ML20150B818
| ML20150B818 | |
| Person / Time | |
|---|---|
| Site: | 05000470 |
| Issue date: | 06/30/1988 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML20150B805 | List: |
| References | |
| NUDOCS 8807120199 | |
| Download: ML20150B818 (185) | |
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{{#Wiki_filter:. (Sh* * ' 5) CESSAR CERTIFICATl2N EFFECTIVE PAGE LISTING CHAPTER 5 Table of Contents Pace Amendment i 11 B 111 iv v vi B vil viii B ix B x xi xii c xiii Text Pace Amendment 5.1-1 5.1-2 B 5,1-3 B 5.1-4 5.1-5 B 5.1-6 B 5.1-7 5.1-8 5.1-9 B 5.1-10 B 5.1-11 B 5.1-12 B 5.1-13 B 5.1-14 B 5.1-15 5.1-16 5.1-17 B 5.1-18 5.1-19 5.1-20 B 5.1-21 B 5.1-22 8807120199 080630 PDR ADOCK 05000470 Amendment C K PNU June 30, 1988
ons u (Shsst 2 of 5) ' CESSAR CERTIFICATICN EFFECTIVE PAGE LISTING (Cont'd) CHAPTER 5 Text (Cont'd) Page Amendment 5.1-23 5.1-24 B 5.2-1 B 5.2-2 B 5.2-3 5.2-4 5.2-5 5.2-6 5.2-7 5.2-8 B 5.2-9 5.2-10 5.2-11 B - 5.2-12 B 5.2-13 B 5.2-14 5.2-15 B 5.2-16 5.2-17 5.2-18 5.2-19 5.2-20 B 5.2-21 B 5.2-22 B 5.2-23 5.2-24 B 5.2-25 B 5.3-1 B 5.3-2 5.3-3 B 5.3-4 B 5.3-5 B 5.3-6 5.3-7 5.3-8 B 5.3-9 B 5.3-10 5.3-11 5.3-12 B 5.3-13 5.3-14 B l 5.3-15 Amendmer.t C June 30, 1988
C E S S A R EnMnc m 2,. EFFECTIVE PAGE LISTING (Cont'd) CHAPTER.5 Text (Cont'd) Pace Amendment 5.3-16 B 5.3-17 B 5.4-1 5.4-2 5.4-3 5.4-4 5.4-5 5.4-6 5.4-7 5.4-8 B 5.4-9 B 5.4-10 B 5.4-11 5.4-12 5.4-13 5.4-14 5.4-15 5.4-16 5.4-17 b 5.4-18 C 5.4-19 C 5.4-20 C 5.4-21 C 5.4-22 C 5.4-23 C 5.4-24 C 5.4-25 C 5.4-26 C 5.4-27 C 5.4-28 C 5.4-29 C 5.4-30 C 5.4-31 C 5.4-32 C 5.4-33 C 5.4-34 C 5.4-35 C 5.4-36 C 5.4-37 C 5.4-38 C 5.4-39 C 5.4-40 C Amendment C June 30, 1988
(Shoot 4 of 5) , j CESSAR !!!Gemu l l EFFECTIVE PAGE LISTING (Cont'd) I CHAPTER 1 Text (Cont'd) Pace Amendment 5.4-41 C 5.4-42 C 5.4-43 B 5.4-44 B 5.4-45 B 5.4-46 B 5.4-47 5.4-48 B 5.4-49 5.4-50 5.4-61 B 5.4-52 B Tables Amendment 5.1.1-1 B 5.1.1-2 B 5.1.1-3 B 5.1.4-1 S.1.4-2 B 5.1.4-3 B 5.2-1 5.2-2 (Sheets 1-3) B 5.2-2 (Sheets 4-5) 5.3-1 5.3-2 5.3-3 5.3-4 5.3-5 B 5.3-6 5.3-7 5.4.1-1 5.4.2-1 B 5.4.7-1 (Sheets 1 and 2) C 5.4.7-2 C 5.4.7-3 C 5.4.10-1 B 5.4.10-2 5.4.13-1 B 5.4.13-2 ! Amendment C June 30, 1988
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CESSAR CEWTIFICATl3N . EFFECTIVE PAGE LISTING (Cont'd) CHAPTER 5 Ficures Antendment 5.1.2-1 5.1.2-2 5.1.3-1 B 5.1.3-2 B 5.1.4-1 5.2-1 3 5.2-2 5.3-1 5.3-2 5.3-3 5.3-4 5.3-5 5.3-6 B 5.4.1-1 B 5.4.2-1 B 5.4.7-1 C 5.4.7-2 C 5.4.7-3 C 5.4.7-4 C 5.4.10-1 B 5.4.10-2 B 5.4.10 3 B 5.4.10-4 B 5.4.10-5 B 5.4.13-1 5.4.13-2 5.4.14-1 B 5.4.14-2 5.4.14-3 B 5.4.14-4 Ametidment C June 30, 1988
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CESSAR !!nirlCATICN TABLE OF CONTENTS , CHAPTER 5 section Egbiect Ence No. 5.0 REACTOR COOLANT SYSTEM AND CONNECTED 5.1-1 SYSTEMS 5.1
SUMMARY
DESCRIPTION 5.1-1 5.1.1 SCHEMATIC FLOW DIAGRAM 5.1-2 5.1.2 PIPING AND INSTRUMENT DIAGRAM 5.1-3 5.1.3 ELEVATION DRAWINGS 5.1-4 5.1.4 NUCLEAR STEAM SUPPLY SYSTEM - 5.1-5 EALANCE OF PLANT INTERFACE REQUIREMENTS 5.2 IEEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY 5,2-1 5.2.1 COMPLIANCE WITH CODES AND CODE CASES 5.2-1 5.2.1,1 Comoliance with 10CFR50.55a 5.2-1 5.2.1.2 Aeolicable Code Cases 5.2-1 5.2.2 OVERPRESSURE PROTECTION 5.2-1 5.2.2.1 Qgg.ign Bases 5.2-1 1 1 5.2.2.2 Desian Evaluation 5.2-2 5.2.2.3 Eloino and Instrumentation Diaarams 5. 2- 2 S.2.2.4 Eculument & Comoonent Description 3.2-2 5.2.2.4.1 Transients 5.2-3 5.2.2.4.2 Environment 5.2-3 5.2.2.4.2.1 Normal Environment 5.2-3 5.2.2.4.2.2 Main Steam Line Break 5.2-3 (One Occurrence) 5.2.2.4.3 Main Steam Safety Valves 5.2-4 5.2.2.4.3.1 Main Steam Safety Valve 5.2-4 Operation i
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CESSAR n!Gcun. l i l l TABLE OF CONTENT 8 (Cont'd) ; l CEAPTER 5 Section Subiect Pace No. 5.2.2.4.3.2 Transients 5.2-4 5.2.2.4.3.3 Environment 5.2-4 5.2.2.4.3.3.1 Normal Enviror. ment 5.2-4 5.2.2.4.3.3.2 Main Steam Line Break 5.2-5 (One Occurrence) 5.2.2.4.4 Safety Injection System Relief 5.2-5 Valves SI-164 and SI-469 5.2.2.4.4.1 Valve Operation 5.2-5 5.2.2.4.4.2 Transients 5.2-5 5.2.2.4.4.3 Environment 5.2-6 5.2.2.4.4.4 Material Specifications 5.2-6 5.2.2.5 Mountina of Pressure-Relief Deviggg 5.2-6 5.2.2.6 Aeolicable Codes and Classification 5.2-6 5.2.2.7 Process Instrurentation 5.2-6 5.2.2.8 System Reliability 5.2-7 5.2.2.9 Testina and Insocction 5.2-7 5.2.2.10 Overoressure Protection Durina Low 5.2-7 Temeeratur_e Conditions 5.2.2.10.1 Design Criteria 5.2-8 5.2.2.10.1.1 Credit for Operator Action 5.2-8 5.2.2.10.1.2 Single Failure 5.2-8 5.2.2.10.1.3 Testability 5.2-8 B 5.2.2.10.1.4 Seismic Design and IEEE 279 5.2-8 Criteria 5.2.2.10.2 Design and Analysis 5.2-8 5.2.2.10.2.1 Limiting Transients 5.2-9 5.2.2.10.2.2 Provision for Overpressure 5.2-10 Protection 5.2.2.10.2.3 Equipment Parameters 5.2-11 5.2.2.10.2.4 Administrative Controls 5.2-12
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Amendment B 11 March 31, 1988
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ke! h h khk kk riflCATION 1 TABLE OF CONTFJfH (Cont'd) C5 APTER 5 Section Subiect Pace No. 5.2.3 REACTOR COOLANT PRESSURE BOUNDARY MATERIALS 5.2-12 5.2.3.1 Material Soecification 5.2-12 5.2.3.2 Comcatibility with Reactor Coolant 5.2-13 5.2.3.2.1 Reactor Coolant Chemistry 5.2-13 5.2.3.2.2 Materials of Construction 5.2-13 Compatibility to Reactor Coolant 5.2.3.2.3 Compatibility with External 5.2-13 Insulation and Invironmental Atmosphere 5.2.3.3 Fabrication and Processinc Ferritic 5.2-14 Materials 5.2.3.3.1 Fracture Toughness 5.2-14 5.2.3.3.1.1 NSSS Components 5.2-14 5.2.3.3.2 Control of Welding 5.2-15 5.2.3.3.2.1 Avoidance of Cold Cracking 5.2-15 5.2.3.3.2.2 Regulatory Guide 1.34 5.2-15 5.2.3.3.2.3 Regulatory Guide 1.71 5.2-16 5.2.3.3.3 Non-Destructive Examination of 5.2-16 Tubular Products 5.2.3.4 Fabrication and Processina of 5.2-16 Austenitic Stainless Steel 5.2.3.4.1 Avoidance of Stress Corrosion 5.2-16 Cracking 5.2.3.4.1.1 Avoidance of Sensitization 5.2-16 5.2.3.4.1.1.1 NSSS Components 5.2-16 5.2.3.4.1.2 Avoidance of Contamination 5.2-19 Causing Stress Corrosion Cracking lii
CESSAR;!nL mn TABLE OF CONThNTS (Cont'd) CHAPTER 5 Section Subiect Pace No. 5.2.3.4.1.2.1 NSSS Components 5.2-19 5.2.3.4.1.3 Characteristics and 5.2-20 Mechanical Properties of Cold-Worked Austenitic Stainless Steels for RCPB Components 5.2.3.4.2 Control of Welding 5.2-20 5.2.3.4.2.1 Avoidance of Hot Cracking 5.2-20 5.2.4 INSERVICE INSPECTION AND TESTING OF REACTOR 5.2-21 COOLANT PRESSURE BOUNDARY 5.2.4.1 Accessibility of InsDection Areas 5.2-21 5.2.5 REACTOR COOLANT PRESSURE BOUNDARY (RCPB) 5.2-22 LEAKAGE DETECTION SYSTEMS 5.2.5.1 Leakace Detection Methqd.2 5.2-22 5.2.5.1.1 Unidentified Leakaoe 5.2-22 5.2.5.1.2 Identified Leakage 5.2-23 5.2.5.1.2.1 Safety Valves Located on the 5.2-23 Reactor Coolant System 5.2.5.1.2.2 Reactor Coolant Pump Seals 5.2-23 5.2.5.1.3 Leakage Through Steam Generator 5.2-24 Tubes or Tubesheet 5.2.5.1.4 Leakage to Auxiliary Systems 5.2-24 5.2.5.2 Control Room Leakace Instrumentation 5.2-24 5.2.5.3 Limits for Reactor Coolant Leakace 5.2-24 5.2.5.4 Maximum Allowable Total Leakace 5.2-24 5.2.5.5 Q1fferentiation Between Identified and 5.2-24 Unidentified Leaka 5.2.5.6 Sensitivity and Ooerability Tests 5.2-25 iv
d CESSAR BEncm3. TABLE OF CONTENTS (Cont'd) CHAPTER 5 Emotion Rubiasi Pace No. 5.3 FEACTOR VESSEL 5.3-1 5.3.1 REACTOR VESSEL MATERIALS 5.3-1 5.3.1.1 Material Soecifications 5.3-1 5.3.1.2 Soecial Process Used for Manufacturinc 5.3-1 and Fabrication 5.3.1.3 Soecial Methods for Nondestructive 5.3-2 Examination 5.3.1.4 Special Controls for Ferritic and 5.3-3 Austenitic Stainless Steels 5.3.1.5 Fracture Touchness 5.3-3 5.3.1.6 Renator vessel Material Surveillanqe 5.3-4 Procram 5.3.1.6.1 Test Material Selection 5.3-4 5.3.1.6.2 Test Specimens 5.3-5 5.3.1.6.2.1 Type and Quantity 5.3-5 5.3.1.6.2.2 Baseline Specimens 5.3-6 5.3.1.6.2.3 Irradiated Specimens 5.3-6 5.3.1.6.3 Surveillance Capsules 5.3-7 l 5.3.1.6.3.1 Charpy Impact Compartment 5.3-7 Assembly 5.3.1.6.3.2 Temperature, Flux and 5.3-8 Tension Compartment Assembly 5.3.1.6.4 Neutron Irradiation and 5.3-8 Temperatures Exposure 5.3.1.6.4.1 Flux Measurements 5.3-8 i 5.3.1.6.4.2 Temperature Estimates 5.3-9 l l l 5.3.1.6.5 Irradiatien Locations 5.3-9 l 5,3.1.6.6 Withdrawal Schedule 5.3-10 l v l
, CESSARannne-TABLE OF CONTENTD(Cont'd) CHAPTER 5 Section Subiect Pace No. 5.3.1.6.7 Irradiation Effects Prediction 5.3-10 Basis 5.3.1.7 Reactor Vessel Fasteners 5.3-11 5.3.2 PRESSURE-TEMPERATURE LIMITS 5.3-11 5.3.2.1 Limit curves 5.3-12 5.3.2.2 ODeratina Procedures 5.3-15 5.3.3 REACTOR VESSEL INTEGRITY 5.3-15 5.3.3.1 Desian 5.3-15 5.3.3.2 Materialu of Construction 5.3-15 5.3.3.3 Fabrication Methods 5.3-16 5.3.3.4 Inscection Reauirementa 5.3-16 5.3.3.5 Shioment and Installation 5.1-16 5.3.3.6* Ooeratina Conditions 5.3-16 5.3.3.7 Inservice Surveillance 5.3-16 5.4 CQEPONENT AND SUBSYSTEM DESIGN 5.4-1 5.4.1 REACTOR COOLANT PUMPS 5.4-1 5.4.1.1 Pumo Flywheel Intearity 5.4-1 5.4.1.2 Descriotion 5.4-3 5.4.1.3 Evaluation 5.4-4 5.4.1.4 Tests and Inscections 5.4-7
*See site-specific SAR B Amendment B vi March 31, 1988
CESSAR !!!Mnc-TABLE OF CONTENTS (Cont'd) CEAPTER 5 Section EMbiect Pace No. 5.4.2 STEAM GENERATORS 5.4-9 5.4.2.1 Desian Bases 5.4-9 5.4.2.2 Descriotion 5.4-11 5.4.2.3 Economizer Intecrity 5.4-12 5.4.2.4 Steam Generator Materials 5.4-12 5.4.2.4.1 Steam Generator Tubes 5.4-13 5.4.2.5 Tests and Insoections 5.4-13 5.4.3 REACTOR COOLANT PIPING 5.4-14 5.4.3.1 Qggign Basis 5.4-14 5.4.3.2 RAEnriotion 5.4-14 5.4.3.3 Materials 5.4-15 5.4.3.4 Tests and Insoections 5.4-15 5.4.4 MAIN STEAM LINE RESTRICTIONS 5.4-15 5.4.5 MAIN STEAM LINE ISOLATION SYSTEM 5.4-15 5.4.5.1 Desian Bases 5.4-15 5.4.5.2 System Desian 5.4-16 5.4.5.2.1 General Description 5.4-16 , 5.4.5.2.2 Component Description 5.4-16 5.4.5.2.3 System Operation 5.4-16 ( 5.4.5.3 Desian Evaluation 5.4-17 1 5.4.5.4 Tests and Inspections 5.4-17 5.4.6 REACTOR CORE ISOLATION COOLING SYSTEM 5.4-17 5.4.7 SHUTDOWN COOLING SYSTEM 5.4-18 l vil
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CESSARsHMnc-i TABLE OF_ CONTENTS (Cont'd) i CEAPTER 5 l Section Subiect Pace No. 5.4.7.1 Desian Bases 5.4-18 5.4.7.1.1 Summary Description 5.4-18 5.4.7.1.2 Functional Design Bases 5.4-18 5.4.7.1.3 Interface Requirements 5.4-19 5.4.7.2 System Desian 5.4-28 5.4.7.2.1 System Schematic 5.4-28 5.4.7.2.2 Component Description 5.4-29 5.4.7.2.3 Overpressure Prevention 5.4-31 5.4.7.2.4 Applicable Coden and 5.4-32 Classifications 5.4.7.2.5 System Reliability Considerations 5.4-33 5.4.7.2.6 Manual Actions 5.4-34 5.4.7.3 Performance Evalustign 5.4-37 5.4.7.4 Prennerational Testing 5.4-38 5.4.8 REACTOR COOLANT CLEANUP SYSTEM 5.4-43 5.4.9* MAIN STEAM LINE AND FEEDWATER PIPING 5.4-43 5.4.10 PRESdURIZER 5.4-43 5.4.10.1 Desian Bases 5.4-43 5.4.10.2 Descriction 5.4-44 5.4.10.3 Evaluation 5.4-47 5.4.10.4 Tests and Insoections 5.4-47 5.4.11 PRESSURIZER RELIEF TANK 5.4-48 5.4.12 VALVES 5.4-48 5.4.12.1 Desic5r. Basis 5.4-48 5.4.12.2 Desian Descriotion 5.4-48
- See Chapter 10 Amendment B vili March 31, 1988
CESSAR naineman TABLE OF CONTENTS (Cont'd) CHAPTER 5 Section Subiect Pace No. 5.4.12.3 Desian Evaluation 5.4-48 5.4.12.4 Tests and Insoections 5.4-49 5.4.13 SAFETY AND RELIEF VALVES 5.4-49 5.4.13.1 resian Basis 5.4-49 5.4.13.2 Descriotion 5.4-49 5.4.13.3 Evaluation 5.4-50 5.4.13.4 Tests and Insoections 5.4-50 5.4.13.4.1 Pressurizer Safety Valves 5.4-50 5.4.13.4.2 Main Steam Safety Valves 5.4-50 5.4.14 COMPONENT SUPPORTS 5.4-51 5.4.14.1 Desian Basis 5.4-51 5.4.14.2 Descriotion 5.4-51 5.4.14.3 Evaluation 5.4-52 APPENDIX 5A OVERPRESSURE PROTECTION FOR COMBUSTION 5A-1 ENGINEERING SYSTEM 80 APPENDIX 5B STRUCTURAL EVALUATION OF STEAM LINE BREAK 5B-1 FOR STEAM GENERATOR INTERNALS APPEh0IX SC STRUCTURAL EVALUATION OF FEEDWATER LINE SC-1 BREAK FOR STEAM GENERATOR INTERNALS APPENDIX SD PRESSURIZED THERMAL SHOCK (PIS) EVALUATION (LATER) B Amendment B ix March 31, 1988 ,
CESGAR n'4inema l I LIST OF TABLES CHAPTER 5 Table Subiect 5.1.1-1 Process Data Point Tabulation l 5.1.1-2 Design Parameters of Reactor Ccolant System 5.1.1-3 Reactor Coolant System Volumes 5.1.4-1 RCP Cooling Water System Data 5.1.4-2 Heat Loads from NSSS Support Structure ) 5.1.4-3 RCS Insulation Heat Loads 5.2-1 Reactor Coolant System Pressure Boundary Code Requirements 1 5.2-2 Reactor Coolant System Materials 5.2-3 Code Case Interpretations 5.3-1 Total Quantity of Specimens 5.3-2 Type and Quantity of Specimens for Baseline Tests 5.3-3 Type and Quantity of Specimens for Irradiation Exposure and Irradiated Tests 5.3-4 Type and Quantity of Specimens Contained In Each Irradiation Capsule Assembly 5.3-5 Candidate Materials for Neutron Threshold Detectors 5.3-6 Composition and Melting Points of Candidate Materials for Temperature Monitors 5.3-7 Capsule Assembly Removal Schedule 5.4.1-1 Reactor Coolant Pump Parameters l l 5.4.2-1 Steam Generator Parameters l l X l
l CESSAR !!ninema
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1 l LIST OF TABLES (Cont'd) l CHAPTER 5 Table subiect 1 l 5.4.7-1 Shutdown Cooling Design Parameters l l 5.4.7-2 Shutdown Cooling System Interface Requirements i for Component Cooling Water j 5.4.7-3 Shutdown Cooling System FMEA 5.4.10-1 Pressurizer Parameters l 5.4.10-2 Pressurizer Tests 5.4.13-1 Pressurizer Safety Valve Parameters 5.4.13-2 Main Steam Safety Valve Parameters P G xi
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V CESSAR R*Encua .
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f 1> LIST OF FIGURES , CHAPTER E Ficure Subiect < 5.1.2-1 Reactor Coolant System Piping and Instrumentation Diagram 5.1.2-2 Reactor Coolant Pump Piping and Instrumentation Diagram 5.1.3-1 Reactor Coolant System Arrangement (Plan) 5.1.3-2 Reactor Coolant System Arrangement (Elevation) 5.1.4-1 Atmospheric Dump Valve Flow Requirements 5.2-1 Pressurizer Pressure During Inadvertcnt Safety) Injection Actuation 5.2-2 Pressurizer Pressure During RCP Start with RCS LT ( , 5.3-1 Typical Surveillance Capsule Assembly y 5.3-2 Charpy Impact Compartment Assec51y ' It ( 5.3-3 Temperature, Flux and Tension Compartment Assembly, , I 5.3-4 Locations of Surveillance Capsule Assemblies 5.3-5 C-E Design Curve of Transition Temperature Increase 5.3-6 Reactor Vessel 5.4.1-1 Reactor Coolant Pump 5.4.2-1 Steam Generator 5.4.7-1 Shutdown Cooling System, Two Train Cooldown 5.4.7-2 Shutdown Cooling System, one Train Cooldown 4 5.4.7-3 Shutdown Cooling System Flow Diagra.n, Shutdown Cooling Mode C 5.4.7-4 Shutdown Cooling System Flow Diagtam, Shutdown Cooling Mode Amendms;st C xii June 3i'., 1988
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Picure Subiect 1 i 5.4/10-1
- Typical Pressurizer s
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5.4.'.'id-2 Typical Pressurizer Level Setpoint Program
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l 5. 4 .10-3 Typical'Tempetature Control Program
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$,.4.10-4 Typical Press'urizer Level Error Proipam 5.4.'10 551 Pressure Control Program 5.4.13-1 Primary Safaty Valve 5.4.13-2 ;
Maln Steam Safety Valve. t ,
' IReactor Coolant System Arrangement and S';pport
, $ {.4.14-A f' Polrts y
ij.4.14-2 Reactor Vessel Supports }\j 5.4.14-3 Steam Generator Supports J , 5.4.14-4 , Reactor Coolant Pump Supports b l-
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1 I The main steam line isolation system components are qualified te serve in the environment specified in Section 3.13.-- 1 5.4.5.3 *esign_3valuatig.g , Design evaluations are listed to correspond with the design ' cases listing. , / , I A. The steamMain Stet.m Isolation generators within 5.0Valves areaster seconda capable of isolat.inej'the receiving a' signal l 8 from the Engineered Safety Teatures Actuation System. In the evcnt of a steam line break, this action prevents continuous scontrolled steam release from mere than one steam generator. Protection is offered for breaks inside or outside the containant. B. The Ma.in Steam Isolation Valves, their operators, and associated circuitry are Seismic category I, and are protecte<1 against missiles and the effect of high-ecyrgy line breaks. 5.4.5.4 Tests and Inseeotions All main staam isolation valvcs are designed, f at ricated, tested, and installed in accordar c.a with the to.ies and . standards identified in the interf ace ~ requiromenta ' described 1.n Section f, 5.1.4. Assurance of operability is discussed in sectior, 3.9.3 of , the Applicant's SAR. 5.4.6 REACTOR CORE ISOLATION COCLING SYSTEM This system is not applicable to a Preseuri:ed Eater Reactor. , 1
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Amondment B 5.4-17 March 31, 1988
CESSAR n3Mication 5.4.7 FEUTD0WN COOLING SYST34 lC 5.6.7.1 -posion Bases 5.4.7.1.1 Summary Description The Shutdown Cooling System (SCS) is used in conjunction with the Main Steam and Main or Emergency Feedwater Systems (see Sections 10.1 and 10.4.9) to reduce the temperature of the Reactor Coolant lC System (RCS) in post shutdown periods from normal operating temperature to the refueling temperature. The initial phase of a cooldown is accomplished by heat rejection from the Steam Generators (SG) to the condenser or atmosphere. After the reactor coolant temperature and pressure have been reduced to approximately 350*F and 400 psia, the SCS is put into operation to reduce the RCS tanperature to the refueling temperature and to lC maintain this temperahurc during refueling. lC The SCS is used in addf tion to. the SG atmospheric steam release capability and thn Emergency F(edwater System to cooldown the F.CS following a small break LOCA (see Section 6.3). The SCS would also be used subsequent to steam and feedline breaks, steam generator tube rupturos. and is used prior to RCP start to maintain flow through the core during plant startup. 5.4.7.1.2 Functi43ul Design Bases The following functional design bases apply to the Shutdown Cooling System: A. No single active failure . prevents at least one complete train of the SCS from being brought on line from the control room, whether this is during normal plant cooldown or following a Design Basis Event. B. The design basec 'iafined in Paragraph 5.4.7.1.1 are met assuming t"a failutte of a single active component during i' shutdown cooling c a single active or limited leakage passive failure of a component during long-term operations (i.e., >24 hours) following a Design Basis Event. Limited leakage passive failure is defined based on maximum flow through a failed valve packing or pump (e.g., SCS pump mechanical seal). i C. The SCS is designed such that the SCS pumps and containment C l spray pumps are functionally interchangeable. D. The SCS shall be designed for a nominal pressure of 900 psig and a temperature of 400'F. 1 l l Amendment C ; 5.4-18 June 30, 1988 l l l
CESSARn%nce E. No single failure allows the SCS to be overpressurized by the RCS. SCS components whose design pressure is less than the hCS design pressure are provided with overpressure protection (see Section 5.4.7.2.3). C F. The SCS reduces the Reactor Coolant System temperature as follows:
- 1. Two Train Cooldown
- a. to 140*F - within 24 hours after reactor shutdown. C
- b. to 130*F - by the time reactor vessel head stud detensioning operations are started (i.e., within approximately 40 hours).
- c. to 120*F - within 96 hours after reactor shutdown.
- 2. One Train Cooldown
- a. to 200*F - within 24 hours after reactor shutdowr, in conjunction with other heat removal systems (e.g., steam generator atmospheric dump valves).
Typical cooldown curves are shown in Figures 5.4.7-1 c.nd 5.4.7-2. G. The components of the shutdown cooling system are designed in accordance with Section 5.4.7.2.4. H. Materials are selected to preclude system performance degradation due to the effects of short and long term corrosion. I. The SCS heat exchangers are sized to remove decay heat 96 hours after shutdown based upon a refueling water temperature of 120*F and a service water temperature of 95'F with an average reactor core burnup of two years. C' J. The SCS is designed so that the SCS pumps can be tested at full-fl;w conditions with the reactor operating at power. 5.4.7.1.3 Interface Requirements Interface requirements that the SCS places on certain aspects of the Balance of Plant are listed by categories below. In addition, General Design Criteria (GDC) and Regulatory Guides C related to the interface requirements are presented. These GDC and Regulatory Guides are listed only to identify regulatory Amendment C 5.4-19 June 30, 1988
CESSAR 88#icari:n criteria considered to be relevant, and are not imposed as interface requirements unless specifically called out as such in a particular interface requirement. Relevant GDC - 1, 2, 3, 4, 10, 34, 35, 36, 37, 38, 39, 40, 50, 54, 56, 57 Relevant Reg. Guides - 1.1, 1.4, 1.26, 1.28, 1.29, 1.31, 1.34, 1.36, 1.44, 1.46, 1.47, 1.48, 1.50, 1.51, lC 1.61, 1.64, 1.68, 1.73, 1.74, 1.75, 1.79, 1.84, 1.85, 1.89, 1.97, 1.148, 8.8. lC A. Power
- 1. Electrical power requirements for the motor-operated valves in the SCS are contained in Table 8.3.1-1.
- 2. The electrical supplies for SCS pumps, valves and instruments shall be as follows:
- a. The SCS pumps and valves shall be capable of being powered from the plant's normal and emergency power sources. Power connections shall be through independent power trains so that in the event of a LOCA, in conjunction with the loss of normal power and a single failure in the emergency electrical supply, the capability of initiating shutdown cooling with a minimum of one subsystem exists.
- b. An independent electrical bus shall supply one SCS l C pump and the valves in the associated heat exchanger train.
- c. The SCS suction line isolation valves (SD-673, C
671, 659, 672, 670, 658 on Figure 6.3.2-1B) shall receive electrical power such that no fault to a single power supply could open the valves to connect the RCS and SCS inadvertently, nor could a fault to a single power supply prevent opening all the valves of at least one suction line during initiation of shutdown cooling.
- d. Two independent instrument power supplies shall be provided for the SCS Instrumentation.
B. Protection From the Effects of Natural Phenomena
- 1. The location, arrangement, and installation of the SCS components shall be such that floods (and tsunami Amendment C 5.4-20 June 30, 1988
CESSAR GEnce and seiches for applicable sites) or the effects thereof per the requirements of Criterion 2 of 10CFR50 will not prevent them from performing their safety functions.
- 2. The location, arrangement, and installation of the SCS components shall be such that winds and tornadoes or the effects thereof per the requirements of Criterion 2 of 10CFR50 will not prevent them from performing their safety functions.
- 3. The location, arrangement, and installation of the SCS components shall be auch that they will withstand the effects of earthquakes per the requirements of Criterion 2 of 10CFR50 without loss of the capability te perform their safety functions.
- 4. Failure of non-seismic systems and structures shall not cause loss of either SCS train. All SCS instruments and associated instrument lines, root valves, and C isolation valves, shall be designed to maintain pressure boundary integrity following a seismic event.
C. Protection From Pipe Failure
- 1. Pipe Break Considerations The SCS, both inside and outside containment, shall ce protected from the sffects of postulated high and moderate energy pipe rupturea.
- 2. Pipe Leakage Considerations No limited leakage passive failure or the effects thereof (such as flooding, spray impingement, steam, temperature, pressure, radiation, or loss of NPS!!) in a connecting system (e.g., Safety Injection System or Containment Spray System) shall preclude the availability of minimum acceptable shutdown cooling capability. Minimum acceptable shutdown capability is defined as that provided by one SCS cooling pump l C and its associated heat exchanger train.
- 3. Design Requirements For all parts of the SCS, appropriate design procedures shall be employed to ensure that a postulated pipe failure does not result in a loss of function of the SCS as follows:
l Amendment C 5.4-21 June 30, 1988 l l l l
CESSAR !!Mem
- a. Protection of the SCS from the consequences of a postulated pipe failure shall be provdided by:
- 1. separation via physical plant layout,
- 2. pipe restraints,
- 3. protective structures,
- 4. watertight rooms,
- 5. isolation capability, or
- 6. other suitable means.
- b. Isolation valves (system and/or containment) used to contain leakage shall be protected from the adverse effects of a pipe failure which might preclude their operation when required.
D. Missiles
- 1. For the portion of the SCS located inside containment, appropriate design procedures shall be used to insure l C that the impact of any potential missile will not lead to a Loss-Of-Coolant Accident or preclude the system from carrying out its specified safety functions.
- 2. For the portion of the SCS located outside containment, appropriate design procedures (e.g., proper turbine orientation, physical separation, or missile barriers) shall be used to insure that the impact of any potential missile does not prevent the system or equipment from carrying out its specified safety functions.
- 3. Appropriate design procedures shall be used to insure that the impact of any potential missile does not prevent the conduct of a safe plant shutdown, or prevent the plant from remaining in a safe shutdown condition.
E. Separation
- 1. Concrete compartments within containment shall serve as protection for that portion of the SCS which is inside the containment and thus could be subjected to crediblo dynamic effects originating within the containment under the conditions of accidents the SCS is required j to mitigate. Separation via physical plant layout, l
pipe restraints, isolation capability, or other l suitable means shall be provided as necessary to guard against damage to the components of the SCS inside the containment from these dynamic effects. Amendment C 5.4-22 June 30, 1988 l
CESSAR!abbmu l 1
- 2. Containment isolation valves, operators, and associated power and control systems located outside the containment that are part of the SCS shall be protected from dynamic effects and loss of Danction resulting from equipment failurcs and pipe ruptures originating in adjacent areas. Protection from such failure and rupture effects shall be by separation, enclosure, restraint, water-tight rooms or other suitable means.
- 3. Adequate physical separation shall be maintained between the redundant piping paths and containment penetrations of the SCS such that the SCS will meet its functional requirements even with a single active failure or a limited leakage passive failure.
F. Independence
- 1. Electrical - See 5.4.7.1.3 (A.2.b.)
- 2. Environmental - See 5.4.7.1.3 (Q)
- 3. Mechanical - See 5.4.7.1.3 (C, D, E)
G. Thermal Limitations
- 1. Component Cooling Water - See 5.4.7.1.3 (P.2) l
- 2. Environmental - See 5.4.7.1.3 (Q)
H. Monitoring
- 1. The safety related instrumentation of the SCS is identified in Tablo 7.5-2.
I. Operational / Controls
- 1. The SCS components shall be powered such that the operational and control requirements of 5.4.7.1.3.A are met.
- 2. The SCS controls shall be designed to meet the design C bases of 5.4.7.1.2.
J. Inspection and Testing
- 1. All SCS ASME, Section III components shall be arranged to provide adequate clearances to permit inservice inspection.
! 2. Manually operated valves which contain reactor coolant or other potentially radioactive liquids during normal plant operations shall be provided with handwheel extensions and shielding, to allow periodic actuation. I Amendment C 5.4-23 June 30, 1988 i
CESSARUnince
- 3. SCS components which contain reactor coolant or other potentially radioactive liquids during normal plant operations, and which require access for periodic pressure tests and nondestructive examination, shall be capable of being flushed prior to testing. C
- 4. System and component arrangement shall allow adequate clearances for performance of inspections identified in Chapter 16.
K. Chemistry / Sampling
- 1. The component cooling water shall contain corrosion inhibitors. The water shall not contain scale-forming compounds. The pH shall be controlled between 8.3 and 10.5. Chloride concentration shall be less than 1.0 ppm.
- 2. The Sampling System shall provide a means of obtaining remote liquid samples from the Shutdown Cooling System for chemical and radiochemical laboratory analysis.
- 3. The sample lines in contact with reactor coolant shall be austenitic stainless steel that is compatible with the fluid chemistry.
- 4. The sample lines shall be sized such that the fluid velocity allows a representative sample and the purge flow rate is high enough to remove crud from the sample lines.
S. Post accident sampling capabilities shall be provided C for the SCS. L. Materials
- 1. Piping and all metallic parts in contact with the system fluid, with the exception of some component internals as required, shall be of austenitic stainless steel.
Selection shall be on the basis of compatibility with design pressure and temperature stress considerations and with the chemistry of the system fluid. Valve packing, gaskets, and diaphragm materials for packless valves shall be compatible with the radiation dose as well as the chemistry of the system fluid. Amendment C 5.4-24 June 30, 1988 l
CESSAR Mce
- 2. Fabrication and erection of system materials shall be consistent with the quality standards of General Design Criterion 1, Appendix A and Appendix B of 10CFR50.
- 3. Care shall be taken to prevent sensitization and to control the delta ferrite content of (1) the welds which join any system fabricated of austenitic stainless steal to the SCS, and (2) the field welds of the SCS.
- 4. Controls shall be exercised during plant construction to astture that contaminants do not significantly contribute to stress corrosion of stainless steel.
M. System / Component Arrangement
- 1. The first isolation valve on the SCS suction lines shall be located as close to the RCS an practicable.
The volume of the SCS suction piping between the RCS and the first isolation valve shall be as small as possible. This requirement minimizes the amount of piping exposed to normal RCS pressure and minimizes the effect of boron dilution during shutdown cooling C initiation.
- 2. The SCS pumps shall be located as close as practicable to the containment.
- a. The elevation of these pumps shall be low enough such that adequate NPSH is available during shutdown cooling when the pumps take suction from the RCS. The required NPSH during cooling is (LATER) feet. shutdown lC
- b. The elevation of these pumpe when used for containment spray functiers, shall be low enough C such that adequate NPSH is available when all pumps take a suction from the IRWST. '
' ~
- 3. The SCS piping and components shall be arranged such that straight piping runs upstream and downstream of l flow measurement device orifices are provided of sufficient length comply with: ASME Fluid Meters; Their Theory and A : ication, Parts 1 & 2.
i
- 4. The SCS suction lines shall be arranged such that no portion is physically above the lowest point of the RCS hot leg piping.
l Amendment C l 5.4-25 June 30, 1988
CESSAR95 Mace
- 5. If the shutdown cooling suction line overpressure relief valves are located at a higher elevation than the SCS pump suction centerline, their set pressure shall be reduced to adequately compensate. The C elevation of SCS relief valves (SD-769, SD-768) shall be located below RCS hot leg centerline.
- 6. Physical identification for safety related SCS equipment shall be provided to allow recognition of safety status by plant' personnel.
lC
- 7. In the event of a limited leakage passive failure in one SCS train during long term cooling, personnel access to the intact train shall not be affected.
- 8. Protection shall be provided from internally generated flooding that could prevent performance of safety related functions.
N. Radioactive Waste
- 1. The In-containment Refueling Water Storage Tank (IRWST) shall be designed to accept relief valve discharge from the shutdown cooling suction line overpressure relief valves. C O. Overpressure Protection
- 1. Thermal relief valves shall be provided in isolated sections of piping in the system to prevent overpressurization due to thermal transients.
P. Related Service
- 1. A fire protection system shall be provided to protect the SCS and shall include, as a minimum, the following features:
- a. Facilities for fire detection and alarming; l b. Facilities or methods to minimize the probability of fire and its associated effects; l
! c. Facilities for fire extinguishment;
- d. Methods of fire prevention such as use of fire resistant and non-combustible materials whenever l
practical, and minimizing exposure of combustible ' materials to fire hazards; 1 l l Amendment C 5.4-26 June 30, 1988 l l
l l CESSAR ;!nincua l l 1 I
- e. Assurance that fire protection systems do not cdversely affect the functional and structural integrity of safety related structures, systems, and components;
- f. Assurance that fire protection systems are designed to ensure that their rupture or inadvertent operation does not significantly impair the capability of safsty related structures, systems, and components; and,
)
- g. The fire protection system piping design and arrangement shall be such as to ensure that the functional and structural integrity of the SCS is adequately protected against the effects of pipe whip, jet impingement, and environmental effects resulting from postulated piping ruptures in the fire protection system.
- 2. Cooling Water System Requirements
- a. The cooling water system design shall be such that cooling water is available to supply the SCS heat exchangers when an irradiated core is present in the reactor vessel er the spent fuel pool. C
- b. For all conditions, cooling water shall be supplied as follows:
Required Value Parameter Per Heat Exchancer i Normal Allowable Delivery Pressure 100 psig l l Maximum Allowable Delivery Pressure 150 psig Required Flow rate 11,000 gpm l Maximum Allowable Flow rate 13,000 gpm
- c. Cooling water piping supplying the shutdown l cooling heat exchangers shall be designed and I fabricated in accordance with ASME B&PVC, Section III, Class 3, as a minimum, and shall be designed as Seismic Category I, Safety Class 3, as a minimum.
l l l Amendment C 5.4-27 June 30, 1988 l t
CESSAR Emince i 1 i
)
I
- d. The cooling water system which services the SCS shall be designed with sufficient redundancy and diversity such that one SCS heat exchanger train will always be supplied cooling water.
- e. The cooling water system which services the SCS shall be designed consistent with the cooling water chemistry.
- 3. Containment Spray System (CSS)
The CSS pumps are designed to be identical to the SCS pumps. These pumps shall be functionally C interchangeable with the SCS pumps for either SCS or CSS service to facilitate maintenance and/or testing activities during normal plant operations. Q. Environmental
- 1. The proper operating environmental conditions for the equipment of one train of the SCS shall be maintained independently of the environment of the other train of the SCS, e.g., failure or isolation of the ventilation capability to one train of the SCS shall not cause the environmental limits of the other SCS train to be exceeded.
- 2. The ventilation system shall control ambient air lC conditions in the proximity of all C-E supplied motor driven or diaphragm operated equipment in the SCS in accordance with the requirements of Section 3.11.
5.4.7.2 System Desien 5.4.7.2.1 System Sobesatic The SCS is shown on the RCS P&ID (Figure 5.1. 2-1) and on the SIS C and SCS P& ids (Figures 6.3.2-1A, 6.3.2-1B, 5.4.7-3, and 5.4.7-4). The pressure and temperature of the RCS system vary from 400 psia and 350*F at initiation of shutdown cooling to pressure and 120*F at refueling conditions. atmospheric SCS design lC parameters are given in Table 5.4.7-1. The SCS suction side pressure and temperature follow RCS conditions. The discharge side pressure is higher by an amount eqaal to the pump head and the temperature is lower at the shutdown cooling heat exchanger outlet. The SCS contains two heat exchangers and two pumps. One SCS pump is capable of meeting safety-grade cooldown criteria. Two SCS C pumps are required to meet normal cooldown design criteria. Amendment C 5.4-2E June 30, 1988
l l CESSARanecum l 1 During initial shutdown cooling, a portion of the reactor coolant flows out the SCS nozzles located on the reactor vessel outlet (hot leg) pipes and is circulated through the SCS heat exchangers by the SCS pumps. The return to the RCS is through SIS direct vessel injection (DVI) nozzles. C Shutdovn cooling flow is measured by orifice meters installed in each train of the SCS discharge piping. The information provided by these flow elements is used by the operator for flow control during SCS operation. The cooldown rate is controlled by adjusting flow through the heat exchangers with throttle valves on the discharge of each heat exchanger. The operator maintains a constant total SCS flow to the core by adjusting the hee exchanger bypass flow to compensate for changes in flow through the heat exchangers. 5.4.7.2.2 Component Description A. Shutdown Cooling Heat Exchangers The SCS heat exchangers are used to remove decay, sensible and SCS pump heat during cooldown, and decay and pump heat $ during cold shutdown. The units are sized to maintain a refueling water temperature of 120*F with the service water C tempersture 95'F at 96 hours after shutdown following an assumed reactor core average burnup of two years. A conservativs fouling resistance is assumed, resulting in an additional area margin for the heat exchangers. SCS heat exchanger parameters are given in Table 5.4.7-1. C l The design temperaturo is based upon the temperature of the I reactor coolant at the initiation of shutdown cooling plus a design tolerance. B. Instrumentation The operation of the SCS is controlled and monitored through the use of installed instruOentation. The instrumentation l provides the capability to monitor cooldown rate and I shutdown cooling flow to detset degradation of flow or SCS C heat removal capabilities. The instrumentation provided for the GCS is summarized in (LATER). ! C. Piping l l All SCS piping is austenitic stainless steel. All piping i joints and connections are welded, except for a minimum l Amendment C 5.4-29 June 30, 1988 l l
CESSAR nuinem number of flanged connections that are used to facilitate equipment maintenance or accommodate component design. D. Valves The location of valves, along with their type, type of operator, position (during the normal operating mode of the plant), type of position indication, and failure position is shown on Figures 6.3.2-1A and 6.3.2-1B. Throttle valves (SD-650, 652, 651, 653) are provided for remote control of the heat exchanger tube side and bypass flow.
- 1. Relief Valves Protection against overpressure of components within the SCS is provided by conservative design of the system piping, appropriate valving between high pressure sources and lower pressure piping, and by relief valves. The SCS suction lines up to and including SD-670 and SD-671 are designed for full RCS lC pressure. Relief valves will be provided as requirad by the applicable codes. All relief valves are of the totally enclosed, pressure tight type, with suitable provisions for gagging.
C
- 2. Actuator Operated Throttling and Stop Valves The failure position of each valve on loss of actuating signal or power supply is selected to ensure safe operation. System redundancy is considered when defining the failure position of any given valve.
Valve position indication is provided at the main control panel, as indicated in Figures 6.3.2-1A and 6.3.2-1B. A momentary push button with appropriate status control on the main control panel and/or manual override handwheel is provided where neccc:ary for , efficient and safe plant operation. All actuator operated valvee have stem leakage controlled by a double packing with a lantern ring leakoff connection. E. Shutdown Cooling System Pumps The function of the SCS pumps is to provide flow through the C reactor core and SCS heat exchangers for normal plant shutdown operation or as required for long term core cooling, i Amendment C 5.4-30 June 30, 1988
CESSAR 8Discum During normal operation the SCS pumps are isolated from the RCS by motor-operated valves. The shutdown cooling and containment spray functions have been evaluated to select a single pump to serve both functions. The flow available with a single SCS pump is sufficient to both maintain an acceptable cooldown rate (75'F/hr maximum) during shutdown cooling operation and C supply the CSS. The design temperature for the SCS pumps is based upon the temperature of the reactor coolant at the initiation of shutdown cooling (350'? nominal) plus a design tolerance resulting in a design temperature of 400*F. The design pressure for the pumps is based upon the system functional design pressure (see Section 5.4.7.1.2). The SCS pumps are vertical, single-stage centrifugal units equipped with mechanical seals backed up by a bushing, with a leakoff to collect the leakage past the seals. The seals are designed for operacion with a pumped fluid temperature of 400*F. The pump motors are specified to have the capability of starting and accelerating the driven equipment, unde; load, to design point running speed within 5 seconds, based upon an initial voltage of 75% of the rated voltage at the motor terminals, and increasing linearly with time to 90% voltage in the first 2 seconds, and increasing to 100% voltage in the next 2 seconds. The pumps are provided with drain and flushing connections to facilitate reduction of radiation levels before maintenance. The pressure containing parts are fabricated from stainless steels the internals are selected for compatibility with boric acid solutions. The punps are provided with minimum flow protection (recirculation lines) to prevent damage when starting against a closed system. The SCS pump data is provided in Table 5.4.7-1. During shutdown cooling, the pumps take suction from the reactor hot leg pipes and discharge through the SCS heat exchangers. The flow is then returned to the RCS through the SIS direct vessel injection nozzles, one SCS pump is aligned to each SCS heat exchanger. 5.4.7.2.3 OverDressure Prevention A. Overpressurization of the SCS by the RCS is prevented in the following ways:
- 1. The shutdown cooling suction isolation valves (SD-673, C 672, 671, 670) are powered by four independent power Amendment C 5.4-31 June 30, 1988 j
CESSAR E%nce,. supplies such that a fault in one power supply or valve will neither line up the RCS to either of the two SCS trains inadvertently nor prevent the initiation of shutdown cooling with at least one SCS train. h
- 2. Interlocks associated with the shutdown cooling suction isolation valves prevent the valves from being opened if RCS pressure exceeds (LATER), and close these valven automatically if RCS pressure should rise above the accumulation pressure of the SCS suction line relief p' valves. The instrumentation and controls which F implement this are discussed in Section 7.6.
- 3. The SCS suction valves inside the containment are designed for full RCS pressure with the second valve forming the pressure boundary and safety class change.
- 4. Alarms on SD-673, 672, 671 and 670 annunciate 9 hen the SCS suction isolation valves are not fully open. Also, ;
if SD-673 and 671 or SD-672 and 670 valves are open and RCS pressure exceeds the maximum pressure for SCS operation, an alarm will notify the operator that a pressurization transient is occurring during low temperature conditions.
- 5. Relief valves are provided as discussed in Section 5.4.7.2.2.
B. Inadvertent overpressurization of the SCS is also precluded by the use of Safety Injection Tank isolation valves (see C Sections 7.6) . 5.4.7.2.4 Applicable Codes and Classifications A. The SCS is a Safety Class 2 System, except for that portion discussed in B. below, which is Safety Class 1. B. The piping and valves from the RCS up to and - SD-671 and 670 are designed to ASME B&PVC Section III,including Class lC j 1. I C. The piping, valves, and components of the SCS, with the exception of those in Section 5.4.7.2. 4 (B) are designed to ASME B&PVC Section III, Class 2. l D. The component cooling water side of the SCS heat exchanger l is designed tc ASME B&PVC Section III, Class 3. I l Amendmen? C 5.4-32 June 30, 1988
~~ ~~ ~ ' ~ ' ~-~ ~ ~ ' ~
CESSARnnLm,. E. The power operated valves are designed to the applicable IEEE Standards. F. The SCS is a Seismic Category I System. 5.4.7.2.5 System Reliability considorations The SCS is designed to perform its design function assuming a single frilure, as described in Secticn 5.4.7.1.2. To assure availability of the SCS when required, redundant components and power supplies are utilized. The RCS can be brought to refueling temperature utilizing one of the two redundant SCS trains. However, with the design heat load, the cooldown would be considerably longer than the specified 96 hour l C time period. The SCS does not utilize any pneumatically operated valves. C The instrumentation, control, and electric equipment pertaining to the SCS is designed to applicable portions of IEEE Standards 279, 308 and 603. lC In addition to normal offsite power sources, physically and , electrically independent and redundant emergency power supply systems are provided to power safety-related components. See Chapter 8 for further information. Since the SCS is assential for a safe (cold) shutdown of the reactor, it is a seismic category I system and designed to remain C functional in the event of a safe shutdown earthquake. For long-term performance of the SCS without degradation due to , corrosion, only materials compatible with the pumped fluid are used. Environmental envelopes are specified for system components to , ensure acceptable performance in normal and applicable accident i environments (see Section 3.11). In the event of a limited leakage passive failure in one train of the SCS, continued core cooling is provided by the unaffected independent SCS train. The limited leakage passive failure will C be identified via appropriats leak detection provisions. (see Section 5. 4. 7.1. 3.C. 2) . Make-up of the leakage is provided by the manual alignment of the SIS to the IRWST or by opening the l Safety Injection Tank isolation valves. The affected SCS train I can then be isolated and core cooling continued with the other train. Amendment C 5.4-33 June 30, 1988
CESSAR n!Mem A limited leakage passive failure is defined as the failure of a pump seal or valve packing, whichever is greater. The leakage is expected to be from a failed SCS pump seal. maximum lC This leakage to the pump compartment will drain to the room sump. From there it is pumped to the waste management system. The sump l C pumps in each room will handle expected amounts of leakage. If leakages are greater than the sump pump capacity, the room will be isolated. 5.4.7.2.6 Manual Actions A. Plant Cooldown Plant cooldown is the series of manual operations which bring the reactor from hot shutdown to cold shutdown. Cooldown to approximately 350'F is accomplished by releasing steam from the secondary side of the steam generators. the RCS pressure falls below the normal operating range, WhenlC the Safety Injection Actuation Signal (SIAS) setpoint can be manually decreased as discussed in Section 7.2.1.1.1.6. When RCS pressure reaches (LATER) psig, the Safety Injection Tank pressure is reduced to (LATER) psig. When RCS pressure C reaches (LATER) psig, the Safety Injection Tank isolation valves are closed. When RCS temperature and pressure decrease below 350*F and the maximum pressure for SCS operation, the SCS may be used. If the SCS suction relief valves are not aligned to the RCS l C before cold leg temperature is reduced to below the maximum RCS cold leg temperature requiring LTOP, an alarm will notify the operator to open the SCS isolation valves (SD-673, 672, 671, 670). The maximum temperature requiring l C LTOP is based upon the evaluation of the applicable P-T curves. This operator action requires that the RCS be depressurized to below the maximum pressure for SCS operation, in order to clear tha permissive SCS interlock ' (see paragraph 5.4.7.2.3, item A.2). Interlocks associated with the six valves on the two SCS suction lines prevent l overpressurization of the SCS. See Section 7.6 and i 5.4.7.2.3 for details. Also, if SD-673 and 671 or SD-672 C l and 670 SCS suction isolation valves are open and RCS l pressure exceeds the maximum pressure for SCS operation, an I alarm will notify the operator that a pressurization transient is occurring during low temperature conditions. Shutdown cooling is initiated using the SCS pumps. The SCS is varmed up and placed in operation as follows (refer to Figures 6.3.2-1A, 6.3-2-1B, 5.4.7-3 and 5.4.7-4): C Amendment C 5.4-34 June 30, 1988
CESSAR nMem
- 1. The SCS suction line isolation valves (SD-673*, 672, 671*, 670, 659*, 658) are opened. C
- 2. The SCS throttle valves (SD-652, 653*) are cracked open.
- 3. The SCS warmup line isolation valves (SD-657*, 656) are opened and the SCS pumps are started to induce recirculation flow through the SCS (flow is limited to (IATER) gpm per pump).
- 4. Once flow has been induced in the SCS, the SCS isolation valves (SD-655*, 654) are cracked open to allow a small amount of flow from the RCS to heat up SCS valves and piping.
- 5. The SCS discharge isolation valves (SD-655*, 654) are then gradually opened, while the warmup line isolation valves (SD-657*, 658) are gradually closed to maintain a constant flow of (LATER) gpm per pump. When complete, the system is in its normal operational mode.
- 6. The SCS throttle valves (SD-652, 653*) and the SCS bypass flow control valves (SD-651*, 652) are adjustcd as necessary to maintain the RCS cooldown rate at 75'F/ hour or less, at a total SCS flow of (LATER) gpm through each subsystem until the refueling temperature of 120*F is attained.
A maximum rate of cooldown (not to exceed 75 ' F/ hour) is maintained by adjusting the flow rate of reactor coolant through the SCS heat exchangers utilizing the GCS throttle valves on the discharge of the heat exchangers (SD-653*, 652) in conjunction with the SCS bypass flow control valves (SD-651*, 650). With the shutdown cooling flow indicators, the operator maintains a total shutdown cooling flow rate by adjusting the amount of coolant which bypasses the SCS heat ' exchangers. When the system is first put into operation, the temperature difference for heat transfer is large and only a portion of C the total flow from the SCS pumps is diverted through the heat exchangers. As cooldown proceeds, the temperature differential decreases and the flow rate through the heat ' exchangers is increased to maintain the maximum permissible C cooldown rate.
*0dd numbered valves are located in SCS Train 1 and even numbered valves are located in SCS Train 2.
Amendment C 5.4-35 June 30, 1988
\
CESSARSEGem l The flow to the SCS heat exchangers is increased periodi-cally until full SCS pump flow through the heat exchangers C is attained. A graph of RCS temperature vs. time af ter shutdown for a normal design basis cooldown is presented in Figure 5.4.7-1. l C Shutdown cooling is continued throughout the entire period of of plant shutdown 120*F or less.to maintain a refueling Whenever water shutdown temperature cooling is inl C operation, shutdown purification flow may be initiated to purify the circulating coolant in the CVCS. B. Plant Heatup Plant heatup is a series of manual operations which bring the RCS from cold shutdown to hot standby. The SCS is used during Prior tocold shutdown plant heatup,tothecontrol SCS reactor coolant temperature. heat exchangers are bypassedl C to maintain flow through the core without the heat removal effect of the heat exchangers. Flow can be initiated to the heat exchangers 12 necessary to control the heatup rate. When the reactor coolant punpa can be run, the SCS pumps are C stopped and the system is isolated for the standby mode. C. Abnormal Operation
- 1. The SCS heat exchangers may be used to supplement the spent fuel pool cooling heat exchangers when more than one third of a spent core is stored in the spent fuel pool. Normally this would be done during refueling when both SCS heat exchangers are no longer needed to maintain reactor coolant at the refueling temperature.
The SCS would be aligned with one heat exchanger train lined up to the spent fuel pool cooling system and the other SCS heat exchanger train lined up for shutdown cooling of the RCS. The SCS heat exchanger train aligned to the spent fuel pool would be in a shutdown cooling lineup for use of SCS pumps, except normal lC the pumps take a suction on the spent fuel pool vice the RCS, and the discharge of the shutdown cooling heat exchanger goes to the spent fuel pool, vice the RCS. , 2. Initiation of shutdown cooling with the most limiting l single failure (loss of one shutdown cooling train) can be accomplished using the procedure under plant cooldown for the operable train (i.e., operating the valves with (*) for SCS train number 1 or the valves without (*) for SCS train number 2). Amendment C 5.4-36 June 30, 1988
l l CESSARnucm D. Design Bases Event Operations Following certain Design Bases Events (feedwater line break, small break LOCA, steam line break, or loss of power), shutdown cooling can be initiated with RCS hot conditionr. which exceed the normal shutdown cooling offsitelC leg initiation temperature of 350*F. However, shutdown cooling will navsr be initiated at conditions which exceed design temperature of the SCS components. the l C 5.4.7.3 Performance Evaluation The design point of the SCS is taken at 96 hours after plant shutdown. At this point, the desigr casis is to maintain a 120*F C refueling temperature with a service water temperature of 95'F. Two SCS heat exchangers and two SCS pumps are assumed to be in operation at the design flow of (LATER) gpm per train. The SCS heat exchanger size is determined at this point, since it requires the greatest heat transfer area due to the relatively small aT between primary fluid and component cooling water. The design input heat load at 96 hours is based on decay heat at 96 hours, assuming an average reactor core burnup of two years. C Additional energy input to the RCS from two SCS pumps running at design flow rate was also included with no credit taken for component energy losses to the external environment. For the cooldown temperature process of 350*F from the shutdown to a refueling cooling temperature initiation of 120'F, the lC heat load evaluated is comprised of the instantaneous decay heat, SCS pump heat input, and the stored sensible heat of the primary l C and secondary liquid and metal masses. Metal mass is assumed to be steel with a specific heat of 0.12 Btu /lb 'F. The temperature of the component cooling water to the SCS heat exchanger is taken C as 120'F initially, gradually decreasing to (LATER) when 120*F refueling temperature is attained. At each time intarval in the cooldown, an iterative process is utilized to analyze transient performance, whereby the heat removal is established by balancing the available heat load with the SCS heat exchanger heat removal capability. The cooldown rate is limited to a maximum of 75'F/ hour throughout the cooldown. The normal two train cooldown curve is shown in Figure 5.4.7-1. With the most limiting single active failure in the SCS, RCS temperature can be brought to 200'F within 24 hours following shutdown using one SCS pump and one SCS heat exchanger assuming C that the RCS pressure and temperature are reduced to SCS initiation conditions by other heat rejection means in 3.5 hours. The single train cooldown curve is shown in Figure 5.4.7-2. Amendment C 5.4-37 June 30, 1988
CESSARUS%am. The CCS is designed utilizing a philosophy of total physical separation of redundant trains such that the system can carry out its safety function assuming a single active failure during both normal and short-term post accident modes and a single active or passive failure during long-term post accident modes (i.e., time periods >24 hr) after event initiation. Total train separation C assures that a single failure in one train cannot preclude the second train from accomplishing its safety functions. A Failure Modes and Effects Analysis for the SCS is presented in Table 5.4.7-3. 5.4.7.4 Preonorational Testino Preoperational tests are conducted to verify proper operation of the SCS. The preoperational tests include calibration of instrumentation, verification of adequate cooling flow, and verification of the operability of all associated valves. In addition, a preoperational hot functional performance test is made on the installed 3CS heat exchangers as part of the precore hot functional test program. See Chapter 14 for further details on these tests. The SCS also undergoes a series of preoperational hydrostatic tests conducted in sccordance with Section III of the ASME Boiler and Pressure Vessel Code. l I I l l Amendment C ' 5.4-38 June 30, 1988 i l l
- I CESSAR E!nincum THIS PAGE INTENTIONALLY BLANK l
l l l l Amendment C 5.4-34 June 30, 1988
CESSAR niacm:n THIS PAGE INTENTIONALLY BIANK l l l ! Amendment C 5.4-40 June 30, 1988
CESSAR EL' Geni:n THIS PAGE INTENTIONALLY BLANK Amend =ent C 5.4-41 June 30, 1988
t CESS EGGMRA CLiTIFICATl3M B Ossi?N t 9 4 i THIS PAGE INTENTIONALLY BLANK 4 4 I l I I 1 Amendment C l l ' 5.4-42 June 30, 1968 l-------- - _ __ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ ._ _ , _ _ _ . _ _
s t CESSAR nEncam,. , i TABLE 5.4.7-1 lC (Sheet 1 of 2)
\
SHUTDOWN COOLING DESIGN PARAMETERS
\
SYSTEM DESIGN PARAMTERS Shutdown cooling system startup Approximate')y 3.5 hours after reactor shutdown or trip Reactor coolant system maximum cooldown rate (at initiation of shutdown cooling), 'F/hr 75 Refueling water temperature, 'F 129 Nominal shutdown cooling flow, gpm/HX 5000 C,
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( COMPONENT DESIGN PARAMETERS Shutdown Coolina Heat Exchancer Dats t Quantity 2 Type 2 Shell and' tt.oe, horizontal U tube Service transfer rate,28tu/hr 'F-ft 350 C Heat Transfer area, ft /HX (LATER) Tube Side Fluid Reactor coolant Design pressure, psig (LATER) C Design temperature, 'F 400 Material Austenitic stainless steel Code ASME Section 111, Class 2 Fouling resistance, hr-ft 2,.F/8tu (LATER) lC Shall Side Fluid Component cooling water Design pressure, psig 150 Design temperature, 'F 250 Material Carbon Stool Code ASME Sect'on III, Class 3 Fouling resistance, hr-ft.2 *F/ Btu (LATER) At 96 hours after shutdown: 0 Tube Side Flow, million lb/hr. 2,47 Inlet temperature, 'F 120 Outlet temperature, 'F 110 Shell Side Flow,millionlb/hr 5.46 Inlet temperature 'F (LATER) Outlet temperature, 'F (LATER) Heat load, million Btu /hr 25.5 C AIcondment C June 30,' 088 >
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'p Safety Classification 2 Cctn ! ASME III, Class 2 C l# tesig Pressure . fLATER)
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7ESSAR E!nine,.uon i I t IAEJJ.1,.Z..2 i / SHlfTDOWN COOLING SYSTEM c\ INTERFACE RQllBf3Fy$ FOR C0HPONENT C00 Lift (M, ER s
's Shutdown \ ' Shutdown Cooling \ ' Cooling Mode ,_ (3.5 hrs.1 , ; , Jf;ML , , l, Supply Temp, 'F (Nax)(A) 120 (LATE'R)
C Outlet Temp, 'F 142 . (LATERL Minimum Flow per SDCHX, g p 52) 11,000 ' h,,000 Total Heat Load 6 212 51 for Both SDCHX, 10 Btu /t'r ,
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NOTES:(1) For maximum sup' ply te peratures lower than those 4 listed, the ufnimum flow listed may be reduced provided that the ! heat removal capability cf the Shutdown Cooling System l is not adverse'ly affecte':. Conversely, for Component Cooling Water supply temperatures lcwer than those , , listeJ, it may be necessary to reduce flow, so that ae heat capacity of the ultinate heat sink is not exceeded. C i (2) Maximum allowable component cooling water flow through each I shutdown cooling heat exchanger is 13,000 gpm. 1 I A
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CESSAR Ennneim. (Sheet 1 of 7) EFFE0TIVE PAGE LISTING CHAPTER 6 Table of Contents Page Amendment i 11 111 iv v C vi C vil viii ix x xi Xii xiii C xiv xv xvi C xvii C xviii xix xx xx1 xxii xxiii xxiv xxv xxvi xxvii xxviii xxix Text i l Page Amendment 6.1-1 6.1-2 6.1-3 6.1-4 6.2-1 6.2-2 6.2-3 l Amendment C June 30, 1988 l
CESSAR anMacm:= (sw..t 2 or 7) EFFECTIVE PAGE LIST MQ (Cont'd) CHAPTER 6 Text (Cont'd) Pace Amendment 6.2-4 6.2-5 6.2-6 6.2-7 6.2-8 6.2-9 6.2-10 6.2-11 6.2-12 6.2-13 6.2-14 6.2-15 10 6.2-16 6.2-17 6.2-18 6.2-19 6.2-20 4 6.2-21 4 6.2-22 4 6.2-23 ! 6.2-24 1 6.2-25 l 6.2-26 6.2-27 6.2-28 6.2-29 6.2-30 6.3-1 C l 6.3-2 C 6.3-3 C 6.3-4 C 6.3-5 C 6.3-6 C 6.3-7 C 6.3-8 C 6.3-9 C 6.3-10 C 6.3-11 C 6.3-12 C 6.3-13 C 6.3-14 C 6.3-15 C 6.3-16 C Amendment C l June 30, 1988 l
CESSAR !!!#lCATION (Shsst 3 of 7) EFFECTIVE PAGE LISTING (dont'd) CHAPTER 6 Text (Cont'd) Pace c.mendment 6.3-17 C 6.3-18 C 6.3-19 C 6.3-20 C 6.3-21 C 6.3-22 4 6.3-23 6.3-24 10 6.2-25 4 6.3-26 6.3-27 10 6.3-28 4 6.3-29 6.3-30 10 6.3-31 6.3-32 4 6.3-33 9 6.3-33(a) 9 6.3-34 9 6.3-35 9 C.3-36 c.3-37 6.3-38 6.3-39 6.3-40 6.3-40(a) 4
- 6. 3 -4 0 (b) 4 6.3-41 6.3-42 1 6.3-43 l 6.3-44 l 6.4-1 Tables Amendment 6.1-1 (Sheets 1 and 2) 8 6.1-2 6.1-3 6.1-4 G.2.1-1 (Sheets 1-3) 6.2.1-2 (Sheets 1-14) 6.2.1-3 (Sheets 1-11) 6.2.1-4 (Sheets 1-11)
Amendrent C June 30, 1988
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CESSAR E!.'ainem:n (Sh .e 4 or 2) i EFFECTIVE PAGE LISTLEQ (Cont'd) CHAPTER 6 Tables (Cont'd) Amendment 6.2.1-5 (Sheets 1-11) 6.2.1-6 (Sheets 1-13) 6.2.1-7 (Sheets 1-13) 6.2.1-8 (Sheets 1-13) 6.2.1-9 (Sheets 1-13) 6.2.1-10 (Sheets 1-7) 6,2.1-11 (Sheets 1-6) (Sheet 7) 10 6.2.1-12 (Sheets 1-8) (Sheet 9) 10 6.2.1-13 (Sheets 1-6) (Sheet 7) 10 6.2.1-14 (Sheets 1-8) (Sheet 9) 10 6.2.1-15 (Sheets 1-6) (Sheet 7) 10 6.2.1-16 (Sheets 1-8) (Sheet 9) 10 6.2.1-17 (Sheets 1-6) (Sheet 7) 10 6.2.1-18 (Sheets 1-8) (Sheet 9) 10 6.2.1-19 (Sheets 1-6) (Sheet 7) 10 6.2.1-20 (Sheets 1-8) (Sheet 9) 10 6.2.1-21 6.2.1-22 6.2.1-23 (Sheets 1 and 2) 6.2.1-24 6.2.1-25A (Sheets 1-3) 6.2.1-25B (Sheets 1-3) 6.2.1-26A (Sheets 1-3) 6.2.1-26B (Sheets 1-3) l 6.2.1-27A (Sheets 1-3) l 6.2.1-27B (Sheets 1-3) 6.2,1-28A (Sheets 1-3) 6.2.1-28B (Sheets 1-3) 6.2.1-29A (Sheets 1-3) 6.2.1-298 (Sheets 1-3) 6.2.1-30 (Sheets 1-3) 6.2.1-31A (Sheets 1-3) 6.2.1-31B (Sheets 1-3) 6.2.1-32A (Sheets 1-3) l Amendment C June 30, 1988 l
. 1 CESSAR n!G"icnic. (Shoot 5 of 7) i l
1 l EFFECTIVE PAGE LISTIliG (Cont'd) CHAPTER 6 Tables (Cont'd) Amendment 6.2.1-32B (Sheets 1-3) 6.2.1-33A (Sheets 1-3) 6.2.1-33B (Sheets 1-3) 6.2.1-34 (Sheets 1-3) 5.2.1-35A (Sheets 1 and 2) 6.2.1-35B (Sheets 1 and 2) 6.2.1-36 6.2.1-37 (Sheets 1-3) 4 6.2.1-38 6.2.4-1 (Sheets 1-3) (Sheet 4) 10 (Sheet 5) 6.3.2-1 (Sheet 1-3) C 6.3.2-2 C 6.3.2-3 C 6.3.2-4 C 6.3.3.2-1 4 6.3.3.2-2 4 6.3.3.2-3 6.3.3.2-4 4 6.3.3.2-5 4 6.3.3.2-6 6.3.3.3-1 6.3.3.3-2 4 6.3.3.3-3 6.3.3.3-4 6.3.3.3-5 4 6.3.3.3-6 10 6.3.3.5-1 (Sheet 1) 10 (Sheet 2) 6.3.3.5-2 (Sheets 1 and 2) 6.3.3.5-3 - 6.3.3.6-1 6.3.3.7-1 1 l ( Amendment C June 30, 1988 I l
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CESSAR !!ninema (Shast 6 of 7) l EFFECTIVE PAGE LISTING (Cont'd) CHAP'"ER 6 l l Ficures ! Amendment 6.2.1-1 (Sheets 1 and 2) 6.2.1-2 6.2.1-3 6.2.1-4 6.2.1-5 6.2.1-6 6.2.1-7 6.2.1-8 6.2.1-9 6.2.1-10 6.2.1-11 (Sheets 1 and 2) 6-2.1-12 (Sheets 1 and 2) 6.2.1-13 (Sheets 1 and 2) 6.2.1-14 (SheatL 1 and 2) 6.2.1-15 (Sheets 1 and 2) 6.2.1-16 (Sheets 1 and 2) 6.2.1-17 (Sheets 1 and 2) 6.2.1-18 (Sheets 1 and 2) 6.2.1-19 (Sheets 1 and 2) 6.2.1-20 (Sheets 1 and 2) 6.2.1-21 6.2.1-22 6.2.1-2? 6.2.1-24 4 6.2.1-25 4 6.2.1-26 4 6.2.4 (Sheets lA, 1C) (Sheet 1B) 10 6.3.2-1A C 6.3.2-1B C 6.3.2-10 C-6.3.2-1D C 6.3.2-1E C 6.3.2-1F C 6.3.2-2 C i 6.3.2-3 C l 6.3.3.2 (Sheets lA-lH) 4 l (Sheets 2A-2H) 4 (Sheets 3A-3H) 4 (Sheets 4A-4H) 4 (Sheets 5A-50) 4 I (Sheets SP) 7 I 1 Amendment C June 30, 1988
CESSAR !!nincuia (Shast 7 of 7) EFFECTIVE PAGE LISTING (Cont'd) CHAPTER 6 Ficures (Cont'd) Amendment (Sheets SQ-5U) 7 (Sheets 6A-6H) 4 (Sheets 6I-6S) (Sheets 7A-7H) 4 (Sheets 8A-8G) (Sheet 8H) 4 (Sheets 9A-9G) (Sheet 9H) 4 (Sheets 10 and 11) 4 6.3.3.3 (Sheets lA-lH) (Sheets 2A-2H) (Sheets 3A-3E) (Sheet 3F) 4 (Sheets 3G and 3H) (Sheets 4A-4H) (Sheets 5A-5H) (Sheets 6A-6H) (Sheet 7) 6.3.3.4-1 9 6.3.3.4-2 6.3.3.4-3 6.3.3.4-4 6.3.3.4-5 4 6.3.3.4-6 6.3.3.5 (Sheets lA-lF) l l I l Amendment C j June 30, 1988 l
)
CESSARSEWic- ; TABLE OF CONTENTS CHAPTER 6 Secticg Subiect Pace No. 6.0 ENGINEERED SAFETY FEATURES 6.1-1 6.1 ENGINEERED SAFETY FEATURE MATERIALS 6.1-1
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6.1.1- METALLIC MATERIALS 6.1-1 6.1.1.1 Materials Selection and Fabrication 6.1-1 6.1.1.1.1 Specifications for Principal ESF Pressure Retaining Materials 6.1-1 6.1.1.1.2 Engineered Safety Features Construction Materials 6.1-1 6.1.1.1.3 Integrity of ESF Components During Manufacture and Construction 6.1-2 6.1.1.1.3.1 Control of Sensitized Stainless Steel 6.1-2
- 6.1.1.1.3.2 Cleaning and Contamination Protection Procedures 6.1-3 6.1.1.1.3.3 Cold Worked Stainless i Steel 6.1-3 1 6.1.1.1.3.4 Non-Metallic Insulation 6.1-4 6.1.1.1.4 Weld Fabrication and Assembly of Stainless Steel ESF Components 6.1-4 6.1.2 ORGANIC MATERIALS 6.1-4 6.1.2.1 Protective Cortinas 6.1-4 6.1.2.2 Other Materials 6.1-4 6.2 CONTAINMENT SYSTEMS 6.2-1 6.2.1 CONTAINMENT FUNCTIONAL DESIGN 6.2-1 l
1
CESSAR ! Enema TABLE OF CONTENTS (Cont'd) CHAPTER 6 Section Subiect Pace No. 6.2.1.1 Centainment Structutg 6.2-1 6.2.1.1.1* Design Bases 6.2-1 6.2.1.1.1.1 Postulated Accident Conditions 6.2-1 6.2.1.1.1.2 Mass and Energy Release 6.2-2 6.2.1.1.1.3 Effects of ESF Systems Energy Removal 6.2-2 6.2.1.1.1.4 Effects of ESF Systems on Pressure Reduction 6.2-2 6.2.1.1.1.5* Containment Leakage Rate Bases 6.2-2 6.2.1.1.1.6 Bases for Analysis of Minimum Containment Pressure 6.2-2 6.2.1.1.2* Design Features 6.2-3 6.2.1.1.3 Design Evaluation 6.2-3 6.2.1.1.3.1 Containment Peak Pressure Analysis 6.2-3 6.2.1.1.3.2 Long-Term Containment Performance 6.2-5 6.2.1.1.3.3 Energy Balance 6.2-5 6.2.1.1.3.4 Accident Chronology 6.2-5 6.2.1.1.3.5* Functional Capability of Containment Normal Ventilation Systems 6.2-5 6.2.1.1.3.6* Protection Against Severe External Loadings 6.2-5 6.2.1.1.3.7* Post-accident Containment Pressure / Temperature Monitoring 6.2-5 6.2.1.2 Containment Subcomoartments 6.2-5 6.2.1.2.1 Design Bases 6.2-5 6.2.1.2.2* Design Features 6.2-6
- See Site-Specific SAR 11
l CESSARE%A - TABLE OF CONTENTS (Cont'd) ) CHAPTER 6 Section Subiect Pace No. ! 6.2.1.2.3 Design Evaluation 6.2-6 6.2.1.3 Mass and Enerav Release Analyses for Postulated Loss of Coolant Accidents 6.2-7 6.2.1.3.1 Mass and Energy Release Data 6.2-8 6.2.1.3.2 Energy Sources 6.2-8 6.2.1.3.3 Description of Blowdown Model 6.2-9 6.2.1.3.4 Description of Core Reflood Model 6.2-10 6.2.1.3.5 Description of Post Reflood Model 6.2-11 6.2.1.3.6 Description of Long Term Cooling Model 6.2-12 6.2.1.3.7 Single Active Failure Analycis 6.2-13 6.2.1.3.8 Metal-Water Reaction 6.2-14 6.2.1.3.9 Energy Inventories 6.2-14 6.2.1.3.10 Additional Information 6.2-14 6.2.1.4 Mass and Enerav Release Analysis l for Postulated Secondary System Eloe Ruotures Inside Containment 6.2-14 l 6.2.1.4.1 Mass and Energy Release Data 6.2-16 6.2.1.4.2 Single Failure Analysis 6.2-16 6.2.1.4.3 Initial Conditions 6.2-16 6.2.1.4.4 Description of Blowdown Model 6.2-17 i 6.2.1.4.5 Energy Inventories 6.2-19 l 6.2.1.4.6 Additional Information 6.2-19 1 4 t l iii l l_
CESSARnnine-TABLE OF CONTENTA (Cont'd) CEAPTER 6 l Section . Sic _tlan Pace No. l l 6.2.1.5 Minimum Containment Pressure bnalysis for Performance Q3cability Studies On Emercency Core Con. lina System 6.2-20 6.2.1.5.1 Introduction and Summary 6.2-20 6.2.1.5.2 Method of Calculation 6.2-20 6.2.1.5.3 Input Parameters 6.2-20 6.2.1.5.3.1 Mass and Energy Release Data 6.2-20 6.2.1.5.3.2 Initial Containment Internal Conditions 6.2-20 6.2.1.5.3.3 Containment Volume 6.2-21 6.2.1.5.3.4 Active Heat Sinks 6.2-21 6.2.1.5.3.5 Steam Water Mixing 6.2-21 6.2.1.5.3.6 Passive Heat Sinka 6.2-21 6.2.1.5.3.7 Heat Transfer to Passive Heat Sinks 6.7 21 6.2.1.5.3.8 Containment Purge System 6.2-21 6.2.1.5.4 Results 6.2-21 6.2.1.6 Testina and Insoection 6.2-22 6.2.1.7 Instrumentation Aeolications 6.2-22 6.2.2* CONTAINMENT HEAT REMOVAL SYSTEMS 6.2-25 6.2.3* SECONDARY CONTAINMENT FUNCTIONAL DESIGN 6.2-25 6.2.4 CONTAINMENT ISOLATION SYSTEM 6.2-25 6.2.4.1 Desian Bases 6.2-25 6.2.4.1.1 Criteria 6.2-25 6.2.4.1.2 Design Requirements 6.2-26 l l
*See Site-Specific SAR l
iv
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CESSAR !!nincan TABLE OF___ CONTENTS (Cont'd) CHAPTER 6 Section Subiect Pace No. 6.2.4.2 System Desian 6.2-27 6.2.4.2.1 General Description 6.2-27 6.2.4.2.2 Applicable Codes, Standards, and Regulatory Guides 6.2-28 6.2.4.3 Desian Evaluation 6.2-28 6.2.4.4 Tests and Inseections 6.2-29 6.2.5* COMBUSTIBLE GAS CONTROL IN CONTAINMENT 6.2-30 6.2.6* CONTAINMENT LEAKAGE TESTING 6.2-30 6.3 SAFETY INJECTION SYSTEM 6.3-1 C 6.3.1 DESIGN BASES 6.3-1 6.3.1.1 Summary Descriotion 6.3-1 6.3.1.2 Criteria 6.3-1 6.3.1.2.1 Functional Design Bases 6.3-1 6.3.1.2.2 Reliability Design Bases 6.3-2 6.3.1.3 Interface Recu(rements 6.3-2 6.3.2 SYSTEM DESIGN 6.3-11 6.3.2.1 System Schematic 6.3-11 6.3.2.2 Comconent Descriotion 6.3-12 ' 6.3.2.2.1 Incontainment Refueling Water Storage Tank 6.3-12 C 6.3.2.2.2 Safety Injection Tanks 6.3-12 6.3.2.2.3 Safety Injection Pumps 6.3-13 6.3.2.2.4 Piping 6.3-15 6.3.2.2.5 Valves 6.3-15 C j *See Site-Specific SAR Amendment C v June 30, 1988
CESSAR Mine-TABLE OF CONTENTS (Cont'd) CHAPTER 6 Section Subiect Pace No. 6.3.2.3 Aeolicable Codes and Classifications 6.3-16 6.3.2.4 Materials Soecifications and Comoatibility 6.3-16 6.3.2.5 System Reliability 6.3-16 6.3.2.5.1 Safety Injection Tanks 6.3-16 6.3.2.5.2 Safety Injection Subsystems 6.3-17 6.3.2.5.3 Power Sources lC 6.3-18 6.3.2.5.4 Capacity to Maintain Cooling Following a Single Failure 6.3-18 6.3.2.6 Protection Provisions 6.3-20 6.3.2.6.1 Capability to Withstand Design Bases Environment 6.3-20 6.3.2.6.2 Missile Protection 6.3-20 6.3.2.6.3 Seismic Design 6.3-21 6.3.2.7 Recuired Manual Actions 6.3-21 6.3.3 PERFORMANCE 6.3-22 6.3.3.1 Introduction and Summarv 6.3-22 6.3.3.2 Larce Break Analysis 6.3-23 6.3.3.2.1 Mathematical Model 6.3-23 6.3.3.2.2 Safety Injection System Assumptions 6.3-23 1 6.3.3.2.3 Core and System Parameters 6.3-24 1 6.3.3.2.4 Containment Parameters 6.3-25 1 6.3.3.2.5 Break Spectrum 6.3-25 I l i l Amendment C vi June 30, 1988
CESSAR 8!!Lmu TABLE OF CONTENTS (Cont'd) CHAPTER 6 l Section Subiect Pace No 6.3.3.3 Small Break Analveis 6.3-27 ) 6.3.3.3.1 Evaluation Model 6.3-27 ' 6.3.3.3.2 Safety Injection System Assumptions 6.3-27 6.3.3.3.3 Core and System Parameters 6.3-28 6.3.3.3.4 Containment Parameters 6.3-28 6.3.3.3.5 Break Spectrum 6.3-28 l 6.3.3.3.6 Results 6.3-29 6.3.3.3.7 Instrument Tube Ruptdre 6.3-30 6.3.3.4 Post-LOCA Lona Term Coolina 6.3-33 6.3.3.4.1 General Plan 6.3-33 6.3.3.4.2 Assumptions Used in the Performance 6.3-33a Evaluation of the LTC Plan 6.3.3.4.3 Parameters Used in the Performance Evaluation of the LTC Plan 6.3-34 6.3.3.4.4 Results of the LTC Performance Evaluation 6.3-35 6.3.3.5 Secuence of Event and Systems Oceration 6.3-36 6.3.3.6 Radioloaical Consecuences 6.3-39 6.3.3.7 Chanter 15 Accident Analysis 6.3-40b 6.3.4 TE3TS AND INSPECTIONS 6.2-41 6.3.4.1 ECCS Performance Tests 6.3-41 6.3.4.2 Reliability Tests and Inscections 6.3-41 6.3.4.2.1 System Level Tests 6.3-41 6.3.4.2.2 Component Testing 6.3-41 vii
CESSAR 8nsincum TABLE OF CONTENTS (Cont'd) CHAPTER 6 Section Subiect Pace No. 6.3.5 INSTRUMENTATION 6.3-41 6.3.5.1 Desian Criteria 6.3-41 6.3.5.2 System Actuation Sianals 6.3-42 6.3.5.2.1 Safety Injection Actuation Signal (SIAS) 6.3-42 6.3.5.3 Instrumentation Durina Operation 6.3-42 6.3.5.3.1 Temperature 6.3-43 6.3.5.3.2 Pressure 6.3-43 6.3.5.3.3 Valve Position 6.3-43 6.3.5.3.4 Level 6.3-44 6.3.5.3.5 Flow 6.3-44 6.3.5.4 Post Accident Instrumentation 6.3-44 6.4* HABITABILITY SYSTEMS 6.4-1 6.5* FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 6.4-1 6.6 INSERVICE INSPECTION OF CLASS 2 &3 COMPONENTS 6.4-1 6.6.1* COMPONENTS SUBJECT TO EXAMINATION 6.4-1 6.6.2 ACCESSIBILITY 6.4-1 APPENDIX 6A CONTAINMENT SPRAY SYSTEM LICENSING - REPORT 6A-1 i APPENDIX 6B IODINE REMOVAL SYSTEM LICENSING REPORT 6B-1 l
*See Site-Specific SAR viii
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LIST OF TABLES CEAPTER 6 Table Subioot j 6.1-1 Principal ESF Pressure Retaining Materials 6.1-2 Engineered Safety Features Structural Materials That Could Be Exposed To Core Cooling Water or Containment Spray In The Event of A LOCA 6.1-3 Coating Materials Used In Containment 6.1-4 Other Organic Materials In Containment 6.2.1-1 Postulated Accidents For Containment Design Peak Pressure / Temperature Determination 6.2.1-2 Data For Containment Peak Pressure / Temperature Analysis 6.2.1-3 Data For Containment Peak Pressure / Temperature Analysis 6.2.1-4 Data For Containment Peak Pressure / Temperature Analysis 6.2.1-5 Data For Containment Peak Pressure / Temperature Analysis 6.2.1-6 Data For Containment Peak Pressure / Temperature Analysis 6.2.1-7 Data For Containment Peak Pressure / Temperature Analysis 6.2.1-8 Data For Containment Peak Pressure / Temperature Analysis
- 6.2.1-9 Data For Containment Peak Pressure / Temperature Analysis l
6.2.1-10 Data For Containment Peak Pressure / Temperature ! Analysis 6.2.1-11 Data For Containment Peak Pressure / Temperature Analyses 102% Powe r/ Slot /8. 78 Sq. Ft./ Loss of Containment Cooling ix l
CESSAR8MMnc-LIST OF TABLES (Cont'd) CHAPTER 6 Section Subiect 6.2.1-12 Data For Containment Peak Pressure / Temperature > Analyses 102% Power / Guillotine /8.78 Sq. Ft./ Locs 1 of Containment Cooling 6.2.1-13 Data For Centainment Peak Pressure / Temperature ) Analyses 75% Power / Slot /8.78 Sq. Ft ./ Loss of Containment Cooling 6.2.1-14 Data For Containment Peak Pressure / Temperature Analyses 75% Power / Guillotine /8.78 Sq. Ft./ Loss of Containment Cooling 6.2.1-15 Data For Containment Peak Pressure / Temperature Analyses 50% Power / Slot /8.78 Sq. Ft./ Loss of Containment Cooling 6.2.1-16 Data For Centainment Peak Pressure / Temperature Analyses 50% Power / Guillotine /8.78 Sq. Ft./ Loss of Containment Cooling . 6.2.1-17 Data For Containment Peak Pressure / Temperature Analyses 25% Power / Slot /8. 78 Sq. Ft ./ Loss of Containment Cooling 6.2.1-18 Data For Containment Peak Pressure / Temperature Analyses 25% Power / Guillotine /8.78 Sq. Ft./ Loss of Containment Cooling 6.2.1-19 Data For Containment Peak Pressure / Temperature Analyses 0% Power / Slot /4. 00 Sq. Ft ./ Los s of Containment Cooling 6.2.1-20 Data For Containment Peak Pressure / Temperature Analyses 0% Power / Guillotine /8.78 Sq. Ft ./ Los s of Containment Cooling 6.2.1-21 Summary of Calculated Energy Releases 6.2.1-22 Initial Conditions For Containment Peak Pressure Analysis 6.2.1-23 Engineered Safety Feature Systems operating Assunptions For Containment Peak Pressure Analysis X
CESSAR nMince,. LIST OF TABLES (Cont'd) CHAPTER 6 Section Subiect 6.2.1-24 Summary of Postulated Pipe Ruptures For Containment Subcompartment Analysis 6.2.1-25A Mass / Energy Relea'se Data For 100 Sq. Inch Hot Leg Guillotine Break For Containment Subcompartment Analysis (Flow From Reactor Vessel Side) 6.2.1-25B Mass / Energy Release Data For 100 Sq. Inch Hot Leg Guillotine Break For Containment Subcompartment Analysis (Flow From Reactor Vessel Side) 6.2.1-26A Mass / Energy Release Data For 600 Sq. Inch Hot Leg Guillotine Break For Containment Subcompartment Analysis (Flov From Reactor Vessel Side) 6.2.1-26B Mass / Energy Release Data For 600 Sq. Inch Hot Leg Guillotine Break For Containment Subcompartment Analysis (Flow From Steam Generator Side) 6.2.1-27A Mass / Energy Release Data For 350 Sq. Inch Discharge Leg Guillotine Break For Containment Subcompartment Analysis (Flow From Reactor Coolant Pump Side) 6.2.1-27B Mass / Energy Release Data For 350 Sq. Inch Discharge Leg Guillotine Break For Containment Subcompartment Analysis (Flow From Reactor Coolant Pump Side) 6.2.1-28A Mass / Energy Release Data For 480 Sq. Inch Discharge Leg Guillotine Break For Containment Subcompartment Analysis (Flow From Reactor Coolant Pump Side) 6.2.1-28B Mass / Energy Release Data For 480 Sq. Inch Discharge Leg Guillotine Break For Containment Subcompartment Analysis (Flow From Reactor Vessel Side) 6.2.1-29A Mass / Energy Release Data For 430 Sq. Inch Discharge Leg Guillotine Break For Containment Subcompartment Analysis (Flow From Steam Generator Side) xi
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LIST OF TABLES (Cont'd) CHAPTER 6 , j.
%N Section glibiect ,
6.2.1-29B Mass / Energy Release Data For 430 Sq. Inch Discharge Leg Guillotine Break For Containment Subcompartment Analysis (Flow Fror Reactor Coolant, Pump Side) s 6.2.1-30 Mass / Energy Release Data For 532 Sq. Inch Suction x. Leg Slot Break For Containment Subcompartment, \ Analysis 1 6.2.1-31B Mass / Energy Release Data For 591 Sq. Inch Suctior. Leg Guillotine Break For Containment Subcompartment Analysis (Flow From Reactor Coolant , [. ,g Pump Side) b 6.2.1-32A Mass / Energy Release Data For Double-Ended Surge g ' Line Guillotine Break For ~ Containment ~ Subcompartment Analysis (Flow From Hot Leg Side) 6.2.1-32B Mass / Energy Release Data For . 9ouble-Ended Sune '
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Line Guillotine Break 'For Ccntaiveent '
- Subcompartment Analysis (Flow From Prestt41Ier Side) \
6.2.1-33A Mass / Energy Release Data For Pressurizer Spray ' *'
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Line Break For Containment Subcompartment Analysis ,; - (Flow From Pressurizer Side) \ t j 1, 6.2.1-33B Mass / Energy Release Data For Pressurizer Spray Line Break For Containment Subcompa.rtme:nt Analysis (Flow From Discharge Leg Side) >
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6.2.1-34 Mass / Energy Release Data For Press. trizer Safety Valve Line Break For Containment Yabcompartment Analysis [, 6.2.1-35A Mass / Energy Release Data For 900.5 Sk Inch Main g Steam Line Guillotine Broa'c At SC tioz zle For - Containment Subcompartment Anal > sis](Flow' From , Ruptured SG Side) y 6.2.1-35B Mass / Energy Release Data For 800.5 Sq.!Iy.:h Main b Steam Line Guillotine Break At SG N>ehle For Containment Subcompartment Analysis (Flow From ; Intact SG Side)
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LIST OF TABLES (Cont'd) i CHA ?TER 6 ! grction i s 3.u M e c t r (
' Primary Side Resistar ce Factors, FLOOD. MOD 2 CODE 6.2.1-36 6.2.1-37 ;
Blowdown And Reficod Mass And Energy Release 0.8 x (> - i -- Double Ended GJillocine Break in Pump Discharge Leg j ( l
\ 6.2.1-38 Cont &dnment Physical Parameters N 6.2.4-1 / containment Isclation System f
6.3.2-1 SafrRy Injection'dystem Component Parameters
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) (.- 6.3.2-2 Safe *Q Injection SIEter.:tFMEA:
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, u y , C . '3 . 2 - 3 f. ( 'Safec;r Injection Pump NPS.'i Requ3.rerants iI (-
6.31.]g/ \ t Saf tif In-Jeccion System Head Loss Mcqd raments
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6.3.3.F 1 T,ime Sequence of Important Events for u Spectrum
'i , s c4 Large LCCAs (Seconds After Break) 6.3,3.2-'$ General System arameters 'ond Daitia . i ,.1ditions
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- 6.3.3.2 3 5 Large Greak Spect mr.
< , t 6 . 3 . 3 . 'M . ( Pekk Citd Temperatures and ' Oxidation Percentage
' f e.t bbe Large BreQ Spectrum 3, '
I / / . ( l6.3.3.2-5 Variables PlotteQ as 'u Function of Time For Each rfa!:qe Break i,n the Spectrum ! 'i \ 6.3.3.2-6 ' Addr.tiona3, Varit.b:es; Plot./dd as a Function of Time ' cFor P.he Wors4. Ltp:.7)i ireah ' x / > 6.3.3.3-1 i
.hf7ty Injecticn Purha hbinur Delivered Flev to RCS ( Assuming Oria Em&rgenc's Ge.nerator Failed)
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'Geceral System Parabetern' s 6.3)3.3-3 mall Brs.n Spectrum ,
<'i 6.3.3.3-4 q Variables 111 cited e1 a Function of Time for Each
" x 1.Tr3e Dre:n'4 in'i the Spectrum l, i, [, ,
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p G ' LI81' QF , TABLES (Cont G) 'l CidPTEI! 6 Section Subiect s 6.3.3.3-5 Fu Al' Rod Perf c t bance ) Summary
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6.3.3.3-6 Times of Intersht for' Small Bretaks (Seconds) / 6.3.3.5-1 Sequen :e of. Events . for Representative Large and Small Break LOCAs , 6.3.3.5-2 Disposition of Normally Operating i Systems for + Large and Small Break LOCA knalyses ei . 3. 3. 5-3 Utilization of Safety Systems for Representative small' Break (0.02 ft'y 6.3.3.5-4 Utilization of Safety Systems for }Representativ e Large Break (0.G DEG/PD) , 6.3.3.6-1 Parameters Used in the Radiolce/ical Consequences i of a LOCA - 6.3.3.7-1 Chapter 15 Limiting Events Which Actuate the Safety Injection bystem i 4 l l t
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e >; . y .: v fs CESSAR E!ninem:n __ v LIST OF FIGURES CHAPTER 6 Ficure S.Mt h M , 6.2.1-1 Normalized Decay Heat Curve 6.2.1-2 Safety Injection < Flow Rate vs Time 6.2.1-3 Safety Injection Flow Rate vs Time 6.2.1-4 Safety Injection Flow Rate vs Time 6.2.1-5 Safety ' Injection Flow Rate vs Time 6.2.1-6 Safety Injection Flow Rate vs Time 6.2.1-7 Safety Injection Flow Rate vs Time 6.2.1-8 Safety Injection Flow Rate vs Time 6.2.1-9 ifety Injection Flow Rate vs Time 6.2.1-10 Steam Line Model 6.2.1-11 102% Power - Slot MSLB, Loss of 1 CHRS, Feedwater Addition vs Time , 6.2.1-12 , 102% Powcr - Guillotine MSLB, Loss of 1 CHRS, l Feedwaterl Addition vs Time l 6.2.1-13 75% Power - Slot MSLB, Loss of 1 CHRS, Feedwater Addition Na Time l 6.2.1-14 75% Power - Guillotine MSLB, Loss of 1 CHRS, Feedwater Addition vs Time . 6.2.1-15 50% Power - Slot MSLB, Loss of 1 CHRS, Feedwater i Addition,vs Time 1 6.2.1-16 50% Power - Guillotine MSLB, Loss of 1 CHRS,
- Feedwater Addition vs Time 6.2.1-17 25% Power - Slot MSLB, Loss of 1 CHP.S , Feedwater t'
Addition vs Time 6.2.1-18 25% Power - Guillotine MSLB, Loss of 1 CHRS, Feedwater Addition vs Time i XV
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CESS AR CE.2lFICATE:N LIST OF FIGURES (Cont'd) CHAPTER 6 Ficure Subject 6.2.1-19 0% Power - Slot MSLB, Loss of 1 CHRS, Feedwater Addition vs Time 6.2.1-20 0% Power - Guillotine MSLB, Loss of 1 CHRS, Feedwater Addition vs Time 6.2.1-21 Main Steam System Nodal Model With Guillotine Break at SG Nozzle Terminal End (For Subcompartment Analysis) 6.2.1-22 Combined Spillage and Spray into Containment 6.2.1-23 Condensing Heat Transfer Coefficients for Static Heat Sinks 6.2.1-24 1.0 x DEG Break in Pump Discharge Leg, Minimum Containment Pressure for ECCS Performance 6.2.1-25 1.0 x DEG Break in Pump Discharge Leg, Containment Atmosphere and Temperature 6.2.1-26 1.0 x DEG Break in Pump Discharge Leg, Containment Sump Temperature 6.2.4-1A Containment Isolation Valve Arrangement 6.2.4.1B Containment Isolation Valve Arrangement 6.2.4.1C Containment Isolation Valve Arrangement 6.3.2-1A Safety Injection System Piping and Instrumentation Diagram 6.3.2-1B Safety Injection System Piping and Instrumentation Diagram C 6.3.2-1C Safety Injection System Flow Diagram, Short-Term < Mode 6.3.2-1D Safety Injection System Flow Diagram, Short-Term Mode 6.3.2-1E Safety Injection System Flow Diagram, Long-Term Cooling Mode Amendment C xvi June 30, 1988
CESSAR 8HMCAMN LIST OF FIGURES (Conted) CHAPTER 6 Section ER)iect 6.3.2-1F Safety Injection System Flow Diagram, Long-Term Cooling Mode 6.3.2-2 Engineered Safeguards System Schematic Diagram C 6.3.2-3 Safety Injection Pump Head and NPSH Curves (rypical) 6.3.3.2-1A 1.0 x Double Ended Slot Break in Pump Discharge Leg-Core Power 6.3.3.2-1B 1.0 x Double Ended Slot Break in Pump Discharge Leg-Pressure in Center Hot Assembly Node 6.3.3.2-lC 1.0 x Double Ended Slot Break in Pump Discharge Leg-Leak Flow 6.3.3.2-lD.1 1.0 x Double Ended Slot Break in Pump Discharge Leg-Flow in Hot Assembly - Path 16, Below Hot Spot 6.3.3.2-lD.2 1.0 x Double Ended Slot Break in Pump Discnarge Leg-Flow in Hot Assembly - Path 17, Above tiot Spot 6.3.3.2-lE 1.0 x Double Ended Slot Break in Pump Discharge Leg-Hot Assembly Quality 5.3.3.2-lF 1.0 x Double Ended Slot Break in Pump Discharge Leg-Centainment Pressure ( 6.3.3.2-lG 1.0 x Doubled Ended Slot Break in Pump Discharge Leg-Maas Added to Core During Reflood 6.3.3.2-lH 1.0 x Double Ended Slot Break in Pump Discharge Leg-Peak Clad Temperature 1 6.3.3.2-2A 0.8 x Double Ended Slot Break in Pump Discharge Leg-Core Power 6.3.3.2-2B 0.8 x Double Ended Slot Break in Pump Discharge i Leg-Pressure in Center Hot Assembly Node ( l 6.3.3.2-2C 0.8 x Double Ended Slot Break in Pump Discharge j Leg-Teak Flow l l t Amendment C xvii June 30, 1988 l
1 1 CESSAR 9!.%ncma ! l LIST OF FIGURES (Cont'd) CHAPTER 6 Section Subieot 6.3.3.2-2D.1 0.8 x Double Ended Slot Break in Pump Discharge Leg-Flow in Hot Assembly - Path 16, Below Hot Spot 6.3.3.2-2D.2 0.8 x Double Ended Slot Break in Pump Discharge Leg-Flow in Hot Assembly - Path 17, Above Hot Spot 6.3.3.2-2E 0.8 x Double Ended Slot Break in Pump Discharge Leg-Hot Assembly Quality 6.3.3.2-2F 0.8 x Double Ended Slot Break in Pump Discharge Leg-Containment Pressure 6.3.3.2-2G 0.8 x Double Ended Slot Break in Pump Discharge Leg-Mass Added. to Core During Reflood 6.3.3.2-2H 0.8 x Double Ended Slot Break in Pump Discharge Leg-Peak Clad Temperature 6.3.3.2-3A 0.6 x Double Ended Slot Break in Pump Discharge Leg-Core Power 6.3.3 2-3B 0.6 x Double Ended Slot Break in Pump Discharge Leg-Pressura in Center Hot Assembly Node 5.3.3.2-3C 0.6 x Double Ended Slot Break in Pump Discharge Leg-Leak Flow 6.3.3.2-3D.1 0.6 x Double Ended Slot Break in Pump Dischsrge Leg-Flow in Hot Assembly - Path 16, Below Hot Spot 6.3.3.2-3D.2 0.6 x Double Ended Slot Break in Pump Discharge ' Leg-Flow in Hot Assembly - Path 17, Above Hot Spot 6.3.3.2-3E 0.6 x Double Ended Slot Break in Pump Discharge Leg-Hot Assembly Quality 6.3.3.2-3F 0.6 x Double Ended Slot Break in Pump Discharge L,'-Containment Pressdre 6.3.3.2-3G 0.6 x Double Ended Slot Break in Pump Discharge _ , , m_.-- Leg-Mass Added to Core During Reflood . xvill
CESSAR nnincucu LIST OF TIGURES (Cont'd) CHAPTER 6 Section subiect 6.3.3.2-3H 0.6 x Double Ended Slot break in Pump Discharge Leg-Peak Clad Temperature 6.3.3.2-4A 0.5 Ft 2 Slot Break in Pump Discharge Leg-Core Power 6.3.3.2-4B 0.5 Ft Slot Break in Pump Discharge Leg-Pressure in Center Hot Assembly Node 2 6.3.3.2-4C 0.5 Ft Slot Break in Pump Discharge Leg-Leak Flow 2 6.3.3.2-4D.1 0.5 Ft Slot Break in Pump Discharge Leg-Flow in Hot Assembly - Path 16, Below Hot Spot 2 6.3.3.2-40.2 0.5 Ft Slot Break in Pump Discharge Leg-Flow in Hot Assembly Path 17, Above Hot Spot 6.3.3.2-4E 0.5 Ft Slot Break in Pump Discharge Leg-Hot Assembly Qua14 y 2 6.3.3.2-4F 0.5 Ft Slot Break in Pump Discharge Leg-Containment Pressure 2 6.3.3.2-4G 0.5 Ft Slot Break in Pump Discharge Lcg-Mass Added to Core During Reflood 2 6.3.3.2-4H 0.5 Ft Slot Break in Pump Discharge Leg-Poak Clad Temperature 6.3.3.2-5A 1.0 x Double Ended Guillotine Break in Pump Discharge LegCore Power 6.3.3.2-5B 1.0 x Double Ended Guillotine Break in Pump Discharge LegPressure in Center Hot Assembly Node 6.3.3.2-5C 1.0 x Double Ended Guillotine break in Pump l Discharge Leg-Leak Flow 6.3.3.2-SD.1 1.0 x Double Ended Guillotine Break in Pump Discharge Leg- Flow in Hot Assembly-Path 16, Below Hot Spot i l l xix l l l
CESSAR!Ence LIST OF FIGURES ( Co.",t ' J ) CHAPTER 6 8ection Subiect 6.3.3.2-SD.2 1.0 x Double Ended Guillotine Break in Pump Discharge Leg-Flow in Hot Assembly-Path 17, Above Hot Spot 6.3.3.2-5E 1.0 x Double Ended Guillotine Braak in Pump Discharge Leg-Hot Assembly Quality 6.3.3.2-5F 1.0 x Double Ended Guillotine Break in Pump Discharge LegContainment Pressure 6.3.3.2-5G 1.0 x Double Ended Guilletine Break in Pump Discharge Leg-Mass Added to Core During Reflood 6.3.3.2-5H 1.0 x Double Ended Guillotine Break in Pump Discharge Leg- Peak Clad Temperature G.3.3.2-5I 1.0 x Double Ended Guillotine Break in Pump Discharge Leg-Mid Annulus Flow 6.3.3.2-5J 1.0 x Double Ended Guillotine Break in Pump Discharge Leg-Qualities Above and Below the Core 6.3.3.2-5K 1.0 x Double Ended Guillotine Break in Pump Discharge Leg-Core Pressure Drop 6.3.3.2-5L 1.0 x Double Ended Guillotine Break in Pump Discharge Lea-Sa f ety Injection Flow Into Intact Discharge Leg 6.3.3.2-5M 1.0 x Double Ended Guillotine Break in Pump Discharge Leg-Water Level in Downcomer During Reflood i 6.3.3.2-5N 1.0 x Double Ended Guillotine Break in Pump Discharge Leg-Hot Spot Gap Conductance 6.3.3.2-50 1.0 x Double Ended Guillotine Break in Pump Discharge Leg-Local Clad Oxidation
- 6.3.3.2-SP 1.0 x Double Ended Guillotine Break in Pump l Discharge Leg-Clad, Centerline. Average Fuel and l
Coolant Temperature for Hottest Node l l XX l 1 l
s CESSAR E!!L"icui:n LIST OF FIGURES (Cont'd) CHAPTER 6 Section Subiect 6.3.3.2-5Q 1.0 x Double Ended Guillotine Break in Dump Discharge Leg-Hot Spot Heat Transfer Coefficient 6.3.3.2-5R 1.0 x Double Ended Guillotine Break in Pump Discharge Leg-Hot Rod Internal Gas Pressure 6.3.3.2-55 1.0 x Double Ended Guillotine Break in Pump Discharge Leg-Containment Temperature 6.3.3.2-5T 1.0 x Double Ended Guillotine Break in Pump Discharge Log-Sump Temperature 6.3.3.2-5U l.0 x Double Ended Guillotine Break in Pump Discharge Leg-Core Bulk Channel Flow Rate 6.3.3.2-6A 0.8 x Double Ended Guillotine Break in Pump Discharge Leg-Core Power 6.3.3.2-6B 0.8 x Double Ended Guillotine Break in Pump Discharge Leg-Pressure in Center Hot Assembly Node 6.3.3.2-6C 0.8 x Double Ended Guillotine Break in Pump Discharge Leg-Leak Flow 6.3.3.2-6D.1 0.8 x Double Ended Guillotine Break in Pump Discharge Leg-Flow in Hot Assembly-Path 16, Below Hot Spot 6.3.3.2-SD.2 0.8 x Double Ended Guilloti.ie Break in Pump Discharge Leg-Flow in Hot Assembly-Path 17, Above Hot Spot 6.3.3.2-6E 0.8 x Double Ended Guillotine Break in Pump Discharge Leg-Hot Assembly Quality 6.3.3.2-6F 0.8 x Double Ended Guillotine Break in Pump Discharge Leg-Containment Pressure 6.3.3.2-6G 0.8 x Double Ended Guillotine Break in Pump Discharge Leg-Mass Added to Core During Reflood xxi
CESSAR EPGncwou LIST OF FIGURES (Cont'd) CHAPTER 6 Section Subiect 6.3.3.2-6H 0.8 x Dcuble Ended Guillotine Break in Pump Discharge Leg-Peak Clad Temperature 6.3.3.2-6I 0.8 x Double Ended Guillotine Break in Pump Discharge Leg-Mid Annulus Flow 6.3.3.2-6J 0.8 x Double Ended Guillotine Break in Pump Discharge Leg-Qualities Above and Below the Core 6.3.3.2-6K 0.8 x Double Ended Guillotine Break in Pump Discharge Leg-Core Pressure Drop 6.3.3.2-6L 0.8 x Double Ended Guillotine Break in Pump Discharge Leg-Safety Injection Flow into Intact Discharge Leg 6.3.3.2-6M 0.8 x Doucle Ended Guillotine Break in Pump Discharge Leg-Water Level in Down Comer During Reflood 6.3.3.2-6N 0.8 x Double Ended Guillotine Break in Pump Discharge Leg-Hot Spot Gap Conductance 6.3.3.2-60 0.8 x Double Ended Guillotine Break in Pump Discharge Leg-Local Clad Oxidation 6.3.3.2-6P 0.8 x Double Ended Guillotine Break in Pump Discharge Leg-Clad, Centerline, Avg Fuel and Coolant Temperature for Hottest Node 6.3.3.2-6Q 0.8 x Double Ended Guillotine Break in Pump Discharge Leg-Hot Spot Heat Transfer Coefficient 6.3.3.2-6R 0.8 x Double Ended Guillotine Break in Pump Discharge Leg-Hot Rod Internal Gas Pressure 6.3.3.2-6S 0.8 x Double Ended Guillotine Break in Pump Discharge Leg-Core Bulk Channel Flow Rate 6.3.3.2-7A 0.6 x Dot?ble Ended Guillotine Break in Pump Discharge Leg-Core Power xxii
C E S S A R E!nincu a LIST OF FIGURES (Cont'd) CHAPTER 6 Section Subiect 6.3.3.2-7B 0.6 x Double Ended Guillotine Break in Pump Discharge Leg-Pressure in Center Hot Assembly Node 6.3.3.2-7C 0.6 x Double Ended Guillotine Break in Pump Discharge Leg-Leak Flow 6.3.3.2-7D.1 0.6 x Double Ended Guillotine Break in Pump Discharge Leg-Flow in Hot Assembly-Path 16, Below Hot Spot 6.3.3.2-7D.2 0.6 x Double Ended Guillotine Break in Pump Discharge Leg-Flow in Hot Assembly-Path 17, Above Hot Spot 6.3.3.2-7E 0.6 x Double Ended Guillotine Break in Pump Discharge Leg-Hot Assembly Quality 6.3.3.2-7F 0.6 x Double Ended Guillotine Breuk in Pump Discharge Leg-Containment Pressure 6.3.3.2-7G 0.6 x Double Ended Guillotine Break in Pump Discharge Leg-Mass Added to Core During Reflood 6.3.3.2-7H 0.6 x Double Ended Guillotine Break in Pump Discharge Leg-Peak Clad Temperature 6.3.3.2-8A 1.0 x Double Ended Guillotine Break in Pump Suction Leg-Core Power 6.3.3.2-8B 1.0 x Double Ended Guillotine Break in Pump Suction Leg-Pressure in Center Hot Assembly Node 6.3.3.2-8C 1.0 x Double Ended Guillotine Break in Pump Suction Leg-Leak Flow 6.3.3.2-8D.1 1.0 x Double Ended Guillotine Break in Pump Suction Lag-Flow in Hot Assembly-Path 16, Below Hot Spot 6.3.3.2-8D.2 1.0 x Double Ended Guillotine Break in Pump Suction Leg-F'ow in Hot Assembly-Path 17, Above Hot Spot 6.3.3.2-8E 1.0 x Double Ended Guillotine Break in Pump Suction Leg-Hot Assembly Quality xxiii
CESSAR anOICATION LIST OF FIGURES (Cont'd) CHAPTER 6 section subieot 6.3.3.2-8F 1.0 x Double Ended Guillotine Break in Pump Suction Leg-Containment Pressurc 6.3.3.2-8G 1.0 x Double Ended Guillotine Break in Pump Sucticn Leg-Mass Added to Core During Reflood 6.3.3.2-8H 1.0 x Double Ended Guillotine Break in Pump Suction Leg-Peak Clad Temperature 6.3.3.2-9A 1.0 x Double Ended Guillotine Break in Hot Leg-Core Power 6.3.3 2-9B 1.0 x Double Ended Guillotine Break in Hot Leg-Pressure in Center Hot Assembly Node 6.3.3.2-9C 1.0 x Double Ended Guillotine Break .i n Hot Leg-Leak Flow 6.3.3.2-9D.1 1.0 x Double Ended Guillotine Break in Hot Leg-Flow in Hot Assembly - Path 16, Below Hot Spot 6.3.3.2-9D.2 1.0 x Double Ended Guillotine Break in Hot Leg-Flow in Hot Assembly-Path 17, Above Hot Spot 6.3.3.2-9E 1.0 x Double Ended Guillotine Break in Hot Leg-Hot Assembly Quality 6.3.3.2-9F 1.0 x Double Ended Guillotine Break in Hot Leg-Containment Pressure 6.3.3.2-9G 1.0 x Double Ended Guillotine Break in Hot Leg-Mass Added to Core During Reflood - 6.3.3.2-9H 1.0 x Double Ended Guillotine Break in Hot l Leg-Peak Clad Temperature 6.3.3.2-10 Peak Clad Temperature vs Break Area I l 6.3.3.2-11 1.0 x DE Guillotine Break in Pump Discharge l Leg-Peak Clad Temperature and Peak Local Oxidation vs Rod Average Burnup xxiv
~ CESSAR EnMem,. LIST OF FIGURES (Cont'd) CHAPTER 6 Section Cubioq] 6.3.3.3-1A 0.5 Ft.2 Cold Leg Break at Pump Discharge-Normalized Total Core Power 6.3.3.3-1B 0.5 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Pressure 6.3.3.3-lc 0.5 Ft.2 Cold Leg Break at Pump Discharge-Break Flow Rate 6.3.3.3-lD 0.5 Ft. Cold Leg Break at Pump Discharge-Inner Vessel Inlet Flow Rate 6.3.3.3-lE 0.5 Ft. Cold Leg Break at Pump Discharge-Inner Vessel Two-Phase Mixture Volume 6.3.3.3-lF 0.5 Ft.2 Cold Leg Break at Pump Discharge-Heat Transfer Coefficient at Hot Spot 6.3.3.3-1G 0.5 Ft. Cold Leg Break at Pump Discharge"Coolant Temperature at Hot Spot 6.3.3.3-lH 0.5 Ft.2 Cold Leg Break at Pump Discharge-Hot Spot Clad Surface Temperature 6.3.3.3-2A 0.35 Ft. Cold Leg Break at Pump Discharge-Normalized Total Core Power 6.3.3.3-2B 0.35 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Pressure 6.3.3.3-2C 0.35 Ft.2 Cold Leg Break at Pump Discharge-Break Flow Rate 6.3.3.3-2D 0.35 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Inlet Flow Rate 6.3.3.3-2E 0.35 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Two-Phase Mixture Volume 6.3.3.3-2F 0.35 Ft. Cold Leg Break at Pump Discharge-Heat Transfer Coefficient at Hot Spot 6.3.3.3-2G 0.35 Ft. Cold Leg Break at Pump Dischargo-Coolant Temperature at Hot Spot XXV
~
CESSAR U!ecm:,. LIST OF FIGURES (Cont'd) CHAPTER 6 section subiect 6.3.3.3-2H 0.35 Ft. Cold Leg Break at Pump Discharge-Hot Spot Clad Surface Temperature 6.3.3.3-3A 0.2 Ft.2 Cold Leg Break at Pump Discharge-Normalized Total Core Power 6.3.3.3-3B 0.2 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Pressure 6.3.3.3-3C 0.2 Ft.2 Cold Leg Break at Pump Discharge-Break Flow Rate 6.3.3.3-3D 0.2 Ft.2 Cold Leg areak at Pump Discharge-Inner Vessel Inlet Flow Rate 6.3.3.3-3E 0.2 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Two-Phase Mixture Volume 6.3.3.3-3F 0.2 Ft.2 Cold Leg Break at Pump Discharge-Heat i Transfor Coefficient at Hot Spot
- 6.3.3.3-3G 0.2 Ft.2 Cold Lag Break at Pump Discharge-Coolant Temperature at Hot Spot 6.3.3.3-3H 0.2 Ft. Cold Leg Break at Pump Discharge-Hot Spot l Clad Surface Temperature 6.3.3.3-4A 0.05 Ft. Cold Leg Break at Pump Discharge-Normalized Total Core Power 6.3.3.3-4B 0.05 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Pressure 6.3.3.3-4C 0.05 Ft. Cold Leg Break at Pump Discharge-Break Flow Rate l 6.3.3.3-4D 0.05 Ft. Cold Lag Break at Pump Discharge-Inner l Vessel Inlet Flow Rate 6.3.3.3-4E 0.05 Ft. Cold Leg Break at Pump Discharge-Inner Vessel TwoPhase Mixture Volume l
l l xxvi l l
CESSAR !!L"icucu LIST OF FIGURES (Cont'd) CHAPTER 6 Section Subieot 6.3.3.3-4F 0.05 Ft. Cold Leg Break at Pump Discharge-Heat Transfer Coefficient at Hot Spot 6.3.3.3-4G 0.05 Ft.2 Cold Leg Break at Pump Discharge-Coolant Temperature at Hot Spot 6.3.3.3-4H 0.05 Ft.2 Cold Leg Break at Pump Discharge-Hot Spot Clad Surface Temperature 6.3.3.3-5A 0.02 Ft. Cold Leg Break at Pump Discharge-Normalized Total Core Power 6.3.3.3-5b 0.02 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Precsure 6.3.3.3-5C 0.02 Ft.2 Cold Leg Break at Pump Discharge-Break Flow Rate 6.3.3.3-5D 0.02 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Inlet Flow Rate 6.3.3.3-5E 0.02 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Two-Phase Mixture Volume 6.3.3.3-5F 0.02 Ft. Cold Leg Break at Pump Discharge-Heat Transfer Coefficient at Hot Spot 6.3.3.3-5G 0.02 Ft.2 Cold Leg Break at Pump Discharge-Coolant Temperature at Hot Spot 6.3.3.3-5H 0.02 Ft.2 Cold Leg Break at Pump Discharge-Hot ' Spot Clad Surface Temperature . 6.3.3.3-6A 0.03 Ft.2 Break at Top of Pressurizer-Normalized Total Core Power 6.3.3.3-6B 0.03 Ft. Break at Top of Pressurizer-Inner Vessel Pressure 6.3.3.3-6C 0.03 Ft. Break at Top of Pressurizer-Break Flow l Rate 6.3.3.3-6D 0.03 Ft. Break at Top of Pressurizer-Inner Vessel Inlet Flow Rate xxvii
CESSAR%nMcus LIST OF FIGURES (Cont'd) CHAPTER 6 Scotion Subiect 6.3.3.3-6E 0.03 Ft. Break at Top of Pressurizer-Inner Vessel TwoPhase Mixture Volume 6.3.3.3-6F 0.03 Ft.2 Break at Top of Pressurizer-Heat Transfer Coefficient at Hot Spot i 6.3.3.3-6G 0.03 Ft.2 Break at Top of Pressurizer-Coolant Temperature at Hot Spot 6.3.3.3-6H 0.03 Ft.2 Break at Top of Pressurizer-Hot Spot Clad Surface Temperature 6.3.3.3-7 Maximum Hot Spot Clad Temperature vs Break Size 6.3.3.4-1 Long Term Cooling Plan 1 2 6.3.3.4-2 Core Flush by Hot Side Injection For 9.8 Ft Cold Leg Break 6.3.3.4-3 Inner Vessel Boric Acid Concentration vs Time 6.3.3.4-4 RCS Refill Time Versus Break Area 6.3.3.4-5 Overlap of Acceptable LTC Modes In Terms of Cold Leg Break Size 6.3.3.4-6 RCS Pressure After Rafill vs Break Area 6.3.3.5-1A Sequence of Events Diagram for Large and Small Break LOCAs 6.3.3.5-1B Sequence of Events Diagram for Large and Small Break LOCAs 6.3.3.5-lc Sequence of Events Diagram for Large and Small Break LOCAs 6.3.3.5-1D Sequence of Events Diagram for Large and Small Break I4CAs xxviii
CESSAR Uninemen LIST OF FIGURES (Cont'd) CHAPTER 6 Section Subiect 6.3.3.5-lE Sequence of Events Diagram for Large and Small Break LOCAs 6.3.3.5-lF Sequence of Events Diagram for Large and Small Break LOCAs XXiX
CESSAR UMema 6.3 SAFETY IRTECTION SYSTEM 6.3.1 DESIGN BASES 6.3.1.1 Spmmary Descriotion The Safety Injection System (SIS) is designed to provide core cooling in the unlikely event of a Loss-of-Coolant Accident (LOCA). The SIS limits fuel damage so as to maintain a coolable core geometry, limits the cladding metal-water reaction, removes the energy generated in the core and maintains the core suberitical during the extended period of time following a LOCA. More specifically, the SIS shall assure that the criteria of 10CFR50.46 are met. The SIS accomplishes these functional requirements by usa of redundant active and passive injection subsystems. The active portion of the SIS consists of four identical Safety Injection (SI) Pumps and associated valves. The passive portion consists of four identical pressurized Safety Injection Tanks (SITS). The SI Pumps can be utilized to achieve safe shutdown by C providing make-up for volume contraction or lost fluid and by providing suf ficient boron to the reactor vessel to achieve and maintain necessary shutdown margin. The SIS is alsci capable of injecting borated water into the reactor vessel to mitigt.te accidents. Safety injection would be initiated in the event of a Steam Generator Tube Rupture, Steam Line Break or a CEA Ejection incident. The system is actuated automatically. The SI Pumps are also used for bleed-and-feed operations in conjunction with the Safety Depressurization System. 6.3.1.2 Criteria 6.3.1.2.1 Functional Design Bases A. The shutoff head and flow rates of the SI Pumps were l C selected to insure that adequate flow is delivered to the reactor vessel to accomplish the functional requirements of l C Section 6.3.1.1. B. Storage of fluid for the SIS is accomplished by the In-containment Refueling Water Storage Tank (IRWST) which l C contains a sufficient amount of borated fluid to accomplish the functional requirements of Section 6.3.1.1. Amendment C 6.3-1 June 30, 1988
CESSAREnecua C. The SIS is designed for Direct Vessel Injection (DVI). The discharge of each SI Pump and Tank is piped directly to a reactor vessel nozzle where the flow is directed into the reactor vessel downcomer region. C D. The SIS is designed so that the SIS pumps can be tested at full-flow conditions with the reactor operating at power. 6.3.1.2.2 Reliability Design Bases A. The safety function defined in Section 6.3.1.1 can be accomplished assuming the failure of a single active component during the short-term mode of operation or a singic active or limited leakage passive failure of a C component during long-term (>24 hr.) post-accident operation. For failure analysis, all necessary supporting systems, including the onsite electrical power system, are considered as part of the safety Injection System. A Failure Modes and Effects Analysis (FMEA) is presented in Table 6.3.2-2. B. Components of the SIS and instrumentation which must operate following a LOCA are designed to operate in the environment of Section 3.11. b C. The SIS is designed to Seismic Category I requirements. 6.3.1.3 Interface Recuirements Below are the detailed interface requirements that the SIS places on certain aspects of the Balance of Plant ( BOP) , listed by categories. In addition, applicable GDC and Regulatory Guides, which C-E utilizes in its design of the SIS, are presented. These GDC and Regulatory Guides are listed only to show what C-E considers to be relevant, and are not imposed as interface requirements, unless specifically called out as such in a particular interface requirement. l l Relevant GDC - 1, 2, 3, 4, 13, 14, 17, 18, 20, 21, 22, 23, 30, C 31, 32, 34, 35, 36, 37, 54, 56, 57 t Relevant Reg. Guides - 1.1, 1.26, 1.28, 1.29, 1.31, 1.36, 1.38, l 1.44, 1.45, 1.47, 1.53, 1.64, 1.68, l 1.68.2, 1.75, 1.79, 1.82, 1.84, 1.89, lC
! 1.97, 1.148 l
A. Power I
- 1. Elec*rical power requirements for the motor-operated C valves in the SIS are contained in Table 8.3.1-1.
Amendment C ( 6.3-2 June 30, 1988 l
CESSARinecunn e
- 2. The electrical supplies for the SIS pumps, valves and C '
instruments shall be as follows:
- a. The SI Pumps and valves shall be capable of being powered from the plant turbine generator (onsite power source), and/or plant startup power source (offsite power), and the emergency generators (emergency power). Fower connections shall be through a minimum of two independent buses so that in the event of a LOCA in conjunction with a single failure in the electrical supply, the flow from at least two safety injection pumps sball be l C available for core protection.
- b. An independent electrical bus of the above shall C supply two SI Punps and associated valves,
- c. Each emergency generator and the automatic sequencers necessary for generator loading shall be designed such that flow is delivered to the reactor vessel within a maximum of 40 seconds lC after an SIAS is generated. The emergency generator design requirements are described in Section 8.3.1. C
- d. The SIS hot leg injection valves shall be powered such that a single electrical failure cannot cause spurious initiation of hot leg injection flow through either hot leg injection line, nor shall a single electrical failure prevent initiation of l C hot leg injection flow through at least one of the hot leg injection lines,
- e. Air for all SIS pneumatic valve operators shall be clean, dry and oil-free.
- f. Provisions shall ce made to remove power from the SIT vent valves (SI-331, 335, 329, 333, 330, 334, C 328, 332) during plant operation. Provisions shall be made to allow restoration of power to these valves from the control room and a remote shutdown location. The two vent valves on each !C SIT shall be powered from separate and independent emergency power sources. Provisions shall be made to remove power from the motors on the SIT isolation valves whenever the RCS pressure exceeds .
( LAl'ER) psig. Amendment C 6.3-3 June 30, 1988 {
1 CESSARinece ! B. Protection from Natural Phenomena
- 1. Design provisions snali S incorporated such that SIS components are capacA= of functioning in the event of the maximum probable flood or other natural phenomenon defined in Criterion 2 of 10CFR$0.
C. Protection from Pipe Failure
- 1. The maximum expected leakage from a moderate energy pipe rupture postulated during normal plant conditions in tha SIS shall be as defined by the methods of Section 3.6. Isolation valves used to contain leakage shall be protected from the adverse effects cf a high or moderate energy pipe rupture which might preclude their operation when required.
- 2. No limited leakage passive failure or the effects thereof (such as flooding, spray impingement, steam, temperature, pressure, radiation, loss of NPSH, or loss of recirculation water inventory), in the SIS during the long term post-accident mode shall preclude the C
availability of minimum acceptable safety injection capability (minimum acceptable capability is defined as that which is provided by the operation of one subsystem).
- 3. The SIS shall be protected from the effects of pipe rupture.
- 4. The SIS shall be protected from the effects of pipe whip.
D. Missiles
- 1. The SIS shall be protected from missiles. C E. Separation
- 1. Adequate physical separation shall be maintained between the redundant piping paths and containment penetrations of the SIS such that the SIS will meet its functional requirements even with the failure of a single active component during t' short-term injection mode, or with a single act; e lC failure or a limited leakage passive failure during the long-term, post-accident mode. lC Amendment C 6.3-4 June 30, 1988
' l CESSAR !!!Mem i I
- 2. The cabling which is associated with redundant channels of vital Class 1E circuits for the SIS shall be physically separated to preserve redundancy and prevent a single event from causing multiple channel malfunctions or interactions between cM nnels.
Associated circuit cabling from redundant channels shall either be separated, provided with isolation devices, or analyzed and/or tested to demonstrate that no credible single failure could adversely affect redundant channels of Class 1E circuits.
- 3. In the routing of SIS Class lE circuits and location of equipment served by these Class 1E circuits, consideration shall be given' to their exposure to potential hazards such as postulated ruptures of piping, flammable material, flooding, and non-flame retardant wiring. Adequate separation or protective measures shall be provided.
- 4. Failures of non-safety grade systems shall not compromise redundancy of the SIS.
F. Independence
- 1. The environmental control system provided for each independent SIS train shall be powered by the same C emergency power associated with that train.
- 2. Power connections for SIS components shall be from a minimum of two independent electrical buses.
See A.2 above.
- 3. Two independent vital instrument power sources shall be provided for the SIS instrumentation.
See A.5 above. G. Thermal Limitations Not Applicable H. Mcnitoring
- 1. Provisions shall be made for the detection, containment, and isolation of the maximum expected leakage from a moderate energy pipe rupture, as discussed in C.1 above.
l Amendment C 6.3-5 June 30, 1988
CESSAR 85nCATl3N
- 2. Process instrumentation shall be available to the operator in the control room to assist in assessing post-LOCA conditions. The type of instrument, parameter measured and instrument range are listed in Table 7.5-2. C I. Operational Controls Not Applicable J. Inspection and Testing
- 1. Inspection and testing requirements for the SIS are contained in Section 16.4.5. lC
- 2. Prior to initial plant startup, SIS flow tests shall be performed. An adequate supply of and the necessary test connections at the water IRWST lC shall be provided.
K. Chemistry / Sampling
- 1. The Sampling System shall provide a means of obtaining remote liquid samples from the SIS for chemical and radiochemical laboratory analysis.
- 2. The sample lines in contact with the reactor coolant shall be austenitic stainless steel or equivalent material compatible with the fluid chemistry.
- 3. Post-accident sampling capabilities shall be provided C for the SIS.
L. Materials
- 1. SIS piping and fittings shall be seismic category I.
- 2. Design and fabrication of the SIS piping and fittings shall conform to ASME Boiler and Pressure Vessel Code (B&PV) Section III, Class 1 or 2.
- 3. Pipes and all parts in contact with the system fluid must be of austenitic stainless steel. Valve packings, gaskets, and valve diaphragm materials shall also be compatible with the chemistry of the water and the radioactive dose at that location.
Amendment C 6.3-6 June 30, 1988
CESSAR 8Hece
- 4. Care shall be taken to prevent sensitization and to control the delta ferrite content of (1) the welds which join any system fabricated of austenitic stainless steel to the SIS, and (2) the field welds on the SIS.
- 5. Controls shall be exercised to assure that contaminants do not significantly contribute to stress corrosion of stainless steel.
- 6. Materials used for the containment and its internal structures shall withstand exposure to all post-accident conditions without causing deleterious or undesirable reactions, or significantly altering existing long-term operating water chemistry. the lC
- 7. The Containment Spray System is interconnected to the RCS during shutdown or during recirculation of the C IRWST water and, therefore, materials used in this system shall be austenitic stainless steel or other compatible material and shall conform to the standards of Section III Class 2, ASME B&PV Code and applicable Code cases.
M. System / Component Arrangement
- 1. To assure that the Engineered Safety Features Systems flow requirements are met, the maximum and minimum acceptable head losses for the piping and fittings along with the required NPSH are as presented in Tables 6.3.2-3 and 6.3.2-4.
C
- 2. The SI Pumps shall be located in the auxiliary building as close as practicable to the containment structure,
- a. The elevation of these pumps shall be low enough such that adequate NPSH is available when the pumps take suction from the IRWST. .
- b. The available NPSH shall be calculated at the pump suction. C
- c. The calculation shall consider concurrent safety injection and containment spray pump operation. Tables 6.3.2-3 and 6. 3. 2-4 provide SI Pump NPSH and head less requirements. The NPSH requirements listed include a 10% margin above that required for proper pump operation.
Amendment C 6.3-7 June 30, 1988
CESSAREEnew,.
- d. Credit shall not be taken for water that could be trapped in volumes which do not drain to the IRWST. C
- 3. SIS components shall be properly supported such that pipe stresses and support reactions are within allowable limits. C-E will provide to each Applicant the design loads at the support / structure interface locations for components that C-E supplies.
- 4. The loadings imposed by the SIS piping on the reactor C vessel nozzles, or by the connecting system piping on SIS nozzles, shall be less than the design loads for these nozzlas. C-E will provide to each Applicant the design loads for all nozzles on those SIS components tnat C-E supplies.
- 5. In the event of a limited leakage passive failure in one SIS train during the long term cooling mode, personnel access to the other intact trains will not be affected. C
- 6. The two SIS check valves in each of the six safety injection lines shall be located as follows: one as close as practicable to the DVI nozzles (4 DVI lines) C and the RCS piping (2 hot leg injection lines), the other as close as practicable to the containment penetration.
- a. Allowance shall be made for valve accessibility and maintenance.
C
- b. The check valve leakage lines shall be connected to the safety injection line immediately upstream of the safety injection check valve closest to the RCS piping. This is necessary to ensure that all of the safety injection piping is borated at all times.
appropriately lC
- 7. Each SIT shall be located inside the containment, outside the biological shield, and as close as possible to the reactor DVI nozzle into which it injects. lC l
- a. The piping run from each tank to the reactor C vessel nozzle shall be as direct as possible with a minimum of bends and elbows.
- b. Long radius ellows or pipe bends shall be used.
C Amendment C 6.3-8 June 30, 1988
CESSAR MMcua
- c. The bottom of the SIT shall be located above the C centerline of the reactor vessel DVI nozzle.
- 8. Manually-operated valves shall be provided with locking provisions as shown in Figures 6.3.2-1A and 6.3.2-13.
- 9. Physical identification of safety related SIS equipment and cabling shall be provided to allow recognition of safety. status by plant personnel.
- 10. In the routing of SIS Class 1E circuits and location of equipment served by these Class 1E circuits, consideration shall be given to their exposure to potential hazards. See E.3 above.
- 11. All SIS ASME B&PV Code Section III components shall be arranged to provide adequate clearances to permit inservice inspection.
- 12. Protection shall be provided from internally generated flooding that could prevent performance of safety related functions.
N. Radioactive Waste
- 1. SIS leakage to the safeguards room will normally drain to the room sump. Provisions shall be provided to accept the maximum leakage rates listed below:
- a. Safety Injection Pump seals: 1000 cc/hr C
- b. Valves backseat leakage: 10 cc/hr/ inch seat diameter across the valve seat: 10 cc/hr/ inch of nomine.L valve size All leakages shall be treated as radioactive waste with a low dissolved solids and organic content.
O. Overpressure Protection Not Applicable Amendment C 6.3-9 June 30, 1988
CESSAR nuincua P. Related Services
- 1. Nitrogen gas shall be supplied to each SIT. This supply shall satisfy the following requirements:
- a. Minimum Required Flow Rate 300 SCFM (at supply pressure)
- b. Maximum Allowable Flow Rate 2490 SCFM (at supply pressure)
- c. Minimum Supply Pressure 630 psig (for normal plant operations)
- d. Maximum Supply Pressure 700 psig (all conditions)
- e. Gas Volume Required for 105,000 SCF 4 Tank Dlowdown
- f. Design Criteria ANSI B31.1
- g. Maximum Water Content 0.1 percent
- h. Minimum Supply Stream 80*F Stagnation Temperature
- 1. Maximum Supply Free Stream 115'F Temperature
- j. No single failure shall allow the compressed nitrogen system delivery pressure to exceed 700 psig.
- 2. An IRWST shall be provided. Baffles and intake screens shall be installed to limit the maximum particle sizo C entering the IRWST to 0.09 inches in diameter in order to prevent flow blockage in the SIS components and -
piping and in the reactor.
- 3. A fire protection system shall be provided to protect the SIS consistent with the requirements of GDC 3, and shall includ:, as a min."'n, the following features:
- a. Facilities for fire detection and alarming.
- b. Facilities or methods to minimize the probability of fire and its associated effects.
Amendment C 6.3-10 June 30, 1988
K - "
\.) ,
CESSAR imincma , V
/ s e,3,
- p. \ g >
+
1
}.i
}i ) y/
- c. Facilities for fire extinguishment. ,,, s. ,
- d. Methods of fire prevention such as usb.' of .
fire resistant and non-combustible mate.rlais whenever practical, and minimizing expbsure of combustible materials to fire hazards. 1
- e. Assurance that fire protection systems d6 not adversely affect the functional and structural g integrity of safety related structures, synt.tms ,
and components. -
- f. Care should be exercised to ensure fi$N '
protection systems are designed to assurp that their rupture or inadvertent operation does not significantly impair the capability of and safety related structures, systems, components. j
\ )
- 4. The SIS containment penetrations shall not be subject to loss of function from dynamic effects (e.g., '
> f missiles, pipe reactions, fluid reaction forces) /
resulting from failure of equipment or piping insi.la,or , g outside the containment. l 5. Where required, bellows shall be provided between , l piping and the containment wall to prevent excessiva ', forces on the piping. Q. Environmental >
- 1. Each SIS safeguards train shall be provided with an independent environmental control system, which 'o. .
b' ( powered by the same emergency power associated wi t.5 that train, such that the safety related equipment in each train operates within the environmental design limits specified in Section 3.11. Af s,# l 6.3.2 SYSTEM DESIGN r System Schematio \ 6.3.2.1 The SIS Piping and Instrumentation Diagram (P&ID) is shown in , Fi gures 6. 3. 2-1A through 6. 3. 2-1F. The major components of this system are four identical Safety Injection Pumps, ,n C In-containment Refueling Water Storage Tank, four identical Safety Injec>. ion Tanks, and associated valves. The ma33r components are described in the following section. Figta e 6.3.2-2 shows the system orientation with respect to the containment, reactor, and RCS. ,
+
Amendment C 6.3-11 June 30, 1985 - I'
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A sum ry of design parameters a$/ codes for t mafx/womp h.its ) 't g 7 is N O n. 'In compogatsp used to Table 6.3.2-providej. Sectica 6.3. rcre protection cpr/iM4 0 ' the bon.plete ('
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N ') / 4.,3.2s2'.2, ,s 4 3afety, Injection Tmiks %1' , The fourMfetpi %/ ject 40n Tbr.ks MNs) discharge tft?.r s s# , , i contents f Onto theg RCS fer21ocirAJT hepressuri' etiv. as a resuic'M a LOCA. T
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/( 6 Eccb.trp 6/ Dup,?.gi.. oral plant pWation e d1' SIT is piped invo,b reactor >'vessel- safetyinjectio(;
is isola't.',dl rom nozzle. the r' Ii reacM e (M. two check yelves ;in sWas. The SITS btchtically k 'y e?ischige }into the reactor dovu:omer if RCS pres'/.hty giecrouen
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I h$ m tor-operated) isolation valves 3 on the SIT disNharde are interbcked wi'c4 tf..e pressurizer pressurCasasurement chary 1ela,? co open thesje valWe's automatically as RCS$ vressure is ';ln('uslased to ,; ' (LATER) psia, and to prevent inadvertent closureg prior to 'or g' durirq The valve is subject to statuV control and fpower sl,Tn ' theaccident. motor will be removed (see Section 7.6)'.' .( 4 ' f
,. i Dudt;.J normal poug ri speration, the valve d m.Tthcug h ach.d open, I redelves a confirBacMy 'SIAS "CTen" siti.ul. ' Durp0 startup and h ,
shutdown operations, prariable t tpoint.is ,used sq :tescribed in , L/ action 7. 2.1.1.1.tt. O ing plant cooldowns,. "JT prsssure will i t-, ' be lowered to 'O.bTER ) p~1g js ' by the opeM.t% when RCS pressure t m.ches J.LATER)'/ psfg. (l Anj {Winloc)r, pressure l 3
,) [ will pr,Went e the SIT ist%attor Vw frp yky\ytessur3zerkingclbseduntilRCS C pressure drops to (LAf ER)' '@Lig f , AltM. ugh the SIT isolation e, ,
,. valves will be closed by the op.m.or tf fche time RCS pressure is
- /C..
i/.g'jdown Izedvertent to (LATER) psig,/ an SIAS will huse the valves to open. repressurization of the SITS during this mode of l oparation, due to a Idaky nitrogen supply valve or by accide:r.tal tripping of a nitrogen sup,:ly valve switch, is prevented by
,Yaving two fail-closed valves in series with ; parate hand
/ qvitches on each SIT nitrogen suppTJ line; Ts air supply f &tuating the / nitrogen supply ' elvea is ccatrci'. lid by solenoid
/ vhlves. Thi t do nitrogen .Jupp1y valve stolenh % s on each SIT
, p's ) ; I at! a corinngl.aci to separy.te electreal bum s 713 Iedundant and
(' physic 11D Ivarated electrical tralris. Thi s is' ,<o Qnsurs that a f aulc in' <vid M t're trains.will not cause A spurious opening of j W,th \}tio ei sufply,nalvss > N / ,[ ,,) ' ~
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The operator will repressurize the SITS when pressurizer pressure ce . 's (I.ATER) psig. Failure to do so will result in an alarm C wns. _surizer pressuro reaches (LA*ER) psig.
/
/ water fractions, gas pressurr., and outlet pi'p The tan. g size /
are sela . to allow three of the four tanks to recover the core befora aificant fuel damage or zirconium-water reaction can occur f.ollowing a LOCA. The volume of water in the tanks is conse:vatively calculated assuming that all water injected iprior to the end of the RCS blowdovn is lost. The SITS contain borated water &* the minimum required boron concentration and are pressurized with nitrogen at a nominal C pressure of (LATER) psig. Redundant level and pressure , instrumentation (described in more detail in Section 6.3.5.3 and Table 7.5-2) are provided to monitor the condition of the tanks. Sufficient visual and audible indication are made available to th." operator such that maintaining the SITS within the required tuchnical specifications during various modes of plant operation is readily accouplished from'the controi room. Provisions have been made for sampling, filling, draining, and correcting boron concentration. Atmospheric vent valves are provided for tank venting. They are lockod closed and the power to each valve is removed during normal operation. This prevents inadvertent SIT venting during normal plant operation. SIT Data is summarized in Table 6.3.2-1. 6.3.2.2.3 Safety Injection Pumps The primary function cf the Safety Injection (SI) Pumps is to C inject borated water into the RCS if a break occurs in the Reactor Coolant Pressure Boundary (RCPB). For small breaks, the RCS pressurs. remains high for a long period nf time following the accident, cnd the SI Pumps ensure that the injected flow is sufficient to meet the criteria given in Section 6.3.1. The SI Pumps are also used for bleed-and-reod operation with the SDS and can be utilirad to achieve safe shutdown by providing makeup for volume contraction and ay providing sufficient boron to achieve and maintain necessary shutdown margins. For long-term core cooling. the SI Pumps are manually realigned for simultaneous het leg and reactor vessel inj ect i.on. This insures flushing and ultimate subcooling of the core independent of break location. For small breaks, the SI Pumps continue injecting into the reactor vessel downcomer to provide makeup for spillage out the break while a plant cooldown is implemented. i Amendment C C.3-13 June 30, 1988
f. CESSAR nE"icarisu _ a During normal operation the SI Pumps are isolated fron the reactor by motor-operated valves. During safety injection the SI Pumps deliver water from the IRWST to the reactor /essel downcomer via DVI nozzles whenever RCS pressure falls bele w pump shutoff head. During the long terre : node of operation, the SI pumps continue to take suction from the IRWST. The SI Pumps are sized such that for breaks, up to a double-ended guilotine break, two SI Pumps shall provide the required minimum C injection flow rate to the core to meet the performance criteria of Section 6.3.1. The SI Pumps are also sized such that, after consideration of spillage directly out through the break, one SI Pump will supply adequate water to the core to match decay heat boiloff rates soon enough to minimize core uncovery and allow small break LOCAs to meet the performance criteria of Section 6.3.1. A typical SI Pump characteristic curve is shown in Figure 6.3.2-3. The effectiveness of the SI Pump during a steam line break is also analyzed to assure that the pumps are adequately sized. Mechanical shaft seals are used and are provided with leakoffs which collect any leakage past the seals. The seals are designed for operation with a pumped fluid temperature of 350'F. The SI Pump motors are specified to have the capability of starting and accelerating the driven equipment, under load, to design point running speed within 5 seconds based on an initial voltage of 75% of the rated voltage at the motor terminals, increasing linearly with time to 90% voltage in the first 2 seconds, and increasing to 100% voltage in the next 2 seconds. The SI Pumps are provided with drain and flushing connections to permit reduction of radioa.:tive contamination before maintenance. The pressure containing parts of the pump are stainless steel with internals selected for compatibility with boric acid solutions. The materials selected are analyzed to ensure that differential expansion during design transients can be accommodated. ' The SI Pumps are provided with minimum flow protection (recirculation lines) to prevent damage resulting from operation against a closed discharge isolation va've. Also, individual SI Pump ultrasonic flow meters provide low flow alarms. , The design temperature of the SI Pump is based on the saturation temperature of the reactor coolant at the containment design pressure plus a design tolerance. The design pressure for the SI Pumps is based on the shutoff head plus maximum containment pressure plus a design tolerance. The pump data is provided in Table 6.3.2-1. Amendment C 6.3-14 June 30, 1988
CESSARimine-6.3.2.2.4 Piping Piping is designed to deliver borate safety injection water from the Safety Injection Tanks and from ,he In-containment Refueling Water Storage Tank via the Safety Injection Pumps, to the reactor vessel nozzles. The major piping sections are (refer to Figures 6.3.2-1A & 1B):
- a. From each SIT to its respective DVI nozzle;
- b. Redundant piping from the IRSWT to the SI Pumps; and,
- c. Redundant piping from each SI Pump discharge to the pump's respective DVI nozzle and one nozzle o1 each shutdown cooling suction line.
C The SIS piping is fabricated of austenitic stainless steel and is designed to ASME Code Section III. Flexibility and seismic loading analyses are performed by each Applicant to confirm the structural adequacy of the system piping. 6.3.2.2.5 Valves The relative location, type, type of operator, position (during the normal operating mode of the plant) and failure position of the SIS valves, is shown in Figures 6.3.2-1A and 6.3.2-lD. C A. Relief Valves Protection against overpressurization of components within the l SIS is provided by conservative design or the system piping, appropriate valving between high-pressure sources and low-pressure piping, and by relief valves. All lines from the ! RCS up to and including the outermost containment isolation C valves are designed for full RCS pressure. Relief valves will be orovided as required by applicable codes. All relief valves are totally enclosed and pressure tight with suitable provisions for gagging. i ig B. Actuator-Operated Throttling and Stop Valves ! The position of each valve on loss of actuating signal or power supply (failure position) is selected to ensure safe operation. System redundancy is considered when defining the failure position of any given valve. Valve position indication is ! provided at the main control panel es indicated in Figures 6.3.2-1A and 6.3.2-1B. A momentary push button with appropriate C status control on the main control panel and/or manual override Amendment C 6.3-15 June 30, 1988 1
I CESSAft nSine-I I handwheel is provided where necessary for efficient and safe plant operation. All actuator-operated valves have stem leakage controlled by a double packing with a lantern ring leakoff connection. The SIS will be adjusted to its design flow during C pre operational testing. C. Check Valves All check valves are the totally enclosed type. Check valves in pump suction lines are of a low pressure drop type with flow resistance characteristics equal to or less than a swing check valve of the same size as the connecting pipe. 6.3.2.3 Applicable Codes and Classification Refer to Section 6.3.2.2 and Table 6.3.2-1. 6.3.2.4 Materials EDecifications and ConDatibility The materials used in the construction of the SIS components are presented along with the component parametero in Table 6.3.2-1. Basically, all materials in contact with reactor coolant are austenitic stainless steel with stellite or equivalent material being used for valve seats. The materials of construction used in both the active and passive components have been evaluated and in each case it has been concluded that the materials selected are both compatible with the most severe environmental condition they will be exposed to and in accordance with all ASME Ccde requirements. 6.3.2.5 SysteL Reliability 6.3.2.5.1 Safety Injection Tanks The Safety Injection Tanko (SITS), containing borated water pressurized by a nitrogen cover, constitute a passive injection syatem because no operator rction or electrical signal is required for operation. reactor vessel nozzle forEachtankisconnectedtoitiassociatedlC DVI by a separate line containing two . check valves which isolate the tank from the RCS during normal I operation. When the RCS pressure falls below the tank pressure, l the check valves open discharging the contents of the tank the reactor vessel. into l C The performance evaluation in Section 6.3.3 demonstrates the adequacy of the quantity of coolant supplied. In order te prevent accidental overpressurization of the Shutdown Cooling System, SIT pressure is decreased to (IATER) psig when reactor l C Amendment C 6.3-16 June 30, 1988
CESSAR !!ninema coolant pressure is Lelow (LATER) psig and, subsequently, thel isolation valves on the tanks are closed. An interlock with pressurizer pressure prevents these valves from being cloced if pressurizer pressure is greater than (LATER) psig. In the l C unlikely event of a LOCA during shutdown cooling, an SIAS will automatically open the SIC isolation valves. Inadvertent repressurization of the SITS during shutdown cooling due to a leaky nitrogen supply valve or the accidental tripping of a valve switch is prevented by having two such fail-closed supply valves in series with separate hand switches. The air supply actuating the nitrogen supply v' ves is controlled by solenoid valves. The two nitrogen supply valve solenoids on each SIT are connected to separate electrical buses via redundant and physically separated electrical trains. This is to ensure that a fault in one of the trains will not cause a spurious opening of both nitrogen supply valves. The motor-operated isolation valves on the SIT discharge are interlocked witn pressurizer pressure to open the valves automatically as system pressure is increased to (LATER) psig. C When RCS pressuro increases to (LATER) psig, the operator will repressurize the SITS. Failure to do so will result in an alam at a pressurizer pressure of (LATER) psig. Further details of lC valve control are provided in Section 7.6. The atmospheric vents on the SIT are locked closed, fail closed and power to their solenoid valve is interrupted during operation 'C with the RCS pressure greater than (LATER) psig. This ensures l that the tank will not be vented during RCS power operation. 6.3.2.5.2 Safety Injection Subsystems The SIS consists of four independent identical safety injection C l discharge lines. The IRWST is the water source for all SI Pumps. Each safety injection line contains one SI Pump and associated . injection valves. Two SI Pumps and tha associated injection valves operate from one emergency power supply, the other two SI Pumps and injection valves from a second, independent, emergency power supply. This provides the automatic operation of one complete, full capacity subsystem in the unlikely event of concurrent loss of offsite power and the failure of an active component, including a standby generator. With the exception of check valves, all valves in the injection paths not receiving a JIAS signal are status controlled. l Amendment C l 6.3-17 June 30, 1988 l l
CESSAR naineme,.
)
I Prevention of flow blockage in small diameter pipes, including the above piping, specific weight in isaccomplishedbycontrolofparticlesizeandlC the injection water through IRWST design. 6.3.2.5.3 Power Sources Independent electrical buses supply power to the SIS equipment. Each bus may receive power from: A. Onsite power B. Offsite power C. Erargency power The safeguards initiation sensors, electrical controls, and electrical indication equipment normally receive power from four 120-volt AC buses. Four 125-volt station batteries with inverters are provided as a backup upon loss of all other sources of power. System reliability is achieved with the following: A. Two electrical buses, with each bus supplying er to two 100% capacity SI Pumps, associated valves anc. associated lC support systems. (Each support system contains two full capacity subsystems, one connected to each bus and one subsystem servicing each independent injection train). B. Two sources of power, normal and standby to both buses, with automatic backup from the emergency generators. C. Two emergency generators, each capable of supplying power for the ainimum safeguards loads. D. The system is designed such that a single electrical failure can neither spuriously initiate unnecessary in9ction flow, nor prevent initiation of required injection flow. A detailed description of the power sources will be given in the C site-specific SAR. 6.3.2.5.4 Capacity to Maintain Cooling Following a Single Failure The SIS is designed the failure of a single to meet its functional active componentrequirements even with l C during the short-term mode of operation or with the single active or limited leakage Amendment C 6.3-18 June 30, 1988
CESSAR E!!Macmu passive failure of a component during the long-tera cooling mode l C of operation. By providing proper redundancy of equipment, even with the single failure noted above, the minimum required SIS equipmen* is available. The SIS is designed utilizing a philosophy of total physical separation of redundant trains such that the system can carry out its safety function assuming a single active failure both during normal and chort-term post-accident modes and a single active or passive failure during long-term, post-accident modes (i.e., time periods >24 hr) after event initiation. C Total train separation assures that a single failure in one train cannot preclude the other trains from accomplishing their safety functions. A Failure Modes and Effects Analysis (FMEA) for the SIS is presented in Table 6.3.2-2. Minimum operability requirements for components of the SIS are as l C delineated in Section 16.3.5. requirements and system failure Consistent with these modes, the minimum SISoperability equipment l C that will operate during postulated accidents is as discussed in - Section 6.3.3. This complement of equipment is required to mitigate the consequences of a LOCA initiated when the reactor is anywhere. from hot shutdown to full power operation, and this complement will result in conservative results for incidents where the SIS is required. other l C t l The following design features are provided in the system in order l to meet the single failure criterion. l A. Redundant SI Pumps. l B. Redundant piping and valving between IRWST and SI Pump l section. C C. Redundant safety injection discharge piping. D. Four injection discharge points into the reactor vessel , downcomer and redundant injection discharge points into the l RCS hot legs. E. Separation of the redundant subsystems of the SIS. No limited leakage passive failure, as defined in Section 3.1.31, or the effects thereof (such as flooding, spray impingement, steam, temperature, pressure, radiation, loss of NPSH), will preclude the SIS from accomplishing C its safety functions. l C l Amendment C 6.3-19 June 30, 1988
CESSARanMc-6.3.2.6 Protection Provisions The SIS is provided with protection from damage that could result from a I4CA by: (a) designing components to withstand the Design Bases Event environment including coolant chemistry, radiation, temperature and pressure resulting from the accident, (b) a seismic design that will withstand the stress imposed by a safe Shutdown Earthquake occurring simultaneously with a LOCA, and (c) protection from missiles in accordance with Section 3.5. lC 6.3.2.6.1 Capability to Withstand Design Bases Environment Components located in the containment, such as remote-operated valvas and instrumentation and control equipment, required for initiation of cafety injection are designed to withstand the I4CA conditions of temperature, pressure, humidity, chemistry and radiation3.11. Section for theThe extended valves period ofthose include time required as detailed associated with fill,in lC drain, and pressure control of the Safety Injection Tanks which receive an SIAS or are required to operate following an accident. The instrumentation includes the wide range level and pressure instrumentation associated with the Safety Injection Tanks. Insofar as practical, SIS components required to maintain a functional status have been located outside containment to eliminate exposure of this equipment to the post-LOCA conditions. The equipment outside containment is designed in consideration of the chemical and radiation effects associated with operation fdlowing a LOCA. (Figures 6.3.2-1A and 6.3.2-1B indicate location of equipment inside or outside of containment). C The design life of the SI Pumps is 60 years, corresponding to the life of the plant. Design pressures and temperatures are in excess of the maximum pressures and temperatures seen by the respective component during the worst of normal operating and design bases conditions. Materials of construction for the pumps are compatible with the expected water chemistry under normal and LOCA conditions. A radiation resistance requirement has been ' placed on the pumps consistent with Section 3.11. ! 6.3.2.6.2 Missile Protection l Protection from possible RCS generated missiles is afforded by locating all components outside the containment except for the IRWST nnd SIT. The SITS are located outside the biological l C shield such that protection from possible Reactor Coolant System generated missiles is provided. l Amendment C 6.3-20 June 30, 1988
CESSAREmincma 6.3.2.6.3 Seissio Design Since operation of the SIS is essential following a Loss-of-Coolant Accident, it is considered Category I for seismic design. The general design basis for Category I equipment is that it must be able to withstand the appropriate seismic loads plus other applicable loads without loss of design functions which are required to protect the public. For the SIS, this means that the components must be able to withstand the stresses resulting from emergency operation following a LOCA, simultaneous with the stresses resulting from the Safe Shutdown Earthquake (SSE) without loss of function. Ref ar to Section 3.7 for details on seismic design and analysis methods. 6.3.2.7 Recuired Manual Actions The short-term injection mode of operation is automatically C initiated by a Safety Injection Actuation Signal (SIAS). Lcng-term core cooling is manually initiated at approximately 2 hours post-LOCA at which time the hot leg injection valves are opened to provide simultaneous hot leg and direct vessel lC injection, which results in a circulation flow through the core. For small pipe creaks, the SI Pumps provide makeup for spillage, lC while the RCS is cooled down and depressurized to shutdown cooling initiation conditions utilizing the steam generator Atmospheric Dump Valves and Emergency Feedwater System. For small LOCA's, the SITS must be vented to allow RCS depressur-ization. This is followad by manual shutdown cooling operation. Amendment C ! 6.3-21 June 30, 1988 '
CESSAR inuncam,. i l l THIS PAGE INTENTIONALLY BLANK
C E S S A R n Sine m a TABLE 6.3.2-1 (Sheet 1 of 3) SAFETY INJECTION SYCTEM COMPONENT PARAMETERS Safety Iniection Punos Quantity 4 Type Multistage, Horizontal, Centrifugal Safety Classification 2 Code ASME III, Class 2 Design Pressure 2050 psig Maximum operating suction Pressure 100 psig .,, Design Temperature 350*F Design Flow Rate - 815 gpm* Design Head 2850 ft Maximum Flow Rate 1130 gpm* Head at MaximtM Flow Rate 1580 ft Materials Stainless Steel, type i 304, 316 or approved alternate C Shaft Seal Mechanical Brake Horsepower 910 1 l
*Does not include by-pass flow l
1 l l l Amendment C l June 30, 1988 l t
CESSAR nnir,cy,2 TABLE 6.3.2-1 (Cont'd) (Sheet 2 of 3) SAFETY INJECTION SYSTEM COMPONENT PARAMETERS Safety Iniection Tanks Quantity 4 Safety Classification 2 Code ASME III, Class 2 Design Pressure, Internal / External 700 psig/100 psig Design Temperature 200*F Operating Temperature 140*F Normal Operating Pressure (LATER) psig Minimum Operating Pressure (LATER) psjg Volume, Total (LATER) ft Liquid 3 Minimum (LATER) ft 3 Nominal (LATER) ft 3 Maximum (LATER) ft Fluid Borated Water, (LATER) ppm Material Clad - Stainless Steel, type 304, 316, or approved alternate Body - Carbon Steel, type SA-516 Gr.' or approved alternate Amendment C June 30, 1988
CESSARanec- l l l l TABLE 6.3.2-1 (Cont'd) (Sheet 3 of 3) SAFETY INJECTION SYSTEM COMPONENT PARAMETERS In-containment Refuelina Water Storace Tank Quantity 1 Safety Classification 2 Code ASME III, Class 2 Design Pressure, Ir.ternal/ External Atmospheric Design Temperature 400*F Operating Temperature 70*F Normal Operating Pressure Atr ospheric Minimum Operating Pressure Atmospherig Volume, Total (LATER) ft C Liquid 3 Minimum (LATER) ft 3 Nominal (LATER) ft 3 Maximum (LATER) ft Fluid Borated Water, (LATER) ppm Material Austenitic Stainless Steel Lined Amendment C June 30, 1988
CESSAR n!Macm. Table 6.3.2-2 SAFETY INJECTION SYSTEM FME4 (LATER) C l l Amendment C June 30, 1988 L
CESSAR anGenian TABLE 6.3.2-3 DAFETY INJECTION PUMP NPSH REQUIREMENTS Flow / Pump NPSH Safety Iniection PuBD4 (CrDB) # feet) Long-Term Cooling Mode 1235 22(1) C NOTES: (1) Based on the properties of saturated water at 300'F. All pumps taking suction from the IRWST at runout flows. Amendment C June 30, 1988 1 T --- - w- +wyi
1 C E S S A R 8!ainca m u TABLE 6.3.2-4 SAFETY INJECTION SYSTEM HEAD LOSS REOUIREMENTS Flow / Pump Required System Safety Iniection PumDs (CDs) Resistance (ft) Long-Term Cooling Mode 1235 1580 (1) C NOTES: (1) Friction and elevation losses between the water level in the IRWST at the start of long-term cooling and the outlet of the reactor vessel nozzle. Two safety injection pumps operating. Amendment C June 3J, 1988
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CESSAR rPLficiri::= c
<33 et 1 c, 33 EFFECTIVE PAGE LISTING CHAPTER 10 Table of Contents Peqe Amendment i C 11 C iii C iv C v C Vi C Vii C Text Page AmendmeAt 10.1-1 A 10.1-2 A 10.2-1 A 10.3-1 A 10.3-2 A 10.3-3 A 10.3-4 A 10.3-5 A 10.3-6 A 10.3-7 A 10.3-8 A 10.3-9 A 10.3-10 A 10.3-11 A 19.3-12 A 10.3-13 A 10.3-14 A ,
10.3-15 10.3-16 A 10.3-17 A 10.3-18 A 10.4-1 A 10.4-2 A 10.4-3 A 10.4-4 A 10.4-5 A 10.4-6 30.4-7 A 10.4-8 A Amendment C Jane 30, 1988
4 C E S S A R 8!?Mic4ritu csnoot 2 or 3) EFFECTIVE PAGE LISTING (Cont'd) CHAPTER 10 Text (Cont'd) Pace Agendment 10.4-9 A 10.4-10 A 10.4-11 A 10.4-12 A 10.4-13 A 10.4-14 A 10.4-15 A 10.4-16 A 10.4-17 A 10.4-18 A 10.4-19 A 10.4-20 A 10.4-21 A 10.4-22 A 10.4-23 A 10.4-24 C 10.4-25 C 10.4-26 C 10.4-27 C 10.4-28 C 10.4-29 C 10.4-30 C
- 10. -31 C 10.4-32 C 10.4-33 C 10.4-34 C 10.4-35 C 10.4-36 C 10.4-37 C 10.4-38 C 10.4-39 C 10.4-40 C 10.4-41 C 10.4-42 C Tables Amendment 10.1-1 A 10.3.5-1 A 10.3.5-2 A 10.4.9-1 C Amendment C June 30, 1988
CESSAR E!?Ricari:n (susot 3 or 3) EFFECTIVE PAGE LISTING (Cont'd) CHAPTER 10 Tables (Cont'd) Paqo Amendment 10.4.9-2 C 10.4.9-3 C 10.4.9-4 C 10.4.9-5 C Ficures Amendment 10.1-1 A 10.1-2 A 10.1-3 10.3.2-1 A 10.4.7-1 A 10.4.7-2 A 10.4.7-3 A 10.4.8-1 A 10.4.9-1 (Sheet 1) C 10.4.9-1 (Sheet 2) C Amendment C June 30, 1988 l
CESSAR 8HFincua. . TABLE OF CONTENTS CHAPTER 10 Section subiect Pace No. 10.0 STEAM AND POWER CONVERSION SYSTEM
- 10.1-1 10.1
SUMMARY
DESCRIPTION 10.1-1 10.2 TURBINE GENERATOR 10.2-1 A 10.3 MAIN STEAM SUPPLY SYSTEM 10.3-1 10.3.1 DESIGN BASES 10.3-1 10.3.2 SYSTEM DESCRIPTION 10.3-2 10.3.2.1 System Performance 10.3-2 10.3.2.2 System Arranaement 10.3-3 10.3.2.3 Pioina, Valve. I&C and Insulation 10.3-6 10.3.2.3.1 Piping 10.3-6 10.3.2.3.2 Valves 10.3-7 C 10.3.2.3.2.1 Main Steam Isolation l Valve (MSIV) and MSIV Byass Valve 10.3-7 10.3.2.3.2.2 Main Staam System Safety Valves 10.3-8 10.3.2.3.2.3 Main Steam Atmospheric Oump Valves (ADvs) 10.3-10 10.3.2.3.3 Instrumentation and Control 10.3-11 ! 10.3.2.3.3.1 Main Steam Isolation ! Valve (M3IV) 10.3-11 10.3.2.3.3.2 AtJospheric Dump Valves (ADVs) 10.3-12 l l l
- Chapter 10 will be updated in future submittals to includa I
baseline data from Chapters 6 and 15 safety analyses and the System 80+ probabilistic risk assessment. Amendment C i June 30, 1988 l l
CESSAR !!Since l l TABLE OF CONTENTS (Cont'd) CHAPTER 10 Section Subiect Pace No. 10.3.2.3.4 Insulation 10.3-13 lC 10.3.3 SAFETY EVALUATION 10.3-13 10.3.4 INSPECTION AND TESTING REQUIREMENTS 10.3-14 A 10.3.5 SECONDARY WATER CHEMISTRY 10.3-15 l 10.3.5.1 Chemistry Control Basis 10.3-15 l 10.3.5.2 Corrosion Control Effectiveness 10.3-16 C l 10.3.6 STEAM AND FEEDWATER SYSTEM MATERIALS 10.3-18 lA l 10.3.6.1 Fracture Touchness 10.3-18 C 10.3.5.2 Materials Selection and Fabriptdgn 10.3-18 10.4 OTHER FEATURES OF STEAM AND POWER A CONVERSION SYSTEM 10.4-1 l l 10.4.1 MAIN CONDENSER 10.4-1 l 10.4.1.1 Desian Bases 10.4-1 l 10.4.1.2 System Description 10.4-1 10.4.1.3 Safety Evaluation 10.4-3 C 10.4.1.4 T.g.sts and Inscections 10.4-4 10.4.1.5 Jnstrumentation Aeolication 10.4-4 10.4.2 MAIN VACUUM SYSTEM 10.4-4 lA 10.4.1.1 Desian Bases 10.4-4 l 10.4.2.2 System Descriotion 10.4-5 C 10.4.2.3 Egfety Evaluation 10.4-5 10.4.2.4 Tests and Insoections 10.4-5 10.4.2.5 Instrument Aeolication 10.4-5 Amendment C 11 June 30, 1988
CESSAR E!nincua TABM OF CONTENTS (Cont'd) CHAPTER 10 Section Dubiect Pace No. 10.4.3 TURBINE GLAND SEALING SYSTEM 10.4-4 A 10.4.4 TURBINE BYPASS SYSTEM 10.4-5 10.4.4.1 Desian Bases 10.4-5 10.4.4.2 System Descriotion 10.4-7 10.i.4.2.1 General Description 10.4-7 10.4.4.2.2 Piping and Instrumentation 10.4-7 10.4.4.2.3 Turbine Bypass Valve 10.4-7 C 10.4.4.2.4 System Operation 10.4-7 10.4.4.2.4.1 System Performance 10.4-8 10.4.4.3 Safety Evaluation 10.4-9 10.4.4.4 Insoection and Testina Recuirements 10.4-9 10.4.4.5 Instrumentation Aeolication 10.4-9 10.4.5 CIRCULATING WATER SYSTEM 10.(-10 10.4.6 CONDENSATE CLEANUP SYSTEM 10.4-10 10.4.6.1 Desian Basis 10.4-10 1 10.4.6.2 System Description 10.4-10 10.4.6.3 Safety Evaluation 10.4-11 C
, 10.4.6.4 Insoection and Testina l Recuirements 10.4-11 l
10.4.6.5 Instrumentation Aeolications 10.4-11 l 10.4.7 CONDENSATE AND FEEDWATER SYSTEMS 10.4-11 lA l I I 10.4.7.1 Dpsian Basis 10.4-11 ' C 10.4.7.2 System Descriotion 10.4-12 Amendcent C iii June 30, 1988 l l 1
CESSAR Enacma i TAELE OF CONTENTS (Cont'd) CHAPTER 10 Section Subiect Pace No. 10.4.7.2.1 System Performance 10.4-13 10.4.7.2.2 System Arrangement 10.4-15 10.4.7.2.3 Piping, Valves, Equipment and C Instrumentation 10.4-16 10.4.7.3 Safety Evaluation 10.4-19 10.4.7.4 Tests and Inspections 10.4-19 10.4.7.5 lastrumentation Aeolications 10.4-20 10.4.8 STEAM GENERATOR BLOWDOWN SYSTEM 10.4-21 lA 10.4.8.1 Desian Basis 10.4-21 10.4.8.2 System Descriotion 10.4-21 10.4.8.3 Safety Evaluation 10.4-24 10.4.8.4 Tests and Insoectiotia 10.4-24 10.4.8.5 Instrunentation Aeolications 10.4-24 10.4.9 EMERGENCY FEEDWATER SYSTEM 10.4-25
'0.4.9.1 . Desian Basia 10.4-25 C 10.4.9.1.1 Functional Requirements 10.4-25 10.4.9.1.2 Design Criteria 10.4-25 10.4.9.2 System Description 10.4-29 10.4.9.2.1 General Description 10.4-29 10.4.9.2.2 Component Description 10.4-31 10.4.9.2.2.1 Emergency Feedwater Pumps 10.4-31 1
1 l l l Amendment C iv June 30, 1988
CESSAR !!%nce TABLE OF CONTENTS (Cont'd) CRAPTER 10 Section subiect Pace No. 10.4.9.2.2.2 Steam-Driven Emergency Feedwater Pump Turbines 10.4-31 10.4.9.2.2.3 Emergency Feedwater Storage Tanks 10.4-32 10.4.9.2.2.4 Emergency Feedwater Cavitating Venturis 10.4-32 10.4.9.2.2.5 Active Valves 10.4-33 1 10.4.9.2.3 Electcical Power Supply 10.4-35 10.4.9.2.4 Emergency Feedwater System operation and Control 10.4-35 10.4.9.3 Safety Evaluation 10.4-37 10.4.9.4 Inseection and Testina Recuirements 10.4-38 C 10.4.9.4.1 EFW System Performance Tests 10.4-38 10.4.9.4.2 Reliability Tests and Inspections 10.4-39 10.4.9.5 Insttyment Recuirements 10.4-39 10.4.9.5.1 Pressure Instrumentation 10.4-40 10.4.9.5.2 Temperature Instrumentation 10.4-40 10.4.9.5.3 Flow Instrumentation 10.4-41 10.4.9.5.4 Level Instrumentation 10.4-41 10.4.9.5.5 Steam-Driven Pumps ' Turbine Speed 10.4-42 l ( l APPENDIX 10A EMERGENCY FEEDWATER SYSTEM RELIABILITY ANAIYSIS 101.-1 l l 1 Amendment C v June 30, 1988 i I
rp _ _ _ CESSAR !!!sinceu LIST OF TABIES CHAPTER 10 Table Subiect 10.1-1 Steam and Power Conversion System Design and Performance Characteristics (LATER)
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10.3.5-1 Operating Chemistry Limits for Secondary Steam Generator Water 10.3.5-2 Operating C.>mistry Limits for Feedwater and Condensate 10.4.9-1 Emergency Feedwater System Component Parameters 10.4.9-2 Emergency Feedwater System Active Valve List 10.4.9-3 Emergency Feedwater System Failure Analysis 10.4.9-4 Emergency Feedwater System Instrumentation and Control 10.4.9-5 Emergency Feedwater System Emergency Power Requirements l Amendment C vi June 30, 1988
l ) CESSAR E! Mince I LIST OF FIGURE 3 CEAPTER 10 Ficure Subiect 10.1-1 Flow Diagram of Steam and Power Conversion System A 10.1-2 Heat Balance for Steam and Power Conversion System (LATER) 10.1-3 Main Steam System Piping and Instrumentation Diagram 10.3.2-1 Atmospheric Dump Valve Flow Requirements 10.4.7-1 Steam Flow Versus Power
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10.4.7-2 Steam Generator Outlet Pressure Versus Power 10.4.7-3 Economizer /Downcomer Flow Split 10.4.8-1 Flow Diagram of Steam Generator Blowdown System 10.4.9 ' Emergency Feedwater System Piping and C Instrumentation Diagram Amendment C vil June 30, 1988 j l l
CESSAR R*OcAMpf
- 4. A constant speed motor driven feedwater pump is utilized for startup and shutdown. Upon loss of main feedwater and reactor trip, this pump starts automatically and maintains steam generator level.
Instrumentation and Control
- 1. "he required accuracy of the feedwater temperature measurement devices is 5'F for any calorimetric measurement.
- 2. Feedwater control valves are capable of manual control at all times.
Insulation
- 1. Non-metallic insulation conforms to NRC Regulatory Guide
. 3 6. The chloride and fluoride content of the non-metallic insulation are acceptable as shown in Rogulatory Guide 1.36. A Tests will be made on representative samples of the non-metall3 thermal insulation to certify that the maximum chloride and flitoride content are not exceeded. All water used in the fabrication of non-metallic thermal insulation is demineralized or distilled water.
- 2. The insulation thickness is selected to minimize the heat load on the containment ventilation and cooling system,g A th3 pal-2 transference of not more than 0.14 Btu-hr -
'F -ft of insulated component surface area is used as a dosign basis for insulation.
10.4.7.3 Safety Evaluation l l The safety-related portion of the F1edwater System is designed in
- accordance with the design bases presented in Section 10.4.7.1 ,
(as long as the applicant conforms with the descriptions and requir sments presented in Section 10.4.7.2) . Any failure in the ! non-safety class portions of the Condensate and Feedwater Systems l does not prevent safe shutdoen of the reactor. Effects of equipment malfunction on the Reacter Coolant System are presented in Chapter 15. 10.4.7.4 Impts and Inscoctions ASME Section III Code Class 2 piping is inspected and tested in accordance with ASME Code Section III and XI. ASME Sections III code class 2 valves are periodically inservice tested for exercising and leakage in accordance with ASME Code Section XI, Subsection IWV. Amendment A 10.4-19 September 11, 1987 I
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CESSAR ;!L"icua I l See the site-specific SAR for additioital test and inspection i requirements. l 10.4.7.5 Instrumentation Apolications Feedwater flow control instrumentation measures the feedwater A flow rate from the condensate and feedwa*.or system. This flow i measurement, transmitted to t e feedvr.cer control system, l regulates the feedwater flow to the steam generators to meet system demands. Refer to Section 7.7.1.1.4 for a description of the feedwater control system. Amendment A 10.4-20 September 11, 1987
CESSAR8Ence 10.4.8 STEAM GENERATOR BLONDOWN SYSTEM 10.4.8.1 Desian Basis The design bases for the Steam Generator Blowdown System are: A. Maintain propor steam generator shell side water chemistry as outlined in Section 10.3.5 by removing non-volatile materials due to condenser tube leaks, primary to secondary tube leaks, and corrosion that would otherwisa become more concentrated in the shell side of the steam generators. B. Process steam generator blowdown for reuse as condensate. C. Enable blowdown concurrent with steam generator tube leak (s) or radioactivity present on the secondary side without release of radioactivity to the environment. D. Process a continuous steam generator blowdown rate of 0.2% or 1% of the full power rain steam flow. A E. Continuously sample the radioactivity of the steam generator blowdown. F. Isolate the blowdown lines leaving the Containment upon e Containment Isolation Signal, Main Steam Isoletion Signal, or Emergency Feedwater Actuation Signal. 10.4.8.2 Sistem DescriDtion A continuous high flow blowdown controls the concentration of impurities in the steam generator secondary side water. A general schematic of a typical blowdown system is shown in Figure 10.4.8-1. Each stean generator is equipped with its own blowdown line with the capability of blowing down the hot leg and/or the economizer regions of the steam oenerator shell cide. The blowdown will be directed into a flash tank where the flashed steam is returned to the cycle via the low pressure feedwater heaters. The liquid portion flows to a heat exchanger where it is coolcd, and then l directed through a blowdown filter where the majar portion of the suspended solids are removed. After filtration, the blowdown fluid is processed by blowdown demineralizers and returned to the condenser. l l Amendment A 10.4-21 September 11, 1987 l
CESSAR!Ence l l l l The final design and layout of the Steam Generator Blowdown l System is described in the site-specific SAR. The following ! requirements are to be met to ensure a reliable system. 1 A. The Steam Generator Blowdown System is designed to I accommodate a continuous blowdown of approximately 1% (172,000 lb/hr) maximum steaming rate (MSR). B. Each steam generator is provided with 2 tubesheet connections, including a 6 inch nozzle for hot leg blowdown, and a 6 inch rozzle for economizer blowdown. The Steam Generator Blowdown System, connected to each steam generator blowdown connection is capable of accommodating a continuous blowdown of approximately 0.5% MSR f 86,000 lb/hr) . C. Makeup systems are capable of previding secondary makeup water at a rate greater than 172,000 lb/hr. D. Steam Generator 3 lowdown System piping and valves are arranged to allow blowdown from either or both blowdown nozzles. E. The Steam Generator Slowdown Processing System is capable of accepting both a total continuous blowdown rate of 0.2% of each steam generator's MSR (17,200 lb/hr/ generator) while A the plant is at power and steam generator chemistry is within normal limits, and a continuous blowdown of up to 1% of each steam generator's MSR (86,000 lb/hr/ generator) while the plant is at power and steam generator chemistry is not within normal limits. F. The thermodynamic conditions at the blowdown nozzlos are as follows: Nossle Flow PJjLt Power Level Fluid Conditin Hot Leg 1% MSR Full Load (LATER) psia, (LATER) quality Economizer 1% MSR Full Load (LATER) psia, 30-40*F subcool Hot Leg 0.2% MSR No Load (LATER) psia, saturated lig. G. Provisions are made to process the continuous steam generator blowdown water to 90% reduced radioactivity levels. H. Blowdown water returned to the steam generator meets the water chemistry requirements outlined in Section 10.3.5. I. The blowdown system piping material is compatible with saturated steam service. Amendment A 10.4-22 September 11, 1987
CESSARUSinem . J. All components, piping and their supports and restraints associated with steam generator blowdown between t? a steam generator and the outer most Containment Isolation Valves or Branch Piping Isolation Valves are Seismic Category I and are designed in accordanca to ASME Code, Section III, Class 2 requirements. K. Blowdown piping exiting containment consists of Redundant Blowdown Line Isolation Valves in accordance with General Design Criteria 54 and 57 and is isolated by a MSIS, a CIAS, and an Emergency Feedwater Actuation Signal (EFAS). L. The Steam Generator Blowdown System piping is used as the means to drain the secondary side of the steam generators. The drain connectione are located such that complete steam I ger erator drainage can be accomplished. M. One nitrogen supply connection is provided on either steam generator blowdown line to provide a purge path following steam generator maintenance. N. The steam generators are designed with the capability to achieve the following high capacity blowdown rates and associated thermodynamic ,:onditions at the blowdown nozzles: Nozzle Flow Rat _e_ Power Level Fluid Condition A Hot Leg 5.5% MSR Full Load (LATER) psia 17.1% quality Hot Leg 8.6% MSR No Load (LATER) psia 5.6% quality Economizer 8.6% MSR No Load (LATER) psia 5.6% quality
- Note Conditions presented assume a blowdown ng system with an equivalent flow resistance ofpipf {/D
= 50. The balance of plant design shall provide a system resistance as follows:
b 50<f /D <60. , acity assumed an equivalent system Maximum resistanceflow of cag/D f = 50. The blowdown system is designed for a very high capacity flow (10% MSR) for a short period of time (2 minutes). O. A system is provided to maintain the steam generators in wet layup with the capability to adequately mix, sample, and add chemicals to them. Amendment A 10.4-23 September 11, 1987 l l
CESSARanGem,. P. In addition to the above described High Japacity Steam Generator Blowdown System, it is recommended that blanked connectors be incorporated in the blowdown and/or main l feedwater piping to allow for cbemical cleaning of the steam I generators should it become necessary in the future. Q. Thermal insulation used on the blowdown system inside containment meets the following requirements: Non-metallic insulation conforms to NRC Regulatory Guide 1.36. The chloride and fluoride content of the non-metallic insulation are acceptable as shown in Regulatory Guide 1.36. Tests will be made on representative samples of the non-metallic thermal insulation to certify that the maximum chloride and fluoride contents are not exceeded. All water used in the fabrication of non-metallic thermal insulation is demineralized or distilled water. The insulation thickness is selected to minimize the heat load on the containment ventilation and cooling system. It is suggested _ythat a_1the g l transference of not more than 0.14 Btu-hr -
'F -ft of insulated component surface area be used as a design basis for insulation.
10.4.8.3 Safety Evaluation The Steam Generator Blowdown System is designed to operate manually and on a continuous basis as required to maintain A acceptable steam generator secondary side water chemistry. The presence of ASME Section III - Class 2 piping and the system containment isolation function require the system to be designated "Nuclear Safety Related". The operation of the system is not required, however, for plant safe shutdown. All blowdown lines which penetrate the containment are isolated automatically upon containment isolation signal, Main Steam Isolation Signal or Emergency Feedwater Actuation Signal. The portion of the system inside the containment and the portion utilized as containment . isolation are designed in accordance with applicable safety class requirements. 10.4.8.4 Tests and Inscoctions ASME Section III Code Class 2 piping is inspected and tested in accordance with ASME Code Sections III and XI. ASME Sections III Code Cless 2 valves are periodically inservice tested for exercising and leakage in accordance with ASME Code Section XI Subsection IWV. 10.4.8.5 Instrumentation Acclications_ This system is non safety-related. Amendment A 10.4-24 September 11, 1987
CESSAR !ErlCATION 10.4.9 EMERGENCY FEEDWATER SYSTEM 10.4.9.1 Desien Basis 10.4.9.1.1 Functional Requirements The Emergency Feedwater (EFW) System provides an independent safety related means of supplying secondary-side, quality feedwater to the steam generator (s) for removal of heat and prevention of reactor core uncovery during emergency phases of plant operation. The EFW System is a dedicated safety system which has no operating functions for normal plant operation. The EFW System is designed to be automatically or manually C initiated, supplying feedwater to the steam generators for any event that results in the loss of normal feedwater and requires heat removal through the steam generators, including the loss of normal onsite and normal offsite AC power. Following the event, the EFW System maintains adequate feedwater inventory in the steam generator (s) for residual heat removal and it is capable of maintaining hot standby and facilitating a plant cooldown (at the maximum administratively controlled rate of 75
'F/hr) from hot standby to shutdown Cooling System initiation.
The Shutdown Cooling System becomes available for plant cooldown when the RCS temperature and pressure are reduced to the entry conditions given in Section 5.4.7. The EFW System is designed to be initiated with operator action following a major loss of coolant accident to keep the steam generator tubes covered for the long term to enhance the closed system containment boundary. Note: Covering the steam generator tubes post LOCA minimizes potential containment bypass leakage should pre-existing primary to secondary leakage be present. 10.4.9.1.2 Design Criteria A. The EFW System and its supporting auxiliaries are provided with emergency power and adequate redundancy, diversity and separation to perform its design basis function in the event of a loss of offsite and normal onsite power, coincident Amendment C 10.4-25 June 30, 1988
CESSARUMem - with: (1) a single active mechanical component failure, or (2) a single active electrical component failure, or (3) the effects of a high or moderate energy pipe rupture. B. All components and piping (upstream of the automatic steam generator isolation valves and essential to the emergency function) are ANS Safety Class 3. Components and piping involved in containment integrity (downstream of the automatic steam generator isolation valves) are ANS Safety Class 2. The line classifications are consistent with the C requirements specified in ANS 51.10. All components and piping essential to the emergency function are designed to Seismic Category I requirements as described in Section 3.7. The seismic category and safety and quality classification of the EFW System components are listed in Table 3.2-1. C. The EFW System is equipped with diverse pump drive mechanisms. This is accomplished by providing one full-capacity motor-driven pump and one full-capacity steam-driven pump in each EFW mechanical train. All controls, instrumentation, and valves which are essential to the emergency operation of the steam-driven pump subtrains are powered by battery backed Class 1E power. The batteries are capable of powering the EFW steam-driven pump subtrains for a station blackout up to (LATER) hours. D. The EFW System coLponents are located in Seismic Category I structures which also protect the components from extecnal environmental hazards such as tornados, hurricanes, floods, and external missiles. Each redundant and diverse subtrain of tne EFW System is physically separated from the others within these structures. E. All essential components are designed to account for, located to protect against, or protected from internal flooding, internal missiles, interactions from earthquakes, or the effects of high or moderate energy line breaks as described in Chapter 3. F. The EFW System is designed so that it can be either manually initiated or automatically initiatsd by the Emergency Feedwater Actuation System (EFAS) described in Section 7.3.1 or the Alternate Protection System (APS) described in Section 7.7 The EFW System is designed to deliver flow to the steam generator (s) within 60 seconds upon receipt of an EFAS signal. G. Each EFW pump is capable of providing 100 percent of the required minimum flow of 500 gpm, to meet the design basis Amendment C 10.4-26 June 30, 1988
CESSAR8!Ence heat removal requirements. Each pump is capable of delivering this flow under the following coincident parameters:
- 1. The maximum steam generator downcomer nozzle pressure is 1217 psia which accounts for the oteam genere. tor design pressure, safety valve uncertainty and feed nozzle losses from the downcomer nozzle to the steam generator steam space. C
- 2. Feed line losses are those existing when full EFW flow is diverted to one steam generator.
- 3. Pump suction is at the minimum suction pressure.
- 4. A margin of 7 percent is added to the required head.
This includes a 2 percent margin for wear and a 5 percent margin for uncertainties. H. The maximum EFW flow to each steam generator is restricted by a cavitating venturi to protect the EFW pumps from damage due to excessive runout flow, and to allow the operator 30 minutes to regulate or terminate EFW flow to prevent RCS overcooling, steam generator overfill or containment overpressurization. I. The EFW System is designed to maintain an emergency feedwater temperature of at least 40'F and no greater than 120*F. J. The EFW System has a safety related condcnsate storage volume of 350,000 gallons to achieve safe cold shutdown, basad on:
- 1. A main feedline break without isolation of EFW flow to the affected steam generator for 30 minutes;
- 2. Refill of the intact steam generator;
- 3. Eight hours of operation at hot standby conditions;
- 4. Subsequent cooldown of RCS within six hours to conditions which permit operation of the Shutdown Cooling System; and,
- 5. Continuous operation of one reactor coolant pump.
Amendment C 10.4-27 June 30, 1988
- 1
CESSARinece This safety related condensate storage volume is sufficient to permit cooldown with the reactor in natural circulation. Notes EFW System storage utilizes two dedicated safety grade supply tanks terraed the Emergency Feedwater Storage Tanks (EFWSTs). Each tank contains 50% of the total required EFW water supply. A low EFWST water level alarm allows the operator 30 minutes to manually align the tank from the other train. C K. A non-safety grade source of condensate can be aligned should the safety related source of condensate be exceeded before Shutdown Cooling System entry conditions are reached. L. The EFW System is controllable in a post-accident environment from either the control room or the remote shutdown station. M. The EFW System piping in the vicinity of tne steam generators is arranged to minimize the potential for destructive water hammer during startup. The EFW piping continuously rises as it penetratos the containment to connect with the downcomer feedwater pipe which enters the steam generator. After the two lines connect, the downcomer feedwater pipe continues to rise to prevent draining into the steam generator with the feedwater flow shut off. It then connects to a 90 degree elbow facing downward, which is attached to the steam generator downcomer nozzle. N. The EFW Syst_ep is desggned to have an unavailability in the , range of 10 to 10 per demand based on analysis using ! methods and data presented in NUREG-0611 and NUREG-0635. Analysis to support this criterion is presented in Appendix 10A. l o. The emergency feedwater supplied to the steam generators is l of the same or better quality than the secondary system , makeup water, except that the requirement on oxygen is
- excluded.
l P. Each turbine-driven pump shall be supplied with steam from a single steam generator, i.e., the one to which it supplies , feedwater. l l Q. Means are provided to permit periodic surveillance testing I of the EFh pumps and valves, and functional testing of the integrated operation of the system. Amendment C 10.4-28 June 30, 1988 l
CESSARineem R. Emergency feedwater is delivered to the downcomer nozzles of the steam generators. S. The EFW System provides double isolation from the Main Feedwater System during normal plant conditions when the .FW System is not required. C T. A four-channel control scheme is provided to preclude inadvertent actuation in the event of a single failure. A four-channel design is provided for the initiation logic, actuation logic, and power. 10.4.9.2 S.y.g.tsm Description 10.4.9.2.1 General Description The EFW System is shown in Figure 10.4.9-1, Sheets 1 and 2. The EFW System is configured into two separate mechanical trains. Each train is aligned to feed its respective steam generator. Each train consists of one Emergency Feedwater Storage Tank (EFWST), ona 100 percent capacity motor-driven pump subtrain, one 100 percent capacity steam-driven pump subtrain, valves, one cavitating venturi, and specified instrumentation. Each pump subtrain takes suction from its respective EFWST and has its respective discharge header. Each subtrain discharge header contains a pump discharge check valve, flow regulating valve, steam generator isolation valve and steam generator isolation check valve. The motor-driven subtrain and steam-driven subtrain are joined together inside containment to feed their respective steam generator through a common EFW header which connects to the steam generator downcomer feedwater line. Each common EFW header contains a cavitating venturi to restrict the maximum EFW flow rate to each steam generator. The cavitacing venturi restricts the magnitude of the two pump flow as well as the magnitude of individual pump runout flow to the steam generator. A cross connection is provided between er.ch EFWST so that either tank can supply either train of EFW. The two EFWSTs are safety grade tanks of seismic design in which each tank contains 50 percent of the total volume specified in Section 10.4.9.1.2.J. A normally locked closed, local manually operated isclation valve is provided for each EFWST to provide separation. A line connected to a non-safety source of condensate is also provided with local canual isolation so that it can be manually aligned for gravity feed to either of the EFWSTs should the EfWSTs reach Amendment C 10.4-29 June 30, 1988
CESSARSEnem,. low level before Shutdown Cooling System entry conditions are reached. A check valve and a normally locked closed, local manually operated isolation valve are provided for separation of the non-safecy source of condensate from the safety related sources. pump discharge crossover piping is provided to enhance system versatility during long term emergency modes, such that a single pump can feed both steam generators. Two normally locked closed, local manually operated isolution valves are provided for C subtrain separation. A flow recirculation line is provided downstream of each pump discharge, which allows: 1) a continueus flow back to the EFWST for pump minimum flow protection; and 2) full or partial flow testing of the pumps. A multi-stage flow restrictive orifice restricts the flow to the minimum required for pump protection. Each pump has adequate flow capacity to continuously recirculate this flow plus provide the required design basis flow to the steam generators. The recirculation lines are adequately sized so that full pump flow can be recirculated through the bypass provided around the flow restrictive orifice for full flow pump testing during power operations. The bypass line contains a manual flow regulating valve in order to vary the pump flow for performance testing. Each steam-driven pump is provided with an atmospheric-discharge, non-condensing turbine. Driving steam is supplied from the Main Steam System upstream of the main steam isolation valves. Each turbine is supplied with steam from the steam generator to which the pump feeds. Each supply line contains a normally closed fail-open air operated steam isolation valve. A bypass is provided around each of these isolation valvas with a flow restricting orifice and a normally closed fail-open air operating bypass isolation valve. The bypass provides a small controlled rate of steam flow to the turbine. This allows the hydraulic control portion of the governor to pressurize at turbine idle speed before the steam isolation valve is opened for full rated ! speed operation. The turbines exhaust to atmosphere through a missile protecud Seismic Category I vent line routed through the roof. Low point drains are provided for collection and return to the condensate system of any liquid that may condense in the supply and exhaust lines. A low point drain, located upstream of the steam supply j isolation valve, provides a continuous blowdown through a flow l restricting orifice in order to keep the supply line warm and prevent water slugs from entering the turbine on an automatic emergency start. A pcwer operated valve is provided in this line so that it can be remotely isolated from the control room should high activity be present. A bypass is p*ovided around each drain orifice should additional drain capacity be required. Amendment c 10.4-30 June 30, 1988
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