters on the instrumentation. Core Makeup Tanks Since both CMTs behaved similarly in this test, their responses are discussed collectively in this section. Natural recirculation flow of [ ]'6' from the CMTs began about [ ]'** after break initiation (Figure 5.2.1-31), coincident with coastdown of the RCP pump. Differential pressure for this flow arose from the higher density of cold water in the CMT, Flow from the CMTs continued until the transition from recirculation to draindown when the CMT balance lines drained between [ ]'6" as indicated by the sudden drop in balance-line levels (Figure 5.2.1-32). Once the baia nte lines drained, flow from the CMT, which was now driven by the liquid head in the tank, increased to [ ]'6" Flow declined slowly as the liquid head in the CMTs dropped. Flow from the CMTs stopped between [ ]'6" because backpressure created by the RNS pump, which started at 405 seconds, was sufficient to balance the driving pressure for the CMTs. here was no further flow from the CMTs for the remainder of the test. i REVISION 1 c:\15%Revi\l5%w-6.non:ltM)81798 5.2.1-12
. - . - . . . . . . ~-. . . - - _
.-. - - . - ..-- ~ . - . - - .. - - -
FINAL DATA REPOIT Liquid levels in the CMTs (Figure 5.2.1-23) remained constant as long as the balance line remained full (Figure 5.2.1-32). Hot water flowed into the CMTs, leading to a small increase in temperature during the natural convection phase. l CMT-1 drained at [ ]'6' CMT-2 drained at [ ]'6' Drainage from the CMTs stopped at about [ ]'6# after startup of the RNS pump. CMT-1 reflooded at [
]# CMT-2 reflooded at [ ]'6' (Figure 5.2.1-23x). Only about [ ]**' percent of the CMTs' water inventory had been injected before the RNS pump halted their flow.
1 Pressures in the CMTs (PT-501 and PT-502) are compared with reactor upper-head pressure (PT-107) in Figure 5.2.1-21. Since the CMT tops are connected to cold legs through the balance lines, the pressures closely follow the reactor top pressure for the entire test. Events that influence the behavior of the reactor top pressure are discussed in the reactor response discussion of Subsection 5.2.1.5. When the CMT-1 and CMT-2 started to drain at about [ ]'6# respive4 saturated steam accumulated at the top of each tank (Figures 5.2.1-33 and 5.2.1-35). n pressure in
. the CMTs continued to drop, the local saturation temperature decreased and the stear., became i
superheated relative to the pressure. As depressurization continued, the saturation temperature became l lower than the tank wall temperature, and the CMT metal became a heat source. His phenomenon was observed in CMT-1 at about [ ]'6' for CMT-2), at which time i steam at the top of the tank superheated. De metal wall continued to transfer heat to the steam until the CMTs reflooded. Cold water in the lower portion of the CMT then rose to the top of the tank (Figures 5.2.1-33 and 5.2.1-35). Reflooding l occurred at about [ ]'6' for CMT-1 and [ ]'6' for CMT-2, as shown by the ! sudden decrease in temperatures (Figures 5.2.1-34 and 5.2.1-36). Dese refills resulted from ! condensation events in the CMTs initiated by cooler water flowing into the CMTs from the balance l lines. These events are described in greater detail in Subsection 7.1. When the CMTs were partially empty, a hot water layer existed in each CMT, separating superheated steam at the top of the tank and subcooled water at the bottom. Thickness of the saturated hot water ( layer varied. For CMT-1, by the time thermocouple TF-509, the thermocouple [ ]'6# below l TF .' 13, reached saturation temperature, all thermocouples above it had already superheated and all thermocouples bei. ,v it were subcooled (Figure 5.2.1-33), suggesting that the saturated hot water layer in CMT-1 was located between [ ]'6' (distance to the next lower thermocouple). Similar results were observed for CMT-2. An- = =!=eors Each accuniulator was filhd with [ ]'6" of ambient temperature water and was pressurized to y r. bout [ ]'6' with nitrogen. Both accumulators injected water at about [ ]'6' when RCS pressure decreased to about [ ]'6' Flow initiation, as indicated by flow (Figure 5.2.1 19) o:us36aevm5m.6.non:strost?9s 5.2.1-13 REVISION 1
FINAL DATA REPCT l and level (Figure 5.2.122) changes at [ ]'6# coincided with the time that DVI pressures reached [ ]'6# (Figure 5.2.1-5). Initial flow for both accumulators varied between [
]'6# until the ADS-1 valve opened. Flow rose over the next [ ]'6# to a maximum of l
about [ l'6# fell to [ ]'6# between [ ]'6# m to [ ]'6#and l then declined as the tanks emptied at [ ]'6# The cause of the reduction in flow from [
]*6# cannot be explained.
CMT flow influenced flow injection from the accumulators, since the CMT and accumulator for each DVI line were connected to a common line. As pressure in the CMT line varied, backpressure in the accumulator injection line affected accumulator injection flow. Although the RNS pump was i l operating during accumulator injection, it did not have a significant effect on accumulator flow j l l because its discharge pressure was reduced by an orifice upstream of the junctions with the lines from l the accumulators. Pressurizer Durir.g normal operation, the pressurizer maintains RCS pressure by control of the electric power to a j heater submerged in water. The pressurizer also provides a surge volume to accommodate primary ! coolant volumetric changes resulting from temperature variations. Initial power to the pressurizer required to maintain system pressure was several kilowatts during the steady-state period prior to test i initiation. Six seconds after the simulated break initiation, power to the heaters was shut off. l Liquid levels in the pressurizer and the surge line are shown in Figures 5.2.1-7 and 5.2.1-7x. As primary coolant flowed through the simulated break in the cold leg, the level in the pressurizer fell immediately because of differential pressure between the pressurizer and the break. The pressurizer emptied completely at about [ ]'6# The surge line level indication oscillated initially and then decreased before the pressurizer emptied. Since it is impossible to drain the surge line before the pressurizer is empty, this effect was caused by steam flashing in the surge line. Hot water draining from the pressurizer flashed into a steam-water mixture in the surge line, since pressure was decreasing rapidly, resulting in a lower mean density for two-phase flow and an apparent level decrease. At about [ ]'6# when the pressurizer emptied, steam volume in the pressurizer surge line was estimated at [ ]*** percent. The surge line drained completely at [ ]'6" Surge line refill began at about [ ]'6# when the ADS-1 valve opened and vented the pressurizer. The pressurizer level increased a few seconds later, after the time delay required to fill the surge line.
'Ihe pressurizer filled completely. Level surges in the surge line and pressurizer during refill were caused by pressure surges resulting from the ADS-2 and ADS-3 valves opening and condensation events.
Passive Residual Heat Removal Heat Exchancer Inlet and outlet flow rates from the PRHR HX for the initial 1000 seconds and from 1000 seconds to test termination are provided in Figures 5.2.1-37 and 5.2.1-37x, respectively. Prior to break initiation, oA1$36RevluS36w-6.non:lt>081798 5.2.1-14 REVISION 1
FINA1, DATA REroRT l l j O both flow rates were 0, since the PRHR isolation valve was closed. When the valve opened at b [ ]'6" natural circulation flow was established in the PRHR HX because of the higher l l ! density of the cooled water in the HX. For about [ ]'6" liquid flowed through the PRHR HX since inlet and outlet flows were about the same during this period (discounting sharp rises due to pressure surges). Liquid flowed from HL-2 through the PRHR HX and returned to the SG-2 cold-leg channel head because of the pressure difference between HL-2 and the SG-2 channel head. When the short tubes of SG-2 drained at about [ l'6' pressures in the SG-2 channel head and Hie 2 equalized, eliminating the pressure necessary to lift the liquid to the high point in the PRHR inlet piping, which was over [ ]'** above the PRHR HX inlet. When this occum:d, water in the inlet line of the PRHR HX drained. l Steam from HL-2 continued to flow into the PRHR HX because of the differential pressure between i the HL-2 water and the saturation pressure in the PRHR HX. However, this steam rate was insufficient to keep the inlet head of the PRHR HX full of water. Since the upstream pressure at the ! top of the PRHR HX was determined by the saturation pressure of water at the local temperature, the L pressure differential was insufficient to maintain flow through the PRHR HX (i.e., the PRHR HX became vapor locked). Between [ ]'6' flow into the PRHR HX was mostly steam since inlet flow measured by the magnetic flow meters was much higher than outlet flow. Between [
]'** the indicated flow rates oscillated at an average of zero flow; then both increased
}
from [ ]'6# retumed to oscillating about [ ]'6'and showed positive intermittent flow rates between [ ]'6' During the periods of
. flow, outlet flow varied between 5 and 10 gpm (discounting short, high flows).
The liquid levels in the PRHR HX inlet header and the entire HX for the initial 1000 seconds and j from 1000 to 2000 seconds are shown in Figures 5.2.1-38 and 5.2.1-38x, respectively. Both the inlet j head and overall HX levels remained nearly full during the first [ ]'6" after the PRHR valve opened, confirming that liquid was flowing through the PRHR HX as indicated by the flow meters. The level the inlet head fell to 0 in by about [ ]'6' and remained at 0 in. until starting to increase slowly at [ ]'** to a height of about [ ]'6' During this period, the PRHR HX fell to a minimum of about [ ]'6' then increased slowly until it was [ ]'6' The levels indicated that the PRHR HX was receiving steam that condensed in the HX at a rate less than the steam outflow rate during the early part of the test ([
]'** ), then steam flow increased, resulting in a level increase. The increase in steam flow at [ ]'*' was caused by SG-2 cold-leg channel heads refill, which occurred at about
[ ]'6' As the SG channel head and tube level increased, differential pressure from the PRHR HX to the SG channel head decreased, reducing the outflow from the PRHR HX and increasing i its fluid level 1 (' Fluid temperature at the inlet of the PRHR HX followed the saturation temperature based on pressure at the top of the reactor (Figures 5.2.1-39 and 5.2.1-39x) for the first [ ]'** fell below I a:usuaeviush4.non:Ib 08D98 5,2,].15 REVISION 1
FINAt. DATA REPORT saturation temperature until [ ]'6' rose to near saturation temperature between [
]'** and then decreased again to reach a temperature nearly [ ]'6' below saturation froc, [ ] to test termination. This temperature profile indicated that saturated water or steam entered the PRHR HX during the initial [ ]'6' and the period between
[ ]*** Condensate entered the PRHR HX between [ ]'*' because the inlet temperature ranged between [ ]*6' If there had been no condensate flow, the PRHR HX inlet would have reached equilibrium with the IRWST (i.e., below [ ]" 6' ). Midpoint temperatures of the inshumented short and long tubes indicated warm condensate (up to [ ]'*" ) in both tubes for about [ ]'6' Th h me upta almated between the Aort and long instmmented tubes, indicating unstable flow through the HX during the period between [ ]'** At [ ]'b' midpoint temperatures converged at [ ]'** indicadng very low or no flow. After a slow increase, short-tube midpoint temperatures rose sharply b f [ ]'6' fell back, rose by [ ]'** then fell back. The pulsas coincided with 6e reflooding of CMT-l'and CMT-2. When each CMT reflooded, water was drawn into the CMT by the collapse of the trapped steam bubble. The PRHR HX re.: named above [ ]'6' until [ ]'6' At[ ]'6' the outlet temperature increased to [ ]'b' and remained at this level until the end of the test. Temperature patterns confirmed flow measurements; i.e., the flow meter oscillated about [ ]'b' and then indicated high flows for several [ nhx In summary, the PRHR HX functioned in natural circulation liquid flow for the initial [
]'** until the transition from recirculation to draindown in the SG. Steam from HL-2 was then condensed, either in the PRHR HX (up to [ ]'6' ) or in the piping. Because of small (or negative) pressure differences, the HX remained partially filled without draining after the transition from recirculation to draindown. Flow through the HX was unstable with steam alternately flowing through the short and long PRHR HX tubes. Steam flowed preferentially through the short tubes during the latter part of the test. Since flow rates through the PRHR HX were small, the effect of this system on the overall test facility response was minimal.
Steam Generators Feedwater to both SGs was shut off [ ]'** after break initiation. Secondary steam flow decreased to [ ]6" (Figure 5.2.1-40) as the secondary-side was isolated. The brief indication of steam flow in SG-1 between [ ]'"' appeared to be an instrumentation anomaly, since the steam header flow meter did not measure any steam flow during this time. Levels in representative short and long tubes of the SGs (Figures 5.2.1-15, 5.2.1-15x, 5.2.1-16, and 5.2.1-16x) were measured by pressure differentials. As the levels in the SG channel heads decreased, o:\l536RevN536w-6.non:lb-081798 5.2.1-16 REVISION 1
p I FINAL DATA Rzroar l p the tubes of the SG drained. The times shown in the Sequence of Events, Table 5.2.1-3 were derived from these figures. , Level measurements were inaccurate after the tubes emptied, because water in the reference high-pressure legs of the differential level transmitters was vaporized by heat from the secondary side of the SGs. This generic problem is discussed in detail in Subsection 2.4. The level data incorrectly indicated that the SGs refilled with water shortly after the tubes emptied. This did not occur because temperature measurements (Figures 5.2.1-41,5.2.1-41x,5.2.1-42, and 5.2.1-42x) indicated that the SG tubes filled with superheated steam. l Liquid level in the cold-leg plenum of each SG fell abruptly at about [ ] for SG-1 and [ ]**# for SG-2 (Figures 5.2.1-17, 5.2.1-17x, 5.2.1-18, and 5.2.1-18x). Later in the test, the SG-2 cold leg started to refill at [ ]'6' because the RNS pump (which started at [ ] ) was pumping about [ ]'6' of water from the IRWST into the RCS. The cold-leg level in SG-1 remained empty, because water from the downcomer tended to flow out the simulated break in CL-3 rather than accumulate in the SG-1 cold-leg plenum. l De SG-2 hot leg plenum level behaved similarly to the cold-leg plenum; however, it emptied about
-[ ] later than the cold-leg plenum. This difference was probably caused by the more rapid loss of coolant from the simulated break in CL-3. The SG-1 hot-leg plenum levels were inaccurate because the range for this instrument was incorrectly calibrated, as discussed in Subsection 2.4. I Fluid temperatures in the tubes (primary side) of the SGs remained at saturation until shortly after drainage (Figures 5.2.1-41 and 5.2.1-42). Steam filled the tubes after drainage and superheated as the RCS continued to depressurize and heat was transferred from the SG secondary coolant structure j (Figures 5.2.1-43 and 5.2.1-44). Pressures in the primary and secondary sides of the SGs were equal from about [ ]'6' (Figure 5.2.1-45). This steam reached thermal equilibrium with j the SG secondary coolant, which were significantly higher in pressure (and temperature) than the primary. Steam in the tubes remained superheated for the remainder of the test (Figures 5.2.1-41x and j 5.2.1-42x), as the SG secondary sides cooled very gradually, reaching a minimum of about [ P' l at test termination. Fluid temperatures in the short tubes of SG-2 decreased sharply to the saturation temperature at times ([ ]'6' ) that correspond to condensation events in the reactor vessel downcomer. These events caused the liquid level in the downcomer to rise, resulting in an increase in the cold-leg levels. De level surges were sufficiently large to fill the SG-1 short tubes briefly during the condensation events at [ ]'6' Level surges consistent with these i are indicated in the channel head level data, with the exception of the transient in SG-1 at I- [ ]'6' Although the temperature decreased, indicating inflow of saturated liquid, there was l no concomitant effect in the cold-leg plenum. This apparent anomaly possibly resulted from a surge of saturated steam rather than liquid for this case, since the plenum was empty. The level surge could
\ have been sufficient to displace the superheated steam with saturated vapor from the cold-leg piping and, thus, would not produce a level indication for the SG cold-leg plenum.
i a:us36aeviush6.non:ius179s 5.2.1-17 REVISION 1 l'
FINAI. DATA REICT Cold Lees and Hot Lees When the RCPs stopped and coasted to no measurable flow in the first [ ]'6# some cold-leg levels indicated an increase and same a decrease (Figures 5.2.1-46 and 5.2.1-46x). The apparent level changes were not real, but were caused by dynamic pressure differentials resulting from flow effects on the pressure measurements because of ap location and/or installation. Therefore, the level indications after the pumps stopped are used as full levels for the purposes of this discussion. The cold legs remained full for the initial [ ] because water supplied by the pressudzer, CVS, and SGs was about equal to break flow. When the pressurizer emptied, however, the levels ic all of the cold legs fell sharply. As the SGs continued to drain, cold-leg levels rose [ ] Injections from the CMTs ar.:1 accumulators did not provide sufficient flow to significantly affect the levels. When the ADS-1 valve opened at [ ]'b' the levels in the cold legs decreased rapidly, and pressurizer refilled (because it was now vented) and depleted the water inventory in the RCS. In addition, the CVS pump shut off at [ ]'6' reducing the water supply to the cold legs. Level indications became erratic, oscillating at short frequencies from belaw 0 to over-range. These were most likely dynamic pressure effects caused by stean. or two-phase flow, and it is probable that the cold-legs were empty from [ ]'6" These differential pressure oscillations were the greatest in CL-3, which contained the simulated break. After [ ]'6# (Figure 5.2.1-46x), cold-leg levels rose slowly to several inches, with the exception of CL-2. Fluid temperature at the reactor flange top and bottom (at the inlet to the RCP) and the saturation temperature based on reactor pressure are compared for CL-1 through CL-4 in Figures 5.2.1-47 through 5.2.1-50x. Temperature patterns for CL-2 through CL-4 are similar and confirm the level indications. From about [ ]'b' thermocouples in the top of the reactor flanges for these three cold legs were at the saturation temperature, indicating the presence of steam at this level (i.e.,1 1/2 in. from the top of the pipe). The lower thermocouple remained below the saturation temperature demonstrating that these thermocouples (located 1/2 in. from the flange bottom) were covered with water. Temperatures in CL-2 through CL-4 from [ l'6' to test termination (Figures 5.2.1-48, 5.2.1-48x,5.2.1-49,5.2.1-49x,5.2.1-50, and 5.2.1-50x) were below the saturation temperature during this entire period. Temperatures show that these cold legs were filled with subcooled water to a minimum of the height of the uuper thermocouples. Temperatures for CL-1 (Figures 5.2.1-47 and 5.2.1-47x) differ from the other cold legs. The thermocouple at RCP-1 was superheated from about [ ]'6' which could only occur if the vertical leg from the SG to the pump were filled with steam. The superheat decreased at times that coincide with condensation events (i.e., [ ]'6# ), as a result of level fluctuations from the pressure imbalances initiated by this event. The period of low level in CL-1 c:\l 536Revi\l 536w.6.non:l b o81798 5.2.1-18 REVISION 1
__q FINAL DATA RsronT 1
, lasted longer (from about [ ]'6' ) than in the other cold legs. The anomalous
\ behavior of CL-1 may be related to its connection to the same SG channel head as CL-3, the leg with the simulated break. Rapid depiessurization resulted in earlier drainage of SG-1 tubes and, therefore,
' lower liquid inventories after about [ ]'6' (when the tubes of SG-1 emptied).
CL-1 and CL-4 pressures (Figures 5.2.1-20) agreed very closely with reactor pressure during the entire test. , Level data for the SG-2 HL 2 elbow (Figures 5.2.1-12 and 5.2.1-12x) indicated that this level fell several inches during the test, but refilled completely at about [. J' 6' Wide oscillations from [ ]'6' were probably caused either by pressure variations or steam bubbles rising through the liquid. HL-1 and HL-2 liquid levels both decreased after about [ ]'6'similarly to the level behavior in the cold legs. He levels in both hot legs decreased, and the pipes emptied at about [ ] The hot legs refilled partially between [ ]'6' altho @ & level indications oscillated widely as a result of dynamic pressure variations. De hot-leg levels rose slowly at about [ ]'6' and reached full level at about [ ]'6' (Figures 5.2.1-12 arid 5.2.1-12x). HL-1 and HL-2 temperatures at [ ]'6' from the reactor vessel flange top and [ ]'6' from the reactor vessel flange bottom (Figures 5.2.1-13,5.2.1-13x,5.2.1-14, and 5.2.1-14x) followed the saturation temperature based on the reactor vessel for about [ ]'*# indicating the presence of saturated liquid or steam. Between [ ]'6' these temperatures for both hot legs (SC-140, SC-141, TF-142, and TF-143) became subcooled, indicating that the hot legs refilled with water. %ese temperatures agreed with the level measurements that showed refilling of the hot legs during this period. Hot-leg temperatures then followed the saturation temperature until about [ ]'6' when the temperatures subcooled slightly, indicating that the pipes were filled with water. The hot legs refilled initially at [ ]'6' because the reactor vessel had been refilled to the level of the hot legs. He level decreased at about [ ]'** (as indicated by an increase in temperature and a decrease in level indication) because the level in the reactor started to decrease after the accumulators completed injection at [ ]'6' In-Containment Refuellne Water Storane Tank De IRWST injection sy stem was not actuated during this test. Water was supplied to the suction of the RNS pump from this tank. Water in the IRWST also acted as a heat sink for steam vented through ADS 1-3 and heat rejected from the PRHR HX. Figures 5.2.1-51 and 5.2.1-51x show the temperature of the water in the IRWST during this test. o:us36aevmsh6.non:H> 081798 5.2.1-19 REVtSION 1
FINAL DATA RF. PORT Break and ADS Measurement System When the break valve opened, pressure upstream of the break orifice and differential pressure across the break orifice were equal and closely followed reactor vessel pressure (Figure 5.2.1-52). CL-3 fluid discharged from the break hole flashed because of reduced pressure. Two-phase mixture exiting the break hole was piped to the break separator. Liquid flow rose steadily for the first [
]'6" then fell by about one-third from [ ]'6' (Figure 5.2.1-2). This constant flow occurred during the period when RCS pressure had become constant; therefore, the driving force at the break orifice was constant. Steam flow decreased slowly without significant oscillations. Steam flow then declined, mirroring the decrease in pressure upstream of the break. It became so small that it was unmeasurable at [ ]'b' (Figure 5.2.1-3).
Liquid flow decreased sharply at about [ ]'6" and then rose slowly from [
]"' The increase in liquid flow during this period resulted from water injections from the CMTs and accumulators. Liquid flow became constant at about [ ]'6* at sout [ ]' 6" until test termination. Since the RNS pump supplied [ ]'6' of water into the DVI lines, about half the water flowed out of the system through the simulated break and half refilled the system.
Since the ADS-4 valves were not actuated during this test, the ADS-4 separators were not used. Nonsafety-Related Systems The nonsafety-related systems (CVS and RNS) functioned in this test as programmed in the control logic (Figures 5.2.1-8 and 5.2.1-8x). CVS injection started at [ ]'6" and continued at a flow rate of [ ]'6* m [ ]'6* The CVS pump was shut off when the pressurizer low-low level reset. Flow was initiated from the RNS when DVI pressures decreased below RNS pump discharge pressure. This flow started at about [ ]'6" and rose to a total of [ (*6' through each DVI). RNS flow remained constant at these rates until test termination. 5.2.1.6 Mass Balance Mass balance for this test was calculated from water inventories in the components before and after the test. Water mass at the completion of the test agreed with pre-test water mass within [ ]'6* percent. Eetails of this mass balance are provided in Appendix E. 5.2.1.7 Conclusions 'Ihe test was performed with minimal problems and is considered acceptable. Although not all of the facility initial conditions met the specified acceptance criteria, the deviations did not impact the quality oA1536Revi\l536w4.non:lb o81798 5.2.1-20 REVISION 1
..a.. . 4 - . (.- . A .,_ _ _.4 2 - 2__. 4 a m . .. a, _ ... . _ . u . , . . ,..
FINA1, DATA REPORT I (~ of the data. The instrumentation problems encountered were not critical to the performance of the C facility mass and energy balances. Facility response to the test was as anticipated for the conditions that were established. The data clearly demonstrated that cooling of the reactor heater rods was maintained throughout the duration of the test. This test was successfully concluded when the CMTs started to refill. Adequate cooling of the core heaters was achieved during the entire test. The capacity of the RNS was sufficiently large to start refilling the system even with a simulated 2-in. break. I 1 I ^(h L) 1 i O v c:us36aevius36w-6.non: stat 79s 5.2.1-21 REVISION 1
FINAL DATA REPORT TABLE 5.2.I.1 MATRIX TEST SB04 INITIAL CONDITIONS . Instrument Specified Initial Actual Initial Condition Condition Comments Parameter No. Pressurizer pressure") FT-6M 370 2 2 psig HL-1 temperature
- SC-141 420 2'F HL-2 temperature
- SC-140 420 2'F SG-1 pressure") FT-301 28515 psig SG-2 pressure"' FT-302 28515 psig 65 5in. Izvel signal was Pressurizer level"' LDP-601 temperature- compensated by TF-605 2613 in. level signal was SG-1 narrow-range LDP-303 temperature- compensated level
- by 17-301 26 3 in. Level signal was SG-2 narrow-range LDP-3N temperature- compensated level"'
by TF-310 1RWST temperature
- TF-709 < 80'F CMT-1 temperature
- TF-529 < 80*F CMT-2 temperature
- TF-532 < 80*F ACC-1 temperature"' TF-403 < 80*F ACC-2 temperature
- TF-4N < 80*F IRWST level
- LDP-701 Level established by fill-line elevation ACC-1 level"3' LDP-401 Level established by standpipe at 37 in.
ACC-2 level J' LDP-402 Level established by standpipe at 37 in. - ACC-1 pressure
- FT-401 23212 psig ACC-2 pressure
- FT-402 232 2 2 psig O
REVISION 1 c:\l 536Revi\l 536w-6.non:I b-081798 5.2.1-22
FINAt. DATA REroni (" TABLE 5.2.I I (Continued) MATRIX TEST SB04 INITLAL CONDITIONS Instrument Specified Initial Actual Initial Parameter No. Condition Condition Comments CMT-1 level
- LDP-507 Full
- CMT-2 level
- LDP-502 Full _ _
Neis: (1) Data for the indicated parameter were recorded in the test procedure as an initial condition for the test. The value was determined by the test engineer from the appropriate control board indicator. (2) Data were not recorded in procedure, but the test engineer verified that specified conditions were achieved while establishing initial conditions. The value of the parameter was determined post-test by calculating the average DAS indication for a time of about 2 minutes before the break valve opened. (3) The bourdon pressure tube local indicator (PI-401 or PI-402) was tubed to the lower portion of the reference leg of the accumulator level transmitter (LDP-401 or LDP-402). As pressure in the accumulator was increased, air inside the bourdon tube was compressed, thereby lowering the reference leg liquid level, resulting in a false indication of measured level. s r u oA1536Revi\l536w-6.non:lball798 5.2.1-23 REVISION 1
1 FINAL DATA REPORT i TABLE 5.2.12 MATRIX TEST SB04 INOPERABLE INSTRUMENTS / INVALID DATA CilANNELS Instrument No. Instrument Type Description of Problem FMM-20l
- Magnetic flow meter Removed from system FMM-202* Magnetic flow Meter Removed from system FMM-203* Magnetic flow meter Removed from system FMM-204* Magnetic flow meter Removed from system FMM 502* Magnetic flow meter Data invalid after 800 seconds because of steam in balance line FMM-503* Magnetic flow meter Dcta invalid after 800 seconds because of steam in balance line FMM-802* Magnetic flow meter Data invalid after steam formed in PRHR HX inlet line at about 120 seconds FMM-804* Magnetic flow meter Data valid until PRHR HX initially drained at 580 seconds; after which the possibility of steam in the outlet line invalidated the data FVM-602 Vortex flow meter Ignore data since ADS-4 did not actuate FVM-603 Vortex flow meter Ignore date since ADS-4 did not actuate HFM-103 Heat flux meter Failed HFM-105 Heat flux meter Failed I'FM-I l2 Heat flux meter Failed l HFM 201 Heat flux meter Failed l
HFM-505 Heat flux meter Data appear erratic HFM-601 Heat flux meter Failed HFM-902 Heat flux meter Data appear erratic HPS-203-1 Heated phase switch Inoperable throughout test through HPS 203-3 HPS-509-1 Heated phase switch Inoperable throughout test through HPS-509-3 LDP-201 Differential pressure Data invalid due to effect of vertical portion of sense line LDP-202 transmitter - level attached to top of pipe; data can show level trends when the LDP-203 pipe is empty or starts to drain, but absolute level indication LDP-204 can not be used LDP-205 LDP-206 o:\t 536Rev l\l 536w-6.non: I b-081798 5.2.1-24 REVISION 1
FINAt. DATA REPOa7 TABLE 5.2.12 (Continued) MATRIX TEST SB04 INOPERABLE INSTRUMENTS / INVALID DATA CHANNELS lastrument No. Instrument Type Description of Problem ! i LDP-207 Differential pressure Inoperable; ranged improperly; data can show trends, but through transminer - level l absolute level indication can not be used LDP-209 LDP-215* Differential pressure Data invalid when tube drained and reference leg started to LDP-216 transmitter - level vaporize LDP-217 LDP-218* , LDP-219' LDP-220 LDP-221 LDP-222* LDP-40l* Differential pressure Data invalid; see note 3 of Table 5.2.1-2 LDP-402* transmitter - level LDP-608 Differential pressure Over-ranged at 625 to 800 seconds transmitter - level LDP-804 ' Differential pressure Data valid until PRHR HX initially drained at 580 seconds; 0 FT-101 transmitter - level Pressure transmitter after which data are suspect because of possible vaporization of the common reference line Data less than 6.1 psig invalid PT-102 Pressure transmitter Data less than 6.2 psig invalid PT 103 Pressure transmitter Data less than 6.2 psig invalid FT-104 Pressure transmitter Data less than 6.4 psig invalid PT-108 Pressure transmitter Data less than 8.4 psig invalid FT 109 Pressure transmitter Data less than 6.3 psig invalid PT-ill Pressure transmitter Data less than 6.0 psig invalid FT-ll2 Pressure transm;tter Data less than 8.8 psig invalid FT-ll3 Pressure transmitter Data less than 6.4 psig invalid FT-20l* Pressure transmitter Data less than 1.1 psig invalid FT-202 Pressure transmitter Data less than 5.9 psig invalid FT-205 Pressure transmitter Data less than 6.1 psig invalid PT 801 Pressure transmitter Not reliable after 475 seconds TF-170 Thennocouple fluid Read low throughout test temperature
\.
o:usmaeviusw6.n=cib.os 79s 5.2.1-25 REVISION 1
FINAI, DATA REPORT TABLE 5.2.I 2 (Continued) MATRIX TEST SB04 INOPERABLE INSTRUMENTS / INVALID DATA CHANNELS Instrument No. Instrument Type Description of Problem TF-203 Thermocouple fluid Read low throughout test temperature TF-50l
- Rermocouple fluid Read low throughout test temperature TF-504
- nermocouple fluid Inoperabley indicated ambient temperature throughout test temperature TFM-703 Thermocouple for HFM-703 Inoperable; indicated ambient temperature throughout test i TH-317-1 Thermocouple heater rod Inoperable; heater rod removed prior to test i through TH-317-4 TW-503 Thermocouple wall Inoperable throughout test temperature 7W-534 Thermocouple wall Inoperable; indicated ambient throughout test temperature TW-552 Thermocouple wall Inoperable; indicated ambient throughout test temperature Note:
- Instruments marked with an asterisk are critical instruments. See Subsection 5.2.1.2 for discussion.
l l l t l l l l o:\l 536Rev i\l 536w-6.non: l b-081798 5.2.1-26 REVISION 1 i
j FINAL DATA REPORT l ? TABLE 5.2.13 MATRIX TEST SB04 SEQUENCE OF EVENTS Time After Break Event
- Data Source * (sec.)
TEST Pushbutton Depressed D
- Break Valve Open Signal D i
l Break Valve Starts To Open D Feed Pump Trips D CMT-1 Outlet Valve Starts To Open D CMT-2 Outlet Valve Starts To Open D PRHR HX Outlet Valve Starts To Open D Reactor Coolant Pumps Trip D CVS Pump Injection Starts (FMM-801) A ' Pressurizer Empty (LDP-601) A HL-1 Pipe Starts To Drain (LDP-205) A SG-1 Hot-Leg Elbow Starts Draining (LDP-207) A
\
Pressurizer Surge Line Empty (LDP-602) A SG-1 Cold-Leg Short Tube Empty (LDP-221) A SG-1 Cold-Leg Long Tube Empty (LDP-219) A SG-2 Cold-Leg Short Tube Empty (LDP-220) A SG-1 Hot-Leg Short Tube Empty (LDP-217) A SG-1 Hot-Leg Long Tube Empty (LDP-215) A CL-3 Channel Heed Empty (LDP-213) A SG-2 Cold-Leg Long Tube Empty (LDP-222) A CL-1 Channel Head Empty (LDP-211) A CMT-1 Recirculation Flow Stops (LDP-509) A CMT-2 Recirculation Flow Stops (LDP-510) A SG-2 Hot-Leg Short Tube Empty (LDP-216) A SG-2 Hot-Leg Long Tube Empty (LDP-218) A
\
oA15xneviush4.non lb481798 5.2.1-27 REVISION 1
FINAL DATA REPORT TABLE 5.2.13 (Continued) MATRIX TEST SB04 SEQUENCE OF EVENTS Time After Break Event
- Data Source * (sec.)
'A' HL-2 Pipe Starts To Drain (LDP-206T A CL-2 Channel Head Empty (LDP-210) A CL-4 Channel Head Empty (LDP-22) A Time Of Minimum Reactor Level Observed During Test A (LDP-127)
ACC-1 Injection Starts (FMM-401) A ACC-2 Injection Starts (FMM-402) A RNS Pump Injection Starts (FMM 803) A CMT-1 Low Level Signal D ADS-1 Valve Starts To Open D Pressurizer Refloods (LDP-601) A HL-2 Pipe Empty (LDP-206) A SG-1 Hot-Leg Elbow Minimum (LDP-207) A ADS 2 Valve Starts To Open D SG-2 Hot-Leg Channel Head Empty (LDP-214) A SG-1 Hot-Leg Elbow Minimum (LDP-208) A ADS-3 Valve Starts To Open D Reactor Pressure Low D IRWST-2 Injection Valve Starts To Open D IRWST-1 Injection Valve Starts To Open D ACC-1 Empty (LDP-401) A ACC-2 Empty (LDP-402) A CMT-1 Starts To Reflood (LDP-507) A CMT-2 Starts To Reflood (LDP-502) A _ _ l Note: (1) Data from the instrument channel in parenthesis were used to determine level, flow, or pressure conditions. (2) D = time data obtained from a software program that monitored the input and output of the facility's PLC. A = time data obtained by reviewing data from the instrument channel listed in the Event Description column. o:\l 536Rev l\l 536w4.non:I b-081798 5.2.1-28 REVISION 1
. . . . . - . - . . . . . ~ . . . - . . ~ . . . ~ . . . . . . . . - ~ - . . . - - . - -
FINAL DATA REPO2T t l The Bar Charts for Table 5.2.13 on pages 5.2.1-29 through 5.2.132 are not included in this nonproprietary document. t l l I 1 I O j l 1 I l l oA1536Revi\l536w4.nmutb-081798 5.2.1 29 REVISION 1
FINA1, DATA Rs.ront O I t m h I lg ! $ k E f i l i i
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8, g~'
, n 3- r 3- r s p , . . .x.+ _p.w... < 4_
s g 1 'F T f gg;, I O Figure 5.2.1 la Primary Loop and Break Piping Layout (Sh.1 of 2) o:\l536Revl\l536w4.non:Ib 061798 5.2.1-33 REVISION l
FINAL DATA REroRT l O 3n-nt O D, c'
~
_/ L i w __ n TS-201 PT ! 203 f y PLUG it y I
- RV2E7 s CL DP
\
215 ,, , _f I J{ T 1b / _ _ _ _ _ _ _ _ . m
\ l RV237 f
)!s'L TS-205 i
' O Figure 5.2.1-1b Primary Loop and Break Pipe Arrangement (Sh. 2 of 2) o:\l 536Rev i\l 536w-6.non: I b-082598 5.2.1-34 REVISION 1
. - . . . .. - - ..- - - . ... . - - - . . . . . . . . . - . . _ - . = - . . - . - . - . - - . . -
FINAL DATA REPORT t I l t l Figures 5.2.1-2 through 5.2.2-53x on pages 5.2.1-35 through 5.2.1-121 are not included in this ! g nonproprietary document. 1 I l I l t-i I i \ l [ I .O i i ? i i l I t 1 i c:\l536Revl\l536w-6.non:Ib 082898 5.2.1-35 REVISION 1 i
. - . - - . - . - - - - - . . - . ~_ - . - . - - -. - . - - . . - ~ - -
l l FmAI. DATA REPC;:T 5.2.2 Effect of a Smaller Break Size (Matrix Test SB24 Comparison with Matrix Test SB04) ; I In this section, the results of Matrix Test SB24 (OSU Test U0024) are compared with those of Matrix ; l Test SB04 (OSU Test UO204). These two tests were identical, except that the break simulated in SB24 was a 1/2-in. pipe break and the break simulated in SB04 was a 2-in. pipe break. Both simulated breaks were located at the bottom of CL-3, and the nonsafety-related systems were operating I l in accordance with the preprogrammed control logic for the AP600. In both tests, failure of one leg of one ADS-4 line was simulated. De transient began with the opening of the valve iso'lating the simulated break and continued through refilling of the system through the RNS. The accumulators injected when DVI pressure decreased below the nitrogen pressure in the accumulators. He ADS was not actuated in this test because j l pressure did not reach the setpoints for actuation. He test was terminated at about 9500 seconds (about 2-1/2 hours) after refilling, and a constant rate of temperature / pressure decline haa been demonstrated. Performance of this test was successful because the core heater rods remained covered with water, and their temperatures remained below or close to the water saturation temperature. l l Subsection 5.2.2.1 provides details related to system configuration and initial conditions. A l description of inoperable instruments appears in Subsection 5.2.2.2, and Subsection 5.2.2.3 references the sequence of events. Subsection 5.2.2.4 describes the test results and evaluation. A comparison of ! O component responses appears in Subsection 5.2.2.5, and a summary of mass balance results appear in j O Subsection 5.2.2.6. He conclusions, as they apply to Matrix Tests SB24 and SB04, appear in Subsection 5.2.2.7. Data plots, (referenced in text as figures), for this section have the same numbers after the section identification numbers as Matrix Test SB04 to facilitate the comparison of data for these two tests. 5.2.2.1 System Configuration and Initial Conditions De test was performed in accordance with an approved written test procedure. There were no special or unique requirements for the test other than those specified in the initial conditions in Table 5.2.2-1. De specified conditions were checked on the control board prior to test implementation. j The test facility was configured in the normal arrangement (described in Section 2) and was identical [ to that of Matrix Test SB04, except for the size of the simulated break. I De appropriate prerequisites were completed, and the initial conditions were satisfied. De required break simulation piping and break instrumentation were installed per Dwg. OSU 600904 and the break ! piping layouts in Figures 5.2.2-la and 5.2.2-lb. A break hole ([ ]*) simulating a 1/2-in, cold-leg break was installed in the bottom of the pipe break spool in CL-3. A 50-percent flow nozzle was installed in the ADS 4-1 line (on HL-1) and a 100-percent flow nozzle was installed in the ADS 4-2 line (on HL-2) to provide the assumed single failure. Flow nozzles that simulate two lines were installed in the ADS 1-3 piping. f oAishnevnish.it.non: b-osi79s 5.2.2-1 REVISION 1 I
l FINAL DATA REPORT 1 ne heater rod power was adjusted prior to break initiation in order to achieve the required hot-leg temperatures. At initiation of the break, the bundle power was set at 600 kW. Power to the heater md bundle was maintained at 600 kW for 140 seconds after the start of the test. After 140 seconds, rod bundle power followed an exponential decay curve. He actual power decay curves are provided in data plots in Appendix F. He differences between the actual and specified power decay are considered acceptable. Pressurizer power was terminated at break initiation. Testing was initiated when test facility conditions, as read from the test facility control board, agreed with the specified initial conditions within acceptable tolerances. All actions were automatic and required no operator action. Table 5.2.2-1 provides a comparison of the specified and actual initial conditions for Matrix Test SB24. He values in this table were averaged over approximately 2 minutes preceding the test. Test initial conditions were achieved for SG pressure, pressurizer pressure, pressurizer level, SG-1 narrow-range level, and SG-2 narrow-range level. Test initial conditions for the hot-leg temperature were found to be acceptable, with the test results not adversely affected. These measurements outside the tolerance band were accepted, since the deviations were insignificant. 5.2.2.2 Inoperable Instruments Table 5.2.2-2 is a list of the instruments considered inoperable or invalid during all or portions of this test. Some of the instruments listed are on the Critical Instrument list (Subsection 3.2, Table 3.2-2) and, therefore, are addressed here. FMM-201, FMM-202, FMM-203, and FMM-204 measured flow (gpm) in each of the four cold legs. A decision was made to continue testing without the availability of these instruments. Replacement flow meters repeatedly failed; their continued use was precluded due to cracking of the ceramic liners from thermal stratification in the loop piping. The necessary boundary conditions for loop flow could be determined from DP-202, DP-203, DP-205, and DP-2%. TF-103 and TF-104 measured CL-3 and CL-4 bottom-of-pipe fluid temperatures entering the reactor downcomer. Both thermocouples were removed to accommodate installation of thermal stratification measurement instrumentation. Both thermocouples were allowed to be inoperable as long as TF-101 and TF-102 were operable. TF-101 and TF-102 were operable during the performance of Matrix Test SB24. FT-201 measured RCS pressure at the top of the 50-1 fong tube. On August 15,1994, it was discovered that the transmitter had an incorrect zero compensation, which resulted in a negative error and negative data at low pressures. He transmitter zero was corrected at that time. PT-201 data obtained during Matrix Test SB24 had the zero correction performed, and the corrected dath appear as PT_201. Since the pressure remained positive during this test, corrected pressure data for PT_201 are valid. o:\l 5Mw Rev i\l 5%w.11.nort i t>O81798 5.2.2-2 REVISION 1
(. FINAL DATA REPORT f TF-504 measured CMT fluid temperature from the long thermocouple rod location near the bottom of ( each CMT. This thermocoaple appears to have measured ambient conditions throughout the test, which would indicate a short somewhere in the thermocouple wiring. With this thermocouple inoperable, the required long thermocouple rod thermocouple availability of "seven out of ten and no ! more than one in succession failed" was met. 1 ? i l Data provided by ADS-4 separator instrumentation were invalid because of the closed position of the I ADS-4 valves and the ADS-4 separator loop seal valves during this entire test. The instruments ' affected are: FMM-602, FMM-603, FVM-602, FVM-603, LDP-611, and LDP-612. Test analysis will l not be affected, since ADS-4 flow did not begin until the valves opened. l Considering these critical instrument failures, sufficient instrumentation was available to allow the I performance of mass balances as demonstrated in Subsection 5.1.1.6 and Appendix E. An energy balance will be performed and reported in the AP600 La-Pressure Integral Systems Test at Oregon i State University Test Analysis Report, WCAP-14292.* l 5.2.2.3 Sequence of Events The sequence of events for Matrix Tests SB24 and SB04 are compared in Table 5.2.2-3 and associated bar graphs based on the chronologic order of event occurrence in Matrix Test SB24. Since the Matrix Test SB24 break area was about [ ]*** of the area of the Matrix Test SB04 break, the rate of flow through the Matrix Test SB24 simulated break was much lower, and, hence, the rate of depressurization was significantly slower. The CVS pump started to inject war at [ ]'6' compared with [ ]'6" for Matrix Test SB04. The accumulators injected war ato the DVI lines at [ ]'6' when DVI pressure fell below the nitrogen pressure. Since the reactor vessel level did not fall sufficiently to uncover the hot legs or cold legs, those lines anu 'he SG tubes remained filled with water. The pressurizer emptied at [ l'6* for Matrix Test SB04, and reflooded at [ ]'b' ADS 1-3 and ADS-4 never ac' Ted because the setpoints for these components were never reached. Cold water circulated from the CMTs by natural circulation. Because the transition from recirculation to draindown did not occur, they continued to provide natural circulation flow. De RNS pump injected water at about [ ] when the pressure in the SG-2 channel head reached RNS pump discharge pressure. De test was terminated at [ ]'b' when steady temperature and pressure decreases had been demonstrated. 5.2.2.4 Test Results and Evaluation Since the ADS was not actuated during this test, the discussion of the test results will not be separated into phases, as was the case for Matrix Test SB04. In general, the slow rate of coolant loss and, therefore, slow depressurization permitted the CVS and RNS to maintain the system at a nearly full condition during the entire transient. He accuraulators injected when their pressures became equal to oA15%wRevnl5%w ll.non:lb-0Bl?98 5.2.2-3 REVISION 1
FINAL DATA REPORT system pressure and the pressurizer emptied and refilled; however, because the rate of depressurization was so much lower, the time scale for these events was thousands of seconds compared with hundreds of seconds in Matrix Test SB04. Specific comparisons of the key events for Matrix Tests SB24 and SB04 are discussed in this section. Liquid flow (Figure 5.2.2-2) from the break separator increased to a peak of slightly less than [
]'6" and then remained between [ ]'6' until the RNS pump was shut off at [ j'6' Comparable barak separator liquid flow rates for Matrix Test SB04 ranged up to a maximum of [ ]'6' and down to a minimum of [ ]'6' with a steady-state rate of [ ]'6' Also, significant steam flows ([ ]'6') were measured for Matrix Test SB04. Steam flow was so low that it could not be measured in Matrix Test SB24 (Figure 5.2.2-3). Steam flow was negligible in Matrix Test SB24 except for the initial 10 seconds after break initiation.
Reactor and DVI pressure (Figure 5.2.2-5) fell much more slowly in Matrix Test SB24 compared with Matrix Test SB04 because of the lower break flow in Matrix Test SB24. In Matrix Test SB04, the RCS reached atmospheric pressure at about [ ]'6' The pressure was still about [ ]'6' when Matrix Test SB24 was terminated at [ ]'6' The level indications for the upper plenum and upper head (Figure 5.2.2-9) showed that steam formed only in the upper head of Matrix Test SB24. The water level gradually decreased in the upper head until the accumulators injected (about [ ]'6") and the RNS pump started (about [ ]'6'), then the level increased slowly to near the top of the reactor vessel. The maximum estimated percent steam in the upper part of the core (LDP-110; Figure 5.2.2-9) of [ ]'6" percent was reached [ ]'6' after break initiation. The steam content gradually l decreased as cold water was injected into the downcomer until it became negligible at about [ j'6" This steam content compares with [ ]*** percent in Matrix Test SB04, as would be ! expected due to the smaller simulated break. Steam percent was estimated from the equation below: Steam percent = level (100% water) -level (with steam) x 100 level (100% water) Coolant in the core remained subcooled in Matrix Test SB24, except for the initial [
]'6' when the coolant was at saturation (Figure 5.2.2-11). In Matrix Test SB04, coolant in the upper core remained at saturation until the RNS actuated at [ ]'6' then returned to saturation at about [ ]'6' and remained saturated until the end of the test. Therefore, l Matrix Test SB24, the core was cooled after the initial [ ]'6' h si@-@ase wa Ris differed from Matrix Test SB04, in which steam was produced in the upper portion of the core for j most of the test. i o A1536w Rev l\l 536w-l l .norr ib o81798 5.2.2 4 REVISION 1
l FINAL DATA REPORT As flow continued through the simulated break, cold water flowed by natural circulation from the s CMTs. Water injected from the pressurizer until it emptied (at about [ j'6*) into HL-2 (Figum 5.2.2-7). He pressurizer in Matrix Test SB04 emptied in about [ ]'6' because of the higher differentiol pressure between the pressurizer and the large break in CL-3. The accumulators injected at about [ ]'6" in Matrix Test SB24 (Figure 5.2.2-19), again, much later than the accumulators in Matrix Test SB04 because of the slow depressurization. l De CVS pump in Matrix Test SB24 started at [ ]'6' when pressurizer instrumentation reached the low low level setpoint, and it shut off at about [ ]'6# when the pressurizer reached normal level (Figure 5.2.2-8). He comparable Matrix Test SB04 starting time was [
]'6"its termination time was [ ]'6# Similarly, RNS injection began at about [
]'6# for Matrix Test SB24, compued with [ ]'** for Matrix Test SB04. The longer times for both CVS and RNS actuation resulted from the slower depressurization rate of Matrix Test SB24.
l ~ The system continued to refill, since RNS flow was greater than the inventory loss through the simulated break. At about [ ]'6' the pressurizer reflooded (Figure 5.2.2-7). Since the i system had refilled and reached a rate of steady decline in temperature and pressure, the test was l terminateo ..; [ ]'6# j 5.2.2.5 Comparison of Component Responses ' Component responses of Matrix Tests SB24 and SB04 are compared in this section. Those components with similar behaviors are not discussed; however, Matrix Test SB24 data for all of the components are provided in the figures. Components with dissimilar behaviors, discussed below, were:
- Reactor
- CMTs
- Pressurizer e Cold legs and hot legs
- SGs
- PRHR HX Reactor In Matrix Test SB24, the level in th .spper head (LDP-115) reached a minimum of [ ]'6# below its full level at about [ ]'b' (Figure 5.2.2-9). De level began to rise when the accumulators injected at [ ' ]'6' and continued to increase when the RNS pump started at
[ -]'6' De level in the upper head during Matrix Test SB04 fell to the upper support
'O plate before the level started to refill at [ ]'b' when its accumulators injected and refilled completely after the RNS started at [ ]'6' a:uswnevlum.ll.natB-081798 5.2.2-5 REVISION 1 er em mir rw w y,-., ,m7 v,--W-^'
FINA1. DATA Rr.ront The downcomer (Figure 5.2.2 4) remained full during Matrix Test SB24; therefore, there was no steam flow from the upper head through the holes in the top downcomer plate as shown by the upper head pressure difference (DP-130) (Figure 5.2.2-24). Since there was no steam in the downcomer, there were no condensation events in Matrix Test SB24. Steam fonnation in the reactor vessel, and core Duid and heater temperatures were discussed in detail in Subsection 5.2.2.4 and, therefore, are not repeated here. l In summary, the reactor level remained close to the top of the upper head during the entire transient and refilled to the top of the reactor vessel after the RNS pump started. A small quantity of steam was produced in the core in the first [ ]'6' then, the core was cooled by subcooled water injected initially by the accumulators and CVS and later by the RNS. Core Makeup Tanks The CMTs in Matrix Test SB24 did not transition to draindown because their balance lines remained l filled with liquid as a result of low How through the simulated break. The CMTs remained filled with liquid during the entire test (Figure 5.2.2-23). Natural circulati on flow (Figure 5.2.2-31) was driven by the density difference between the colder water in the CMTs and the water in the system. As hot water from the cold leg enten:d the CMT, water temperature in the tanks rose, decreasing the density and, hence, the flow rate. At about [ ]'6' flow through CMT-1 halted because the density driving force became inadequate to provide the necessary driving head. Flow from CMT-1 stopped at about [ ]'6' because of this density effect combined with flow from the RNS pump. Cold Legs and Hot Lees The hot and cold legs in Matrix Test SB24 remained filled (Figures 5.2.2-12 and 5.2.2-46) with subcooled water (Figures 5.2.2-47 through 5.2.2-50) compared with Matrix Test SB04 in which these legs drained and filled with superheated steam. CL-2 and CL-1 coolant temperatures were cooler than those in CL-3 and CL-4. CL-2 was cooler because cold water from the CVS was injected in the SG-2 channel head; the cooler temperatures in CL 1 may have been the result of heat losses to the environment by the stagnant coolant in this line. Steam Generators The response of the SGs in Matrix Test SB24 differed significantly from that in Matrix Test SB04. Since the loss of inventory through the simulated break was low enough that coolant makeup from the CVS and accumulators kept the hot and cold legs full, water in the SGs did not drain from either the channel heads (Figures 5.2.2-17 and 5.2.2-18) or the tubes (Figures 5.2.2-15 and 5.2.2-16). Pressure I in the primary side of the SGs remained higher than secondary pressure throughout the test (Figure 5.2.2-45). l l o:\l 536w Rev l\l 536w. l l .non: l t>081798 5.2.2-6 REVISION 1 l
FINAI. DATA Rt.ronT The temperature of the water in the tubes was about [ ]'6" below the saturation temperature based on stactor pressure during the initial [ ] (Figures 5.2.2-43 and 5.2.2-44). Later, water in the tubes became even more subcooled relative to the reactor pressure. This confirmed that RCS pressure was being maintained by the saturation pressure in the reactor vessel. The temperamres of the water in the tubes and the water in the shell of each SG were nearly identical, indicating that thermal quilibrium was reached, and each SG was slowly cooling through heat losses to the environment. Only oc thermocouple, TF-218, located at the top of the long tube in SG-2, deviated. This thermocouple Indicated a higher temperature than either the shell side fluid ! temperatures or the other water temperatures in the tubes. One possible explanation would be that the level in the shell side had decreased sufficiently so that the top of the tube was uncovered and its cooling was delayed relative to the rest of the tubes. Passive Residual Heat Removal Heat Exchanner PRHR provided natural circulation cooling throughout Matrix Test SB24, as shown by continuous flow (Figure 5.2.2-37), full level in the inlet header and PRHR HX (Figure 5.2.2-38), and temperature distribution in the PRHF. HX (Figure 5.2.2-39). Flow decreased from the initial rate of about [ ]'6" at the end of the test because the inlet temperature decreased as RCS temperature declined. His reduced the density difference across the PRHR HX and, hence, the pressure ! differential for natural circulation flow. Although the IRWST temperature rose, the outlet temperature of the PRHR HX was not significantly affected. De PRHR HX provided natural circulation cooling during the entire Matrix Test SB24 because inventory loss through the simulated break was slow enough that the accumulator and nonsafety-related system injections were sufficient to keep the system full so that natural circulation through the PRHR HX was maintained. 5.2.2.6 Mass Balance ne mass balance for Matrix Test SB24 was calculated from the water inventory at the start and conclusion of the test. De final water inventory agreed with the initial inventory within [ ]** percent. Details of the individual component inventories and their mass balance analysis are provided in Appendix E. 5.2.2.7 Conclusions he test was performed with minimal problems and is considered acceptable. Although not all of the facility initial conditions met the specified acceptance criteria, the deviations did not impact the quality of the data. De instrumentation problems encountered were not critical to the performance of the facility mass and energy balances. l l o:\l5hRevlu5h-ll.non:tb 081798 5.2.2-7 REVISION 1
FINAL, DATA REPORT l Facility response to the test was as anticipated for the conditions that were established. The data clearly demonstrated that cooling of the reactor heater rods was maintained throughout the duration of the test. Matrix Test SB24 was significantly less stressful to the plant because the 1/2-in. simulated break limited inventory loss. With this slow rate of loss, the CVS and RNS were able to keep the hot legs, cold legs, and SGs full of water. The level decreased in the reactor about [ ]'6' and then refilled when the accumulators injected and the RNS actuated. The CMTs did not inject in this test because the balance line remained full. However, natural circulation continued through the CMTs through the test. The core remained covered with water, and the core heater rods were cooled by subcooled water during almost the entire test. Some steam was produced in the core during the initial 60 to 70 seconds; however, beyond this initial period, core heat was removed by single-phase subcooled water. l Core heat was removed by the nonsafety-related systems' cold water injection, natural circulation through the CMTs, and natural circulation in the CMTs. There were no condensation events during Matrix Test SB24. Since the downcomer remained full of water, there was no steam volume to subcool and suddenly condense as in tests with larger simulated breaks. O l 4 O oA15%wRe M15%w.ll.non:lts481798 5.2.2-8 REVISION 1
i FINAt. DATA Rzroar TABLE 5.2.2 I MATRIX TEST SB24 INITIAL CONDITIONS Instrunnent Speelfled Initial Actual Initial Parasmeter No. Condition Condition comunents
- - a' Pressurizer FT-604 370
- 2 psig pressure"'
HL-1 SC-141 420 2 2'F temperature"' HL 2 SC-140 420 2*F temperaturen) SG-1 pressure"' FT-301 285 t 5 psig SG-2 pressure") FT302 285 5 psig Pressurizer LDP 601 65 5in. Level signal was level"' temperature-compensated by TF-605 SG-1 na: Tow- LDP-303 26 3in. Level signal was range level"' temperature-compensated by TF-301 SG-2 narrow- L"P-304 26 2 3 in. Level signal was range level") temperature-compensated by TF-310 IRWST TF-709 < 80*F temperature"' CMT-1 TF-529 < 80'F temperature
- CMT-2 TF-532 < 80'F Accepted temperature
- ACC-1 TF-493 < 80'F temperature
- ACC-2 TF-404 < 80*F temperature
- IRWST level
- LDP-701 Level established by fill-line elevation - -
\ o:\l5hRevi\l5h-II.non:Ib-061798 5.2.2-9 REVISION I
l l FINAt. DATA REPORT l TABLE 5.2.21 (Continued) l MATRIX TEST SB24 INITIAL CONDITIONS Instrument Specified Initial Actual Initial Parameter No. Condition Condition Comments
===, ==== Lb.C ACC-1 level'" LDP-401 Level established by standpipe at 37 in.
ACC-2 level'" LDP-402 Level established by standpipe at 37 in. ACC-1 Irr-401 232 2 2 psig Pressure was [ t2 pressure > jan" low; condition acceptable ACC-2 PT-402 232 2 psig Pressure was [ r2 pressure > jmn' low; condition acceptable CMT-1 levele2> Full LDP-507 CMT-2 level Full LDP-502 _ _ Note: (1) Data for the indicated parameter were recorded in the test procedure as an init:J condition for the test. The value was determined by the test engineer from the appropriate control board indicator. (2) Data were not recorded in procedure, but the test engineer verified that specified conditions were achieved while establishing initial conditions. The value of the parameten was determined post-test by calculating the average DAS indication for a time of about 2 minutes before the bicak valve opened. (3) 'Ihe bourdon pressure tube local indicator (PI-401 or PI-402) was tubed to the lower ponion of the reference leg of the accumulator level transmitter (LDP-401 or LDP-402). As pressure in the accumulator was increased, air inside the bourdon tube was compressed, thereby lowering ti.e reference leg liquid level, resulting in a false indication of measured level, i l l l l l O 1 o:\l 536w Rev l \l 536w. l l .non: I b-081798 5.2.2-10 REVISION 1 l i i
FINAL DATA REPORT TABLE 5.2.2 2 MATRIX TEST SB24 INOPERABLE INSTRUMENTS / INVALID DATA CHANNELS Instrunnent No. Instrument Type Description of Problem FMM-20l
- Magnetic flow meter Removed from system FMM-202* Magnetic flow meter Removed from system FMM-203* Magnetic flow meter Removed from system FMM-204* Magnetic flow meter Removed from system FVM-602 Vortex flow meter Ignore data since ADS-4 did not actuate FVM-603 Vortex flow meter Ignore data since ADS-4 did not actuate HFM-103 Heat flux meter Failed HFM-105 Heat flux meter Failed HFM-112 Heat flux meter Failed HFM-Il3 Heat flux meter Failed HFM-201 Heat flux meter Failed HFM-505 Heat flux meter Data appear erratic Q HFM-510 Heat flux meter Failed HFM-601 Heat flux meter Failed HFM-703 Heat flux meter Failed HPS-203-1 Heated phase switch Inoperable throughout test through HPS-203-3 HPS-509-1 Heated phase switch Inoperable throughout test through HPS-509-3 PT_101 Pressure transmitter Data less than 6.1 psig invalid PT_102 Pressure transmitter Data less than 6.2 psig invalid IFT_103 Pressure transmitter Data less than 6.2 psig invalid S PT_104 Pressure transmitter Data less than 6.4 psig invalid PT_108 Pressure transmitter Data less than 8.4 psig invalid PT_109 Pressure transmitter Data less than 6.3 psig invalid PT_Ill. Pressure transmitter Data less than 6.0 psig invalid e PT_Il2 Pressure transmittei Data less than 8.8 psig invalid PT_ll3 Pressure transmitter Data less than 6.4 psig invalid odl5hRevl\l5h-ll.non:lt>081798 5.2.2-11 REVISION I
FINAL, DATA REroaT TABLE 5.2.2-2 (Continued) MATRIX TEST SB24 INOPERABLE INSTRUMENTS / INVALID DATA CIIANNELS Instrument No. Instrument Type Description of Problem Pressure transmitter Data less than 6.1 psig invalid Fr_101 Pressure transmitter Data less than 1.1 psig invalid I'T_20l* Pressure transmitter Data less than 5.9 psig invalid FT_202 Pressure transmitter Data less than 6.1 psig invalid Irr_205 Pressure Transmitter Data less than 8.4 psig invalid FT_108 TF-103 Thermocouple fluid temperature Removed from system TF 105 Thermocouple fluid temperature Removed from system TF-170 Thermocouple fluid temperature Read low throughout test TF-203 Thermocouple fluid temperature Read low throughout test TF-504 Thermocouple fluid temperature Inoperable; indicated ambient throughout test TFM 103 'Ihermocouple for HFMs inoperable throughout test TFM-105 TFM-ll3 TFM-201 TFM 703 TH-317-1 Thermocouple heater rod Inoperable; heater rod removed prior to test through TH-317-4 TW-503 Thermocouple wall temperature Inoperable throughout test Note: Instruments marked with an asterisk are critical instruments. See Subsection 5.2.2.2 for discussion. l l O l cA1536wRevl\l536w ll.non:lt4)81798 5.2.2-12 REVISION 1 i
FINA1. DATA REPORT Table 5.2.2 3 on pages 5.2.213 through 5.2.215 is not included in this nonproprietary document. .d 4 4 4 4 1 W-c \l 536wRev l\1536w.l l . rat ib-082598 5.2.2-13 REVISION 1
FINAL DATA REPORT O n s g 5 I l 1 g i a i i i i ! ; i R
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O o:\l 536wRev l\l 536w- I I .non: l t>O81798 5.2.2-16 REVISION 1
FINAL DATA REPORT O
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RV237 h F !' i TS-205-4 O Figure 5.2.2-Ib Prhnary Loop and Break Pipe Arrangement (Sh. 2 of 2) oA1536w Rev i\l 536w-l l .noit i b-081798 5.2.2-17 REVISION 1
l FINAL DATA REPORT Figures 5.2.2 2 through 5.2.2-53 on pages 5.2.218 through 5.2.2-69 are not included in this nonproprietary document. 1 i 1 9 oA1536wRevi\f 536w.II22:IM82798 5.2.2-18 REVISION ]
- - . . . . , . . . - . - - . - - - . - . . - . ~ . - . . ~ - - - . . . . - . . _ ~ . _ -- -. -
Fmrt. DATA REPORT ( 5.3 Core Makeup Tank / Cold Leg Balance Line Breaks Matrix Tests SB10 (OSU Test U0110) and SB09 (OSU Test U0009) characterized the thermal-hydraulic phenomena and system response to a break in the horizontal portion of the core makeup tank-1 (CMT-1) balance line Both tests were performed using the same break location Matrix Test SB10 simulated a double-ended guillotine (DEG) break; Matrix Test SB09 simulated a single-ended 2-in. break. Tests covered the entire transient from the initial break through long-term cooling, roughly spanning 22,000 to 29,000 seconds. In both tests, the equipment was configured to simulate failure of one of the two automatic depressurization system (ADS)-4 lines. Results of Matrix Tests SBIO and SB09 appear in Subsections 5.3.1 and 5.3.2, respectively. n/
'w l
l l l 1 n%/ i oA15hRevl\l5h-5.non:lWl798 5.3-1 REVISION 1
FINAL DATA REPORT 5.3.1 Reference Double-Ended Guillotine Line Break (Matrix Test SB10) His section pmvides the thermal-hydraulic test results from Matrix Test SB10 for a DEG CMT-1 balance line break. De simulated break was located between cold leg-3 (CL-3) and the vertical portion of the CMT-1 balance line, and was configured to simulate the failure of one of the two ADS-4 lines. Facility responses to the break are documented by data plots, referenced as figures in text. Matrix Test SB10 is the reference test for CMT-1 balance line breaks. De test performed met the specified initial conditions. Any exceptions to the initial conditions are discussed in Subsection 5.3.1.1. Matrix Test SBIO is considered successful. Cooling of the heater bundle was maintained throughout this event. The transient began from the break and continued through ADS actuation and CMT, accumulator, in-containment refueling water storage tank (IRWST), and primary sump injection. Matrix Test SB10 was performed on June 24 and 25,1994. Test duration was about 8.5 hours. Subsection 5.3.1.1 provides details related to system configuration and initial conditions. Subsection 5.3.1.2 provides the test sequence of events. Subsections 5,3.1.4 and 5.3.1.5 provide an overall summary of the system response and component-specific responses for Matrix Test SB10, respectively. A description of types and effects of inoperable instruments appears in Subsection p 5.3.1.3. A summary of the mass balance results appears in Subsection 5.3.1.6. The conclusions, as V they apply to Matrix Test SB10, appear in Subsection 5.3.1.7. 5.3.1.1 System Configuration and Initial Conditions Test performance followed an approved written test procedure. Initial conditions for Matrix Test SB10 are specified in Table 5.3.1-1. A select set of initial instruments (identified in Table 5.3.1-1) were checked on the control board prior to test implementation. De control board contained provisions to display compensated level indications to confirm initial conditions. Prior to pressurization, the accumulator levels were estabhshed by filling the tanks until a visual check showed water overflowing from the standpipe. This method ensured that accumulator levels were the same for all tests, independent of instrument indication. Additional details relative to the filling of tanks and lines are provided in Subsection 2.7. The required break simulation piping and break instrumentation were installed according to the piping and instrumentation diagram (P&ID) (Appendix G Dwg. OSU 600904) and the break piping layouts in Figures 5.3.1-1 and 5.3.1-2. De ADS 4-1 and ADS 4-2 lines were connected and piped to the 8-in. ADS 4-1 separator. The break spool and insert that simulated a DEG break in the AP600 were installed in the balance line connecting CL-3 and CMT-1. Balance-line flow from CL-3 was directed into the break separator; the remainder of the balance line up to the CMT was connected to the 5-in. O ADS 4-2 separator. A 50-percent flow nozzle was installed in ADS 4-1 on hot leg 1 (HL-1), and a 100-percent flow nozzle was installed in ADS 4-2 on HL-2 to provide the assumed single failure. oA15mmevmsh-sam: b os179s 5.3.1-1 REVISION I L.. . . . . .
FINAL DATA RErcDT Flow nozzles that simulate two lines were installed in the ADS 1-3 nozzles. Break valve TS-202 isolated the CMT-1 side of the break. Break valve TS-203 isolated the CL-3 side of the break (Appendix G, Dwg. OSU 600904). Liquid break flow from the cold-leg side of the break was measured by break separator liquid flow meter FMM-905. Break steam flow from the cold-leg side of the break was measured by break separator steam flow meter FVM-905 installed in the [ ]'** steam line of the break separator. The [ ]'** steam line from the break separator was isolated throughout the test (Dwg. OSU 600901). Fill and vent was perfonned according to the approved operating procedure. Instruments were checked for required calibration. With TS-202 and TS-203 closed, RCS-901 and RCS-902 were opened to pressurize the CMT-1 balance line. After the appropriate prerequisites were completed and the test facility reached specified initial conditions, RCS-901 and RCS-902 were closed to maintain the < 80 F initial condition at the top of CMT-1. As final valve alignments were established, the CMT discharge valves were placed in the AUTO and CLOSED position, limiting CMT-1 pressure to that obtained while the vessel was arriving at full pressure and temperature. RCS 529 and RCS-530 were placed in the OPEN and AUTO positions 1 minute after the TEST pushbutton was pressed. The CMT-1 balance line ponion up to the isolation valves then came to reactor coolant system (RCS) pressure. Testing was initiated when test facility conditions, as read from the test facility control board and local indication, agreed with specified initial conditions within acceptable tolerances. The transient continued through ADS actuation and CMT, accumulator, IRWST, and sump injection. All actions were automatic and required no operator action, except isolation of two valves: CSS-902 and CSS-906. These were simultaneously closed when total steam flow through CSS-901 was less than [ ] The steam flow was read on FVM-901 and FVM-902. These valves were isolated to provide more accurate disd.arge flow values. Heater rod bundle power was adjusted prior to break initiation to achieve the required hot-leg temperatures. Pressurizer power was terminated 6 seconds after initiation of the break. Table 5.3.1-1 provides a comparison of the specified and actual conditions for Matrix Test SB10. Initial conditions for the test were established and recorded in the procedure. Refer to Subsection 2.7 for pretest operations. Table 5.3.1-1 shows the initial conditions recorded from the operator's p nel and the average of the same parameters for about 2 minutes prior to the break valve opening from DAS. There were three condition parameters out of specification in Table 5.3.1-1, none of which should invalidate this test, that will be discussed here. O l REVISION 1 ouSMwRevN5%w 5.non:Ib-o81798 5.3.1-2
FINAI. DATA REPORT
=
] f] HL-1 temperature (SC-141) was [ ]'6' percent above the required , V temperature band. HL-2 temperatures (SC-140) was [ ]'6' percent above , the required temperature band. This was within the accuracy requirements of the instrumentation system. t Accumulator levels (LDP-401 and LDP-402) exceeded the planned levels. The explanation for this situation is defined in note 3 of Table 5.3.1-1, ne condition was considered acceptable because the level was set by a standpipe at 37 in. !
- Accumulator-1 (ACC-1) pressure (FT-401) was [ ]'6" percent below the required pressure band. ACC-2 pressure (PT-402) was [ ]'6' percent below the required pressure band. De accumulators were pressurized to the required pressure, as indicated on !
PI-401 and PI-402 prior to test actuation. The loss of pressure between tank pressurization and test actuation was possibly due to cooling of the nitrogen gas in accumulator. Test analysis starting with the recorded lower accumulator and overpressure should still be possible. The reactor heater power controller was programmed using the algorithm defined in Subsection 2.3.2 to obtain the scaled decay power heat input rate. 5.3.1.2 Inoperable Mstrr Table 5.3.12 is a list of instruments considered inoperable or invalid during all or portions of this test. Some of the instruments listed are on the Critical Instrument List (Subsection 3.2, Table 3.2-2) and, therefore, are addressed here. FMM-201, FMM-202, FMM-203, and FMM-204 measured flow (gpm) in each of the four cold legs. A decision was made to continue testing without the availability of these instniments. Replacement flow meters repeatedly failed; their continued use was precluded due to cracking of the ceramic liners from thermal stratification in the loop piping. The necessary boundary conditions for loop flow could be determined from DP-202, DP-203, DP-205, and DP-206. FMM-401 measured ACC-1 injection flow into direct vessel injection line 1 (DVl-1). After injection, the meter exhibited erroneous results due to nitrogen in line; therefore, any results after 490 seconds are invalid. FMM-501, FMM-504, FMM-802, sad FMM-804 provided accurate data when sensing liquid, but became inaccurate when sensing two-phase or steam flow. FMM-701 measured IRWST-1 injection flow. When the primary sump valves opened, the flow meter indicated a negative flow as water flowed from the primary sump to the IRWST. The meter was not designed to measure reverse flow, so this measurement was invalid; however, total IRWST flow was V measured by FMM-702. c:\1$36wRevl\l536w-5.non:Ib.081798 5.3.1-3 REVISION 1
FINAL DATA REPORT FMM-905 measured break separator flow from the separator to the primary sump. When the primary sump level was equal to or higher than the break separator pipe connection, FMM-905 indicated a negative flow as water flowed from the primary sump to the break separator. The meter was not designed to measure reverse flow, so this measurement was invalid. Steam generator (SG) tube level data (LDP-215, LDP-218, LDP-219, and LDP-222) were biased by vaporization of the water in the transmitter reference leg after the SG tubes started draining. However, data provide accurate indication of the time when the tubes were empty. LDP-401 and LDP-402 measured ACC-1 and ACC-2 levels, respectively. Due to air trapped in the transmitters sense lines, date fmm these transmitters were invalid. The initial level of the tank, however, was established by a standpipv so it was constant test to test. The drain rate can be calculated using FMM-401 and FMM-40., respectively. Alternately, a pressure correction may be applied directly to the level indication i LDP-401 and LDP-402. 17T-201 measured RCS pressure at the top of the SG-1 long tube. On August 15,1994, it was discovered that the transmitter had an incorrect zero compensation which resulted in a negative error l and negative data at low pressures. The transmitter zero was corrected at that time. PT-201 data obtained during Matrix Test SB19 had the zero correction performed on it; the corrected data appear as PT._201. Negative data and corrected negative data can be used to determine trends by are considered inaccurate. PT_201 data are considered unreliable for values less than [ ]** but a sufficient amount of other pressure data are available. 1 TF-501 and TF-504, located on the long thermocouple rod location near the bottom of each CMT, appeared to have measured ambient conditions throughout the test, which would indicate a short j somewhere in the thermocouple wiring. With these thermocouples inoperable, the required long thermocouple rod thermocouple availability of seven out of ten and no more than one in succession l failed was met. LDP-802 measured PRHR HX wide-range level. The PRHR HX may have been refilling, but a more l logical reason is that LDP-802 was slowly losing its reference leg due to a low saturated pressure in the HX tubes. 1 Data provided by ADS-4 separator instrumentation prior to the ADS 4-1 and ADS 4-2 valves c, :ning l at 745 seconds were invalid due to the closed position of the ADS-4 valves and the ADS-4 separator j loop seal valves. The instruments affected were: FMM-602, FMM-603, FVM-602, FVM-603, LDP-611, AND LDP-612. Test analysis will not be affected, since ADS-4 flow did not begin until I the valves opened. l Considering these critical instrument failures, sufficient instrumentation was available to allow the performance of mass balances as demonstrated in Subsection 5.3.1.6 and Appendix E. An energy 1 l oA1536wRevi\l536w 5.non:Ib 081798 5.3.1-4 REVISION 1 l
FINAt. DATA REFC2T balance will be performed and reported in the AP600 Low-Pressure Integral Systems Test at Oregon J State University Test Analysis Report WCAP-14292.* 5.3.1.3 Sequence of Events ; l
'Ihe chronologic sequence of events is shown in Table 5.3.1-3. The actual time of event occurrence is listed along with the planned time of event occurrence. The table also provides the source of the actual time values. A D in the Data Source column indicates the recorded time was obtained from a software program that monitored digital events in the plant, including pump stants and stops, valve limit switch actuations, and alarms. An A in the Data Source column indicates the time was obtained by reviewing data from the data acquisition system (DAS). Although data from the DAS are digital, analog events such as pressure, flow, and temperature were monitored.
Tables 5.3.1-3 also depicts a bar graph representation of the sequence of events for Matrix Test SBIO sorted in chronologic order to provide a quick visual reference of the timing of events. The results of this timing will be described in more detail in the following sections. 5.3.1.4 Test Results and Evaluation This section contains an overall description of the events that occurred during Matrix Test SB10, including references to specific instrumentation channels cross-plotted and used in the test results
./ evaluation.
The simulated loss-of-coolant accident (LOCA) event resulted in interactions between different systems and simulated components in the facility. For this reason, the event was subdivided into the following three phases to characterize the system's thermal-hydraulic phenomena and component effects.
- Initial Depressurization Phase: simulated break initiation to ADS-1 actuation ADS Phase: ADS-1 actuation to start of IRWST injection IRWST Injection Phase: start of IRWST injection to end of test The behavior of specific components is discussed in Subsection 5.3.1.5.
Initial Depressurization Phase Prior to initiation of the event, ell applicable systems achieved normal operating conditions and initial boundary conditions, as described in Subsection 5.3.1.1. Time zero represents when the signal to open was sent to break valves TS-202 and TS-203. The DEG break flow arrangement was initially monitored by two sets of instmments. Set one was composed of DP-215 and PT-203. Set two was composed of DP-216 and PT-2%. Immediately following the opening of the break valve, subcooled liquid flowed from the cold leg to the break oA15hRevi\l5h-5.non:Ib 081798 5.3.1-5 REVISION 1
FINAi, DATA REPORT separator, and, as a result, RCS pressure decreased (Figures 5.3.1-3 and 5.3.1-4). Differential pressure indicated (DP-216) a lower pressure than the full reactor pressure due to a delay in opening the warming lines, as described in Subsection 5.3.1.1. The reactor and downcomer levels reflected the drop in core water inventory through the break (Figures 5.3.1-5 and 5.3.1-6). The reactor level (LDP-127) dropped immediately following the break opening, whereas the level in the downcomer (LDP-116 and LDP-140) remained constant from [ ]'6" (Figure 5.3.1-5). The reactor level fell to its minimum level at [ ]'6* (LDP-ll3; Figure 5.3.1-6), whereas the downcomer level dropped about [ ]'6" from a full condition (LDP-116; Figure 5.3.1-5). The upper head and upper portion of the upper plenum started to void immediately after the reactor coolant pumps (RCPs) tripped. The plenum voided more rapidly than the upper head because the upper head fluid was separated from the upper plenum by the upper support plate. Fluid could only exit the upper head to the downcomer via [ ]'6# bypass holes in the upper flange of the core barrel or by gravity-drain to the upper plenum via eight holes in the upper support plate. One-half second after the signal to open the break valves, an S signal was generated, causing the following planned events to occur during the first [ ]'6" after the break signal: the pressurizer heater breaker was manually opened at [ ]'6' the feedwater pumps tripped at [ ]'6" the CMT-1 and CMT-2 outlet valves opened at [ j'6# the passive residual heat removal heat exchanger (PRHR HX) outlet valve opened at [ ]'6' and the RCPs tripped at [ ]'6' Liquid from the pressurizer (LDP-601) and CMT-2, injected into the reactor as the reactor depressurized (Figures 5.3.1-7 and 5.3.1-8). Depressurization in the reactor vessel and pressurizer occurred at the same rate (Figures 5.3.1-9 and 5.3.1-10) as a result of the decrease in liquid inventory I and the flashing that occurred during the break. Core heater groups 1 and 2 provided a total initial input of 600 kW up to [ ]'6# KW-102 and KW-104 (group 2) indicated a sharp drop for 8 seconds immediately proceeding the break initiation point, however, this momentary drop in indicated power did not show any adverse effects on the overall decay heat profile. At [ ]'6' the heater controller continued to follow the scaled power decay rate for the remainder of the transient. The decay power curves appear in Appendix F. The total power comparison results indicated excellent agreement between planned decay rates and measured values. The pressurizer heater breakers were manually opened by the operators after verifying the S signal, the break valves opened, and pressurizer heater power (KW-601) changed from 1.5 to 0.0 kW. In [ ]'6' the pressurizer level had dropped to a zero level, and the core and downcomer levels continued to drop due to the large inventory loss from the DEG break and slower rate of ( injection (Figure 5.3.1-7). CMT-2 responded rapidly by injecting its inventory into the DVI line upon opening its isolation valve at [ ]'6" after break initiation (Figure 5.3.1-8). As the CMT-2 began discharging, natural circulation started to replace inventory in CMT-2 from CL-1. A CMT-2 injection flow rate of [ ]'6' continued until about [ ]'6' when the level in the balance line fell, allowing CMT-2 flow to increase to [ ]'6# (Figure 5.3.1-11). At about l o:\l $36w Rev i\l 536w-5.non: I b-081798 5.3.1-6 REVISION 1
i FmAr. DATA REPORT [ ]'6' when the reactor pressure dropped to [ ]'6' ACC-1 and ACC-2 injected to the downcomer through the DVI lines because pressure in the accumulators exceeded reactor pressure. Reactor depressurization occurred at such a high rate that the accumulators injected prior to ADS-1 actuation. De ADS-1 actuation sequence was controlled by the CMT low-level indication ([ ]'6' for OSU) plus a [ ] delcy. Injection by the accumulators slowed the injection flow rate from CMT-2 (Figure 5.3.1-11) due to the higher localized pressure at the junction of the CMT-2 and DVI lines. he higher localized pressure effect actually slowed and suspended CMT flow, as flow from CMT-2 dropped from [ ]'6' and remained at [ ]'6" until accumulator injection decreased from peak flow. His suspension of CMT flow was coincident to the combined ACC-1 and ACC-2 injection approach of peak injection rates. CMT-1 showed no change in level because the break occurred on the horizontal section of the balance line before the top of CMT-1, leaving the top of CMT-1 at atmospheric pressure. CMT-1 injection flow was precluded because pressure in the DVI line was higher than the head of water in CMT-1. Subcooled liquid flow through the PRHR HX began [ ]'6" after the break valves opened and a natural circulation mode was established. The flow rates appear in Figure 5.3.1-12. He initial flow spikes, (FMM-804) were due to the opening of valve RCS-804. FMM-802 continued to show high flow-rate spikes (over-range values) from [ ]'6' because two-phase flow entered the PRHR HX at about [ ]'6' as the hot legs began to drain. Magnetic flow meters used at OSU did not provide accurate indications during two-phase or reverse flow conditions; therefore, indicated flow-rate values from FMM-802 during these periods are not accurate. However, the PRHR HX outlet flow rate, though erratic at times, is correct, since steam entering the PRHR HX was condensed and TF-804 reflected subcooled water ([ l'6') being returned to the cold-leg channel head of SG 2 (Figure 5.3.1-68). As the reactor vessel and pressurizer continued to depressurize and their levels dropped, pressure and liquid levels in the RCS legs dropped. Due to rapid depressurization in the core and ' upper plenum, fluid in the upper plenum flashed while the fluid level in the upper plenum dropped to the hot-leg elevation. The cold-leg levels dropped to zero (LDP-201 through LDP-204) just before the hot legs started to drain (LDP-205 and LDP-206; Figures 5.3.1-13 and 5.3.1-14). Both the cold-leg and hot-leg levels began to drop at about [ ]'6' and reached [ ]'6" The level in the HL-1 elbow (LDP-207) was incorrectly ranged, did not provide accurate values, was inconsistent with other data (e.g., LDP-206), and was subject to high flow due to draining of SG-2 and the pressurizer (Figure 5.3.1-14). As the hot legs drained, this volume was replaced with two-phase fluid, and the HL-1 fluid had a void fraction very close to that shown in the upper plenum (Figure 5.3.1-64). The void fraction in HL-2 was only slightly lower due to the selective removal of vapor from the hot leg by the PRHR HX inlet line. I e:ushneviusw-5.nalbO81798 5.3.1-7 REVISION 1
FmAi, DATA REPORT Using data from the level channels and the calibrated range of the instruments, a core steam percent for each channel was calculated. The equation used to calculate steam percent was: Steam percent -(1 e mpensated level) 100 instrument range In [ ]'6" of the break opening, the primary side of SG-1 showed a rapid drop in level (Figure 5.3.1-16). He SG-1 level decrease continued until [ J'6* when LDP-211 and LDP-213 indicated a drop to zero level (Figure 5.3.1-17). The level in the SG-1 hot-leg channel head (LDP-209) was incorrectly ranged and did not provide accurate values. LDP-207 and LDP-209 were considered inoperable for this test. As the primary-side pressure dropped, any remaining SG water was transformed into superheated steam (Figure 5.3.1-19), indicated by thermocouples in the tubes remaining well above the saturation temperature. A parameter identified as TSAT was included in this report to represent the saturation temperature as a function of reactor pressure (PT-107). Figure 5.3.1-16 presents the indication of a rapid drop in U-tube levels (from [ l'6') followed by an indication of a rapid increase in level (suggesting a refill of the SG U-tubes from [
]*6"). This indicated that refill did not occur, because both the cold and hot legs were drained before [ ]'6" and the SG U-tube levels were drained at [ ]*6' when DP-211 and DP-212 dropped to 0 after a positive differential pressure (flow) indication (Figure 5.3.1-18). Once the SG U-tubes drained, they did not refill during the remainder of the test.
This false indication was due to the arrangement of LDP reference legs, as described in Subsection 2.4.1. He response described previously for SG-1 is consistent for SG-2. Voiding in the hot legs resulted in draining in the SG U-tubes and subsequent draining of the SG channel heads (LDP-211 and LDP-213; Figure 5.3.1-17). The connection from hot-leg to cold-leg sides was broken in SG-1 about [ ]'6' after the break. About [ ]'6* later, the recirculation mode ended and draindown of the U-tubes started in SG-2 (DP-211 and DP-212; Figure 5.3.1-18). From [ ]'6' DP-211 and DP-212 indicated a differential pressure of about [ ]'6' of water. After RCP trip, DP-211 and DP-212 showed 0 differential pressure, as expected, and a very short duration of natural circulation occurred from [ ]'6* As the SGs drained, the levels dropped, and a positive differential pressure existed from [
]'6' reflecting drainage of the cold-leg sides of the SGs. The cold-leg sides drained about
[ J'6# ahead of the hot legs. When DP-211 indicated [ ]*6* drain flow ended, and the U-tubes were drained for SG-1. SG-2 drained at about [ ]'6* based on indications from DP-212 (Figure 5.3.1-18). During initial depressurization of the primary loop, secondary-side pressures increased, but primary pressure decreased (Figure 5.3.1-20). He initial differential pressure (primary pressure > secondary pressure) reflected heat transfer from the primary to the secondary side. The primary and secondary pressures approached a common pressure of [ ]'b' This common-pressure period lasted [ ]'6' At about [ ]'6' the primary and secondary pressures o:us36wRevh1536w.5.non ib-o81798 5.3.1-8 REVISION 1
1 l FINAL DATA REPORT diverged (primary pressure < secondary pressure), allowing the secondary side to transfer heat into the t ( primary side via the SG U-tubes. Heat transfer to the primary side combined with drairiage of the SGs l caused superheating in the SG U-tubes. He U-tubes remained at superheated conditions for the rest of the test (Figure 5.3.1-19) due to the slow decay of SG pressure and energy ia the secondary side. A comparison of fluid temperatures from the primary side of the U-tubes to the bulk fluid j- temperatures on the secondary side of the SGs is provided in Figure 5.3.1-21. Both primary-side (SG-1 and SG-2 short and long tube thermocouples) and secondary-side (SG-1 and SG-2 downcomer I thermocouples) temperatures reached a common value of [ ]'6# During this same time, primary-and secondary-side pressures reached a common pressure of about [ ]'6' defined as the pressure plateau. As the reactor continued to depressurize, primary temperatures diverged from saturated conditions, as indicated by the TSAT parameter (Figure 5.3.1-21). He temperature difference between the primary-side and to secondary-side thermocouples indicated that the secondary side was transferring heat to the primary-side and keeping the primary-side U-tubes superheated. From break initiation to about [ ]'b' a natural circulation flow developed from CL-1 to CMT-2 (FMM-502). The CMT-2 level remained stable (LDP-502; Figure 5.3.1-8). At about [ ]'6# the natural circulation mode ended and the draindown mode started in the CMT-2 balance line, as indicated when flow through FMM-502 dropped to zero and the CL-1/CMT-2 balance line level suddenly dropped. When the transition from recirculation to draindown mode occurred in
/% the CMT-2 balance line, the following resulted. First, the level in the CMT decreased because the liquid draining into the core was no longer being replaced by water from CL-1 (Figure 5.3.1-8).
Second, the CMT-2 balance line level decreased as the region in the upper portion of the balance leg voided. He CMT-2 balance line emptied at about [ ]'6' (Figure 5.3.1-8). Third, CMT-2 injection flow to the vessel increased from [ ]'6' (Figure 5.3.1-11). I l As the system continued to depressurize beyond [ ]'6' CMT-2 continued to drain to the reactor vessel at a slower rate due to the injection of ACC-1 and ACC-2. LDP-502 in Figure 5.3.1-8 and FMM-504 in Figure 5.3.1-11 indicate that, as accumulator injection increased localized pressure at the DVI line junction, the CMT-2 flow rate was reduced, and CMT-2 injection was suspended from [ ]'6' Once accumulator injection was completed, as indicated by ACC-2 injection flow rate, flow from CMT-2 retumed at a rate of about [ ]'6' At[ ]'6' CMT-2 had drained to the low level setpoint, and a 15-second delay timer automatically actuated before the ADS-1 valve opened. At this time, reactor pressure was [ ] and the vessel level had already reached its lowest point of [ ]'6' (Figure 5.3.1-5). The vessel level increased due to accumulator injection but both the hot legs and cold legs were drained. Both the upper head and the upper portion of the downcomer were superheated (Figure 5.3.1-22). Figures 5.3.1-23 and 5.3.1-24 indicate that saturated steam was occupying the bot legs even though they began draining after [ ]'6' Superheated steam continued to exist in the SG O tubes, since the SGs acted as heat sources; secondary pressure and temperatures remained high with little heat loss (Figures 5.3.1-20 and 5.3.121). o:u5hRevnl5h 5.non:ltr081798 5.3.1 9 REVISION 1
FrNA1. DATA REPORT ADS Phase he purpose of the ADS is to depressurize the RCS to allow injection of the accumulators at intermediate pressure and to allow injection of the IRWST at low pressure for long-term cooling. l Depressurization is first accomplished by an inventory loss through three parallel valves: ADS-1, ADS-2, and ADS-3. These valves open in sequence after one of the CMTs reaches a low level setpoint, providing a flow path from the top of the pressurizer, through the ADS 1-3 separator, to the ADS 1-3 sparger located in the IRWST (Dwg. OSU 600203). Opening of the ADS 4-1 and ADS 4-2 valves occurs when either CMT reaches its low-low level setpoint. The ADS-4 valves provide the final depressurization of the RCS by supplying a flow path from the hot legs to the primary sump via the ADS 4-1 and 4-2 separators. The ADS period was initiated [ ]'6 after the break with the opening of the ADS-1 valve. ADS-1 flow was indicated by a transmitter measuring differential pressure due to flow (FDP-605). FDPS were located upstream of each of the ADS valves to measure flow through the valves. The opening of ADS-2 and ADS-3 valves occurred in the next [ ]'** (FDP-604 and FDP-606; Figure 5.3.1-25). The ADS-1 valve opening resulted in a peak differential pressure of [ ]'** after [ ]'** Superheated steam remaining in the pressurizer (TF-602 and TF-605; Figure 5.3.1-63) was released to the ADS 1-3 separator where the steam was allowed to separc and was measured by a vortex flow meter (FVM-601). Water was measured by
- Gr,netic flow meter (FMM 601) prior to injection / discharge into the IRWST via the :;parger dispersion tube (Figures 5.3.1-26 and 5.3.1-27). Once these ADS valves opened, they remained open throughout the remainder of the transient. He ADS-2 valve opened at [ ]'6" with a differential peak pressure of [ ]'6" occurring [ ]'** after opening. The peak differential pressure decreased rapidly to about [ ]'6" The ADS-1 and ADS-2 valve openings reduced RCS pressure from [ ]'** over a [ ]'6' period. ADS actuation increased the rate of primary system depressurization. as measured by reactor upper head pressure (Figure 5.3.1-9), and resulted in a higher injection flow ([ ]'6' to about [ ]'** peak) from the accumulators (FMM-401 and FMM-402; ligure 5.3.1-11). ADS actuation forced flow from the bottom of the pressurizer to the top and out through a single line to the ADS valves, through the ADS 1-3 separator, and into the IRWST via the sparger. Steam flow through the pressurizer caused the pressurizer wide-range level and the pressurizer surge-line level (Figure 5.3.1-7) to indicate a false-high signal due to momentum effects.
When the ADS-1 valve opened (Table 5.3.1-3), the pressurizer surge line and the pressurizer initially became filled with a two-phase mixture. This sudden surge of flow through the pressurizer was reflected in the rapid increase in surge line and pressurizer wide-range level indications. LDPs indicate accurate level values in steady-state or nonflowing conditions; LDP-601 and LDP-602 measurements during ADS-1 actuation include the differential pressure due to flow. The increased injection flow of the accumulators and CMT-2 caused a high influx of liquid flow through the reactor. From [ ]'6" the surge line and pressurizer became partially filled with steam and 0:\l 536w Rev N 536w-5.non: I b-081798 5.3.1-10 REVISION 1
~ - . . . . - - - .. -. - - - - - - . - - - - . - . - - - - - . -
FINA1.. DATA REPORT liquid consistent with the accumulator injection period, to the extent that liquid flow was measured from the ADS 1-3 separator (Figure 5.3.1-27). As steam flow ended at [ ]'6' (end of accumulator injection), liquid flow through the ADS 1-3 separator decreased from [ ]'6' over [ ]'6' During this period, the pressurizer was filled with [ ]'6" percent steam [ and [ ]*' percent liquid supported by an [ J percent steam vapor in the surge line (Figure 5.3.1-65). At about [ ]"# the ADS-4 valves opened, causing a rapid decrease in vessel pressure and, i subsequently, permitting slow draindown of the pressurizer into the hot legs. At about [ !
] the pressurizer emptied and the surge line was filled to a [ ]"# level. A more 1 detailed explanation of flow conditions in the pressurizer is provided in the pressurizer components l description of Subsection 5.3.1.3.
Rapid injection of cold water from the accumulators (at [ ]"' into the event) subcooled the upper region of the downcomer annulus (which contained some superheated fluid, { measured by TF-168, until the accumulators injected) and continued the temporary refill of the core ! region to the top of the DVI line (Figures 5.3-1.22 and 5.3.15). The combined opening of the ADS-1 and ADS-2 valves and the accumulators' injection caused the collapse of the steam bubble at the top I of the downcomer [ ]'6' after the break. Collapse of the steam bubble caused a differential pressure across the bypass plate, superheated steam in the upper head, and saturated steam or subcooled liquid in the downcomer. To compensate for this sudden imbalance, a backflow occurred, as reflected in the sharp temperature decrease from [ ]'6' at [ ]"' (TF-120; (*
\
Figure 5.3.1-22) This condensation event is also reflected in rapid changes in thermocouple TF-168 (Figure 5.3.1-22) and differential pressure transmitters DP-114 and DP-130 (Figure 5.3.1-28). Figures 5.3.1-28 and 5.3.1-29 provide comparison plots of the differential pressure values across the upper support plate (DP-il4), and between the upper head and downcomer (DP-130). Dese differential pressure changes reflected rapid steam condensation events ia which the subcooled water entered the downcomer region from the DVI, and immediately condensed the superheated ceam in the downcomer, causing a large decrease in pressure and differential pressure, causing a flow reversal in the downcomer region. Fluid from the core and downcomer regions reversed and rapidly filled the downcomer to the upper head bypass plate. TF-168 was located in the downcomer above the cold. leg elevation and aligned at 270*az in line above HL-2 (Appendix H). Thermocouple TF-168 was used to help confirm such condensation events. Shortly after the event took place, ti'e top of the reactor retumed to superheated conditions. His generic condensation event was studied and is explained in more detail in Subsection 7.1. The rapid decrease in reactor pressure due to ADS-2 actuation resulted in a sharp increase in the accumulator injection rate over the period of [ ]'*' (Figure 5.3.1-11). During the ADS-2 actuation period, the peak total injection flow occurred based on the combined sources of the CMT-2, ACC-1, and ACC-2, as measured at the DVI lines. DVI flows were monitored by flow meters FMM 205 and FMM-2% for DVI-l and DVI 2 lines, respectively (Figure 5.3.1-30). The peak injection flow rate of [ ]"# per side occurred at [ ]'** after ADS-2 valve opening. v The fluid inventory in the reactor core region recovered to a level slightly above the bottom of the hot oA15hRevnl5h-5.non:lb 081798 5.3.1-11 REVISION 1
FINAL DATA REroar legs, superheated steam occupied the upper head, and saturated steam filled the hot legs and cold legs. Stored energy of the SG continued to provide heat into the primary system via the SG U-tubes and maintained the superheated steam in the U-tubes of both SGs. When reactor pressure decreased to [ ]'6" at [ ]'6' the signal to open RCS-711 and RCS-712 was automatically initiated. The two parallel IRWST injection lines contained a set of check valves to preclude backflow fmm the RCS to the IRWST. These check valves (RCS-713 and RCS-715 in line with FMM-701, and RCS-714 and RCS-716 in line with FMM-702) permitted injection into the DVI line when reactor pressure decreased to the equivalent of the IRWST liquid head. Based on the design configuration described previously, IRWST injection can occur without operator or automatic actuation once valves RCS-711 and RCS-712 open. At[ ]'6" when RCS pressure was [ ]'6' the opening of the ADS-3 valve caused the continued decrease in RCS pressure. The primary portion of the accumulators' injection continued through [ ]'6' As the acctimulator-injection surge was completed, CMT-2 injection flow resumed at a peak flow rate of [ ]'6' however, the core and downcomer liquid levels decreased due to the decreased flow (Figures 5.3.1-11 and 5.3.1-5). At [ ]'6' the CMT-2 level was about [ ]'6" ADS-4 valves opened when the level in at least one of the CMTs dropped to the Chit low-low limit ([ ]'6') and at least [ J'6' had passed since one of the CMT levels dropped to the low level setpoint of [ ]'6* This criteria was achieved at [ ]'6" when the liquid level in CMT-2 decreased to [ ]' 6' At[ J'6* both ADS-4 valves were open, directing flow to the ADS ?-l separator to be separated and measured by FVM-603 and FMM-603 (Figures 5.3.1-26 and 5.3.127). Saturated liquid from the ADS-4 lines was then discharged into the primary sump. Also at [ ]'6' t& pressurizer liquid level indicated a significant decrease, and discharge through the ADS 1-3 valves ended (Figures 5.3.1-7 and 5.3.1-27). FMM-603 indicated that, as the ADS-4 valves opened, flow through the ADS 41 separator was initiated and gradually increased to a peak flow rate of about [ J'6' at about [ J'6" (Figure 5.3.1-27). IRWST Inlection Phase The IRWST injection phase began when RCS pressure decreased to the pressure corresponding to the water elevation of the IRWST. IRWST flow entered the DVI lines through DVI l and DVI-2 beginning at [ ]*** respectively (FMM 701 and FMM-702; Figure 5.3.1-31). l IRWST injection flow steadily increased from an average flow rate of [ ]'6' to a peak IRWST l flow rate of [ ]'** per side at [ ]'6" l l At[ ]'6' the CMT-2 level was about [ ]'6" and CMT-2 injection flow was [ ]'6" l All injection flow from the accumulators had terminated; the accumulators had injected their water l inventory. CMT-1 was full with subcooled water (< 80*F) at atmospheric pressure. Because the l balance line was broken, CMT-1 was unable to inject until the pressure at the junction of the CMT-1 1 0:\l536wRevi\l536w 5.non:lb-081798 5.3.1-12 REVISION 1
FINAI DATA REPORT outlet and DVI-l lines was below the liquid head level of CMT-1. CMT-1 flow through DVI-l began at [ ]'** (FMM-501; Figure 5.3.1-11). CMT-1 liquid injection flow was confirmed when the CMT-1 level decreased (LDP-507; Figure 5.3.1-8). The CMT-1 balance line level (LDP-509) was over-ranged and does not accurately indicate the level; the break was in this line and the LDP was measuring the differential pressure between the RCS and the atmosphere (open side of the break).
'Iherefore, this level can not be used as an indication of level in the CMT balance line. After peak IRWST injection flow was achieved at [ ]'*' flow continued to decrease slowly based on the IRWST head.
As IRWST and CMT-1 injection continued, both the downcomer liquid level and the core liquid level increased (Figure 5.3.1-5). At [ ]'** vessel and downcomer collapsed liquid levels appeared to be equal. Figure 5.3.1-22 shows that, at [ ]'*' based on the sharp temperature excursions, a third localized steam bubble collapse occurred. The upper head remained filled with superheated steam (TF-120 and TF-171; Figure 5.3.132). The outlet plenum contained saturated steam and liquid, as measured by TF-169 (upper reactor fluid temperature at 90 degrees orientation) and TF-170 (upper reactor fluid temperature at 270 degrees orientatien) (Figure 5.3.1-32). Bulk fluid temperature in the core region remained at the saturated temperature or subcooled throughout the transient (TR-001; Figure 5.3.1-33). During early depressurization, little temperature variation occurred across the rod bundle. Injection from the CMT-2, accumulators, and IRWST had not only cooled the core but contributed to a temperature gradient in the core. At [ ]'**a [ ]*** temperature gradient across the heater rod bundle was created and maintained until about [ ) 2e Also during this period, the pressurizer remained filled with both saturated steam and liquid, with no indicated flow through the pressurizer or flow out of the ADS 1-3 valves. The SG U-tubes remained filled with superheated steam that decreased in temperature from about [ ]'*' The ; PRHR HX continued to condense saturated steam from HL-2 and returned an indicated flow of [ J'*' of [ ]'*' water to the bottom of the cold-leg plenum of SG-2 (Figure 5.3.1-12).
'Ihe ADS-4 valves' opening caused the continued decrease in RCS pressure and allowed the IRWST to inject. Liquid levels in both the downcomer and core regions increased to the point where both the hot and cold legs refilled with liquid inventory due to IRWST injection. At about [ ]'6' the downcomer level was [ ]'6'(LDP-116 ar,d LDP-140; Figure 5.3.1-5); however, the core collapsed liquid level was [ ]'** (LDP-127; Figure 5.3.1-5) As IRWST and CMT-1 injection continued, the downcomer level became stable at about [ ]'** and, at about [ ]'*'
the core collapsed liquid level (LDP-127) reached a level to refill the hot legs. The hot legs filled at about [ ]'** when a drop in HL-1 and HL-2 temperatures was indicated (Figures 5.3.1-23 and 5.3.1-24).
/~'N Temperatures in the cold legs confirmed that the cold legs were beginning to refill. Beginning at about [ ]'** cold-leg bottom temperatures (measured by CL-1, CL-2, CL-3, and CL-4 at oA1536wRev l M 536w-5.non: ll>.081798 5.3.1-13 REVISION 1
FINAL DATA REPORT the reactor flange bottom) dropped from superheated to saturated conditions (Figures 5.3.1-34 to 5.3.1-37). Once the cold legs reached a full condition (about [ ]'6"), the cold-leg top temperatures (SC-105, SC-106, SC-101, and SC-102) exhibited a constant temperature. A temperature gradient existed from the top to the bottom of the cold-leg pipe (Figures 5.3.1-34 to 5.3.1-37). The hot-leg temperature became subcooled at about [ ]'6# suggesting a full condition (Figures 5.3.1-23 and 5.3.1-24). Refilling the cold legs with liquid covered the CMT-2 balance line opening. As the CMT-2 balance line partially filled, the level in the balance line increased (LDP-510; Figure 5.3.1-8). Refill of CMT-2 continued until a level of [ ]'6" was reached at [ ]'** As CMT-2 refilled, no flow was discharged. Figure 5.3.1-8 shows the liquid level of CMT-1 to the break (LDP-507 and LDP-509). 'Ihe CMT-1 balance line level (LDP-509) was not responding as an LDP due to high flow through the break. Pressure in the variable leg was higher than the reference leg and was not providing a true level indication; therefore, data from this channel (LDP-509) should not be used. IRWST and CMT-1 injection continued at a slowly decreasing rate with a total combined flow of [ ]'6" at [ ]'6# (FMM-205 and FMM-206) with the majority of this flow coming from the IRWST (Figure 5.3.1-30). At [ ]'6' CMT-2 refilled to a level of [ ]'6" and remained full until $mt [ ]'6* At[ ]*6# after the break event, flow from CMT-2 entered the DVI line (FMM-504; Figure 5.3.1-38). The effect of draindown on the CMT-2 liquid level, as measured by LDP-$G2. is shown in Figure 5.3.1-39. Figure 5.3.1-39 also shows the continued draindown of CMT-1 (LDP-507) until about [ ]'6" whm CMT-1 gtid CMT-2 emptied at [ ]'6* Duriag the period from [ ]'6' the downcomer region reached a maximum over-ranged liquid level of [ ]'6" and the core reached a maximum liquid level of [ ]'6# which is about at the [ ]'b' (Figure 5.3.1-40). Both levels decreased after [ ]'6' as flow exiting through the ADS-4 valves (FMM-603; Figure 5.3.1-45), to the primary sump slightly exceeded combined injection flow from the IRWST and CMTs, as measured by the sum of FMM-205 and FMM-206 (Figure 5.3.1-41). With the core heaters still generating heat and with little ficw oui of the vessel, the bulk core liquid temperature increased from subcooled (average temperature of [ ]'6") so that the top layers of the core became saturated at [ ]*6' (TR-001-8; Figure 5.3.1-46). Superheated steam continued to exist in the upper head; however, the temperature decreased from [ ]'6" (Figure 5.3.1-42). Beginning at about [ ]'6" temperatures in both the cold and hot legs increased from subcooled ([ ]'6* ) and approached a steady-state saturated temperature of [ ]'6' (Figures 5.3.1-82 and 5.3.1-84). It should be noted that the saturated temperature at local pressure TSAT value was calculated based on PT-107. For Matrix Test SB10, the PT-107 value had a slight zero offset causing a minimum value indication of [ ]'6# where the saturated temperature should not be lower than 212 F. It is not clear when the o A1536wRev !M 536w-5.non: I b-081798 5.3.1-14 REVISION I
FINAL DATA REPORT l l l zero affect occurred in the transient; therefore, a correction was not applied to the PT-107 data. Also ! during this period, the liquid level in the upper head drained to a level just below the upper support plate. This draindown was indicated by a sharp increase in temperature measured by thermocouple TF-171. Thermocouple TF-171, located directly above the upper support plate, indicated a superheated temperature of [ ]'b' (Figure 5.3.1-42). The top of the downcomer annulus also voided and filled with superheated steam, as indicated by the temperature increase to [ ]'6' (TF-168; Figure 5.3.1-44). An oscillation was seen beginning at [
]'6" (FMM-603; Figure 5.3.1-45 and FMM-701 and FMM-702; Figure 5.3.1-43). These oscillations may be due to repeated momentary relief of steam buildup in the upper head or SG through the hot legs and ADS-4 valves since all other vent paths were filled with saturated liquid (e.g.,
cold legs). The U-tube thermocouples in the SGs continued to indicate [ ]'b'superheated steam at [ ]'6" These tubes were superheated for the remainder of the test ([
]'6').
At[ ]'b' pressure in the DVI lines was less than pressure corresponding to the water elevation of the primary sump. Flow from the primary sump (FMM-901 and FMM-902) entered the DVI line (Figure 5.3.1-43). Flow was initially limited through the double-check valve arrangement (see valves CSS-921 through CSS-924, Dwg. OSU 600206). At [ l'6" when the IRWST reached the low-low setpoint, the programmable logic controller (PLC) automatically opened l CSS 909 and CSS-910, which allowed the flow rate to increase to a total combined peak flow rate of 3 [ ]'6" (FMM-901 and FMM-902; Figure 5.3.1-43). However, FMM 702 indicated a continuing flow of [ ]'6' and FMM-701 indicated a negative or zero flow at the seme time (Figure 5.3.1-43). A comparison of flows measured by FMM-205 and FMM-206, which represent total flow in the DVI lines, indicated a total combined flow of [ ]'b' from each line) (Figure 5.3.1-41). Therefore, the indicated negative value for FMM-701 represented a reverse-flow condition in which flow from the primary sump entered the IRWST. As previously discussed, magnetic flow meters used at OSU cannot measure reverse flow and will read negative if the flow is in the reverse direction of the normal flow path. This accounts for the negative flow value for FMM-701; however, the negative value does not reflect an accurate measure of reverse flow. Liquid levels in both the downcomer and the core fell until about [ ]'b' The downcomer levels (LDP-116 and LDP-140) stabilized with the cold leg at an elevation of [ ]' 6' (cold-leg centerline), and the hot legs drained (LDP-127) with an elevation of [ ].'6' The level ~ in the hot legs cannot be directly confirmed by the temperatures in the hot legs, cold legs, upper head, and downcomer (Figures 5.3.1-44,5.3.1-82, and 5.3.1-84) because all of the temperatures were just above the saturation temperature. At about [ ]'b' the system reached an equilibrium condition with the fluid temperature in the core, downcomer, upper head, cold legs, and hot legs at between [ J'6" and the primary sump injection mode in progress. The primary sump injection mode occurred when an equal volume of liquid was discharged through the ADS-4 valves, then returned to the vessel through the DVI lines (Figures 5.3.1-43 and 5.3.1-45). At this time, [ ]'6" of flow exited through the ADS-4 oil 5hRevi\l536w 5.non:Ib-081798 5.3.1-15 REVISION 1
FINAL DATA REPORT
~
valves to the primary sump (FMM-603; Figure 5.3.1-45). An equivalent combined flow of [ J'b' was returned by the primary sump drain valves to the DVI line (FMM-901 and FMM-902; Figures 5.3.1-41 and 5.3.1-43). The test was terminated after at least 2 hours of stable sump injection was achieved. 5.3.1.3 Component Responses Reactor Discussion of rea: tor response includes the regions: reactor core, upper plenum, upper head, and downcomer annulus. The regions will be discussed in this order. Level response of the downcomer annulus was measured by two wide-range and several narrow-range level transriitters (Figure H-2). Wide-range channels were positioned 90 degrees from one another Data from these two channels were in excellent agreement. Both channels were also in agreement with their associated narrow-range channels. When the TEST pushbutton was pressed, the reactor controller was in auto-local, controlling hot-leg average temperature at 420 F (the reactor controller automatically controls that temperature by varying the demand signal to the heaters). At time zero, the programmable logic controller (PLC) sent a signal to open the break valve ud then 0.5 second later signaled the reactor controller to shift control to auto-remote with total power demand initially at 600 kW (the setpoint is generated by an algorithm programmed into the controller, and the controller automatically controls the demand to the heaters to control the setpoint kW). The power algorithm programs full power (600 kW) for the first 140 seconds and then lets power decay at an exponential rate that simulates the decay heat input of the AP600 nuclear reactor following a trip from full power (Appendix F). When the signal to open the break occurred, the S signal was initiated and system pressure dropped; however, since the coolant everywhere in the reactor vessel was at or below the saturation temperature for system pressure, no boiling or flashing occurred. [ ]'b' after break initiation, the RCPs tripped, and the upper head reached saturation teraperature of [ ]'6* As the core level and pressure dropped due to loss of inventory out the break, the entire upper head, upper plenum, and upper annulus regions reached saturation temperature (TF-120, TF-171, TF-169, and TF-168; Figures 5.3.1-22 and 5.3.1-32). At 140 seconds, heater input power was shifted from 600 kW and then let power decay at an exponential rate that simulated the decay heat input of the AP600 reactor following a trip from full power. Actual power decay curves are provided in Appendix F. Figures 5.3.1-5 and 5.3.1-22 show the collapsed liquid levels in the core /downcomer region and the temperatures in the upper head /downcomer region, respectively, during the Matrix Test SB10 event. l At[ ]'b' the indicated level in the core reached a minimum level, even though 100 percent l of the inventory from the pressurizer, and primary sides of the SGs, and 28 percent of CMT-2 had I been added. Reactor core response was evaluated using temperature data from thermocouples located in the core. Data from thermocouples mounted on a thermocouple rod at the center of the core were l l o:\l 536m Rev nl536w-5.nott i b-081798 5.3.1-16 REVISION 1 l l \ - _ - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
. _ - - - . -~ .-. - .- - -- - - - - - --.- -
FINAL DATA REPORT p used to assess axial temperature distribution. Dese thermocouples were given a TR designator. Data ( from the top thermocouple of six heater rods were used to assess radial temperature distribution at the top of the core. Dese thermocouples were given a TH designator (Dwg. OSU 600007 and 600008). , The overall core collapsed liquid level showed that the core was covered throughout the whole e. vent Hermocouple rod temperatures / core-center thermocouple and heater rod temperatures showA no long-duration temperature spikes indicative of a core uncovery (TR-001; Figures 5.3.1-33 and 5.3.1-46) (TH-101-4 through TH-505-4; Figures 5.3.1-47 and 5.3.1-48). i ne break created a depressurization in the RCS. Temperatures at the center of the core became saturated within [ ]'6" of the break. As the test progressed, RCS pressure decreased as the : result of the break and actuation of ADS 1-3, and ADS-4. The core cooled and maintained saniration ! 1 temperature as the RCS depressurized (Figures 5.3.1-33 and 5.3.1-46). Subcooled temperatures at the ) bottom of the core first appeared at about [ J'6' shortly after ADS-1 actuated. Up to this l time, core temperatures were equal to or slightly below the saturation temperature as indicated by l TSAT. At about the time of ADS-2 actuation, the rate of RCS depressurization increased l significantly, resulting in a more rapid saturation temperature decrease. Core cooling continued with l DVI, so the bottom of the core started to subcool. ' Elevation in the core at which subcooling existed increased as the test progressed. From [
]'6' TR-001-7 and TR-001-8 remained at saturated conditions (elevation [ ]'6' and D [ ]'6# respectively). The remainder of the tnermocouples were subcooled (Figure 5.3.1-33).
Heating elements in the heater rods simulated the active fuel of the AP600. De top of the heating elements in the heater rods was located at 47 in. from the bottom of the reactor vessel. Thus, by the { end of the test, fluid in the center of the core was subcooled [ ]'6' percent of the length of the active core (TR-001-1 through TR-001-8; Figure 5.3.1-46). Data from the thermocouples installed in the tops of six heaters staggered radially across the core indicated that the design for a " flat" heat-flux profile was achieved in the core (Figures 5.3.1-47 and 5.3.1-48). De center heaters were generally at a temperature only [ ]'6# lower than the midcenter and outer heaters. The lower-center temperatures were probably due to a slightly higher flow rate through the center of the core. There were no notable temperature excursions, although spikes did occur as in the unheated thermocouple rod temperatures. Core steam percent may be calculated by using data from three level transmitters that measure core level: LDP-109, LDP-110, and LDP-138 (Appendix H). LDP-138 is a wide-range transmitter that I spanned the entire core area, from the top of the lower core plate to the bottom of the upper core plate. LDP-109 and LDP-110 are narrow-range transmitters whose summed span equals that of the wide-range transmitter. He narrow-range transmitters are mcunted in the same position as LDP-138 (0*az) and share the same taps. Here is close agreement between core steam percent calculations using core level data from the wide-range (LDP-138) and narrow-range level transmitters (LDP-109 and LDP-110). LDP-108 and LDP-109 c:uswRevmsw.5.non:H>481798 5.3.1-17 REVISION 1
FrNAL DATA Rr. PORT have equal spans, which together cover the same span as wide-range level transmitter LDP-138. . Therefore, the sum of their calculated core steam percent should equal the calculated steam percent of ! wide-range channel LDP-138 (Figure 5.3.1-87). The maximum steam percent in the core occurred at about [ ]'6# the same time as the I minimum core level. A secondary peak steam percent occurred at the minimum downcomer level at ! [ ]'6' From this time on, the core steam percent was constant or decreased, but never increased. In other words, core levels measured by the wide-range transmitter or either of the narrow-range transmitters never decreased after [ J'6" but remained constant or increased (Figure 5.3.1-87). In summary, the core showed indications of an increasing core steam percent immediately after the break valve opened. Temperatures in the core quickly reached saturation. Core cooling maintained temperatures at saturation as the RCS depressurized and lowered saturation temperature. Maximum core steam percentage are occurred after ADS-1 actuation and at the same time as minimum core level. At maximum core steam percentage and throughout the test, there were no excursions of temperature from either the heated or unheated rod thermocouples. The bottom of the core started cooling at the time of ADS-2 actuation and remained subcooled. The upper head and the upper plenum are separated by the upper support plate. Both the upper plenum and upper head began to drain immediately after the break. The upper plenum reached its minimum level at [ ]'6' about the same time as the maximum core steam percent (Figure 5.3.1-87). We upper head took longer to void (about [ ]'6#) because of its restrictive flow. In order to empty, the upper head had to gravity-drain either through holes in the upper support plate to the upper plenum and/or through bypass holes in the core barrel flange to the downcomer. The flow paths were sufficiently restrictive to delay :he upper head void well past the time that the upper plenum reached its minimum level. Differential pressure across the upper support plate and the flow bypass holes was monitored by DP-ll4 and DP-130, respectively (Dwg. OSU 600101). A positive value of differential pressure indicates that flow is moving in the direction of flow during normal operation before the RCPs trip. When the pumps were running, flow was from the downcomer, through the bypass holes into the upper head, continuing from the upper head into the upper plenum via holes in the upper support plate. A negative value for differential pressure indicates that flow is reversed from this description. The direction of flow into and out of the upper head was indicated by data from DP-114 and DP-130 (Figures 5.3.1-28 and 5.3.1-29). At the time of the break, differential pressure across the bypass holes was [ ]'6" and differential pressure across the upper support plate was [ ]'6" Thus, the flow was per design, from the downcomer to the upper plenum. At [ ]'6" flow reversed, and both differential pressures were negative. This flow reversal was due to the pressurization of the reactor core barrel above the downcomer pressure. o:ushRevN5h-5.non:lb-081798 5.3.1-18 REVISION 1
FmAL DATA REPORT p Differential pressure across the bypass holes remained negative for the remainder of the test. After (/ [ ]*** flow was always from the upper head into the downcomer annulus (Figures 5.3.1-28 and 5.3.1-29). By the time the upper head emptied at about [ ]'b' differential pressures across the bypass holes and upper support plate were at their first maximum negative value. A second maximum negative differential pressure occurred following ADS-4 opening. Both differential pressures steadily decreased in magnitude as the steam-flow rate to the downcomer decayed. Note: Data for differential pressure across the bypass holes given by DP-130 must be reduced by a value between [ ]'6' The magnitude of the correction factor is a function of water level in the upper head above the reference leg tap for DP-130. The variable leg tap for DP-130 is located in the downcomer. When this' tap is uncovered, the correction factor for DP-130 is [ ]' 6" This unusual correction factor is a result of mounting a differential pressure transmitter in a veitical position while measuring differential pressure in a column that does not stay full of water. A similar correction applies to DP-114; however, the correction factor is [ ]'6" of water. He full explanation for this correction factor is given in Subsection 2.4. Note: Data from DP-il4 and DP-130 provide a method to check the validity of data from LDP-Il6, LDP-140, and LDP-127. The method to use DP-130 and DP-114 as data verifiers can be found in Subsection 2.4. Both accumulators injected subcooled water that started ccre liquid level recovery (LDP-127; Figure 5.3.15). At [ ]'6" with the vessel drained below the cold-legs and hot legs levels, the upper head (TF-120 and TF-171) became superheated; however, the upper plenum became saturated (Figures 5.3.1-22 and 5.3.1-32). Even though the downcomer bulk fluid temperature remained subcooled, the upper [ ]*** of the downcomer region filled with superheated steam i (TF-168; Figure 5.3.1-22). The combination of voiding in the upper portion of the downcomer and the direct connection to the superheated upper head through the bypass holes caused the downcomer to fill with superheated steam. Eight downcomer fluid thermocouples (TF-147 to TF-150 and TF-164 to TF-167), mounted [
]'6" above the top of the DVI nozzles, allowed monitoring of the temperature in the downcomer.
TF-147 to TF-150 were mounted at 180*az above the DVI-2 penetration; TF-164 to TF-167 were mounted at 0*az above the DVI-1 penetration (Dwg. OSU 600101). The thermocouples were mounted at one of two elevations. TF-147, TF-148, TF-164, and TF-167 were located about [ ]'6"above the top of the DVI line penetration to the downcomer. TF-149, TF-150, TF-166, and TF-167 were located about [ ]b' above the top of the DVI line penetration to the downcomer. From [ ]'6# data from the eight thermocouples revealed temperatures in the downcomer were below saturated conditions (Figure 5.3.1-80). De thermocouples uncovered after the accumulators completed injection. He downcomer level at the time was just at the DVI nozzles, the k minimum downcomer level. The thermocouples were not covered again until about [ ]'6# o \l5hRevl\l5h-5.norrit>061798 5.3.1-19 REVISION 1
FINAL DATA RF. PORT Some temperature fluctuations may be due to mixture of saturated steam generated from fluid in the downcomer and superheated steam flowing from the SGs, through the cold legs, and to the break at the cold-leg balance line. The upper head remair. f wrheated throughout most of the transient, except when condensation events caused a sudden m - temperature; details of these events are described in Subsection 7.1. As ADS actuation occurred, core liquid levels continued to recover to a level of [ ]at [ ]' bd A second drop in the core level occurred just prior to the ADS-4 valves opening, even after the emptying of the accumulators and the single-injection feed from CMT-2. A recovery of the core liquid level began when IRWST injection started in the [ ]'** period. Recovery of the core and downcomer liquid levels peaked at [ ]'6# when both the downcomer became full and the hot legs and cold legs filled to an elevation equivalent of [ ]'6# (LDP-127; Figure 5.3.1-40). As IRWST draindown slowed, and the level in the primary sump increased, sump injection was initiated (at about [ ]'b'). As sump injection progressed, the upper head, core region, and upper downcomer cooled to a steady-state saturated temperature condition of [ ]'6* (Figures 5.3.1-44 and 5.3.1-46). Break and ADS Measurement System The BAMS provided a means to collect, measure, and redirect flow to the primary and secondary sumps for subsequent return to the reactor vessel via the DVI lines. Elevation of the separators (break separator, ADS 1-3, ADS 4-1, and ADS 4-2) and their cross-connecting pipe were modeled to simulate the physical conditions in the reactor cavity and containment areas. Prior to initiation of the break, all loop seals were filled in accordance with test procedures. Only the ADS 1-3 loop seal was heat traced. On the break opening, break flow from the cold-leg balance line was directed to the break separator, and flo" '.om the pipe attached to CMT-1 was directed to the ADS 4-2 separator (Figure 5.3.1-27). L.,vels in the break separator and primary sump (LDP-905 and LDP-901) increased immediately following the break (Figures 5.3.1-49 and 5.3.1-50). Steam separated in the break separator was directed through in-line vortex flow meters where the steam flow was measured before being discharged from the building. Figures 5.3.1-51,5.3.1-52, and 5.3.1-26 illustrate steam flows at various locations and times during the transient. FVM-901 and FVM-902 indicated a combined steam flow discharge rate of [ ]'b'about [ ]'b' after the break (Figure 5.3.1-51). Since these two parallel vortex flow meters were the last flow meters before the steam was released, all released flow from the BAMS system was measured. For Matrix Test SB10, the operator closed valves CSS-902 and CSS-906 (Appendix G, Dwg. OSU 600901) when the combined peak flow through FVM-901 and FVM-902 was less than e [ ]'6# At about [ ] this occurred, and, from that point on, flow exiting the building was directed through FVM-901. The majority of steam flow occurred from [ l l oA1536wRevl\l536w-5.non:Ib-081798 5.3.1 20 REVISION 1
FINAI. DATA Rrront I.
] with the largest portion flowing out of the break and to the break separator (FVM-905 and FVM-906; Figure 5.3.1-51). After initial release through the break, the only additional steam flow occurred from the ADS 1-3 separator during ADS valve openings and went into the IRWST through the sparger (FVM-601; Figure 5.3.1-26). Steam flow of [ ]'6' occurred over a [ ]
interval through FVM-701 following ADS 2 valve opening to ti:e IRWST (Figure 5.3.1-52). His is l probably due to air being pushed out of the IRWST as the level hereased. Some measurable steam i flow occuned from the transfer of hot liquid to the primary sump (FVM-903). These values were small companed with steam produced through the break. A total of [ ]'6' of steam exited the primary system and building through the break and ADS measurement system (BAMS) as a result of this transient. l. Figure 5.3.1-49 shows the liquid level changes in the IRWST due to ADS 1-3 valve opening (LDP-701), in the break separator due to break flow (LDP-905), and in the primary sump fed by separators (LDP-901). He secondary sump level (LDP-902) was unaffected until about [
]'6' when overflow from the primary sump occurred (Figure 5.3.1-50). The break separator level changed rapidly (LDP-905; Figure 5.3.1-49), when the break opened, which compares well with measured flove from the break separator to the primary sump (FMM-905; Figure 5.3.1-27). The initial l peak level in the break separator occurred from [ ]'** with a level of [ l'6'
! The primary sump level increased immediately following the break, initially due to flow from the break separator. No flow to the primary sump occurred from the ADS 4-1 separator (FMM-603) and ADS 4-2 separator (FMM-602) early in the transient. Flow from the ADS 1-3 separator to the IRWST caused the IRWST level to increase from [ ]'6
- at [ ]'6' when IRWST injection started (Figures 5.3.1-27 and 5.3.1-49).-
When the primary sump level reached about [ ]'6" (about [ ]'6'), sump liquid covered the loop seal pipe from the break separator (Figures 5.3.1-49 and 5.3.1-50). Once this occurred, any
- liquid added to either the primary sump or the break separator was reflected in the liquid levels (LDP-901 and LDP-905). In effect, both levels increased together until the primary sump reached the L
overflow level of [ j'** above the centerline of the DVI line. He secondary sump received overflow from the primary sump. Overflow to the secondary sump first occurred at about [ ]' as the primary sump level reached [ ]'6' As primary sump injection was initiated at [ ]'** the level in the sump decreased slightly. He primary sump inventory was replaced with a [ ]'6' flow from the ADS 4-1 separator (FMM-603; Figure 5.3.1-45). Core Makeun Tanks [ nermal-hydraulic behavior in the CMTs was quite asymmetrical throughout Matrix Test SBIO due to the break location; therefore, data from each CMT will be described separately. l lO r. l a:ushneviusw.5.non:masi?98 5.3.1-21 REVISION 1 l l
FINA1. DATA REPORT CMT-2 Related Response 'Ihe system was filled with subcooled liquid at [ ]"' prior to break initiation. Isolation valves in the discharge lines were opened [ ]"" after the break, and natural circulation flow started in the loop formed by CMT-2, CMT-2 discharge, DVI line, pressure vessel, CL-1, and cold-leg balance line. CMT-2 injection flow (FMM-504) and the axial temperature profile (TF-504 to TF-532) are illustrated in Figures 5.3.1-11, 5.3.1-53, and 5.3.1-54. With the onset of natural circulation flow through the CMT, subcooled water started to flow through the cold-leg balance line into CMT-2, as shown by a rapid temperature increase (TF-532) and a discharge flow rate of [ ]"' (FMM-504). Within [ ]"' after break initiation, the top [ ]'6' percent of CMT volume was at temperature of [ ]"' The CMT transitioned from recirculation to draindown at [ l'6' as indicated by sharp decrease in the balance line level (LDP-510; Figure 5.3.1-8). During this transition from recirculation to draindown, the CMT-2 injection flow rate increased to [ ]"' (Figure 5.3.1-11). As cold-leg levels dropped, saturated steam filled the cold-leg balance line and, at about [ ]"' became slightly superheated, as shown by TF-536 and the thermocouple at the top of CMT-2 (TF-546; Figures 5.3.1-55 and 5.3.1-56). During accumulator injection and ADS 1-3 actuation ([ ]"*), the CMT-2 level remained constant but the space above the [ ]"* water level became superheated ([ ]'6' increase over saturation temperature). Figures 5.3.1-55 and 5.3.1-56 show the interaction between the water layer and the upper head of the CMT. The upper portion of the CMT-2 dome was exposed to [ ]'6' liquid (TF-546), and the inside wall surface temperature quickly changed from [
]'6' (TW-556; Figure 5.3.1-55). The outside wall temperatures also increased but at a slower rate, beginning at [ ]"' and increasing to a peak of [ ]"" over [ ]"'
(TW-554). The outside wall temperature on the CMT-2 dome at a corresponding radial direction (TW-548) showed excellent agreement with TW-554 (Dwg. OSU 600502). Quickly heating the metal mass at the top of the CMT allowed storage of thermal energy that was retumed to the system by heating up the saturated steam entering the dome of the CMT, beginning at [ ]'6' (TF-556 and TF-558; Figure 5.3.1-55) and continuing until about [ ]'6' when they became subcooled (Figure 5.3.1-53). The upper layers of saturated steam in the CMT were superheated ([ ]"") due to exposure to higher wall temperatures and rapid depressurization during ADS-1 and ADS-2 valve actuation. Characterizati on of the early stages of the CMT-2 superheating effect was performed by review of the thermocouples along the axial length of the CMT, upper CMT dome inside-wall temperatures, and a combination of fluid and wall temperatures at the top layers (Figures 5.3.1-57, 5.3.1-58, and 5.3.1-59, respectively). All fluid thermocouples above the [ ]"' percent volume layer indicated temperatures above the saturation temperature from [ ]"' however, most of these thermocouples (TF-516, TF-518, TF-522. TF-526, and TF-530) reflected superheated conditions at [ ]'6" about [ ]'6" after the ADS 2 valve opened and rapid REVISION 1 oats %wRevl\l5Mw.5.non:Ib481798 5.3.1-22
FrNAL DATA REPORT depressurization of the system occurred. Dermocouples TF-556 and TF-564, which are located very Os close to the same elevation as TF-516 (except closer to the walls), exhibited an earlier departure from saturation at [. ] with a temperature of [ ]'6" (Figure 5.3.1-57). Upper-dome inside-wall temperatures indicated a peak wall temperature of [ ]*' (Figure 5.3.1-57) and slowly decreased to about [ ]'6' Both the upper-dome and upper-inside-wall temperatures reached peak temperature values of [ ]'6' between [ ]'6' (Figure 5.3.159). Herefore, a [ ]'6' difference (fluid temperatures > wall temperatures) existed between the wall and the saturated steam volume above the water surface at the [ ]'6' point. This common effect of CMT steam heating is described in detail in Subsection 7.2. As accumulator injection was completed and prior to ADS-4 actuation, the upper portion of the cold legs superheated to a temperature of [ ]'6' (Figures 5.3.1-34 to 5.3.1-37). On ADS-4 actuation at [- ]'6' he cold legs dropped to saturation conditions but the upper CMT-2 temperatures remained [ ]'6# above saturation temperature due to early dome metal heating. De effect of upper dome heating and subsequent superheating of saturated steam in the upper portion of the CMT had less effect on the lower portion of the CMT. As the CMT continued to drain, the lower inside surface reached saturation temperature late in the transient due to increased surface area to be heated and lack of cold-leg flow entering the CMT during the transition from recirculation to draindown. At about [ ]'6' the upper layers of the CMT dropped to slightly above saturation temperature and remained at these conditions until the CL-1 balance line began to refill (Figure 5.3.1-8). Temperatures O in the top of CMT-2 (TF-532; Figure 5.3.1-53) indicated superheated steam had re-entered CL-1 k_.) (SC-105; Figure 5.3.1-34) from the reactor vessel via the bypass holes (TF-168; Figure 5.3.1-22) from [ J'6' or from the SGs. Because the cold leg refilled with superheated steam, steam also filled the CMT-2 balance line. During this same period, the hot and cold legs refilled with water due to injection from the IRWST and CMT-1. At about [ ]'6' CL 1 refilled and sealed off the entrance of the CMT +
- balance line with saturated liquid and continued to fill the balance line. This isolated the superheated steam in CMT-2 by the liquid seal and the check valve located in the discharge line of CMT-2 (Dwg.
OSU 600206). As IRWST and CMT-1 injection continued and the core / hot-leg levels increased, liquid continued to fill the balance line. De trapped steam contained in CMT-2 decreased in temperature due to heat loss through the CMT walls. nis decrease in steam temperature lowered the pressure which caused more liquid to fill the balance line. At about [ ]'6' the balance line became filled and some liquid flowed into the top of CMT-2 (LDP-510; Figure 5.3.1-8). As more of the liquid entered the top of CMT-2, the slightly superheated steam continued to condense, reducing the local pressure and increasing flow into the CMT. At [ ]"' the CMT level was about [ ]'6' and all CMT temperatures sharply decreased from saturated conditions to [ ]'6" (Figure 5.3.1-53). His sudden temperature decrease combined with the sharp increase in CMT-2 level (from [ ]'6') suggested that the steam bubble existing in CMT-2 collapsed. Balance line flow showed a small flow of about [ ]'6' As a possible Q condensation event occurred, a localized low pressure was created due to the collapsed bubble causing a[ ]'6' increase of flow at [ ]'6' through FMM-502. This possible condensation o:u5hRevlu5h-5.non:lt> 061798 5.3.1-23 REVISION 1
FINA1. DATA REPOIT event was also evident at other locations in the system. Sudden refill due to the condensation event subcooled the CMT (Figure 5.3.1-53). As pressure in CMT-2 and the RCS achieved equilibrium, refill ended. After the termination of refill, CMT-2 remained stable with subcooled water with only a [ ]'6' difference from bottom to top of the liquid layer (Figure 5.3.1-53). There was no flow to the reactor vessel due to the higher localized line pressure in the vessel caused by IRWST injection flow. From [ ]'6' the cold leg balance line slowly filled with saturated steam that entered and condensed in CMT-2 at about [ ]'6' (Figure 5.3.1-54). This additional flow into CMT-2 causeda[ ]'6' inemase in the top layers of the CMT and only a slight increase in the CMT level. CMT-2 injection began at [ ]'6" as IRWST injection decreased to the low-low level. CMT-2 emptied at [ ]'6" and the remaining liquid and vapor stratified (Figure 5.3.1-54). CMT 1 Related Reso(>nse The system was filled with subcooled liquid (69 F) priol to break initiation. On break initiation, balance-line flow was directed to the break separator, and any liquid in the horizontal portion of the balance line drained into either the ADS 4-1 separator or CMT-1. The break isolated CMT-1 with the top open to atmospheric pressure, but the initiation of injection ficw was controlled by reactor pressure. Due to high pressure in the reactor vessel /DVI line, no flow occurred even after the isolation valves in the discharge lines opened at the [ ]'6" mark. CMT-1 remained isolated and the liquid in the tank remained isothermal at 69'F (Figures 5.3.1-60 and 5.3.1-61). About [
]*6* after ADS-4 actuation ([ ]'6'), pressure in the reactor vessel and DVI line was less than CMT-1 head elevation. A discharge flow ([ ]'6" peak at [ ]'6') from CMT-1 occurred (Figures 5.3.1-8 and 5.3.1-11). CMT-1 flow was limited by DVI line pressure due to parallel injection of the IRWST. CMT-1 draindown continued until [ ]'6' when the tank emptied. The tank remained empty for the remainder of the transient.
Accumulators The accumulators were located below the CMTs in the system and provided water injection inte the DVI lines by an expansion of nitrogen gas volume stored in the accumulator at an average pressure of [ ]'6' set at initial conditions. Water from the accumulators was expelled into the injection l line when primary system pressure dropped below [ ]'6' Accumulator injection started before ADS-1 actuation with a low flow rate of [ ]*6' per accumulator; however, when ADS 1 actuated, the injection rate sharply increased (Figure 5.3.1-11). The accumulator injection rate reached a peak injection flow rate of [ ]'6'immediately following ADS-2 valve opening. Accumulator injection lasted about [ ]'6' h & mtied at [ ]*6' (Figure 5.3.1-62). O l cA1536wRev l\l 536w-5.non: I b-081798 5.3.1-24 REVISION 1 J
_ _ . _ _ _ . . _ _ _ . _ . _ _ _ _ _ _ _ _ . . . - . _ _ _ _ _ _._-.._ ___ m_. _ .. FINAL DATA Rt. roar l
. I Pressurizer The pressurizer immediately drained to the reactor vessel after the break occurred and was completely
' drained in about [ ]** (LDP-601 and LDP-602; Figure 5.3.1-7). Water in the pressurizer flashed due to loss of system pressure, and the temperature of the water (TF-602) dropped from a value of [ ]** before the break to [ ]*# (Figure 3.5.1-63). He pressurizer was superheated until about [ ]'6' then remained at saturation temperature.
When ADS-1 actuated at [ ]'6' temperatures in the upper portion of the pressurizer i decreased as the steam flow from the reactor vessel entered and exited the top of the pressurizer. Just I prior to ADS actuation, the hot legs (LDP 206) and the upper plenum (LDP-ll3) were filled witn [ ]*' percent steam (Figure 5.3.1-64). The hot-leg elbow /SG-2 channel heads were partially filled with two-phase fluid (LDP-208 and LDP-214). When the ADS-1 valve opened at [ ]** the pressurizer surge line and the pressurizer initially filled with a two-phase mixture. The combination of the ADS valve actuation with the injection of the accumulators and CMT-2 caused a high influx of liquid flow through the reactor, which discharged two-phase flow through the pressurizer and exited through the ADS valves. A steam percentage versus time plot of the pressurizer and surge line confirmed the filling of the pressurizer with liquid from [ ]'6' (Figure 5.3.1-65). Just prior to [ ]'6# steam percentage in the pressurizer and surge line was [ ]** percent, reflecting an empty condition (Figure 5.3.1-7). The surge line and pressurizer refilled with two-phase flow, decreasing the steam percentage to about [ ]'6" percent at [ l p(/ actuation caused a small increase in liquid and steam flow through and into the ADS 1-3 separator
]'6' ADS-1 valve (Figure 5.3.1-66). Figure 5.3.1-66 shows a comparison of the liquid and steam flow exiting through the ADS valves and ADS 1-3 separator downstream from the ADS 1-3 valves.
At[ ]** when the larger ADS-2 valve opened, system pressure decreased and ACC-1 and ACC-2 injection flow increased rapidly from [ ]"' which caused a continued decrease in the pressurizer steam percentage. This decrease in steam percentage indicated that the pressurizer was being filled with an increasing amount of liquid supported by a two-phase steam environment in the smaller-diameter surge line. The measured flow of liquid into the ADS 1-3 separator from the pressurizer during this period also supports this conclusion. Steam flow peaks were consistent with ADS-1 opening ([ ]*'), ADS-2 opening ([ ]'6#), and ADS-3 opening ([ ]'6'). De liquid flow rate for flow exiting the separator showed a peak flow rate of [ ]'6# at about [ ]'6# that tapered off to 0 at [ ]'6" MDS4 vain opening). The drop in the steam flow rate (FVM-601; Figure 5.3.1-66) at [ ]** suggested that two-phase flow through the pressurizer ended, and only liquid was collected in the pressurizer. At[ ]'6# steam percentage in the pressurizer reached a minimum value of [ ]** percent, supported by a surge line with a steam percentage of [ ]'b' percent (Figure 5.3.1-65). From [
]'b' accumulator injection decreased rapidly but still provided adequate driving force to transfer liquid from the pressurizer to the ADS 1-3 separator for a period of time. As flow c:U5hRevn15h.5.non:Ib-081798 5.3.1-25 REVISION 1
I FINAI, DATA REPORT from CMT-2 continued into the reactor, the partially filled pressurizer retained the liquid, and the two-phase environment in the surge line maintained this level. At[ ]'6" ADS-4 valves opened and caused a further decrease in reactor pressure and diverted steam filling the hot legs to the ADS 4-1 separator. As system pressure decreased and IRWST injection flow started, the steam bubble in the surge line became unstable, and liquid from the pressurizer began to drain slowly into the hot legs through the surge line. System depressurization due to ADS-4 actuation ended both steam and liquid flow to the ADS 1-3 separator (Figure 5.3.1-66), and the steam percentage in the pressurizer increased as the liquid volume decreased (Figure 5.3.1-65). Draining of the pressurizer from [ J'6" was reflected by an increase in liquid content in the HL-2 elbow (Figure 5.3.1-64) and surge line (Figure 5.3.1-65). At [ ]'6" the pressurizer was completely voided, and the surge line was filled with [ ]'** percent water and [ ]'6* percent steam (Figure 5.3.1-65) with an indicated liquid level of about [ ]'** (Figure 5.3.1-7). This condition remained stable from [
]'6" when primary sump injection caused a drop in surge line level from [ ]'6" From [ ]'** through the end of the transient, the pressurizer remained voided (Figune 5.3.1-67).
Passive Residual Heat Removal Heat Exchanger At the initiation of the event, the PRHR HX was filled with subcooled liquid. After [ ]'** RCS-804 (Figure 5.3.1-12) opened, and quickly became over-ranged with a high flow rate of [ ]'** after about [ ]'b' Figures 5.3.1-68 and 5.3.1-69 show the temperature profile for the PRHR HX throughout the transient. Within [ ]'** the PRHR HX inlet temperature had increased from [ l'6* while the outlet temperature line was about [ ]' 6* As the primary system began to drain (at about 120 seconds), the hot legs filled with two-phase fluid and a portion of this fluid flowed from HL-2 to the PRHR HX (Figure 5.3.1-12). PRHR HX inlet and outlet lines contained FMM-802 and FMM-804, respectively, which can produce erratic or inaccurate data for two-phase flow conditions. Therefore, the PRHR HX inlet flow rates appeared erratic and sometimes over-ranged after the line became filled with steam after ([ ]* 6*h Since flow out of the PRHR HX remained subcooled for the entire test (Figures 5.3.1-68 and 5.3.1-69), data from FMM-804 (Figure 5.3.1-12) were not affected by fluid phase. However, the measured flow rate through the HX outlet line was small and oscillating. By [ ]'6* flow oscillated rapidly between negative [ ]'** and positive [ ]*6* making it difficult to j determine how much mass was passing through the HX. FMM-802 (Figure 5.3.1-12) measured flow from HL-2 to the PRHR HX. As HL-2 became voided, PRHR HX inlet flow data will become invalid. It cannot be determined precisely when the hot leg voided sufficiently to affect the inlet flow measurement. Although the PRHR HX inlet temperature reached saturation temperature at [ ]'** (Figure 5.3.1-68), the state of the fluid was o:\l536witevl\l536w.5.non:Ib-081798 5.3.1-26 REVISION I
FINAI, DATA REPCaT uncertain. The saturated temperature of the fluid allowed it to be liquid, steam, or two-phase. However, steam percentage calculations using the SG-2 hot-leg elbow indicated that voiding of the hot legs was well established when ADS-2 actuated at [ ]'6" at which time the rate of voiding in the hot leg increased (Figure 5.3.1-14). As FMM-802 was exposed to two-phase and/or steam flow, the output was no longer valid; therefore, flow values are not valid after [ ]'6# ne two-phase mixture was condensed and subcooled in the PRHR HX and retumed at a flow of about [ ]'6# of [ ]'6" liquid to the bottom of the cold-leg channel head in SG-2. During the initial depressurization and prior to ADS-1 actuation, a significant variation in the inlet flow was observed, which was caused by the variation in the void fraction in the fluid in HL-2. When the ADS actuated, the reactor and the hot leg became subcooled by cold flow from the accumulators. The driving head for flow in the PRHR HX decreased and almost stopped at about [ ]'6* Subcooled fluid in the hot leg never filled the PRHR HX supply line as system pressure decreased (Figure 5.3.1-70). He PRHR HX levels (LDP-801 and LDP-802) showed that the inlet header was drained from [
]'6' and the HX tubes drained to about [ ]'6' (Figure 5.3.1-70). This time span is coincident with draindown of the cold legs and hot legs. Spikes were also observed in the PRHR HX outlet flow at [ ]'6' which cre coincident with condensation events that occurred at various locations in the system (Figure 5.3.1-12). TF-808 (on the shoit tube at mid-g length) showed a sharp increase from [ ]'6# beginning at [ ]'6# and remained at this temperature until [ ]'6" (Figure 5.3.1-68), shortly after ADS-4 actuation
([ ]'6'). De level measured by LDP 802 (Figure 5.3.1-70) was about [ ]'6' which suggested that TF-808 was uncovered. TF-809 remained subcooled at [ ]'6" (Figure 5.3,1-68), possibly because TF-809 was in the long leg of the PRHR HX and the length of cooling path was longer compared with the short leg (TF-808). At about [ ]'6' the inlet temperature of the short tube increased and the inlet temperature of the long tube decreased (response of the two tubes reversed). The short tube was the one with the significant temperature difference between inlet and outlet; long-tube temperatures correlated to IRWST temperatures at their thermocouple location. Long-tube temperatures indicated no significant temperature across the tube for the rest of the test, except to reflect IRWST temperature at the long tube thermocouple location. From [ ]'6' the PRHR HX refilled to about [ ]'6' of the total liquid level; however, the upper head level partially filled before the main body of the HX, which reached a level of[ ]'6" (Figure 5.3.1-70). He level increase in the PRHR HX at [
]'6' is not clearly understood, however, one explanation may be a condensation event.
Because the inlet channel head completed filling at the same time that the tube level started increasing (Figure 5.3.1-70), flow reversal through the HX was a possible explanation for both level events. His would have required pressure in the SG-2 cold-leg channel head to be greater than the pressure in v HL-2; however, the pressure in SG-2 was never greater than HL-2. Backward flow through the HX o \l5hRevl\l5h-5.non:lb 081798 5.3.1-27 REVtSION 1 ( ... .
FINAL DATA REPORT would have resulted in the outlet temperature of the HX (from the SG 2 cold-leg channel head) being greater than the inlet temperature (from HL-2). This was never the case because the outlet temperature was always subcooled; the inlet temperature reached saturation temperature at [ J'6* (Figure 5.3.1-68). Data from the fluid thermocouples in the HX provided no further evidence that there was any flow reversal through the HX and no insight into the reflooding of the tubes at [ ]'6* A condensation event that occurred at [ ]'6# caused the level to drop to [ ]'6" l (Figure 5.3.1-70), and FMM-804 reflected a discharge flow of [ ]'6' over [ ]'6# l (Figure 5.3.1-12). After this event, the levels in both the inlet head and HX body refilled slowly from [
]'b' as the hot legs and cold legs refilled. j l
At[ ]'6' when teactor pressure had decreased and flow injection from the IRWST had l slowed, the PRHR HX began to inject condensed water back into the vessel via the SG-2 channel head j (LDP-802; Figure 5.3.1-71), (TF-803; Figure 5.3.1-69), and (FMM-804; Figure 5.3.1-72). As IRWST flow ended and primary sump injection initiated, the PRHR HX continued to condense l
- steam generated in the core, but at a slower rate since the HX was transferring heat to the air instead j of to the IRWST water.
In summary, data indicated that the PRHR HX provided a cooling flow to the RCS that continued through the entire test. Outlet flow and level of the HX oscillated, probably from a chugging effect as steam from HL-2 condensed in the HX. He low flow rates suggest that the HX was more effective l early in the transient. ) I Steam Generators l One long tube and one short tube in the SG were instrumented for level and temperature measurement I (Dwg. OSU 600301). Transmitters measured the level in the hot-leg and cold-leg sides of the tubes; I I thermocouples in the tubes measured temperature at the top of the tubes and at mid-elevation of both the hot-leg and cold-leg sides. Seventy seconds after the break valve opened, the SG tubes started to drain in both SGs I (Figures 5.3.1-16 and 5.3.1-17 contain data for SG-1). Both SGs, including the hot-leg and cold-leg channel heads, were essentially drained before ADS-1 actuation at [ ]'6# The sequence of tube and channel-head draining is provided in Table 5.3.1-3. Before the break occurred, steam production in the secondary side of the SG acted as a heat sink to the primary side. Feedwater and steam isolation occurred automatically after the break was initiated (Table 5.3.1-3). He isolation of steam and feedwater " bottled up" the SGs. Until the RCS o:\l 536w Rev i\l 536w-5.non:l tro81798 5.3.1-28 REVISION 1
FINAL DATA REPORT depressurized below the pressure of the SG's secondary side, the only energy loss from the SGs was Q ambient heat loss through the insulation. When RCS pressure decreased below secondary steam pressure, the SG secondary side became a heat source to the RCS. The SGs acted as a heat sink until pressures in both the SG and primary side reached a common pressure. After the break, pressure of the secondary side increased toward the primary-side hot-leg pressure (Figure 5.3.120) which, at the same time, was dropping. Temperature on the primary side was higher than the secondary side, indicating that heat transfer occurred from the primary to the secondary side. After [ J'6# pressure on the SG secondary side and the U-tube primary side reached a common value of [ ]'6' and the temperatures reached a common value of [ ]'6' (Figure 5.3.1-21). During the first [ ]'6' the SG U-tubes partially emptied and broke the flow through the SG. Transition from recirculation to draindown of the SG and the large heat source of the SG caused the U-tubes to become superheated, and they remained in this condition throughout the transient. With both the hot leg and cold-leg channel heads empty, superheated steam filled the entire primary side of the SG, which caused fluid temperatures in the region of the RCP inlets to superheat (Figures 5.3.1-34,5.3.1-35,5.3.1-36, and 5.3.1-37). As described in the cold-leg response subsection, the SGs supplied superheated steam to the cold legs via the differential pressure between the hot legs and the cold legs. Temperature in both the primary and secondary side continued to decrease at a rate of about [ ]'** for the first [ ]'6' The SG U-tubes liquid level instruments (Figure 5.3.1-14) were not providing a tnie indication of level V due to loss of fluid in the reference and variable legs; therefore, the SG U-tube level "alues are not applicable beyond [ ]'6' (This generic instrument measurement effect is described in Section 2.4.) In summary, the SGs completely drained by ADS-1 actuation. Steam in the tubes superheated when RCS pressure decreased below secondary-side steam pressure. Superheating of all tubes was completed at [ ]'6' before ADS-1 actuation, and the tubes remained superheated for the rest of the test. 'Ihe SGs supplied superheated steam to the cold legs, with the differential pressure between the hot legs and cold legs supplying the head for flow. Cold Lees and Hot Lees The upper head became saturated at about [ ]'6" following the break. In [ ]*** the upper head drained, and the cold legs (initially full on single phase liquid) showed an initial drop in levels (Figure 5.3.1-13). HL-1 and HL-2 indicated a similar draining at [ ]'6" (Figure 5.3.1-14). 'Ihe pressurizer drained about [ ]'6' after the break; the SG U-tubes and channel head drained by [ ]'** As the levels in the cold legs dropped below the top of the pipe, flow through the balance line to CMT-2 stopped due to the transition from recirculation to draindown. As the SG drained, pressure in both the reactor vessel and the secondary side of the SG reached a common pressure and temperature ([ ]'6'); little or no heat transfer occurred. After about 140 seconds, saturated steam in the SG U-tubes superheated. As the SG continued to respond as a heat source, superheated steam filled the primary side of the SG just above o:ushneviush 5.non:st>4sms 5.3.1-29 REVISION 1
s FINAt. DATA REPORT the top layer of the partially drained cold legs (TF-201, TF-202, and TF-204; Figures 5.3.1-34, 5.3.1-35, 5.3.1-36, and 5.3.1-37). Based on temperatures in the upper portion of the downcomer, superheated steam entered and occupied the top layers of the cold legs beginning at [ J'6" (Figure 5.3.1-22). Superheated steam conditions were maintained in the cold legs due to either metal temperatures in the reactor i vessel upper head via the bypass holes or superheated steam flow from the SGs. Both sources were superheated at that time. The instability of the cold-leg temperatures, beginning as early a [ ]'6' indicated some voiding was occuning in the cold legs. The liquid level in the cold l I legs was initially replaced with two-phase saturated fluid with some of the upper layer approaching superheat conditions. In addition, from [ ]'6# the top layer of the cold legs exhibited peak temperatures of l [ ]'6# the same as the downcomer temperature (TF-168; Dgure 5.3.1-22). This suggested that, l during this period, superheated steam was filling the very top layer of the cold legs. With the open connection between the upper head and the SG through the voided cold legs, DP-130 (Figure 5.3.1-28) indicated steam entered the cold leg from the upper head. Over the period from [
]'6' however, DP-201, DP-204, DP-207 and DP-208 all indicated a positive differential pressure equal to [ ]'6' of water, suggesting that some steam flow may have entered the cold legs from the SGs.
The formation of superheated steam in the cold legs can be explained by the rapid depressurization of the RCS; however, the cold legs remained superheated for the duration of the test (Figures 5.3.1-34 to 5.3.1-37). This could have occurred only if a source of thermal energy was available to the cold legs. The only possible sources of thermal energy to the cold legs were:
. Heating of steam in the cold legs by cold-leg piping
. Addition of superheated steam from the SGs
. Addition of superheated steam from the upper head via the bypass holes in the core barrel flange Superheating of the cold legs from the upper head is not likely. The temperature of the upper head is lower than the fluid temperatures of the cold legs at the reactor vessel fiange (Figures 5.3.1-34 to 5.3.1-37).
If the steam in the cold legs was maintained at a superheated condition by the walls of the cold legs, the temperature on the inside wall of the pipe would have been greater than the cold-leg fluid temperature. 'Ihere were no inside-wall thermocouples in the cold-leg piping to measure temperature on the inside wall of the cold leg. The only cold-leg wall thermocouples were part of the heat flux meters (HFM) mounted on the outside wall of the piping. These thermocouples were given a TFM designation. A thermocouple was meanted on each cold leg at the top of its flanged connection to the reactor vessel (Dwg. OSU 600101, Sh. 2). The recorded temperature of the thermocouples mounted at oA1536w Rev l\l 536w-5.non: I b.081798 5.3.1-30 REVISION 1
~.- -. - -- - . - . . = . - . FINAL. DATA REPORT the top of the reactor vessel flange was greater than the recorded temperature of the thermocouples mounted on the side of the pipe, so, the hotter flange thermocouples were used for evaluation. Figures 5.3.1-73,5.3.1-74, and 5.3.1-75 show the effects of vessel / piping metal on the cold-leg temperatures and support the previous discussion. Early in the transient, with the reactor temperatures in the vessel and cold legs at [
]'6' (TF-171 and TFM 111; Figures 5.3.1-73 to 5.3.1-75). TF-171 was located [ ] above the upper support plate and measured localized fluid temperatures but was significantly influenced by the metal plate temperature based on the distance from the plate. TFM-11I was located on the exterior of the CL-1 piping flange, roughly [ ]'6' from the downcomer/ cold-leg interface. TFM-Ill indicated a lower external temperature than the interior wall temperature, based on through-wall heat conduction.
As the upper head and plenum drained in [ ]'6' the upper support plate and cold-leg piping metal temperatures dropped slightly to [ ]'6# respectively (Figure 5.3.1-73). The reactor vessel continued to depressurize via ADS 1-3 valve opening, allowing saturated steam in the upper head to reach superheated conditions. His superheated steam filled the top of the downcomer via the bypass holes until it reached the top of the cold leg. He hotter metal temperatures in the cold legs maintained the upper layer of the cold legs at superheated conditions. The ADS-4 valve openings continued to reduce reactor pressure and permitted the initiation of IRWST injection flow, which p started refill of the loops and cooling of the superheated steam in the cold legs at [ ]'6' (SC-105; Figure 5.3.1-74). Cold-leg temperatures remained slightly above saturated conditions until the cold legs refilled with subcooled water beginning at [ ]'6' De cold-leg fluid temperature (SC-105) and piping metal temperatures (TFM-Ill) subcooled at [ ]'6' after the hot legs and cold legs completely refilled (Figure 5.3.1-75). Both the cold and hot legs drained at about [ ]'b' The hot legs were filled with saturated steam from the reactor vessel or possibly the SGs (SC-140, SC-141, SC-205, SC-206; Figure 5.3.1-76). The loops voided until after accumulator injection, ADS actuation, and IRWST injection provided adequate inventory for refill (LDP-201, LDP-202, LDP-203, LDP-204, LDP-205, LDP-206; Figures 5.3.1-13 and 5.3.1-14). De effect of hot-leg piping metal temperatures (TFM-109 and TFM-113) also had a significant effect on the temperature profiles in the hot legs (Figure 5.3.1-76). Figure 5.1.3-76 shows that, early in the transient, externally mounted thermocouples on the hot legs (TFM-ll3 and TFM-109 located on the top of the HL-1 and HL-2 flanges, irspectively) indicated temperatures of [ ]'6' These temperatures remained essentially unchanged even though fluid temperatures in the hot legs were slightly superheated and decreased along the saturation line. Because the metal temperatures remained in the mid-[ ]' levels, hot-leg temperatures remained saturated (from [ ]'6') while the hot legs were voided (Figures 5.3.1-77, 3 ' 5.3.1-78, and 5.3.1-79). De response of TF-171 (which measures the fluid temperature locally above the upper support plate) showed an early temperature increase above the saturation temperature line, o:ushRevlu5h-5.non:lb-08198 5.3.1-31 REVISION 1
FINAL DATA Rr.roRT suggesting that the upper support plate, having been exposed to the higher fluid temperature, helped to superheat any saturated steam passing through and into the upper head. Figure 5.3.1-80 provides a plot comparing the output of several thermocouples in the downcomer annulus at an elevation equal to the hot-leg centerline (TF-147, TF-148, TF-166, and TF-167), some 3 in. below the hot-leg centerline (TF-149, TF-150, TF-164, and TF-165), and two located about [ ]'b' below the DVI line (TF-134 and TF-152). All temperatures except those below the DVI line exhibited similar subcooled temperatures until about [ ]'6# when the thermocouples approached saturated temperatures. At [ ]'6' when the downcomer annulus reached its minimum level (LDP-140 and LDP-116; Figure 5.3.1-5), the annulus thermocouples reached supes ' 4ed temperatures, which existed in the upper downcomer region and cold legs (TF-168; Figr : 5.3.1-22). Immediately following ADS-4 actuation at [ ]'6# the downcomer started torFil and the thermocouples were quenched and returned to subcooled conditions. This confirms that the cold legs were voided, and the minimum downcomer annulus liquid level was below the bottom of the DVI line. Thermocouples in the core region exhibited relative uniform temperature, except TF-170, which was at the same vertical elevation as TF-169. These thermocouples penetrated the vessel and the core barrel before entering the fluid. A seal was designed and installed around thermocouple TF-170 as it penetrated the core barrel; however, the 40 F difference between TF-169 and TF-170 suggested that the seal may not be as effective as intended (Figure 5.3.1-76). TF-155 indicated the downcomer temperature in the general proximity of the TF-170 thermocouple penetration (Figure 5.3.1-76). The temperature profile and the significant temperature difference in the downcomer indicated that some cooler downcomer fluid was bypassing the seal and affecting the output of TF-170. Therefore, the results of TF-170 should not be directly used for comparison with TF-169. The hot legs showed initial voiding at about [ ]'b' and appeared to be empty at [
]'6" Up to [ ]'6' the average fluid temperatures were about [ ]'6' higher than the saturation temperature as indicated by TSAT (Figure 5.3.1-77). It should be noted that TSAT (PT-107) may have a small offset. With the hot legs empty and the hot-leg piping exhibiting temperatures in excess of [ ]'6* the slightly superheated steam continued to follow the saturation curve (Figure 5.3.1-78).
Between [ ]'6' the hot legs and cold legs began refilling and reached the filled condition at about [ ]'6' Figure 5.3.1-78 shows the transition of both the top and bottom of the hot legs from superheated to saturated to subcooled from [ ]'6" This figure also indicates that heating of the upper support plate and superheating of the upper head had little direct effect on the temperature profiles of the hot legs during this period. This was not the case for the cold legs, as described earlier in this section. At [ ]'6' sharp changes in temperature occurred due to a condensation event in CMT-2. Condensation events a:u5hRevnl536w-5.non:ltK)81798 5.3.1-32 REVISION 1 l
FIN 41. DATA REPORT such as in CMT-2 were reflected throughout many subsystems. He hot legs remained subcooled until about [ J (Figure 5.3.1-79). Levels in both the hot and cold legs remained stable until about [ ]'b' when flow from the IRWST and CMT-1 decreased, and CMT-2 flow (from refill) initiated, caused a slight drop in the vessel level (LDP-127, LDP-116, and LDP-140; Figure 5.3.1-40). As primary sump injection started at [ ] cold-leg levels dmpped slightly to just above the bottom of the pipe ([ ]*' relative to LDP-127), resulting in a partially filled cold leg (LDP-201, LDP-202, LDP-203, and LDP-204; Figure 5.3.1-81). Cold leg temperatures remained subcooled from about [
]*' (Figure 5.3.1-82). As the sump recirculation mode occurred, saturated water at about
[ ] injected, causing the cold leg to reach saturated temperature conditions. Similarly, hot-leg levels remained partially filled ([68.0 in'.) relative to LDP-127) (LDP-205 and LDP-206; Figure 5.3.1-83). Ievels in the hot leg supported liquid flow to the primary sump via the ADS-4 valves and primary sump injection from [ ]'6' (end of the test). Hot-leg temperatures remained subcooled from [ ]"# as the core heating with reduced injection flow caused the hot-leg temperatures to increase (Figure 5.3.1-84). The hot legs remained at a uniform temperature of about [ ]"# for the remainder of the test. In-Ce=* '- at Refueline Water Storane Tank Hermal response of the IRWST was influenced by the addition of heat from the PRHR HX and the direct injection of hot liquid from the ADS 1-3 separator. Initially, a [ ]"' axial temperature gradient existed from the top to the bottom of the IRWST, with the bottom at [ ] (Figure 5.3.1-85). Opening of the PRHR HX outlet valve caused a slight localized temperature increase in the upper one-third of the tank. As ADS-1-3 valves opened, the upper layers of the tank increased from [
]'6' locally. This localized temperature increase created a stable temperature difference of
[ ]# from the middle to the top layers, after the conclusion of the ADS phase. With continued heating from the PRHR HX, the upper half of the fRWST increased to an . average temperature of [ ] (TF-710, TF 709, and TF-707; Figure 5.3.1-85). After [ ]"' the IRWST drained below the PRHR HX tubes, and about [ ]*' of [ ] water remained in the tank (Figure 5.3.1-86). Thermocouples TF-705, TF-707, TF-709 and TF-710 were exposed to tank ambient air temperatures (Figure 5.3.1.86). From [ ]*' through the end of test, these thermocouples increased to an average temperature of about [ ]*'
- (Figure 5.3.1-86). As primary sump injection occurred at about [ ] backflow of hot water from the primary sump entered the IRWST (FMM 701; Figure 5.3.1-43).
oA15mneviusw.5.non:ium98 5.3.1-33 REVISION 1
FINAt. DATA REPORT 53.1.6 Mass Balance Mass balance results for Matrix Test SB10 were calculated based on water inventory before and after the test event and are provided in Appendix E. Inventory at the end of the test was within [ ]"# percent of the inventory at the beginning of the test. 53.1.7 Conclusions The test was performed with minimal problems and is considered acceptable. Although not all of the facility initial conditions met the specified acceptance criteria, the deviations did not impact the quality of the data. 'Ihe instrumentation problems encountered were not critical to the performance of the facility mass and energy balances. Facility response to the test was as anticipated for the conditions that were established. The data clearly demonstrate that cooling of the reactor heater rods was maintained throughout the duration of the test. O 1 i l l O oA1536%Revnl536w-5.non:ll@81798 5.3.1-34 REVISION 1 l
FINAL DATA REromT TABLE 5.3.1 1 MATRIX TEST SB10 INITIAL CONDITIONS Specined Actual Initial Instnament Initial Condition l Parameter No. Condition Comunents}}