ML20247G913
| ML20247G913 | |
| Person / Time | |
|---|---|
| Site: | 05200002, 05000470 |
| Issue date: | 03/30/1989 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML20247G537 | List:
|
| References | |
| NUDOCS 8904040345 | |
| Download: ML20247G913 (216) | |
Text
{{#Wiki_filter:- CESSAR EHErica,, (Sheet 1 of 6) O EFFECTIVE PAGE LISTING CHAPTER S Table of Contents 1 pace Amendment I i l 11, E lii iv v E I vi E vil viii B ix E x E xi xii E xiii Text p_ age Amendment i 5.1-1 5.1-2 B l 5.1-3 B 5.1-4 5.1-5 E 5.1-6 B l 5.1-7 5.1-8 5.1-9 B 5.1-10 D 5.1-11 E 5.1-12 B 5.1-13 B 5.1-14 D 5.1-15 5.1-16 D 5.1-17 E 5.1-18 5.1-19 5.1-20 B 5.1-21 E 5.1-22 D wmm mik K
"i Amendment E December 30, 1988
CESSAR ;!Mmi:n (Sheet 2 of 6) O EFFECTIVE PAGE LISTING (Cont'd) CHAPTER S Text (Cont'd) Page Amendment 5.1-23 D . 5.1-24 B 5.2-1 D , 5.2-2 E I 5.2-3 5.2-4 5.2-5 5.2-6 E 5.2-7 E 5.2-8 E 5.2-9 E 5.2-10 E 5.2-11 E 5.2-12 E 5.2-12a E 5.2-12b E 5.2-13 B 5.2-14 D 5.2-15 D 5.2-16 5.2-17 D 5.2-18 5.2-19 D 5.2-20 D 5.2-21 B 5.2-22 E 5.2-23 D 5.2-24 D ; 5.3-1 B i E.3-2 i 5.3-3 D l 5.3-4 E 5.3-5 E 5.3-6 E 5.3-7 E 5.3-8 E 5.3-9 E 5.3-10 D 5.3-11 E 5.3-12 E 5.3-13 E 5.3-14 E l E I 5.3-15 Amendment E December 30, 1988
CESSARHEncm. (Shoot 3 of 6) i EFFECTIVE PAGE LISTING (Cont'd) CHAPTER-5 Text (Cont'd) Pace Amendment , i i 5.3-16 E , 5.3-17 E l 5.3-18 E l l -5.3-19 E 'l 5.3-20 E 5.3 D 5.3-22 E 5.4-1 D q 5.4-2 D -l 5.4-3 D 5.4-4 D 1 5.4-5 D I 5.4-6 5.4-7 5.4-8 B l O 5.4-9 5.4-10 D E 5.4-11 , 5.4-12 D i 5.4-13 D 5.4-13a D 5.4-14 5.4-15 5.4-16 D j 5.4-17 D i 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 E 5.4-26 C : 5.4-27 C 5.4-28 E 5.4-29 E 5.4-30 C l 5.4-31 C > 5.4-32 E O 5.4-33 5.4-34
-5.4-35 C
E E Amendment E December 30, 1988
CESSAR EHLbn=2 (Shest 4 of 6) l EFFECTIVE PAGE LIS.J'ING (Cent'd) CHAPTER 5 Text ( C(. '. ' d ) Page Amendment 5.4-36 C 5.4-37 E 5.4-38 C 5.4-39 C 5.4-40 C 5.4-41 C 5.4-42 C 5.4-43 E 5.4-44 E 5.4-45 B 5.4-46 B 5.4-47 5.4-48 E 5.4-49 D 5.4-50 5.4-51 D 5.4-52 B 5.4-53 Tables Amendment 5.1.1-1 E ! 5.1.1-2 E 5.1.1-3 D 5.1.4-1 5.1.4-2 D 5.1.4-3 D 5.2-1 D 5.2-2 (Sheet 1) D 5.2-2 (Sheet 2) D 5.2-2 (Sheet 3) D 5.2-2 (Sheet 4) D 5.2-2 (Sheet 5) D 5.2-3 D 5.2-4 E 5.3-1 E l S.3-2 E 5.3-3 E 5.3-4 E 5.3-5 D 5.3-6 5.3-7 E Amendment E j December 30, 1988 j i
CESSAR E!nincou,. (Sheet 5 of 6) O EFFECTIVE PAGE LISTING (Cont'd) CHAPTER S Tables (Contid) Amendment 5.4.1-1 E 5.4.2-1 E 5.4.7-1 (Sheet 1) E 5.4.7-1 (Sheet 2) E 5.4.7-2 C 5.4.7-3 (Sheet 1) E 5.4.7-3 (Sheet 2) E 5.4.7-3 (Sheet 3) E 5.4.7-3 (Sheet 4) E 5.4.7-3 (Sheet 5) E 5.4.7-3 (Sheet 6) E 5.4.7-3 (Sheet 7) E 5.4.7-3 (Sheet 8) E 5.4.7-3 (Sheet 9) E 5.4.10-1 E 5.4.10-2 5.4.13-1 D O 5.4.13-2 Ficures Amendment l 5.1.2-1 5.1.2-2 b.1.3-1 E 5.1.3-2 E 5.2-1 E 5.3-1 E 5.3-2 E 5.3-3 E 5.3-4 E 5.3-Sa E 5.3-5b E 5.3-6 E 5.3-7 E 5.4.1-1 B 5.4.2-1 E 5.4.'7-1 C 5.4.7-2 C 5.4.7-3 (Sheet 1) C 5.4.7-3 (Sheet 2) C 5.4.10-1 E O 5.4.10-2 5.4.10-3 B B Amendment E December 30, 1988 l
CESSAR EMUncm:u (Sheet 6 of 6) O EFFECTIVE PAGE LISTING (Cont'd) CHAPTER 5
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Fictures (Cont'd) Amendment
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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 D 5.4.14-4 O l O l l Amendment E December 30, 1988
CESSAR !!'Unc ww O TABLE OF CONTENTS CHAPTER 5 Section Subiect Pace No. L 5.0 REACTOR COOLANT SYSTEM AND CONNECTED 5.1-1 SYSTEMS 5.1
SUMMARY
DESCRIPTION 5.1-1 l 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 BALANCE OF PLANT INTERFACE REQUIREMENTS 5.2 INTEGRITY 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 10 CFR 50.55a 5.2-1 5.2.1.2 Annlicable Code Cases 5.2-1 5.2.2 OVERPRESSURE PROTECTION 5.2-1 5.2.2.1 Desian Bases .5.2-1 5.2.2.2 pesian Evaluation 5.2-2 l 5.2.2.3 Pinina and Instrumentation Diaorams 5.2-2 5.2.2.4 Eauipment &-Component Description 5.2-2
$.2.2.4.1 Transients 5.2-3 5.2.2.4.2 Environment 5.2-3 l 5.2.2.4.2.1 Normal Environment 5.2-3 l 5.2.2.4.2.'2 Main Steam Line Break 5.2-3.
! (One Occurrence) 5.2.2.4.3 Main Steam Safety Valvos 5.2-4 5.2.2.4.3.1 Main Steam Safety Valve 5.2-4 Operation 1
CESSARHiece TABLE OF CONTENTS (Cont'd) CHAPTER 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 Environment 5.2-5 5.2.2.4.3.3.2 Main Steam Line Break f.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 Mountino of Pressure-Relief Devices 5.2-6 5.2.2.6 Applicable Codes and Classification 5.2-6 5.2.2.7 Process Instrumentation 5.2-6 5.2.2.8 System Reliability 5.2-7 5.2.2.9 Testina and Inspection 5.2-7 5.2.2.10 overpressure Protection Durina Low 5.2-7 Temperature 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 B 5.2.2.10.1.3 Testability 5.2-8 5.2.2.10.1.4 Seismic Design and IEEE Standard 5.2-8 279 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 5.2-12 5.2.2.10.2.4 Administrative Controls 5.2-12a 5.2.2.11 Pressurized Thermal Shock 5.2-12a Amendment E 11 December 30, 1988
CESSAR !!KneAm. O TABLE OF CONTENTS (Cont'd) CHAPTER 5 Section Subiect Pace No. 5.2.3 REACTOR COOLANT PRESSURE BOUNDARY MATERIALS 5.2-12a 5.2.3.1 Material Specification 5.2-12a 5.2.3.2 Compatibility with Reactor Coolant 5.2-13 l 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 Environmental Atmosphere 5.2.3.3 Fabrication and Processina 'erritic r 5.2-13 Materials l
l O 5.2.3.3.1 Fracture Toughness- 5.2-13 5.2.3.3.1.1 NSSS Components 5.2-13 5.2.3.3.2 Control of Welding 5.2-14 5.2.3.3.2.1 Avoidance of Cold Cracking 5.2-14 5.2.3.3.2.2- Regulatory Guide 1.34 5.2-15 5.2.3.3.2.3 Regulatory Guide 1.71 5.2-15 5.2.3.3.3 Non-Destructive Examination of 5.2-15 Tubular Products 5,2.3.4 Fabrication and Processina of 5.2-15 Austenitic Stainless Steel 5!2.3.4.1 Avoidance of Stress Corrosion 5.2-15 Cracking 5.2.3.4.1.1 Avoidance of Sensitization 5.2-15 5.2.3.4.1.1.1 NSSS Components 5.2-16 l 5.2.3.4.1.2 Avoidance of Contamination 5.2-18
,) Causing Stress Corrosion
\_/ Cracking lii i
_ - - - - - _ _ _ _ _ _ - - - - - - - - - _ - - ]
CESSAREPecucu O TABLE OF CONTENTS (Cont'd) CHAPTER 5 Section Subject Pace No. 5.2.3.4.1.2.1 NSSS Components 5.2-18 5.2.3.4.1.3 Characteristics and 5.2-19 Mechanical Properties of Cold-Worked Austenitic Stainless Steels for RCPB Components 5.2.3.4.2 Control of Welding 5.2-19 5.2.3.4.2.1 Avoidance of Hot Cracking 5.2-19 5.2.4 INSERVICE INSPECTION AND TESTING OF REACTOR 5.2-20 COOLANT PRESSURE BOUNDARY 5.2.4.1 Accessibility of Inspection Areas 5.2-20 5.2.5 REACTOR COOLANT PRESSURE BOUNDARY (RCPB) LEAKAGE DETECTION SYSTEMS 5.2-21 0 5.2.5.1 Leakace Detection Methods 5.2-21 5.2.5.1.1 Unidentified Leakage 5.2-21 ; 5.2.5.1.2 Identified Leakage 5.2-22 I 5.2.5.1.2.1 Safety Valves Located on the 5.2-22 Reactor Coolant System 5.2.5.1.2.2 Reactor Coolant Pump Seals 5.2-22 5.2.5.1.3 Leakage Through Steam Generator 5.2-23 Tubes or Tubesheet 5.2.5.1.4 Leakage to Auxiliary Systems 5.2-23 5.2.5.2 Control Room Leakace Instrumentation 5.2-23 5.2.5.3 Limits for Reactor Coolant Leakace 5.2-23 5.2.5.4 Maximum Allowable Total Lenka,gg 5.2-24 S.2.5.5 Differentiation Between I_dentified and 5.2-24 l Unidentified Leaks 5.2.5.6 Sensitivity and Operability _Testg 5.2-24 G, iv
CESSAR inMacm:n Oi V TABLE OF CONTENTS (Cont'd) CHAPTER 5 Section Subiect Pace No.
- 5. 3 - REACTOR VESSEL 5.3-1 5.3.1 REACTOR VESSEL MATERIALS 5.3-1 i
i 5.3.1.1 Material Specifications 5.3-1 5.3.1.2 Special Process Used for Manufacturing 5.3-1 and Fabrication 5.3.1.3 Special Methods for Nondestructive 5.3-2 Examination 5.3.1.4 Snecial Controls for Ferritic and 5.3-3 Austenitic Stainless Steels 5.3.1.5 Fracture Touchness 5.3-3 5.3.1.6 Reactor Vessel Material Surveillance 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 5.3-7 5.3.1.6.3 Surveillance Capsules 5.3-7 5.3.1.6.3.1 Charpy, Flux and Compact Tension 5.3-8 Compartment Assembly 5.3.1.6.3.2 Temperature, Flux, Tensile and- 5.3-8 Charpy Compartment Assembly 5.3.1.6.4 Neutron Irradiation and 5.3-9 Temperatures Exposure l 5.3.1.6.4.1 Flux Measurements 5.3-9 5.3.1.6.4.2 Temperature Estimates- 5.3-10 5.3.1.6.5 Irradiation Locations 5.3-10 l ~/ 5.3.1.6.6 Withdrawal Schedule 5.3-11 Amendment E v December 30, 1988
CESSAR Ennncua I O TABLE OF CONTENT 8(Cont'd) i I CHAPTER 5 g ( l Section Sub_ ject Pace No. 5.3.1.6.7 Irradiation Effects Prediction 5.3-11 Basis 5.3.1.7 Reactor Vessel Fasteners 5.3-12 5.3.2 PRESSURE-TEMPERATURE LIMITS 5.3-12 5.3.2.1 P-T Limit Curves 5.3-13 E 5.3.2.1.1 Material Properties 5.3-13 5.3.2.2 Operatina Procedures 5.3-20 5.3.3 REACTOR VESSEL INTEGRITY 5.3-21 5.3.3.1 Desian 5.3-21 5.3.3.2 Materials of Construction 5.3-21 5.3.3.3 Fabrication Methods 5.3-21 5.3.3.4 Inspection Requirements 5.3-21 5.3.3.5 Shipment and Installation 5.3-21 5.3.3.6* Operatina Conditions 5.3-22 5.3.3.7 Inservice Surveillance 5.3-22 5.4 COMPONENT AND SUBSYSTEM DESIGN 5.4-1 5.4.1 REACTOR COOLANT PUMPS 5.4-1 5.4.1.1 Pump P1vvheel Intecrity 5.4-1 5.4.1.2 Description 5.4-3 5.4.1.3 Eva1uatio_11 5.4-4 5.4.1.4 Tests and Inseg_glions 5.4-7
*See site-specific SAR Amendment E i l vi December 30, 1988 l
CESSARn!Mcamu O TABLE OF CONTENTS (Cont'd) CHAPTER 5 Section Subject Page No. 5.4.2 STEAM GENERATORS 5.4-9 5.4.2.1 Desian Bases 5.4-9 5.4.2.2 Description 5.4-11 5.4.2.3 Economizer Intecrity 5.4-12 5.4.2.4 Steam Generator Materials 5.4-13 5.4.2.4.1 Steam Generator Tubes 5.4-13 5.4.2.5 Tests and Inspections 5.4-13a 5.4.3 REACTOR COOLANT PIPING 5.4-14 5.4.3.1 Desian Basis 5.4-14 5.4.3.2 Description 5.4-14 5.4.3.3 Materials 5.4-15 5.4.3.4 Tests and Inspections 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 5.4.5.4 T,e g_t s_ _ n p d I n s p e c t i o n s 5.4-17 5.4.6 REACTOR CORE ISOLATION COOLING SYSTEM 5.4-17 5.4.7 SHUTDOWN COOLING SYSTEM 5.4 -18 vii
CESSAR EHEncari:n O TABLE OF CONTENTS (Cont'd) CHAPTER 5 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 Codes and 5.4-32 Classifications 1 5.4.7.2.5 System Reliability Considerations 5.4-33 l 5.4.7.2.6 Manual Actions 5.4-34 5.4.7.3 Performance Evaluation 5.4-37 5.4.7.4 Preonerational Testina 5.4-38 . 5.4.8 REACTOR COOLANT CLEANUP SYSTEM 5.4-43 i 5.4.9* M7.IN STEAM LINE AND FEEDWATER PIPING 5.4-43 l l 5.4.10 PRESSURIZER 5.4-43 l 5.4.10.1 Desian Bases 5.4-43 5.4.10.2 Description 5.4-44 5.4'.10.3 Evaluation 5.4-47 5.4.10.4 Tests and Inspections 5.4-47 5.4.11 PRESSURIZER RELIEF TANK 5.4-48 5.4.12 VALVES 5.4 48 5.4.12.1 Desian B3_s.is 5.4-48 ) 5.4.12.2 Denian Description 5.4-48
*See Chapter 10 9
Amendment B viii March 31, 1988
CESSAR E!!ac m. o O 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 Inspections 5.4-49 5.4.13- SAFETY AND RELIEF VALVES 5.4-49 i 5.4.13.1 Desian Basis 5.4-49 5.4.13.2 Description 5.4-49
]
l 5.4.13.3 Evaluation 5.4-50 5.4.13.4 Tests and Inspections 5.4-50 ; i 5.4.13.4.1 Pressurizer Safety Valves 5.4-50 5.4.13.4.2 Main Steam Safety Valves ti . 4 -5 0 5.4.14 COMPONENT SUPPORTS 5.4-51 5.4.14.1 Desian Basis 5.4-51 , 5.4.14.2 Description 5.4-51 j l 5.4.14.3 Evaluatial 5.4-52 l APPENDIX SA OVERPRESSURE PROTECTION FOR COMBUSTION SA-1 ENGINEERING SYSTEM 80 APPENDIX 5B STRUCTURAL EVALUATION OF STEAM LINE BREAK 5B-1 FOR STEAM GENERATOR INTERNALS , APPENDIX SC STRUCTURAL EVALUATION OF FEEDWATER LINE SC-1 BREAK FOR STEAM GENERATOR INTERNALS l E l l Amendment E ix December 30, 1988 L______-__-____-.
CESSARUnince l 9 LIST OF TABLES CHAPTER 5 Table Subiect 5.1.1-1 Process Data Point Tabulation 5.1.1-2 Design Parameters of Reactor Coolant 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 5.2-2 Reactor Coolant System Materials 5.2-3 code case Interpretations 5.2-4 Results of the Inadvertent Safety Injection Actuation Transient Analysis (for a Water-Solid E RCS) 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 l 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 5.4.2-1 Steam Generator Parameters l Amendment E l l x December 30, 1988
)
CESSAR Unincue. O LIST OF TABLES (Cont'd) CHAPTER 5 Table Subiect 5.'4.7-1 Shutdown Cooling Design Parameters 5.4.7-2 Shutdown Cooling System Interface Requirements for Component Cooling Water 5.4.7-3 Shutdown Cooling System FMEA 5.4.10-1 Pressurizer Parameters 5.4.10-2 Pressurizer Tests 5.4.13-1 Pressurizer Safety Valve Parameters 5.4.13-2 Main Steam Safety Valve Parameters i O l O xi
CESSAR MEncari:n 1 O l LIST OF FIGURES l CHAPTER 5 Eiqure 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) 5.2-1 System 80+ RCP Start Transient E 5.3-1 Typical Surveillance Capsule Assembly 5.3-2 Charpy, Flux and Compact Tension Compartment Assembly 5.3-3 Temperature, Flux, Tensile and Charpy Compartment Assembly E 5.3-4 Locations of Surveillance Capsule Assemblies 5.3-Sa System 80+ P-T Limit Curves, EOL (60 years) - Heatup 5.3-Sb System 80+ P-T Limit Curves, EOL (60 years) - Cooldown 5.3-6 System 80+ Allowable Heatup & Cooldown Rates 5.3-7 Reactor Vessel 5.4.1-1 Reactor Coolant Pump 5.4.2-1 Steam Generator 5.4.7-1 Shud a en Cooling System, Two Train Cooldown 5.4.7-2 Shu'. Dwn Cooling System, One Train Cooldown S.4.7-3 Shutdown Cooling System Flow Diagram, Shutdown Cooling Mode Amendment E xii December 30, 1988
CESSAR inL"icam. O LIST OF FIGURES (Cont'd) CHAPTER 5 Ficure Eyhiect 5.4.10-1 Typical Pressurizer 5.4.10-2 Typical Pressurizer Level Setpoint Program 5.4.10-3 Typical Temperature Control Program 5.4.10-4 Typical Pressurizer Level Error Program 5.4.10-5 Pressure Control Program 5.4.13-1 Primary Safety Valve 5.4.13-2 Main Steam Safety Valve 5.4.14-1 Reactor Coolant System Arrangement and Support i Points ] () 5.4.14-2 Reactor Vessel Supports 5.4.14-3 Steam Generator Supports 5.4.14-4 Reactor Coolant Pump Supports V xiii 1 I _ _ _ _ _ _ - - - - - - I
! CESSAR anece,. O v , i 5.0 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS l l 5.1
SUMMARY
DESCRIPTION The reactor is a pressurized water reactor-(PWR) with two coolant' loops. The reactor coolant system (RCS) circulates water in a ] closed cycle, removing heat from the reactor core and internals l and transferring it to a secondary (steam generating) system. . l The steam generators provide the interface between the reactor ! coolant (primary) system and the main steam (secondary) system. ] The steam generators are vertical U-tube heat exchangers with an I integral economizer in which heat is transferred from the reactor coolant to the main steam system. Reactor coolant is prevented from mixing with the secondary steam by the steam generator tubes and the steam generator tube sheet, making the RCS a closed system thus forming a barrier to the release of radioactive j materials from the core of the reactor to the containment building. The arrangement of the RCS is shown in Figures 5.1.3-1 and 5.1.3-2. The major components of the system are the reactor vessel; two parallel heat transfer loops, each containing one steam generator and two reactor coolant pumps; a pressurizer O connected to one of the reactor vessel outlet pipes; and associated piping. All components are located inside the containment building. Table 5.1.1-1 shows the principal pressures, temperatures, and flowrates of the RCS under normal steady-state, full-power operating conditions. Instrumentation provided for operation and control of the system is described in Chapter 7. System pressure is controlled by the pressurizer, where steam and water are maintained in thermal equilibrium. Steam is formed by energizing immersion heaters in the pressurizer, or is condensed by the pressurizer spray to limit pressure variations caused by contraction or expansion of the reactor coolant. The average temperature of the reactor coolant varies with power level and the fluid expands or contracts, changing the pressurizer water level. l The charging pumps and letdown control valves in the chemical and volume control system (CVCS) are used to maintain a programmed pressurizer water level. A continuous but variable letdown purification flow is maintained to keep the RCS chemistry within prescribed limits. A charging nozzle and a letdown nozzle are provided on the reactor coolant piping for this operation. The i chaiging flow is also used to alter the boron concentration or
) correct the chemical content of the reactor coolant.
l 5.1-1 l
CESSARHRL - O Other reactor coolant loop penetrations are the pressurizer surge line in one reactor vessel outlet pipe; the four safety injection inlet nozzles, one in each reactor vessel inlet pipe; two outlet nozzles to the chutdown cooling system, one in each reactor ! l vessel outlet pipe; two pressurizer spray nozzles; vent and drain connections; and sample and instrument connections. f Overpressure protection for the reactor coolant pressure boundary is provided by four spring-loaded ASME Code safety valves connected to the top of the pressurizer. These valves discharge to the in-containment refueling water storage tank, where the steam is released under water to be condensed and cooled. If the steam discharge exceeds the capacity of the in-containment refueling water storage tank, it is relieved to the containment atmosphere. Overpressure protection for the secondary side of the steam generators is provided by spring-loaded ASME Code safety valves located in the main steam system upstream of the steam line isolation valves. Components and piping in the RCS are insulated with a material compatible with the temperatures involved to reduce heat losses and protect personnel from high temperatures. Principal parameters of the RCS are listed in Table 5.1.1-2. Table 5.1.1-3 lists RCS volumes. Shielding requirements of the surrounding structures are described in Section 12.3. Reactor coolant system shielding permits limited personnel access to the containment building during power operation. The reactor vessel sits in a primary shield well. This and other shielding reduces the dose rate within the containment and outside the shield wall during full power operation to acceptable levels. 5.1.1 SCHEMATIC FLOW DIAGRAM The principal pressures, temperatures, and flow rates at major components are listed in Table 5.1.1-1. These parameters are referenced to Figure 5.1.2-1, the piping and instrument diagram, by numbered locations. Instrumentation provided for operation and control of the RCS is described in Chapter 7 and is indicated on Figure 5.1.2-1. O l Amendment B 5.1-2 March 31, 1988
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- n 5 6 0 o 1i o 2 0
4 6
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- p
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_ s 2 5 / / 0 _ s 2 6 N N 0 _ e 1 r P e _ t . ae _ tl _ w sb e o - a - t l yc r 1 a F di _ e - e R al t t2 r c- ep _ e n . , u w im t p m i1 e t o rp sa O _ a o . r a l t g r P5 u r F r e rt _ a s e h m , oo P a . sa p s/ ue FN sm _ t g ei m l t _ ai rs eF ab oa *
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CESSAR !!aL4mu o U ! TABLE 5.1.1-2 DESIGN PARAMETERS OF REACTOR COOLANT SYSTEM Design Thermal Power Mwt (Including Net Heat Addition from Pumps) 3817 Thermal Power, Btu /hr 1.303 x 10 10 (Developed by the RCS) , Design Pressure, psia 2500 Design Temperature (Except Pressurizer), *F 650 6 Coolant Flow Rate, lb/hr 165.8 x 10 E' Cold Leg Temperature, Operating, 'F 558 Average Temperature, Operating, 'F 587 3 Hot Leg Temperature, Operating, 'F 615 Normal Operating Pressure, psia 2250 System Water Volume, Ft3 (Without Pressurizer) 13,100 E l Pressurizer Water Volume, Ft3 (Full Power) 1200 0 Pressurizer Steam Volume, Ft3 (Full Power) 1200 1 l l O 1 1 l Amendment E December 30, 1988
1 CESSAREMMnc=< j i. TABLE 5.1.1-3 REACTOR COOLANT SYSTEM VOLUMES Component Volume (ft3 ) Reactor Vessel 5829.9 lD Steam Generators 2,800 each lB Reactor Coolant Pumps 134 each Pressurizer 2400 lB Piping - Hot leg 135.3 each Cold leg 214.1 each Surge Line (nominal) 43.6 O l j O Amendment D. September 30,.1988 ;
CESSAR EEnneuio G v 5.1.2 PIPING AND INSTRUMENT DIAGRAM Figure S.1.2-1 is the piping and instrument diagram . of the RCS. The. entire' system is located within the containment.- Fluid systems which are connected to the reactor coolant system and which are included within the limits of the reactor ' coolant pressure boundary, as defined in, ANSI- 51'.1-198 3 ' and 10 CFR 50.2 B
.(v), are identified and the appropriate piping and instrument I
I diagrams in other sections are referenced. Figure 5.1.2-2 is the piping and instrument diagram for the reactor coolant pumps. ; O l l 4 Amendment B 5.1-3 March 31, 1988
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OVERSIZE DOCUMENT PAGE PULLED SEE APERTURE CARDS NUMBER OF OVERSIZE PAGES FILMED ON APERTURE CARDS APERTURE CARD / NARD COPY AVAILABLE FROM QECORDS AND REPORTS MANAGEMENT BRANCH
m- q C E S S A R EEMnc m ., j O 5.1.3 ELEVATION DRAWINGS Reactor coolant system plan and elevation drawings are provided as Figures 5.1.3-1 and 5.1.3-2. Major components of the RCS are usually surrounded by concrete structures, which provide support plus shielding and missile protection. Elevation drawings, illustrating principal dimensions of the RCS in relationship to the surrounding building structures, will be presented in the site specific SAR. l O I O 5.1-4
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CESSARn'#,cmu O i 5.1.4 NUCLEAR STEAM SUPPLY SYSTEM. - BALANCE OF PLANT INTERFACE REQUIREMENTS
- B i Below are detailed the interface requirements that the Nuclear !
Steam Supply System (NSSS) places on certain aspects of the Balance of Plant, listed by categories. In addition, applicable { General Design Criteria (GDC) and Regulatory Guides, which C-E i utilizes in its design of the Reactor Coolant System (RCS), are presented. These GDC and Regulatory Guides . are listed only to show what C-E considers to - be relevant, and are not imposed as l interface requirements, unless specifically called out as such in a particular interface requirement. Relevant GDC: 1, 2, 3, 4, S, 14, 15, 26, 27, 28, 30, 31, 32, 33, 34, 35, 36, 37, 38, 39, 40, 41, 42, 43, 54, 55, 56, 57. Relevant Reg. Guides: 1.1, 1.2, 1.4, 1.14, 1.24, 1.26, 1.29, 1.31, 1.34, 1.36, 1.38, 1.43, 1.44, 1.45, 1.47, 1.49, 1.50, 1.54, 1.61, B 1.64, 1.65, 1.71, 1.73, 1.74, 1.79, 1.83, 1.84, 1.85, 1.89, 1.97, 1.99, 1.100, 1.133, 1.147, 1.150,- 1.151, l E O 1.153. A. Power See Chapters 7 and 8 for power information. B. Protection from Natural Phenomena
- 1. The containment shall remain functional for the full range, per GDC 2, of natural phenomena (earthquakes, uornadoes, tornado missiles, flooding conditions, hurricanes, winds, snow, and ice) and external environmental conditions.
- 2. The steam piping and associated supports from the steam generators up to and including the Main Steam Isolation Valves (MSIVs), the ADVs and their associated isolation valves, and any auxiliary steam supply systems up to the isolation valves which connect upstream of the MSIVs shall be seismic category I and designed to ASME B&PV Code, Section III, Class 2 requirements.
- BOP Interface Requirements (IR's) are being replaced by E detailed descriptions for the System 80+ Standard Design. For continuity, the IR's will be retained until the SFD's have been
% completed. B Amendment E 5.1-5 December 30, 1988
CESSAR !!ancmou 9
- 3. The valves, piping, and associated supports of the Feedwater System from and including the Main Feedwater Isolation Valves (MFIVs) to the steam generator feed nozzles shall be Seismic Category I and designed to ASME B&PV Code Section III, Class i 2 requirements.
- 4. All components and piping of the Emergency Feedwater System between the steam generators and the containment isolation valves shall be Seismic Category I and designed to ASME B&PV Code Section III, Class 2 requirements.
- 5. All components, piping and associated supports in the condensate storage facilities for Emergency Feedwater shall be Seismic Category I and designed in accordance with ASME B&PV Code Section III, Class 3.
- 6. All components and piping associated with steam generator blowdown between the steam generator and the containment isolation valves shall be Seismic Category I and designed to ASME B&PV Code Section III, Class 2 requirements.
C. Protection from Pipe Failure
- 1. The following valves shall be protected against internally generated missiles or the effects resulting from a high energy pipe rupture (e.g.,
pipe whip, jet impingement and steam environment) such that these events will not prevent the valves from performing their requisite safety functions.
- a. MSIVs
- b. Secondary Safety Valves.
- c. Atmospheric Dump Valves (ADVs).
- d. MSIV Bypass Valves.
- c. MFIVs.
- f. Blowdown Isolation Valves.
B
- 2. The MSIVs shall be supported such that the valve body and actuator will not be distorted or displaced as a result of pipe break thrust loadings to such a degree that the valve cannot close.
Amendment B 5.1-6 March 31, 1988
CESSAR inMncui:,. O !
- 3. Feedwater piping shall be . routed, protected ar.d l restrained such that'in the case of a rupture of a feedwater line or any other system pipeline, a single failure criteria will not be exceeded with ;
regard to safe shutdown of the plant.
- 4. A containment shall- be provided to limit the release of energy and radioactivity to the environs in the event of a rupture of the RCS and to protect the public health and safety. l j
- 5. The containment, including penetrations, shall not be subject to loss of function from dynamic j effects (e.g., missiles, pipe reactions, fluid l reaction forces) resulting from failure of RCS equipment or piping within the containment.
- 6. The design pressure and temperature of the-containment shall, as a minimum:
i
- a. Be equal to the peak pressure and temperature l resulting from either (1) complete blowdown l of the reactor coolant through any rupture of the RCS piping, up to and including a postulated double-ended severence of the largest reactor coolant pipe or, (2) a compaete blowdown of the unisolated steam generator plus attached steam lines up to the respective main steam isolation valves through any rupture of the steam line piping, up to and including a postulated double-ended severance of the largest main steam. line pipe, assuming a sequence of events for either case which leads to the peak transient accumulation of energy in the building atmosphere. To meet this end, a spectrum of loss-of-coolant accidents (LOCA) and main steam line breaks (MSLB) have been analyzed.
They shall be used by the applicant to establish the design pressure and temperature of the containment. (Refer to Sections 6.2.1.3 and 6.2.1.4).
- b. Take into account all credible post-blowdown energy additions to the- containment atmosphere, such as core residual heat, thin and thick structural metal stored energy, steam generator reverse heat transfer, O metal-water reactions and other possible chemical reactions loss-of-coolant accident..
resulting from a 5.1-7
CESSARH5 Gem O
- 7. Compartments within the containment including the reactor vessel cavity shall be designed for the maximum pressure differential between the compartment and the remainder of the containment based on the maximum RCS pipe break that can occur in the compartment as defined in Section 3.6.
D. Missiles
- 1. The RCS, which is a potential source of missiles, shall to the extent possible, be either surrounded by barriers or restrained to prevent missiles from reaching other parts of the RCS, the containment lines, the secondary steam and feedwater piping or the engineered safeguards systems. See Section 3.5 for additional discussion of missiles.
- 2. A containment structure shall be provided to protect the RCS from loss of function due to missiles generated outside the containment, including those resulting from equipment failure, and weather induced forces such as tornadoes and hurricanes.
E. Egparation
- 1. Adequate physical separation shall be maintained between the redundant electrical and instrumentation systems used for emergency control and safe shutdown of the reactor, and between the multiple instrumentation channels in the Plant !
Protection System.
- 2. Each MSIV shall have two physically separate and electrically independent closure solenoids in order to provide redundant means of valve operation. A Main Steam Isolation Signal (MSIS) shall be provided to each solenoid.
- 3. Redundant feedwater system isolation valves in .
cach feedwater line meeting the single failure criteria shall be provided in piping a interconnecting the steam generators to preclude blowdown of both steam generators following a pipe rupture.
- 4. Each ADV shall be provided with an isolation valve in the piping which connects each ADV to the main steam lines.
5.1-8
CESSAR EEnncamn G V F. Independence
- 1. The provisions of General Design Criteria 54 and 57 for containment isolation valves shall be met.
- 2. The feedwater system piping, Emergency Feedwater System piping, and main steam piping and all of their associated supports and restraints shall be designed so that a single adverse event, such as a ruptured feedwater line, emergency feedwater line, main steam line inside containment, or a closed isolation valve can occur without:
l a. Initiating a Loss-of-Coolant incident.
- b. Causing failure of the other steam generator #s safety class steam and feedwater lines, MSIVs, safety valves, MFIVs blowdown line isolation valves, or ADVs.
- c. Reducing the capability of any of the Engineered Safety Features systems or the Plant Protective System.
CI d. Transmitting excessive loads to the containment pressure boundary. I e. Compromising the function of the plant , control room, f '
- f. Precluding orderly cooldown of the RCS.
- 3. An electrica.1 or mechanical malfunction of one solenoid shall not prevent a MSIV from closing.
lB l
- 4. The ADV control circuits shall be designed or i precautions taken, such that no single electrical failurewouldresultintheopeningofvalveswitg a total combined capacity greater than 1.9 x 10 i
lb/hr at 1000 psia.
- 5. Each MFIV actuator shall be physically and lB electrically independent of the other such that failure of one will not cause failure of the other.
- 6. No single active or passive component failure, single passive or active electrical component
/O failure, or power supply failure shall preclude adequate operation of the Emergency Feedwater Amendment B 5.1-9 March 31, 1988
I CESSAR Hnincua, j Oii System (acceptable guidelines for implementing ! these criteria can be found in ANS-58.9-1981), l assuming the following events:
- a. Loss of normal feedwater with or without e concurrent loss of normal onsite or offsite !
AC power.
- b. Minor secondary system pipe breaks with or without a concurrent loss of normal onsite or offsite AC power.
- c. Steam generator tube. rupture with or without a concurrent loss of normal onsite or offsite AC power.
- d. Major secondary system pipe breaks with or without a concurrent loss of normal onsite or offsite AC power,
- e. Small LOCA with or without a concurrent loss of normal onsite or offsite AC power.
- 7. The ability of the Emergency Feedwater System to perform its design function considering a power supply failure, a single active or passive mechanical component failure, a single active or passive failure of an electrical component, or the effects of a high or moderate energy pipe rupture shall be demonstrated. Acceptable guidelines for implementing these criteria can be found in ANS-58.9-1981.
B. Deleted. D
- 9. Blowdown piping exiting containment shall have redundant blowdown line isolation valves which i shall be actuated by an Emergency Feedwater l Actuation Signal (EFAS).
G. Thermal Limitations
- 1. The Component Cooling Water System (CCWS) shall lD provide cooling water to each RCP as shown in Figure 5.1.2-2.
- 2. RCP heat load and flow data presented in Table 5.1.4-1 shall be utilized in the design of the CCWS.
O Amendment D 5.1-10 September 30, 1988 l l
CESSARHMLm,, I o 3. The maximum and minimum temperature of the component cooling water during normal operation shall be 105'F and 65'F respectively. B 4 Following the events stated in Section 5.1. 4. F. 6, the Emergency Feedwater System shall maintain adequate inventory in the steam generator (s) for residual heat removal and be capable of the following:
- a. Maintaining the NSSS at hot standby with or without normal offsite and normal onsite power available.
i
- b. Facilitating NSSS cooldown at the maximum administratively controlled rate of 75'F/hr from hot standby .to shutdown cooling initiation with or without normal offsite or onsite power availabla. '(The Shutdown Cooling System becomes available for plant cooldown when the RCS temperature and pressure are reduced to 350*F and 400 psia.) approximatelylD
- 5. The Emergency Feedwater System shall be available to deliver flow to the steam generator (s) automatically upon receipt of an EFAS as follows:
- a. Within 60 seconds when' normal offsite or lD normal onsite powe-.is available.
b. Within normal 60 seconds offsite .when power areboth not normal onsite and lD available. I
- 6. Deleted. l
- 7. Deleted. E
- 8. Deleted.
l l i Amendment E 5.1-11 December 30, 1988 j
i CESSAR M5Mena O
- 9. Each MSIV leak flow shall not exceed 0.001 percent of nominal flow at 1200 psia in the forward ;
direction and shall not exceed 0.1 percent of B nominal flow at 1200 psia in the reverse direction. l l
- 10. No single MSIV bypass valve or bypass valge line i shall have a capacity greater than 1.9 x 10 lb/hr )
of saturated steam at 1000 psia. l I
- 11. No single turbine bypass valvg shall have a capacity greater than 1.9 x 10 lb/hr at 1000 psia. l
- 12. The total reverse leak rate of feedwater check valves to each steam generator shall not exceed l 1000 cc/hr. I H. M_on itorinct
- 1. Means shall be provided for detection of reactor )
coolant leakage into the secondary side of the 1 steam generators and cooling water systems ! associated with components containing reactor ) l coolant. l
- 2. Applicant supplied component designs and RCS construction procedures shall ensure that RCS leakage from known sources will not exceed 10 gpm; j from steam generator tubes will not exceed 1.0 gpm; and from unknown sources will not exceed 1 gpm, to minimize in-plant airborne and surface activity levels and activity releases to the environs at system normal operating temperature and pressure.
- 3. The required accuracy of the feedwater temperature B measurement devices shall be 5.5'F for any calorimetric measurement.
I. Operational / Controls
- 1. A power-operated MSIV capable of establishing shutoff under conditions of design pressure, design temperature, and flow conditions resulting from a break just upstream or downstream shall be provided in each main steam line outside of containment.
O Amendment B i 5.1-12 March 31, 1988 l 1
CESSARMuh . _m V) B l l
- 2. The MSIV and MSIV bypass valve shall be either a fail close valve or a valve that is shown by the applicant to close upon receipt of a MSIS.
- 3. The full open to close stroke time of each MSIV B and MSIV bypass valve shall be 5.0 seconds or less upon receipt of an MSIS.
- 4. The ADVs shall be fail close and shall be capable of being remote manually positioned to control the plant cooldown rate.
IB
- 5. In the combined event of either a steam line break or steam generator tube rupture and the loss of power operation of the ADVs, personnel access to the manual operators of the intact valves on the other steam generator shall be possible.
IB
- 6. Redundant feedwater system isolation valving shall I
be provided in both the economizer feedlines and the downcomer feedlines such that the following criteria are met when the effects of single (s
\
failure criteria are imposed:
- a. Complete termination of forward feedwater flow is assumed within 5.0 seconds after lB receipt of an MSIS.
- b. Abrupt complete termination of reverse feedwater flow with the existence of a reverse flow condition. Check valves are considered to be an acceptable means of achieving the above.
- 7. The economizer and downcomer feedwater line isolation valves (MFIVs) in each main feedwater l line shall be remote-operated and be capable of ;
maintaining leak rate of less than 1000 cc/hr j main l under the feedwater line pressure, l temperature and flow resulting from the transient l conditions associated with a pipe break on either side of the valves. IB
- 8. Personnel access to the isolation valve upstream of the ADV shall be possible at all times during operation, f3
]
Amendment B 5.1-13 March 31, 1988 l __ ___-
CESSAR Uninema O
- 9. If the isolation valves upstream of the ADVs are electrically controlled and operated, the valve operator and control systems shall be designed to the same IEEE standards as applied to the ADVs.
J. Insocction and Testina
- 1. All ASME B&PV Code, Section III, Class 1 and 2 valves shall be designed, fabricated and installed such that they are capable of being periodically tested in accordance with ASME Code, Section XI.
- 2. Adequate clearances shall be provided for inservice inspection of the Reactor Coolant Pressure Boundary and the ASME B&PV Code Section III, Class 2 portions of the Main Steam, Main Feed, Emergency Feed, and Blowdown systems' piping, in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.
- 3. Biological shielding and all other insulation, if installed around the Reactor Coolant Pressure Boundary, shall be designed to afford access for inservice inspection as defined by Section XI of the ASME Boiler and Pressure Vessel Code.
- 4. The pressurizer manway shall be accessible for internal examination of the pressurizer.
K. Chemistry /Samplina
- 1. A sampling system which provide a means of obtaining remote liquid samples from the RCS for chemical and radiochemical laboratory analysis shall be provided. The sampling system shall be designed to allow for the following tests:
corrosion product activity levels, dissolved gas, fission product activity, chloride concentration, coolant pH, conductivity levels and boron concen-tration. The pressurizer steam space sample lines shall contain 7/32" x 1" orifice as close to the pressurizer as possible. The sample system shall be as shown on Figure 5.1.2-1.
- 2. A system or systems shall be provided to maintain the steam generator secondary water chemistry I within Section 10.3.5 specifications during plant D O{
Amendment D l I 5.1-14 September 30, 1988
CESSAR EE"icm:n O v operation. The system or systems shall incorporate steam generator blowdown, chemical addition, and monitoring.
- 3. Provisions shall be made '.o allow sampling of the RCS during Shutdown Coolirig System operation.
- 4. Provisions shall be made to allow sampling of the RCS during startup.
L. Materialg
- 1. The materials used for the containment and its internal structures shall be compatible with both the normal operating environment and the most i severe thermal, chemical, and radiation l
environment expected during post-accident conditions (refer to Section 3.11 for the environmental parameters). Consideration shall be given to compatibility with spray water chemistry l and recirculating water chemistry to ensure that containment materials will withstand this exposure without causing deleterious or undesirable
+
reactions, or significantly altering the existing water chemistry of recirculating ECCS water. i
- 2. The following elements and components shall not come in contact with surfaces which will later be in contact with reactor coolant, at any stage of manufacture, assembly or inspection. These are:
(a) lead or lead compounds, (b) mercury or mercury compounds, (c) halogen containing solvents or other halogen compounds.
- 3. The use of the following materials shall be minimized on surfaces normally in contact with reactor coolant:
- a. sulfonated cutting oils,
- b. zinc metal or zinc compounds,
- c. magnesium metal, j
- d. asbestos,
- e. aluminum, V)
I 5.1-15 1
CESSAR nainemo. O
- f. copper acid etchants,
- g. penetrants.
If the above materials are intended to be used, the use shall first be approved by C-E.
- 4. The sample lines in contact with the reactor coolant, including welds shall be designed such that the material is compatible with the reactor D
coolant water chemistry described in Section j 9.3.4.
- 5. Construction materials or protective coatings containing low melting point elements, particularly lead, mercury and sulfur, shall not be used if they could come in contact with the secondary systems. This is required to reduce to a minimum the potential for stress corrosion cracking of Inconel material in the steam generators.
- 6. The secondary system piping shall be designed to allow cleaning for the removal of foreign material and rust prior to operation and to prevent introduction of this material into the steam generator. Chemical cleaning or hand cleaning may be employed. During chemical cleaning, no fluid shall enter the steam generators. Suitable bypass piping shall be provided if required.
- 7. Non-metallic insulation used on the Reactor Coolant Pressure Boundary shall conform to Regulatory Guide 1.36. The chloride and fluoride l content of the non-metallic insulation shall be in the acceptable region as shown in Regulatory Guide 1.36. Tests shall be made on representative samples of the non-metallic thermal insulation shall be demineralized or distilled water.
- 8. No contaminants, except for cutting oils, shall be left on any RCS component surface except for the time required to perform and evaluate the particular fabrication or inspection operation.
- 9. Field welding of the RCS piping assemblies and components shall be done in accordance with a welding procedure or procedures by welders qualified to ASME Section IX requirements.
Amendment D 5.1-16 September 30, 1988
CESSARinnnc-l' M. System / Component Arranctement
- 1. The pressurizer and surge line shall be. located entirely above the reactor coolant loops. The surge line shall be continuously rising from the hot leg nozzle to the pressurizer, thus ensuring that the line contains no water traps. B
- 2. The pressurizer surge line shall be sized and arranged to minimize the flow resistance. The surge line L/D shall not exceed 210, assuming a 12 lE inch, Schedule 160 pipe. The L/D statement above includes the effective L/D of all piping elbows but does not include the surge line entrance and B exit losses. The surge line must continuously rise from the top of the hot leg nozzle to the pressurizer. lE ,
- 3. The maximum acceptable pressure drop through the pressurizer spray line piping is 19 psi at a total flow rate of 375 gpm and at a water temperature of 565*F. This requirement is for the piping only, allowance does not have to be made for elevation (n) losses, the valves, or for the entrance _and exit nozzles.
- 4. Flooding of the reactor cavity from systems other than the reactor coolant and safety injection 3
systems shall be precluded to prevent immersion of the reactor vessel during operation. This is normally accomplished by routing only reactor coolant and safety injection system piping inside h the reactor cavity, by minimizing drainage paths to the reactor cavity, and/or providing gravity drainage paths out of the cavity below the bottom head of the vessel. The combined reactor cavity and ICI chase may be designed without gravity I drainage paths below the- hot and/or cold leg pipe penetrations, thereby allowing the reactor cavity to flood in the event of a breach of the reactor coolant pressure boundary inside the cavity.
- 5. The RCS sample piping shall be designed so that the overall transient time from the loop to the containment wall is approximately 90 seconds to permit the decay of short-lived radionuclides (high energy nuclides such as N-16).
1 1 O 1 Amendment E l 5.1-17 December 30,_1988
CESSAREBUncm O\
- 6. The RCS and main steam piping, MSIVs, primary and secondary safety valves and their discharge piping and ADVs shall be arranged and supported such that the limiting loads are not exceeded for normal and relieving conditions.
- 7. Following a secondary line break, either all steam paths downstream of the MSIV's shall be shown to be isolated by their respective control systems following a MSIS actuation signal, or the results of a blowdown through a non-isolated path shall be shown to be acceptable. An acceptable maximum steam flow from a non-isolated steam pagh is 10%
of the main steam rate (MSR) (1.9 x 10 lb/hr 0 1000 psia saturated steam). It is not required that the control systems for downstream valves nor the downstream valves themselves be designed to IEEE 279 and IEEE 308 or ASME Code, Section III and Seismic Category I criteria respectively.
- 8. The MSIVs for each steam generator shall be arranged such that a maximum of 2000 cubic feet (total for two steam lines per steam generator) is contained in the piping between each steam generator and its associated MSIVs. This volume shall include all lines off of the main steam line up to their isolation valves.
- 9. The main steam lines shall be arranged such that a maximum of 14,000 cubic feet is contained between the MSIVs and the turbine stop valves. This volume shall include all lines off of the main steam line up to their isolation valves.
- 10. The main steam lines shall be headered together prior to the turbine stop valves but not upstream of the MSIVs, and a crossconnect line shall be provided which will maintain steam generator pressure differences within the following limits for all normal and upset conditions,
- a. 0-15% power operation pressure difference to be 1 psi.
- b. 15-100% power operation pressure difference to be 3 psi.
- 11. No automatically actuated valves shall be located i upstream of the MSIVs except as required for ,
supply to steam driven emergency feedwater pumps. Provisions shall be made to prevent blowdown of 5.1-18
I CESSAR Ence i g 4 V both steam generators through the emergency feedwater supply headers in the event of a steamline break. The6 maximum all wable flow rate per valve is 1.9 x 10 lb/hr.
- 12. There shall be no isolation valves in the main steam lines between the steam generators and the secondary relief valves.
- 13. The main steam safety valves shall be arranged such that any condensate in the line between the safety valves and main steam line drains back to the main steam line.
- 14. All valves in the main steam line outside of containment up to and including the MSIVs shall be located as close as practical to the containment wall.
- 15. A 90* or 45' clbow facing downward shall be attached to each feedwater nozzle. Such a precaution will aid in the prevention of water n hammer.
- 16. The MFIVs shall be located outside of the containment building as close to the containment wall as possible.
- 17. The MFIVs for each steam generator shall be arranged such that a maximum of 500 cubic feet of fluid is contained in the piping between each steam generator and its associated isolation valves. This volume shall also include the volumes between the redundant MFIVs. This volume shall include the volumes up to their respective isolation valves of all lines off of the main feedwater lines downstream of the MFIVs for which a mechanism exists for getting the fluid into the main feedwater line (e.g., gravity, flow or flushing).
- 18. The Emergency Feedwater System connection shall be located in the downcomer feedwater line between j the MFIVs and the steam generator downcomer l nozzle. Emergency feedwater flow shall be directed to the downcomer nozzle only. A safety Class 2 check valve shall be located in the main
, , feedwater piping upstream of this interface to l ( 4 prevent back flow of emergency feedwater to other L/ portions of the Main Feedwater System. 5.1-19
CESSAR 8Bi*imeu O.
- 19. Protection shall be provided from internally generated flooding that could prevent the performance of safety related functions.
- 20. The spray line piping shall be arranged to I preclude the potential for waterhammer. )
i
- 21. The piping from the pressurizer to the main and I auxiliary spray systems shall be routed t B minimize the length which could be exposed to l pressurizer steam and shall minimize horizontal j piping subjected to thermal stratification in the !
line. (Minimize horizontal run near the top of 1 the pressurizer.)
- 22. Loop seals shall not be utilized in safety valve ,
inlets. j N. Ra. slioloaica1 Waste j 1 i
- 1. Actuator-operated valves in the Reactor Coolant Pressure Boundary shall be supplied with double packing with lantern ring and leakoff connection j unless they are diaphragm (packless) type. l Leakoffs shall be piped to the reactor drain tank i or other radwaste collection system.
- 2. Provisions shall be made to process the continuous steam generator blowdown water. If separately j provided, the radioactive steam generator blowdown i processing system shall include filtration and ion l exchange or equivalent processes. With design operating conditions in the steam generator, the 3' blowdown water radioactivity will decrease by 90%.
O. Overpressure Protection
- 1. Each primary safety valve inlet line shall be I designed to pass 125 percent of the minimum required safety valve capacity of 460,000 lb/hr with a maximum pressure drop of 50 psi. This pressure drop of 50 psi is for piping and nozzle !
losses. (Pressure loss factor for pressurizer l nozzle is K = 0.23 based on 6" Schedule 160 pipe.)
- 2. Each primary safety valve discharge line shall be designed to pass 125 percent of the minimum B required safety valve capacity (460,000 lb/hr per valve). The safety valve discharge flow analysis Amendment B 5.1-20 March 31, 1988 u _.
i CESSAR !!3H?icui:,i I i l chould include the effects of line heat loss with the resulting two-phase flow and two-phase pressure drop in the discharge line. The safety ; valve discharge lines will collect into a common ' line and be routed to the in-containment refueling , water storage tank. For the common discharge ? line, the minimum safety valve flow is 1,840,000 3 ( lbm/hr (total flow for four valves) while the total maximum flow is 2,300,000 lbm/hr (1.25 x , l 1,840,000 lbm/hr) . The in-containment refueling j water storage tank design pressure is 15 psia. lE ] NOTE: To assure proper safety valve operation, the safety valve back pressure shall not B exceed 700 psig.
- 3. Each main steam line shall be provided with ASME Codo, springloaded secondary safety valves between the containment and the isolation valves.
- 4. The total relieving capacity of the secondary safety valves shall be equally divided between the main steam lines.
O, 5. The total secondary safety galve capacity shall be sufficient to pass 19 x 10 lb/hr at the maximum valve set pressure.
- 6. The maximum steam flow per secondgry safety valve shall be no greater than 1.9 x 10 lbs/hr at 1000 psia.
- 7. Deleted.
E I Amendment E 5.1-21 December 30, 1988
l CESSAREna m O'
- 8. The design pressure, temperature, and flow rating of the main steam piping and valves shall be greater than or at least equal to the design pressure, temperature, and flow rating of the steam generator secondary side.
P. [Lelated Service
- 1. The pressure and thermal transients described in Section 3.9.1.1 shall be utilized in the design of those portions of the RCS not within the CESSAR design scope.
- 2. The systems or portions of reactor coolant pressure boundary outside of the CESSAR design scope shall be Safety class I unless the ;
conditions of 10 CPR 50.55A are met.
- 3. A fire protection system shall be provided to protect the RCS consistent with the of GDC 3 and 5 and, shall include as requirements a minimum, lD the following features:
- a. Facilities for fire detection and alarming.
- b. Facilities for methods to minimize the probability of fire and its associated effects,
- c. Facilities for fire extinguishment,
- d. Methods of fire prevention such as use of fire resistant and non-combustible materials whenever practical, and minimizing exposure .
of combustible materials to fire hazards. I
- e. Assurance that fire protection systems do not adversely affect the functional and structural integrity of safety related structures, systems, and components.
- f. Fire protection systems shall be designed to assure that their rupture or inadvertent i operation does not significantly impair the i capability of safety related structures, l systems, and components.
- 4. Systems shall be provided for the detection of )
reactor coolant leakage from unidentified sources. l Amendment D 5.1-22 September 30, 1988 l
CESSAREn h a 1 m I
- 5. If air-operated ADVs are used, a safety related control air system shall be provided to supply air to the ADV actuators should the normal air supply be unavailable.
- 6. Air for the ADV and MFIV pneumatic valve operator l shall be clean, dry and oil-free. The air shall be delivered at the point of use under system full flow conditions at a pressure of 70 psig minimum to 105 psig maximum. Pneumatic lines and fittings i shall have a minimum design pressure of 150 psig.
l l.
- 7. The containment structure shall be designed and sized to accommodate the Reactor Coolant System arrangement shown in Figures 5.1.3-1 and 5.1.3-2.
Q. Environmental
- 1. For the NSSS components supplied by the site operator one of the following options shall be D i followed.
O a. Demonstration of other environmental qualification envelopes for any or all of ' these buildings not to exceed the qualifi-cation envelopes of Section 3.11. l
- b. Exclusion of specific components from extreme environmental conditions by suitable physical separations or environmental control system techniques,
- c. Use of the same environmental qualification conditions being employed by C-E supplied NSSS components. )
1
- 2. The containment pressure and temperature transients resulting from the LOCA shall meet criteria specified in Section 6.2.1.5.
- 3. A containment ven*.ilation system shall be provided ,
to handle- the total -RCS heat losses to containment. Table 5.1.4-2 lists the heat loads from NSSS support structures -to containment. Table 5.1.4-3 lists typical loads through the NSSS insulation to containment. These values will be confirmed by each site operator since the final value depends on system insulation efficiency. lD i Y) Amendment D 5.1-23 September 30, 1988 E . . - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
CESSAREnGem O R. Mechanical Interaction Between Components l
- 1. The following components shall be designed to j withstand the loads arising from the various ;
normal operating and design basis events.
- a. The main steam piping, supports and l restraints. !
- b. The steam generator steam and feedwater nozzles.
- c. The MSIVs and the MSIV bypass valves and j supports.
- d. Main Steam Safety Valves,
- e. The main feedwater piping, supports and restraints.
- f. MFIVs and supports.
- g. Blowdown piping, supports and restraints.
- h. Blowdown isolation valves.
- 2. Structures shall be provided to mate with component supports to restrain and support RCS components. The loading conditions specified in l Section 3.9.3.1 shall be utilized in the design of l the supporting structures. The loads et the B support / structure interface locations under normal, upset, emergency, faulted, and test i conditions, will take into account the local characteristics of the specific structures at the support / structure interfaces.
- 3. The loadings imposed by connecting system piping on RCS nozzles under normal, upset, emergency, faulted, and test conditions shall be less than the design loads for these nozzles. C-E will confirm using the loads developed by the Applicant that the piping nozzles are within Code allowable stress limits. j O
Amendment B 5.1-24 March 31, 1988
CESSAR E!nincuiu O TABLE 5.1.4-1 RCP COOLING WATER SYSTEM DATA Heat Load. Flow Pressure Component Number (Each) - 8tu/hr (Each) - com Drop (osi) RCP Oil Coolers 4 400,000 60 10 RCP High Pressure Cooler 4 187,00* 75 20 RCP Seal Cooler 4*** 31',800 17.6 7.2 RCP Motor 011 Cooler 4 153,000 48 20** RCP Motor Air Cooler 4 1,610,000 300 20**
- If injection water is . lost the heat load increases to 810,000 Btu /hr.
The flow of 75 gpm is sufficient for the' loss of' injection water
' condition.
** The combined pressure drop including piping will not exceed 20 psig.
*** Four sets of two coolers per pump. Data list applies to each of the four two-cooler combinations.
O O
CESSAR EE"ioun
,~ '% j' IABLE 5.1.4-2 HEAT LOADS FROM NSSS SUPPORT STRUCTURES
- Heat Load Per Heat Load Per Compom nt Component (BTU /hr) Plant (BTU /hr)
Reactor Vessel Support (nozzle pad and vertical column) 62,500 62,500 D Structural Interface at Reactor Vessel Column Base Plate 13,750 13,750 B Steam Generator Lower Support System (skirt and sliding base) 27,500 55,000 D Reactor Coolant Pump Support System (support skirt and 4 vertical and 2 horizontal columns at the skirt only) 40,000 160,000 Pressurizer Upper Key - Support Structure 13,250 13,250 l (3 ( ) Steam Generator Upper Key - Support Structure 14,500 29,000 lD Reactor Coolant Pump Snubbers 4,000 16,000 Steam Generator Snubbers 12,000 24,000 i lD RCP Upper Horizontal Supports 4,000 16,000 1
- For C-E supplied components.
l l
,-m, '%s Amendment D September 30, 1988
~
3 CESSAR !!nincuia q l LJ TABLE 5.1.4-3 RCS INSULATION HEAT LOADS Heat Load Per Heat Load Per ComDonent CoEDonent (BTU /hr) Plant (BTU /hr) Reactor Vessel Closure Head 45,200 45,200 Reactor Vessel and Botton Head 187,400 187,400 Pressurizer 134,420 134,420 lB Steam Generator (primary head only) 56,900 113,800 lD l Reactor Coolant Pump Casing 15,500 62,000 Reactor Coolant System Piping 213,200 213,200 O l l o Amendment D September 30, 1988
CESSAREEMace O 5.2 LN1JLGERY OF REACTOR COOLANT PEE 88URE BOUNDARY This section discusses the measures employed to provide and maintain the integrity of the Reactor Coolant Pressure Boundary (RCPD) throughout the facility's design lifetime. The RCPB is defined in accordance with ANSI /ANS 51.1-1983. Included are all lB pressure containing components such as pressure vessels, piping, pumps, and valves which are: A. Part of the Reactor Coolant System, or B. Connected to the Reactor Coolant System, up to and including the following:
- 1. The outermost containment isolation valve in piping which penetrates the containment;
- 2. The second of two valves normally closed during reactor operation in piping which does not penetrate the containment.
5.2.1 COMPLIANCE WITH CODES AND CODE CASES 5.2.1.1 Compliance with 10 CFR 50.55a The codes and component classifications are listed in Table 5.2-1 . and are in accordance with the provisions of 10 CFR 50.55a. l I 5.2.1.2 Applicable Code Cases j j Reactor Coolant Pressure Boundary components are fabricated in accordance with the ASME Code, Section III. Code Case interpretationsSAR. siite-specific applied to these components will be listed in the lD Unless otherwise stated in the site-specific SAR, Combustion lD Engineering, Inc., will comply with Regulatory Guides 1.84, 1.85 and 1.147 in determining suitable ASME Code Cases. Code Cases not included in the Regulatory Guides 'may be used with specific ln i authorization from the Commission under 10 CFR 50.55a. I 5.2.2 OVERPRESSURE PROTECTION
]
5.2.2.1 Desian Bases I Appendix SA presents .the design bases for sizing the ! overpressurization protection system. The loss of load transient l which is used to size the primary safety valves is not intended to be used as a design transient for any other NSSS equipment. Amendment D - 5.2-1 September 30, 1988 1 L____________.______._________._.__ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ ._ _ _ _ _ . _ _ _ _ _ _ _ _ ._____n
e u CESSAR nai?icariou O 5.2.2.2 Reg {gn EvaluatioD Section 15.2 of Chapter 15 provides the functional . design \ evaluation of the overpressurization protection system. analysis, the adequacy of the overpressure protection system to In this maintain secondary and primary operating pressures within 110% of design is clearly demonstrated for the loss of load analysis. The analytical model used in the analysis is discussed in Chapter 15. Chapter 15 lists the assumptions used in the turbine trip lE analysis. These assumptions are chosen so that they tend to maximize the required pressure relieving capacity of the primary j and secondary valves. The analysis demonstrates that sufficient relieving capacity has been provided so that when acting in conjunction with the reactor protective system the safety valves will prevent the pressure from exceeding 110% of the design pressure. 5.2.2.3 Eiping_and Instrumentation Diagrams l The piping and instrumentation diagram showing the primary safety valves and their discharge lines is given in Figure 5.1.2-1.. The piping and instrumentation diagram showing the reactor drain tank is given in Figure 9.3.4-1, Sheet 2. The secondary safety valves are discussed in Section 10.3. E 5.2.2.4 Equipment & Component Description The primary safety valves are direct acting, spring loaded, stainless steel valves with enclosed bonnets. These valves are mounted on the top cf the pressurizer. For further description of these valves, refer to Section 5.4.13. A schematic drawing of the primary safety valve is given in Figure 5.4.13-1. Valve parameters are given in Table 5.4-13. i Primary safety valve operation is characterized by a sharp pop at the set pressure. This sharp opening is produced by two stages of reaction working together to produce a continuous action. The initial lift is produced when the steam pressure under the disc exceeds the spring force. The escaping steam reacts against the upper guide ring and pushes the disc up to a high lift. The reaction of the deflected steam against the underside of the disc lifts it higher on an accumulation of pressure. The valve reaches a lift in excess of full bore lift within an accumulation of 3 percent above the set pressure. As the system pressure drops, the valve disc settles to a moderate lift, and closes sharply with a blowdown of approximately 18.5 percent of the set pressure. Amendment E i 5.2-2 December 30, 1988
1 l CESSAR HL"icari:. O O : 5.2.2.4.1 Transients The primary safety valves will be designed to withstand the following transients without failure or malfunction: 1 A. 650*F to 375*F in 50 seconds and return to 650*F in 2000 I seconds for 8 cycles (Loss of secondary pressure). ] B. 100*F to 400*F and return to 100*F at a rate of 100*F/hr with concurrent pressure changes from 400 psig to 2250 psig l and returning to 400 psig in step changes for 300 cycles. ! (Plant leak test). ! C. i 10'F step change from 653*F for 1.5 x 10 6 cycles. (Plant loading and unloading, i 10% step load, normal plant variation). D. 75*F to 653*F and return to 75'F at a rate of 200*F/hr with I pressures at saturation levels for 750 cycles. (Plant heat up and cool down). Note: Heat up and cool down are separate transients, each ( beginning at steady state conditions. f (' E. Pressurize to 1. 5 times set pressure at 100
- F-200
- F for 15 l cycles plus number of hydros conducted prior to valve shipment. (Hydrostatic test).
F. 720 cycles from closed to full open to closed. (Turbine Trip). 5.2.2.4.2 Environment l The primary safety valves are designed to operate in the following environmental conditions. 1 5.2.2.4.2.1 Normal Environment A. 122*F maximum. B. Relative humidity of 95% at 60*F to 80*F. 1 C. Fixed moisture content equivalent to 95% relative humidity ) at 80*F, up to 122'F. j 5.2.2.4.2.2 Main steem Line Break (one occurrence) 350*F maximum Superheated steam / air mixture for 12 minutes O followed by saturated steam / air mixture. 5.2-3 .. L __ ___ _ _ __ _
CESSAR E!Encaren O 5.2.2.4.3 Main Steam Safety Valves The main steam safety valves are direct acting, spring loaded, carbon steel valves. The valves are mounted on each of the main steam lines upstream of the steam line isolation valves, and outside containment. A schematic drawing of the main steam safety valves is given in Figure 5.4.13-2. The valve parameters are given in Table 5.4.13-2. For a description of overpressure protection equipment and components for the main steam system j refer to Section 10.3.2. 5.2.2.4.3.1 Main Steam Safety Valve Operation The operation of these valves is similar to the primary safety valves, Section 5.2.2.4.2. 5.2.2.4.3.2 Transients The main steam safety valves will be designed to withstand the following transients without failure or malfunction. A. 565'F to 75'F ic 60 seconds for 8 cycles (loss of secondary pressure). B. Pressure changes form 0 psig to 1375 psig, at a temperature i range of between 100*F to 200*F for 300 cycles (secondary l side leak test). l l 6 C. i 10*F step change from 553*F, 1.5 x 10 cycles (normal plant variations). D. 75'F to 565'F and return to 75'F at a rate of 100*F/hr with pressures at saturation levels for 750 cycles (plant heatup and cool down). No' e: Heat up and cool down are separate transients, each beginning at steady state conditions. E. Pressurize to 1.5 times set pressure at 100*F - 200*F for 15 cycles plus number of hydros conducted prior to valve j shipment (hydrostatic test). F. 720 opening and closing cycles to full stem movement . (turbine trip). j l' 5.2.2.4.3.3 Environment The main steam safety valves are designed to operate in the i environmental conditions outlined below. ! l 5.2-4 l 1
CESSARSinh a O 5.2.2.4.3.3.1 Normal Environment A. 104*F maximum. B. Relative humidity 95% at 60*F to 80*F. C. Fixed moistbre content equivalent to 95% relative humidity at 80*F, up to 104*F. 5.2.2.4.3.3.2 Main Steam Line Break (One Occurrence) A. 330*F maximum for 3 minutes. B. Relative humidity of 100%. 5.2.2.4.4 Safety Injection System Relief Valves SI-169 and SI-469 These relief valves are direct acting, spring loaded, stainless steel valves with enclosed bonnets. The design parameters of these valves are: set pressure 2485 psig 4 rated flow 15 gpm water chemi 'ry 0 - 4 weight percent boric acid throat area .023 in desio:, temperature 650*F 5.2.2,4.4.1 Valve Operation As the set pressure is reached, the disc raises off the nozzle seat. This lift continues until the valve is fully open at 10 r,>ercent accumulation. The lift decreases as pressure drops until the seat and disc contacts and seals closed. The valve is fully closed at a maximum of 10% below set pressure (10% blowdown). l 5.2.2.4.4.2 Transients ! These relief valves will be designed to withstand the following j transients without failure or malfunction. j i A. 60*F to 400*F in 5 seconds, 400*F to 60*F in 15 minutes for ] 83 cycles. B. 60*F to 350*F in 15 minutes, 350*F to 60*F in 2.9 hours for 750 cycles. 5.2-5
CESSARE!" Gem O' C. 120*F to 60*F in 5 seconds, 60*F to 120*F in 15 minutes for 990 cycles. 5.2.2.4.4.3 Environment l l These relief valves are designed to operate in the following j environmental conditions. j A. 122*F maximum. B. 95% relative humidity at 60*F to 80*F. C. Fixed moisture content equivalent to 95% RH at temperatures f above 80*F. ( 5.2.2.4.4.4 Material Specifications Material specifications for the primary safety valves are given in Table 5.4.13-1. Material specifications for the main steam safety valves are given in Table 5.4.13-2. Typical materials used for these relief valves are: Body ASME SA 351 GR. CF 8M Disc Stellite No. 6B Nozzle ASME SA 479 Type 316 with Stellite Seat Inlet Stud ASME SA 193 GR. B6 5.2.2.5 Mountino of Pressure-Relief Devices See site-specific SAR. 5.2.2.6 Applicable Codes and Classification , I The applicable codm and classifications for the overpressurization protection system are contained in Table 3.2-1. The applicable codes and classification for the secondary safety valves are identified in Section 10.3.2. E 5.2.2.7 Process Instrumentation Process instrumentation for the overpressurization protection equipment that is associated with the Reactor Coolant System is shown in Figure 5.1.2-1 and described in Chapter 7. l discharge is I Instrumentation associated with pressurizer relief Amendment E 5.2-6 December 30, 1988 I
CESSARn%nceu l O I described in Section 5.4.11. Process instrumentation for secondary in the site-specific system overpressurization SAR. protection will be identified l D 5.2.2.8 System Reliability Reliability of the main steam system reliefs is discussed in the interface Section 5.1.4. The primary safety valves are passive, , spring actuated mechanismo, and cannot fail closed if setpoint l pressure is exceeded. The operational reliability of the primary safety valves is assured by: A. Stringent compliance with ASME III and XI Code for safety l valves. B. Conservative design criteria. C. Selection of a vendor with proven experience and expertise. D. Accounting for thermal cycling during valve operation. E. Technical Specifications. 5.2.2.9 Testing and Inspection Testing and inspection of the primary and secondary valves is governed by ASME Section XI. 5.2.2.10 overpressure Protection During Low Temperature Conditions Overpressure protection of the RCS during low-temperature i conditions is provided by the relief valves located in the Shutdown Cooling System (SCS) suction lines. Section 5.4.7 provides a description of the SCS. The SCS is shown on the SCS Flow Diagram (Figure 5.4.7-3, Sheets schematically 1 and 2), lE the RCS P&ID (Figure 5.1.2-1) and on the Safety Injection System (SIS) P&ID (Figures 6.3.2-1A and 6.3.2-1B). The electrical schematic for the SCS isolation valves is provided in the SIS lD P&ID (Figure 6.3.2-1B). The SCS relief valves are shown on Figure 6.3.2-1B and described in Section 5.4.7.2.2. Alignment of the SCS relief valve to the RCS is specified by plant procedures to ensure RCS overpressure protection for all temperatures below the LTOP temperature, T The P-T Limit D curves are shown in Figures 5.3-Sa and 5.3- NOP. For temperatures . above the LTOP temperature, overpressure protection is provided by the pressurizer safety valves described in Section 5.2.2.4 or administrative procedures. Amendment E 5.2-7 December 30, 1988
1 CESSAR EMU"lCATl2N O i 5.2.2.10.1 Design Criteria A discussion follows of the criteria considered in the design of the overpressure mitigating system to provide low temperature overpressure protection (LTOP) for the RCS. 5.2.2.10.1.1 Credit for Operator Action l No credit is taken for operator action for 10 minutes after the operator is made aware that a transient is in progress. It is D assumed that the operator is aware of the transient when it commences. 5.2.2.10.1.2 Single Failure In the LTOP mode, each SCS relief valve is designed to protect the reactor vessel given a single failure in addition to the failure that initiated the pre ,sure transient. The event initiating the pressure transient is considered to result from , either an operator error or equipment malfunction. The SCS relief valve system is independent of a loss of offsite power. Each SCS relief valve is a self actuating spring-loaded liquid relief valve which does not require control circuitry. The valve opens when the RCS pressure exceeds its setpoint. The redundant SCS suction line trains between the RCS and SCS relief valves meet the single failure criteria as described in Section 5.4.7.1.2 and Table 5.4.7-3. No single fai]ure of an isolation valve or its associated interlock will prevent one relief valve from performing its intended function. 5.2.2.10.1.3 Testability Periodic testing of the SCS isolation valves is defined in the Technical Specification, Section 16.3/4.5.2. 5.2.2.10.1.4 Seismic Design and IEEE Standard 279 Criteria B The SCS suction line relief valves, isolation valves, associated interlocks, and instrumentation are designed to Seismic Category I requirements as discussed in Sections 3.2.1, 5.4.7.2.4 and D i Table 3.2-1. The interlocks and instrumentation associated with the SCS suction isolation valves satisfy the appropriate portions of IEEE Standards 279, 308 and 603 criteria as discussed in Sections 5.4.7, 7.6.1, 7.6.1.1.1, 7.6.2.1.1 and 7.6.2.2.1 and E Table 7.6-1. Design and Analysis d 5.2.2.10.2 In demonstrating that the SCS relief valves meet the criteria l listed in Section 5.2.2.10.1, the following additional ( information is provided. l Amendment E 5.2-8 December 30, 1988
CESSAREMMema A U 5.2.2.10.2.1 Limiting Transients Transients during the low temperature operating mode are more severe when the RCS is operated in the water-solid condition. Addition of mass or energy to an isolated water-solid system-produces increased system pressure. The severity of the pressure transients depends. upon the rate and total quantity of mass or energy addition. The choice of the limiting LTOP transients is based on evaluations of potential transients for System 80 plants and their applicability to the System 80+ plant. lE The most limiting transients initiated by a single operator error or equipment failure are: A. An inadvertent safety injection actuation (mass addition). lE B. A reactor coolant pump start when a positive steam generator to reactor vessel AT exists (energy addition). lE The most limiting transients are determined by conservative analyses which maximize mass and energy additions to the RCS. In addition, the RCS is assumed to be in a water-solid condition at the time of the transient; such a condition has been noticed to exist infrequently during plant operation since the operator is-instructed to avoid water-solid conditions whenever possible. Table 5.2-4 shows the results of the inadvertent safety injection actuation transient analysis for a water-solid RCS, when the RCS E in the LTOP mode. The mass addition due to the simultaneous operation of four safety injection pumps and one charging pump was considered, along with the simultaneous addition of energy from decay heat and the pressurizer heaters. Figure 5.2-1 shows the result of the transient analysis of reactor coolant pump start when a steam generator to reactor lE vessel AT of 100*F exists. This AT is the maximum allowed by technical specification during the LTOP mode. In addition to considering the energy addition to the RCS from the steam generator secondary side, energy addition from decay heat, the j reactor coolant pump and all pressurizer heaters were also l included. In this analysis the steam generators were assumed to be filled to the zero power, normal -water level. For ! conservatism, the secondary water, both around and above the , l U-tubes, was assumed to be thermally mixed in order to maximize l l the energy input to the primary sid?. This assumption is l L conservative since as a result of the temperature distribution i within the steam generator during the transient, the water ; inventory above the tubes is practically isolated thermally froin the heat transfer region. Therefore the heat transfer rate, and r thus the primary side pressure, is not sensitive to the secondary w side water level as long as the tubes are covered. Amendment E 5.2-9 December 30, 1988 i _ _ _ _ _ _ _ - _ _ . _ _ _ _ _ _- __ __n
l
.CESSAR snWicari:n l l
On the basis of experience, the AT value of 100*F used in the l analysis is much larger than any AT that might be expected during plant operation. This maximum allowable AT of 100*F will prevent pressurizer pressure from exceeding the minimum P-T limit allowed for the lowest system temperature during the LTOP mode of , operation. (See Figures 5.3-Sa and 5.3-5b). During RCS cooldown f E l using the Shutdown Cooling System, coolant circulating with the reactor coolant pumps operating serves to cool the steam generator to keep the temperature difference between the reactor vessel and the steam generator minimal. Procedures for System 80 have directed the operator to maintain the AT below approximately 20*F. LTOP transients have not been analyzed for the simultaneous startup of more than one reactor coolant pump (RCP). Such operation is procedurally precluded since the operator starts only one RCP at a time and a second RCP is not started until system pressure is stabilized. Additionally, there is an LTOP transient alarm that should indicate that a pressure transient is occurring. Accordingly, the second RCP would not be started. The operator not start an RCP if the AT exceeds 100*F. However, as mentioned above, administrative procedures for System 80 have ensured that the AT is maintained below approximately 20*F. E With similar administrative controls on System 80+, AT margins will be even greater that for System 80. The results of the analyses provided in Figure 5.2-1 and Table 5.2-4 show that the use of either SCS relief valve will provide sufficient pressure relief capacity to mitigate the most limiting LTOP events identified above. 5.2.2.10.2.2 Provision for Overpressure Protection During heatup, the RCS pressure is maintained below the LTOP pressure until the RCS cold-leg temperature exceeds the LTOP disable temperature. During cooldown, the RCS pressure is E maintained below the LTOP pressure once the RCS cold-leg temperature reaches the LTOP enable temperature. An LTOP enable temperature is defined in Branch Technical l Position RSB 5-2, "Overpressurization Protection of Pressurized l Water Reactors While Operating at Low Temperatures," to Standard l Review Plan Section 5.2.2, " Overpressure Protection," issued November 1988 as Revision 2. The definition is based on measuring the degree of protection provided by the low temperature overpressure protection system (LTOP System) against violations of the P-T Limits in terms of the RT of the reactor vessel beltline material at either the 1/4t F 3/4e location, gi Amendment E 5.2-10 December 30, 1988 i
CESSAR nnince 1 (a depending upon which PT Limit curve is most. limiting. The LTOP enable temperature in Branch Technical Position RSB 5-2 is defined as "the water temperature corresponding to a metal temperature of at least RT 90*F at the reactor vessel beltline location (either 1/N or+ 3/4t) that is controlling in the Appendix G [to Section III of the ASME Code] limit calculations." Figures 5.3-Sa and 5.3-5b represent the P-T Limits at the end-of-(plant) life of 60 years. At this time in plant life, the calculated maximum RT I for the reactor vessel beltline is 79'F for the 1/4t location,N nd 69'F for the 3/4t location. Given the implicit assumptions in the P-T Limit analysis described in Section 5.3.2.1.1.D.1, the 3/4t location is the controlling l beltline P-T Limit for heatup, and the 1/4t location is controlling for cooldown. Therefore, in accordance with the definition of the LTOP enable temperature given in Branch Technical Position RSB 5-2, T = 169'F, and T f r cooldown b b *for F + heatup 90*F = is 79'F + 90*F 159'F. I LTOP Figure 5.3-Sa shows that, for heatup, a temperature of 309'F is required before the pressurizer safety valves can be used for overpressure protection without violating any P-T Limits. Figure l , 5.3-5b shows that, for cooldown, the safety valves cannot be used
' below a temperature of 225'F. For the temperature interval 140*F for heatup, as defined by Branch Technical Position aboveT{0P RSB 5- and for the 66*F above T for cooldown, E administrative controls would be required IgPprotect some P-T Limits, particularly the design limit of 100*F/hr. Rather than relying on administrative controls during these temperature intervals, the T enable and disable temperatures for System 80+ are defined g oEhe intersection of the controlling P-T Limit (100*F/hr) curve and the pressurizer safety valve setpoint. ,
During heatup, if the SCS suction isolation valves (SD-673 and '
-671 or SD-670 and -672) are open and RCS pressure exceeds the LTOP pressure, an alarm will notify the operator that a pressurization transient is occurring during low temperature conditions. Either SCS relief valve will terminate inadvertent pressure transients occurring while RCS temperature is below the LTOP disable temperature. For temperatures above the LTOP disable temperature, that corresponding to the intersection of the design P-T limit curve for heatup with the pressurizer safety valve setpoint, overpressure protection is provided by the pressurizer safety valves.
I (_) Amendment E 5.2-11 December 30, 1988
CESSAR nL"lCAT12N O During cooldown, whenever the RCS cold leg temperature is below the LTOP enable temperature that corresponding to the intersection of the design P-T Limit curve for cooldown with pressurizer safety valve setpoint, the SCS relief valves provide the necessary overpressure protection. If the SCS is not aligned E to the RCS before the cold-leg temperature is decreased below the LTOP enable temperature, an alarm will notify the operator to open the SCS suction isolation valves (SD-673, -670, -671, and
-672). However, the SCS cannot be aligned to the RCS until the RCS pressure is below the LTOP enable pressure.
The LTOP conditions described above are within the SCS operating range. Technical Specification Section 16.3/4.4.8.3 requires the SCS suction line isolation valves to be open when operating in the LTOP mode. Also, this Technical Specification ensures that appropriate action is taken if one or more SCS relief valves 're out of service during the LTOP mode of operation. Either SCS relief valve will provide sufficient relief capacity to prevent any pressure transient from exceeding the isolation interlock setpoint (see Figure 5.2-1 and Table 5.2-4). lE 5.2.2.10.2.3 Equipment Parameters The SCS relief valves are spring-loaded liquid relief valves with sufficient capacity to mitigate the most limiting over-pressurization event. Pertinent valve parameters are as follows: Parameter Nominal Setpoint 550 psia
- E Accumulation 10%
Capacity 4000 (010% acc) gpm Since each SCS relief valve is a self actuating spring-loaded liquid relief valve, control circuitry is not required. The valve will open when RCS pressure exceeds its setpoint. I l The SCS relief valves are sized, based on an inadvertent safety injection actuation signal (SIAS) with full pressurizer heaters operating from a water-solid condition. The SIAS assumes simultaneous operation of four SIS pumps and one charging purap with letdown isolated. The resulting flow capacity requirement E for water is 3934 gpm. The analysis in Section 5.2.2.10.2.1 assumed that either SCS relief valve relieved water a t this rate.
- Pressure measured at the valve inlet. E l
l l Amendment E 5.2-12 December 30, 1988
CESSAREHL . ) O v - The design relief capacity of each of two SCS relief valves (shown in P&ID Figurc 6.3.2-1B) as supplied by the _ valve ; manufacturer meets the minimum required relief capacity of I 4000 gpm which contains sufficient margin in relieving capacity l E for even the worst transient. The SCS relief valves are Safety Class 2, designed to Section III of the ASME Code. 5.2.2.10.2.4 Administrative Controle. Administrative controls necessary to implement the LTOP provisions are limited to those controls necessary to open the SCS isolation valves. ! During cooldown, when the temperature of the RCS is above that corresponding to the intersection of the controlling P-T Limit I and the pressurizer safety valve setpoint, overpressure protection is provided by the pressurizer safety valves, and no 1 administrative procedural controls are necessary. Before entering the low temperature region for which LTOP is necessary, i RCS pressure is decreased to below the maximum pressure required for LTOP. The LTOP pressure is less than the maximum pressure allowable for SCS operation. Once the SCS is aligned, no further
-~ specific administrative procedural controls are needed to ensure proper overpressure protection. The SCS will remain aligned E whenever the RCS is at low temperatures and the reactor vessel head is secured. As designated in Table 7.5-2, indication of SCS isolation valve position is provided.
During heatup, the SCS isolation valves remain open at least until the LTOP enable temperature. Once the RCS temperature has reached that temperature corresponding to the intersection of the !
- controlling P-T Limit and the pressurizer safety valve setpoint, j
- overpressure protection is provided by the pressurizer safety l l valves. The SCS can be isolated and no. further administrative !
procedural controls are necessary. 5.2.2.11 Pressurized Thermal Shock The System 80+ reactor vessel meets the requirements of 10 CFR 50.61, " Fracture Toughness Requirements For Protection Against Pressurized Thermal Shock Events." The calculated RT which satisfies the screening criteria in 10 CFR 50.6h)is(2). 109'F 5.2.3 REACTOR COOLANT PRESSURE BOUNDARY MATERIALS 5.2.3.1 M_aterial Specification A list of specifications for the principal ferritic materials, h) V austenitic stainless steels, bolting and weld materials, which are part of the reactor coolant pressure boundary is given in Table 5.2-2. Amendment E 5.2-12a December 30, 1988
CESSAR EE.Sacma O Studies have shown that the irradiation induced mechanical property changes of SA-533B and SA-508 materials can depend significantly upon the amount of residual elements present in the D compositions, namely; copper, nickel, phosphorous, and vanadium. It has also been found that residual sulfur affects the initial toughness of SA-533B and SA-508 materials. Specific controls are placed on the residual chemistry of reactor vessel materials and g the as-deposited welds used to join these materials to limit the maximum predicted increase in the reference temperature (RT n' which is discussed in Section 5.3.1.6) and to limit the extenED'6f the reactor vessel beltline. The beltline is defined by Appendix G of 10 CFR 50. Materials used in the reactor vessel beltline and the B as-deposited welds contain no greater than the following low percentages of residual elements: , Copper (in welds) 0.03 Copper (in forgings) 0.06 Phosphorous 0.015 Nickel (in forgings) 1.00 Sulfur 0.015 Nickel (in welds) 0.10 Vanadium 0.030 E O' O Amendment E 5.2-12b December 30, 1988 f l l
CESSAR 8HEricari;u 73 5 1
\,_/
5.2.3.2 Compatibility with Reactor Coolant 5.2.3.2.1 Reactor Coolant Chemistry j Controlled water chemistry is maintained within the RCS. Control of the reactor coolant chemistry is the function of the CVCS which is described in Section 9.3.4. Water chemistry limits applicable to the RCS are given in Section 9.3.4. 5.2.3.2.2 Materials of Construction compatibility with l Reactor Coolant The materials of construction used in the RCPB which are in contact with reactor coolant are designated by an "a" in Table 5.2-2. These materials have been selected to minimize corrosion and have previously demonstrated satisfactory performance in other existing operating reactor plants. Metallic materials in contact with reactor coolant shall be restricted in cobalt content to as low a level as practical for B all stainless steel or nickel base alloy components with a large wetted surface area. Cobalt based alloys shall be avoided except
/~N, in cases where no proven alternative exists.
b 5.2.3.2.3 Compatibility with External Insulation and , Environmental Atmosphere
]
1 The possibility of leakage of reactor coolant onto the reactor vessel head causing corrosion of the pressure boundary has been ] i investigated by C-E. l Tests have shown that reactor coolant system leakage onto surfaces of the reactor coolant pressure boundary will not affect the integrity of the pressure boundary. The insulation supplied by C-E is of the stainlest steel reflective type, which minimizes insulation contamination in the event of a chemical solution spillage. In local areas around stainless steel and the nickel based alloy nozzles in the reactor . vessel head, small sections of non-metallic insulation are used. l However, the quantity of leachable halogens will be limited in accordance with Regulatory Guide 1.36. 5.2.3.3 Fabrication and Processina of Ferritic Materialg 5.2.3.3.1 Fracture Toughness 5.2.3.3.1.1 NSBS Components V) Fracture toughness requirements for Reactor Coolant Pressure Boundary components are established in accordance with the ASME Amendment B 5.2-13 March 31, 1988
CESSAR E!asCATCN J O Boiler and Pressure Vessel Code, Section III. Fracture toughness testing of base, weld and heat affected zone materials will be conducted in accordance with the ASME Code. Data from these tests will be available after the required testing has been performed and may be examined upon request at the appropriate manufacturing facility. C-E complies with 10 CFR Part 50 Appendix G, " Fracture Toughness Requirements" as enacted May 1983. D Consideration is given to the ef fects of irradiation on material toughness properties in the core beltline region of the reactor vessel to assure adequate fracture toughness for the service lifetime of the vessel. Refer to Section 5.3.1.6 for discussion concerning prediction of irradiation effects and the material surveillance program. In addition, C-E complies with the guidance of Regulatory Guide 1.2, " Thermal Shock to Reactor Pressure Vessels." Testing and measuring equipment for fracture toughness tests for the reactor vessel, steam generators, pressurizer, piping and reactor coolant pumps are calibrated in accorance with Subarticle NB2360 of the ASME Code, Section III. 5.2.3.3.2 Control of Welding 5.2.3.3.2.1 Avoidance of Cold Cracking C-E complies with the recommendations of Regulatory Guide 1.50, Control of Preheat Temperature for Welding of Low Alloy Steel, May 1973, as discussed below. Paragraph C.1.b implies that the qualification materials are an infinite heat sink that would instantaneously dissipate the heat input from the welding process. The qualification procedure consists of starting the welding at the minimum preheat temperature. Welding is continued until the maximum interpass temperature is reached. At this time, the test material is permitted to cool to the minimum preheat temperature and the welding is restarted. Preheat temperatures utilized for low alloy steel are in accordanco with Section III of the ASME Code. The maximum interpass temperature utilized is 500*F. The paragraph C.2 recommendation is considered an unnecessary extension of procedures which apply to low-alloy steel welds, meeting ASME Code Sections III and IX requirements. The recommendations of Regulatory Guide 1.50 are met by complying i with paragraph C.4. The soundness of all welds is verified by , ASME Code acceptable examination procedures. . Amendment D 5.2-14 September 30, 1988 i
CESSAR En9c 1,3u
/\
l ) L/ With regard to Regulatory Guide 1.43, major components are fabricated with corrosion resistant cladding on internal surfaces exposed to reactor coolant. The major portion of the material protected by cladding from exposure to reactor coolant is SA-533B D Class 1 or SA-508, Class 2 or 3. Cladding of SA-508, Class 2 forging material is performed using low-heat-input welding , processes controlled to minimize heating of the base metal. I Low-heat-input welding processes are not known to induce underclad cracking. 5.2.3.3.2.2 Regulatory Guide 1.34 l Regulatory Guide 1.34 recommends controls to be applied during welding using the electroslag process. The electroslag process is not used in the fabrication of any RCPB components. ) ' Therefore, the recommendations of this guide are not applicable. 5.2.3.3.2.3 RGgulatory Guide 1.71 C-E complies with the recommendations of Regulatory Guide 1.71 except for the differences indicated below. Performance qualifications for personnel welding under conditions l (A)
'v of limited accessibility are conducted and maintained accordance with the requirements of ASME B&PV Code Sections III in I and IX. A requalification is required when (1) any of the essential variables of Section IX is changed, or (2) when I authorized personnel have reason to question the ability of the welder to satisfactorily comply with the applicable requirements.
Production welding is monitored for compliance with the procedure ) parameters, and welding qualifications are certified in accordance with Sections III and IX. Further assurance of acceptable welds of limited accessibility is afforded by the welding supervisor assigning only the most highly skilled personnel to these tasks. Finally, weld quality, regardless of accessibility, is verified by the performance of the required , non-destructive examinations. I 5.2.3.3.3 Non-Destructive Examination of Tubular Products The non-destructive examination requirements imposed by C-E for tubular products are those specified by Section III of the ASME D code. 5.2.3.4 Fabrication and Processinc of Austenitic Stainless Steel 5.2.3.4.1 Avoidance of Stress Corrosion Cracking
) -
5.2.3.4.1.1 Avoidance of Sensitization Amendment D 5.2-15 September 30, 1988
CESSARn' ace O 5.2.3.4.1.1.1 NSSS Components Fabrication of RCPB components is consistent with the recommendations of Regulatory Guide 1.44 as described in items A through D, below, except for the criterion used to demonstrate freedom from sensitization. The ASTM A 708 Strauss Test is used : in lieu of the ASTM A 262 Practice'E, Modified Strauss Test, to demonstrate freedom from sensitization in fabricated, unstabilized, stainless steel. A. Solution Heat Treatment Requirements All raw austenitic stainless steel material, both wrought and cast, used in the fabrication of the major NSSS components in the RCPB, is supplied in the annealed condition as specified by the pertinent ASTM or ASME Code: 1900-2050*F for 1/2 to 1 hour per inch of thickness and water quenched to below 700'F. The time at temperature is determined by the size and type of component. Solution heat treatment is not performed on completed or partially-fabricated components. Rather, the extent of chromium carbide precipitation is controlled during all stages of fabrication as described below. , B. Material Inspection Program Extensive testing on stainless steel mockups, fabricated using production techniques, has been conducted to determine the effect of various welding procedures on the susceptibility of unstabilized 300 series stainless steels to sensitization-induced intergranular corrosion. Only those procedures and/or practices demonstrated not to produce a sensitized structure are used in the fabrication of RCPB components. The ASTM standard A 708 (Strauss Test) is the criterion used to determine susceptibility to intergranular corrosion. This test has shown excellent correlation with a form of localized corrosion peculiar to sensitized stainless steels. As such, ASTM A 708 is utilized as a go/no-go standard for acceptability. As a result of the above tests, a relationship was established between the carbon content of 304 stainless steel and weld heat input. This relationship is used to avoid weld heat-affected-zone sensitization as described l l below. l C. Unstabilized Austenitic Stainless Steel The unstabilized grades of austenitic stainless steels with ! carbon content of more than 0.03% used for components of the i 5.2-16
CESSAR EHL"icareu [ 't \ / RCPB are 304 and 316. These materials are furnished in the solution annealed condition. Exposure of completed or partially-fabricated components to temperatures ranging from ' 800*F to 1500*F is prohibited.
]
Duplex, austenitic stainless steels containing more than 5FN delta ferrite (weld metal, cast metal, weld deposit L overlay), are not considered unstabilized since these alloys do not sensitize, that is form a continuous network of chromium-iron carbides. Specifically, alloys in this ' category are: CF8M Cast stainless steel (delta ferrite CF8 SFN to 33FN) 308, 309 Singly and combined stainless steel 312, 316 weld filler metals (delta ferrite controlled to SFN-20FN. deposited) D In duplex, austenitic/ferritic alloys, chromium-iron carbides are precipitated preferentially at the ferrite /austenitic interfaces during exposure to temperatures ranging from 800-1500*F. This precipitate O) ( morphology precludes intergranular penetrations associated with sensitized Type 300 series stainless steels exposed to oxygenated or fluoride environments. D. Avoidance of Sensitization Exposure of unstabilized austenitic Type 300 series stainless steels to temperatures ranging from 800 to 1500'F will result in carbide precipitation. The degree of carbide precipitation, or sensitization, depends on the temperature, the time at that temperature, and the carbon content. Severe sensitization is defined as a continuous grain boundary chromium-iron carbide network. This condition induces susceptibility to intergranular corrosion in oxygenated aqueous environments, as well as those containing fluorides. Such a metallurgical structure will rapidly fail the Strauss Test ASTM A 708. Discontinuous precipitates (i.e., an intermittent grain boundary carbide network) are not susceptible to intergranular corrosion in a PWR environment. Weld heat affected 2one sensitized austenitic stainless ! steels (which will fail the Strauss Test, ASTM A 708) are j avoided by careful control of: 1
,m l C) t Amendment D 5.2-17 September 30, 1988
1 CESSAR U"Dicari:n i 1 0 Weld heat input to less than 60 kJ/in
- Interpass temperature to 350'F maximum i Carbon content Homogeneous or localized heat treatment in the temperature <
range 800-1500*F is prohibited for unstabilized austenitic stainless steel with a carbon content greater than 0.03% used in components of the RCPB. .When stainless steel safe ends are required on component nozzles, fabrication techniques and sequencing requirn that the stainless steel piece be welded to the compor.ent after final stress relief. This is accomplished by welding an Inconel overlay on the end of the nozzle. Following final stress relief of the component, the stainless steel safe end is welded to the Inconel overlay, using Inconel weld filler metal. 5.2.3.4.1.2 Avoidance of Contamination Causing Stress Corrosion Cracking 5.2.3.4.1.2.1 NSSS Components Specific requirements for cleanliness and contamination protection are included in the equipment specifications for components fabricated with austenitic stainless steel. The provisions described below indicate the type of procedures utilized for NSSS components to provide contamination control during fabrication, shipment, and storage as required by Regulatory Guide 1.37. Contamination of austenitic stainless steels of the 300 type by compounds which can alter the physical or metallurgical structure and/or properties of the material is avoided during all stages of fabrication. Painting of 300 series stainless steels is prohibited. Grinding is accomplished with resin or rubber-bounded aluminum oxide or silicon carbide wheels which were not previously used on materials other than austenitic alloys. Outside storage of partially-fabricated components is avoided and in most cases prohibited. Exceptions are made for certain components provided they are dry, completely covered with a waterproof material, and kept above ground. Internal surfaces of completed components are cleaned to produce an item which is clean to the extent that grit, scale, corrosion products, grease oil, wax, gum, adhered or embedded dust or extraneous materials are not visible to the unaided eye. Cleaning is effected by either solvents (acetone or isopropyl alcohol) or inhibited water (hydrazine). Water will conform to the following requirements: i 5.2-18 i
[ CESSARMnk m,. 1 /.( , 1 Halides. Chloride (ppm) 0.60
-Fluoride (ppm) 0.40 Conductivity '(pahos/cm) 5.0 pH 6.0-8.0 Visual clarity No turbidity, oil,;or. sediment To prevent halide-induced intergranular corrosion which ' could occur. in aqueous. environment with significant quantities of dissolved oxygen, . flushing water is inhibited ' via additions of hydrazine. Results of-tests have proven.theseLinhibitors to be completely' effective. Operational.- chemistry specifications restrict concentrations-of halide.and oxygen, both prerequisites of intergranular attacks (refer to Section 9.3.4).
5.2.3.4.1.3 Characteristics and Mechanical' Properties of: Cold-Worked Austenitic Stainless Steels for RCPB. Components Cold-worked austenitic stainle'ss steel is not. utilized for l components of the RCPB. 5.2.3.4.2 Control of Welding 5.2.3.4.2.1 Avoidance of Hot Cracking l A. NSSS Components
- 1. Regulatory Guide 1.31 l
In order: to preclude microfissuring. in 'austenitic stainless steel welds, RCPB components are consistent with the recommendations ' of Regulatory Guide 1.31 as-follows: The delta ferrite content of eachLlot and/or. ' heat i of: weld- filler metal. used for . welding- of'austenitic stainless steel code components shall.be' determined for each process to'be used in production. Delta ferrite ,I determinations for consumable inserts, electrodes,! rod D' I or wire filler metal' used with the gas tungsten ~ arc welding process, and deposits made with the plasma arc. welding process - may . be determined by either of the alternative methods of magnetic measurement or. chemical-. ! analysis described in Section'III of-the ASME Code. ;
. Amendment DJ ]
5.2-19 September 30, 1988
.]
CESSAREnac-l 0\ J Delta ferrite verification should be made for all other i processes by tests using the magnetic measurement I method on undiluted weld deposits described by Section D III of the ASME Code. The average ferrite content i shall meet the acceptance limits of SFN to 20FN for weld rod or filler metal.
- 2. Regulatory Guide 1.34 Regulatory Guide 1.34 is discussed in Section 5.2.3.3.2.2.
- 3. Regulatory Guide 1.71 Regulatory Guide 1.71 is discussed in Section 5.2.3.3.2.3.
5.2.4 IN-SERVICE INSPECTION AND TESTING OF REACTOR COOLANT PRESBURE BOUNDARY 5.2.4.1 Accessibility of Inspection Areas Class 1 components and supports are designed to meet the acc6ss ' requirements of Section XI of the ASME Boiler and Pressure Vessel Code. I In the case of the reactor vessel, all internals except the flow l' baffle are removable. Their removal makes the entire inner surface of the vessel, as well as the weld zones of the internal load-carrying structural attachments, available for the surface and volumetric inspections. The closure head is available for ,
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inspection whenever it is removed, and its removal also makes available the vessel closure flange, the flange to shell weld, closure stud holes and ligaments, and the closure studs and nuts. i Each control element drive mechanism is removable as a unit through a closure at the top of its housing. For interim inspections of the reactor vessel primary coolant nozzle to chell welds and inner radii, the two outlet nozzles are accessible from inside the reactor vessel without removal of the ! vessel internals. The inlet nozzles are accessible either from lB outside the vessel or from inside after removal of the vessel internals. Manways are provided for those inspections which must be made internally on the steam generators and pressurizer. Access holes are provided in the support skirt of the steam generators to allow examination of the tube sheet support stay cylinder welds. The steam generators are capable of being examined in accordance < with Regulatory Guide 1.83. , Amendment D 5.2-20 September 30, 1988 l
CESSAR nn"lCATION o V The reactor coolant pumps may be disassembled, if necessary, for ! inspection. Additional provisions for access, and details of the inservice inspection program are included in the site-specific SAR. lB 5.2.5 REACTOR COOLANT PRESSURE BOUNDARY (RCPB) LEAKAGE DETECTION SYSTEMS Means for the detection of leakage from the Reactor Coolant Pressure Boundary are provided to alert operators to the i existence of leakage above acceptable limits, which may indicate I an unsafe condition for the facility. The leakage detection systems are sufficiently diverse and sensitive to meet the criteria of Regulatory Guide 1.45 for leaks from identified and unidentified sources. See Section 5.1.4 for interface requirements. 5.2.5.1 Leakage Detection Methods i 5.2.5.1.1 Unidentified Leakage See site-specific SAR for details of determining unidentified B leakage. Interface requirements are contained in Section'5.1.4. l In addition to the methods for detecting unidentified leakage discussed in the site-specific SAR, amethodfordetectinglargelB volume leaks, which is available as an integral part of the RCS and CVCS, is the reactor coolant inventory method. Leakage from the Reactor Coolant System can be determined by net level changes ) in the pressurizer and in the volume control tank since the i' Reactor Coolant System and the Chemical and Volume Control System are closed systems. Since letdown flow and the reactor coolant j pump seal controlled bleedoff flow are collected and recycled back into the Reactor Coolant System by the Chemical and Volume Control System (CVCS), the net inventory in the Reactor Coolant System and CVCS under normal operating conditions will be i l constant. Transient changes in letdown flow rate or Reactor l Coolant System inventory can be accommodated by changes in the volume control tank level. Makeup flow rate provides a means of . detecting leakage from the Reactor Coolant System through I measurement of the net amount of makeup flow to the system. The net makeup to the system under no-leakage steady state conditions should be zero. The makeup ' flow rates and the integrated makeup flow rates from the Reactor Makeup System and the refueling water tank are continuously monitored and recorded. Analysis of the integrated makeup flow recorders ' over a period of steady state operation can provide detection of abnormal leakage. An O (d increasing trend in the amount of makeup required will indicate an abnormal leak which is increasing in rate. Leaks occurring i Amendment B 5.2-21 March 31, 1988 5 _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . __ _ _ __-____-____U
CESSARnnL a O suddenly will be indicated by a step increase in the amount of makeup which does not decrease as would be the case for a purely transient condition. 5.2.5.1.2 Identified Leakage The amount of identified leakage from the Reactor Coolant System can be determined by adding up the amounts from all identified paths described below. Indicators and alarms associated with all of the identified leakage paths are provided in the control room. See Section 5.1.4 for interface requirements. 5.2.5.1.2.1 Safety Valves Located on the Reactor Coolant System The primary safety valves, located at the top of the pressurizer, are headered together and routed to the In-Containment Refueling Water Storage Tank (IRWST). Upstream of the headers, each safety valve is monitored for seal leakage by an in-line D Resistance-Temperature Detector (RTD). Indirect indication of code safety valve leakage a decrease in pressurizer pressure and pressurizer is provided level byas lE monitored by safety grade instrumentation. Positive indication of safety valve leakage will be provided in the control room. Monitoring will be provided by an Acoustic Leak Monitoring System ( ALMS) consisting of an accelerometer (acoustic sensor) mounted downstream of each valve. A plant D annunciator alarm will be provided to alarm if the valve is not fully closed. The ALMS is part of the NSSS Integrity Monitoring System (see Section 7.7 for a detailed description). 5.2.5.1.2.2 Reactor Coolant Pump Seals Instrumentation is provided to detect abnormal seal leakage. The reactor coolant pumps are equipped with two stages of seals plus i a vapor or backup seal as described in Section 5. 4.1. 2. During normal operation, the Reactor Coolant System operating pressure is decreased through the two seals to approximately CVCS volume control tank pressure. The vapor or backup seal prevents leakage to the containment atmosphere and allows sufficient pressure to be maintained to direct the controlled seal leakage to the volume control tank. The vapor or backup seal is designed to withstand full Reactor Coolant System pressure in the event of failure of any or all of the two primary seals. Since, in the event of the leakage into containment a flow condition would exist through the vapor seal, the pressure would > decrease downstream of the middle seal due to seal differential Amendment E 5.2-22 December 30, 1988
CESSARHMema '
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pressure. The reactor coolant pump seal pressure indicator would give this indication. The vapor seal pressure indicator would show a decrease in pressure and an increased level in the reactor drain tank would be indicated. Seal leakage through the tubes of the reactor coolant pump seal cooler to the Component Cooling Water System would be indicated by increased temperature of the component cooling water return line from the reactor coolant pumps, by increased level in; component cooling water surge tanks, l and by increased radiation; monitor readings in the Component ! Cooling System. l 5.2.5.1.3 Leakage Through Steam Gen 6rator Tubes or Tubesheet An increase in radioactivity indicated by condenser air removal system monitors and blowdown system monitors will indicate reactor coolant leakage through steam generators tubes to the secondary side. Routine analysis of steam generator water samples would also indicate increasing leakage of reactor coolant into the secondary system. 5.2.5.1.4 Leakage to Auxiliary Systems f% Chapter 11 describes the design basis for process monitors used in all potentially contaminated auxiliary systems and their sensitivity. i 5.2.5.2 Control Room Leakace Instrumentation Indication of safety valve leakage is provided by the Acoustic .' Leak Monitoring System (ALMS). The signal from the accelerometer sensor is monitored continuously, along with the RMS value of the signal computed. These signals are compared to a pre-established baseline value and alarmed in the main control room via the Nuplex 80+ Data Processing System (DPS) and Discrete Indication D and Alarm System (DIAS) displays (See Section 7.7). The site operator may provide other additional metheds for leak detection as input to DPS and DIAS. See the site-specific SAR for actual methods employed to provide control room indication of leakage to the containment. Interface lB requirements are contained in Section 5.1.4. 5.2.5.3 Limits For Reactor Coolant Leakace l i The limits for both tcc:.a1 and unidentified leakage are described I in the technical specifications. p A Amendment D 5.2-23 September 30, 1988 t
l CESSAR MSincan 5.2.5.4 Maximum Allowable Total Leakace . The maximum allowable total identified leakage will be as stated in the Technical Specifications, Chapter 16. 5.2.5.5 Differentiation Between Identified and Unidentified Leaks Identified leakage from the Reactor Coolant System would be into contained systems such as the secondary side of the. steam generator, the Safety Injection System, the Component Cooling Water System and the Chemical and Volume Control System. Leakage lD into these systems will be monitored and detected by: (1) radioactivity monitoring of the flow streams in those systems; (2) level indicators in the surge or compensating tanks of those flow systems; or, (3) makeup flow and/or integrated makeup indicators in the Reactor Coolant System Makeup System portion of the Chemical and Volume control System (as described in Section 5.2.5.1). These methods may not pin point the exact location of lB a leak, but (1) temperature indicators in these systems will locate a leak source in that an abnormally high reading is indicative of leakage at the point, or (2) local sample points will allow the detection of specific points of leakage by looking for a point of relatively high radioactivity concentration as ! compared to the remainder of the system. The systems used to detect unidentified leakage from the Reactor Coolant System to the containment are described in the site-specific SAR and in Section 7.7. l8 5.2.5.6 Sensitivity and Operability Tests l A description of the Reactor Cooleat System Makeup Control System and its instrumentation is discuased in Section 9.3.4 and Chapter
- 7. A description of the monitors used in auxiliary systems for leakage detection is found b Chapters 9 and 11.
l A description of the sensitivity and operability of the pressurizer pressure and level instrumentation is contained in Chapter 7. The temperature measurement channels downstream of the primary safety valves are described in the site-specific SAR. lB Descriptions of the unidentified leakage detection systems, l l including the containment airborne particulate and containment sump level monitoring systems, and their sensitivities will be provided in the site-specific SAR. A description of tests to lD demonstrate the sensitivity and operability of these systems will also be provided in the site-specific SAR. B l$ ! Amendment D 5.2-24 September 30, 1988 1
CESSAR !!Sinema (O TABLE 5.2-1 ! l REACTOR COOLANT SYSTEM PRESSURE BOUNDARY CODE REQUIREMENTS Components Codes and Classes I Reactor Vessel, Steam Generators ASME Boiler and Pressure Vessel ! (primary side), Pressurizer Code, Section III, Nuclear Power 1 Plant Components, Class 1. I D I Reactor Coolant Pump (structural ASME Boiler and Pressure Vessel portions necessary to assure the Code, Section III, Nuclear Power i integrity of the reactor coolant Plant Components, Class 1. i pressure boundary) Reactor Coolant Pump Auxiliaries ASME Boiler and Pressure Vessel 4 Code, Section III, Nuclear Power j Plant Components, Class 3. { Lube oil system designed for ' Seismic Category I requirements. p Pressurizer Spray and Safety Valves ASME Boiler and Pressure Vessel t j Code, Section III, Nuclear Power
'v Plant Components, Class 1.
Piping and Valves ASME Boiler and Pressure Vessel Code, Section III, Nuclear Power Plant Components, Class 1. Steam Generators (Secondary Side) ASME Boiler and Pressure Vessel Code, Section III, Nuclear Power Plant Components, Class 2. Control Element Drive Mechanisms ASME Boiler and Pressure Vessel Code, Section III, Nuclear Power Plant Components, Class I. NOTES: Code addenda requirements for System 80+ plants will comply with the requirements of 10 CFR 50.55a. j Codes listed above are construction codes. In addition, all these components are designed and constructed to meet the test and inspection requirements of the ASME Boiler and Pressure Vessel Code, Section XI, Rules for Inservice Inspection.
/G, Code requirements for Safety Depressurization System valves, which V meet the definition of the Reactor Coolant Pressure Boundary, are given in Section 6.7.
D Amendment D l September 30, 1988
CESSAR Ennncua 1 f G TABLE 5.2-2 (Sheet 1 of 5) REACTOR COOLANT SYSTEM MATERIALS Component Material Specification l Reactor Vessel ' Forgings SA-508 Class 1, 2 and 3 Cladding Weld deposited austenitic stainless steel with 5FN-18FN delta ferrite or NiCrFe o alloy (equivalent to SB-168) Reactor vessel head SB-166 CEDM Nozzles Vessel internals (a) Austenitic Stainless Steel and NiCrFe alloy Fuel cladding (a) Zircaloy-4 1 tO b Q Instrument nozzles SB-166 Control element drive j mechanism housings ) Lower Type 403 stainless steel according to Code Case N-4-11 with end fittings to be lD SB-166 and/or SA-182 Type 348 stainless steel l Upper SA-479 and SA-213 Type 316 stainless steel with end fitting of SA 479 Type 316 and vent valve seal of Type 316 and vent valve seal of Type 440 stainless steel seat Closure head bolts SA-540 B24 or B23 Pressurizer Shell SA-533 Grade A or B Class 1 or SA-508 Class 3 Cladding (a) Weld deposited austenitic stainless steel with 5 FN-18FN delta ferrite or NiCrFe D alloy (equivalent to SB-166) t O Amendment D September 30, 1988
CESSAR 8!ainem O' TABLE 5.2-2 (Cont'd) j (Sheet 2 of 5) REACTOR COOLANT SYSTEM MATERIALS Cgmoonent Material Specification Forged nozzles SA-508 Class 1, 2 or 3 l0 ) Instrument nozzles SB-166 Surge and safety valve SA-182 nozzle safe ends Studs and nuts SA-540-824 or B23 Steam generator Primary Head SA-533 Grade B, Class 1 or SA-508 Class 1, 2 or 3 8 1 Nozzles SA-508 Class 2 or 3 Safe ends SA-508 Class 1, 2 or 3 lD Primary head cladding (a) Weld deposited austenitic stainless steel with 5FN-18FN delta ferrite l0 Tubesheet SA-508 Class 2 or 3 Yubesheet stay SA-508 Class 2 or 3 Tubesheet cladding (a) Weld deposited NiCrFe alloy (equivalent to SB-168) Tube (a) NiCrFe Alloy'(SB-163) Secondary shell and head SA-533 Grade A, Class 1 SA-516 Grade 70 lD Secondary nozzles SA-508 Class 1 or Class 2 Secondary nozzle safe ends SA-508 Class 1 Secondary instrument nozzles SA-106 Grade B Studs and nuts SA-540 Grade B24 or B23, or SA-193 , I Grade B7 or SB-637 UNS N07718 l Amendment D September 30, 1988 l
CESSAR Ennneuia O TABLE 5.2-2 (Cont'd) (Sheet 3 of 5) REACTOR COOLANT SYSTEM MATERIALS ' Component Material Specification Reactor Coolant Pumps Casing (a) SA-508 Class 2 or 3 or austenitic stainless steel B Cladding Weld deposited austenitic stainless steel with 5FN-18FN delta ferrite l0 Internals SA-487 CA6NM, SA 336 Grade F8 or austenitic stainless steel B Reactor Coolant Piping l Pipe (30 in, and 42 in.) SA-516 Grade 70 rs Cladding (a) Weld deposited austenitic stainless steel with 5FN-18FN delta ferrite l Piping nozzles and safe ends D Nozzle forgings SA-105, SA-541 Class 1, 2 or 3, or SB-166 Nozzle safe ends SA-182 or SB-166 Valves SA-351 CF8M or SA-182 rx I Amendment D September 30, 1988
CESSAR 2PHnceu O TABLE 5.2-2 (Cont'd) (Sheet 4 of 5) REACTOR COOLANT SYSTEM MATERIALS WELD MATERIALS FOR REACTOR COOLANT PRESSUf;E BOUNDARY COMPONENTS Material Specification Base material Weld Material
- 1. SA-533 SA-533 a. SFA 5.5,(b) E-8018-C3, E-8018-G Gr. B C1.1 Gr. B C1.1 b. MIL-E-18193, B-4
- 2. SA-508 SA-533 a. SFA 5.5, E-8018-C3, E-8018-G C1.2 Gr. B C1.1 b. MIL-E-18193, B-4
- 3. SA-508 SA-508 a. SFA 5.5, E-8018-C3, E-8018-G C1.1 C1.2
- 4. SA-516 SA-516 a. SFA 5.1, E-7018 Gr. 70 Gr. 70
- 6. SA-182 SA-516 a. SFA 5.1, E-7018 F1 Gr. 70
- 6. SA-105 SA-351 a. SFA 5.14, ERNiCr-3 Gr. 11 CF8M
- 7. SA-182 SA-351 a. SFA 5.11, ENiCrFe-3 F1 CF8M
- 8. SA-105 SA-182 a. SFA 5.14, ERNiCr-3 Gr. 11 F316
- 9. SB-166 SA-182 a. Root SFA 5.14, ERNiCr-3 Remaining SFA 5.11, ENitrFe-3 D F316
- 10. 58-167 SA-182 a. Root SFA 5.14, ERNiCr-3 F304 Remaining SFA 5.11, ENiCrFe-3
- 11. SA-516 SA-351 a. SFA 5.1, E-7018 Gr. 70 CF8M b. MIL-E-18193, B-4 l0
- 12. SA-182 SA-182 a. SFA 5.1, E-7018 F1 F316
- 13. SB-166 SA-533 a. SFA 5.14, ERNiCr-3 Gr. B C1.1 i
Amendment D September 30, 1988
l CESSAR Eniincim. l A I U TABLE 5.2-2 (Cont'd) l 1 ($heet 5 of 5) j REACTOR COOLANT SYSTEM MATEh_ is WELD MATERIALS FOR REACTOR COOLANT PRESSURE BOUNDARY COMPONENTS Material Specification Bese material _ Weld Meterial
- 14. SA-182 SB-167 a. SFA 5.14, ERNiCr-3
- 15. SA-516 SA-508 a. SFA 5.5,(b) E-8016-C3 -
Gr. 70 C1.2
- 16. Austenitic a. SFA 5.9, ER-308 )
stainless SFA 5.9, ER-309 j steel SFA 5.9, ER-312 ) cladding i
- 17. Inconel Inconel a. ENiCrfe-3
' ERNiCr-3
- 18. SA-508 SA-508 a. SFA 5.5,(b) E-8018-C3, E-8018-G C1. 3 C1. 3 b. MIL-E-18193, B-4
- 19. SA-508 SA-533 a. SFA 5.5, E-8018-C3, E-8018-G C1. 3 Gr. B C1.1 b. MIL-E-18193, B-4 D
- 20. SA-508 SA-508 a. SFA 5.5, E-8018-C3, E-8018-G Cl. 3 C1. 2
- 21. SA-508 SA-516 a. SFA 5.5,(b) E-8018-C3 l C1. 3 Gr. 70 Notes: a. Materials exposed to reactor coolant.
I b Special weld wire with low residual elements of copper, nickel and phosphorous as specified for the reactor vessel core D beltline region. )
)
l Amendment D September 30, 1988
CESSAR !!!L"icm:n t
\
TABLE 5.2-3 CODE CASE IETERPRETATIOE8
- 1. N-4-11 Requirements for Special Type 403 Modified Forgings and Bars D
- 2. N-71-14 Additional Materials for Subsection NF, Classes 1, 2, 3 and MC Component Supports Fabricated by Welding, Section III, Division 1 Note: Code Case Interpretations applied to the reactor coolant pressure boundary components are in a'cordance with Regulatory Guides 1.84 and 1.85.
t I l l l l O Amendment D September 30, 1988
CESSAR E!MiriCATIEN l TABLE 5.2-4 RESULTS OF THE INADVERTENT SAFETY INJECTION ACTUATION TRANSIENT ANALYSIS" (FOR A WATER-SOLID RCS) Number of Maximum SCS Relief Maximum Relief Number of Number of Valves RCS Pressure Flowrate Char _ginc i'1mps SIS Pumps Operational (psia) (com) 1 1 1 567 1260 1 2 1 580 2162 1 3 1 592 3055 1 4 1 608 3934 O i I NOTE: a. Including the considerations of energy addition from decay heat and all pressurizer heaters. l
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1 1 O Amendment E l December 30, 1988
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490 - - 480 - l 470 - - 460 - - 450 i l i 1 - I - 1 i i i i i i i i i i I O 2 4 6 8 10 12 14 16 18 20 TIME, SECONDS Amendment E December 30, 1988 O ru Figure SYSTEM 80+ RCP START TRANSIENT
CESSARn % mu (n V 1 a' 5.3 REACTOR VESSEL 5.3.1 REACTOR VESSEL MATERIAL 8 5.3.1.1 Material Specifications The principle ferritic materials used in the reactor vessel are listed in Table 5.2-2. These materials are in accordance with the ASME Boiler and Pressure Vessel Code, Section III. 5.3.1.2 Special Process Used for Manufacturing and Fabrication fi The reactor vessel is fabricated in accordance with the ASME Boiler and Pressure Vessel Code, Section III. No special { manufacturing methods that could compromise the integrity of the j vessel are used. ; The reactor vessel is a vertically mounted cylindrical vessel with a hemispherical lower head welded to the vessel and a removable hemispherical upper closure head. The construction is l basically that of forged rings, forged hemispherical heads, q forged flanges on the closure head and vessel, and forged B . nozzles. The internal surfaces that are 3n contact with the (mL] reactor coolant are clad with austenitic stainless steel. , 1 The reactor vessel consists basically of a vessel flange, three ! shell sections (upper, intermediate and lower) and a bottom head. l , The vessel flange is a forged ring with a machined ledge on the l inside surface to support the core support barrel, which in turn supports the reactor internals and the core. The flange is drilled and tapped to receive the closure studs and is machined to provide a mating surface for the reactor vessel closure seals. , Each shell consists of one 360 degree forged ring. The bottom , l head is constructed of a single hemispherical forging. The three l shell sections, the bottom head forging and vessel flange forging B l are joined together by welding, along with four inlet nozzle forgings and two outlet nozzle forgings to form a complete vessel j assembly. The closure head is fabricated separately since it is joined to the reactor vessel by bolting. The closure head consists of-a head flange and a dome. The head flange is a forged ring. The flange is drilled to match the vessel flange stud hole locations, and the lower surface of the flange is machined to provide a mating surface for the vessel closure seals. The dome is constructed of a single hemispherical forging. The dome and flange are welded together to form the closure head, and the B C
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control element drive mechanism (CEDM) nozzles are welded into the head to complete the assembly. Amendment B 5.3-1 March 31, 1988 L - -_ _ - _
1 C E S S A R EM Ur"icarcu O! 5.3.1.3 BDecial Methods for Nondestructive Examination Prior to, during, and after fabrication of the reactor vessel, , nondestructive tests based upon Section III of the ASME Boiler l and Pressure Vessel Code are performed on all welds and forgings as indicated. The nondestructive examination requirements including calibration methods, instrumentation, sensitivity, reproducibility of data, and acceptance standacds are in accordance with requirements of the ASME B&PV Code, Section III. (See Table 5.2-1). Strict quality control is maintained in critical areas such as calibration of test instruments. All full-penetration, pressure-containing welds are 100% radiographer to the standards of Section III of the ASME Boiler and Pressure Vessel Code. Weld preparation areas, back-chip areas, and final weld surfaces are magnetic-particle or i dye-penetrant examined. Other pressure-containing welds, such as l used for the attachments of nonferrous nickel-chromium-iron I mechanism housings, vents, and instrument housings to the reactor . t vessel and head, are inspected by liquid-penetrant tcsts of the root pass, the lesser of one-half of the thickness or each ' 1/2-inch of weld deposit, and the final surface. Additionally, the base metal weld preparation area is magnetic-particle examined prior to overlay with nickel-chromium-iron weld metal. All forgings are inspected by ultrasonic testing, using longitudinal beam techniques. In addition, ring forgings are tested using shear wave techniques. All carbon-steel and low alloy forgings and ferritic welds are also subjected to magnetic-particle examination after stress relief. All vessel bolting material receives ul'rasonic and I magnetic particle examination during the manufactuiAng process. The bolting material receives a straight-beam, radial-scan, ultrasonic examination with a search unit not exceeding i square-inch area. All hollow material receives a second ultrasonic examination using angle-beam, radial scan with a search unit not exceeding i square inch in area. A reference specimen of the same composition and thickness containing a notch (located on the inside surface) 1 inch in length and a depth of j 3% of nominal section thickness, or 3/8-inch, whichever is less, < is used for calibration. Use of these techniques ensures that no . materials that have unacceptable flaws, observable cracks, or l sharply defined linear defects are used. f i The magnetic-particle inspection is performed both before and after threading of the studs. 5.3-2
CESSARnaincua 1 A Q Upon completion of all post-weld heat treatments, the reactor vessel is hydrostatically tested, and all accessible ferritic weld surfaces, including those used to repair material, are magnetic-particleinspectedinaccordancewithSectionIIIofthelD ASME Boiler and Pressure Vessel Code. 5.3.1.4 Special Controls for Ferritic and Austenitic Stainless Steels Special controls for ferritic and austenitic stainless steels are as follows: A. itegulatory Guide 1.31, Control of Ferrite Content in Stainless Steel Weld Metal,.is addressed in Section 5.2.3.4. D B. Regulatory Guide 1.34, Control of Electroslag Weld Properties, is addressed in Section 5.2.3.3. C. Regulatory Guide 1.43, Control of Stainless Steel Weld Cladding of Low-Alloy Steel Components, is addressed in Section 5.2.3.3. O D. Regulatory Guide 1.44, Control of the Use' of Sensitized Stainless Steel, is addressed in Section 5.2.3.4. E. Regulatory Guide, 1.50, Control of Preheat Temperature for Welding of Low-Alloy Steel, is addressed in Section 5.2.3.3. F. Regulatory Guide 1.71, Welder Qualification for Areas of Limited Accessibility, is addressed in Section 5.2.3.3. G. Regulatory Guide 1.99, Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials, is B addressed in Section 5.3.1.6.7. 5.3.1.5 Fracture Touchness > Iri accordance with 10 CFR 50 Appendix G, Paragraph IV A, the l reactor vessel beltline materials have minimum upper-shelf lD energy, as determined from Charpy V-notch tests on unirradiated specimens in accordance with Paragraphs NB-2322.2(a) of the ASME Code, of 75 ft-lbs. Charpy impact tests will be performed on transversely (weak direction) oriented specimens from the beltline forgings to establish RT as requhed by 10 CFR 50, Appendix G. NDT O l Amendment D 5.3-3 September 30, 1988
1 CESSAR Mai"icaritu ! l O 5.3.1.6 Reactor Vessel Material Surveillance Program I t The irradiation surveillance program for System 80+ will be 1 conducted to assess the neutron-induced changes in the RT (reference temperature) and the mechanical properties of N l reactor vessel materials. Changes in the impact and mechanical 1 properties of the material will be evaluated by the comparison of pro- and post-irradiation test results. The capsules containing the surveillance test specimens used for monitoring the neutron-induced property changes of the reactor vessel materials will be irradiated under conditions which represent, as closely as practical, the irradiation conditions of the reactor vessel. ASTM E-185-82, Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels, and 10 CFR 50 B Appendix H, Reactor Vessel Material Surveillance Program Requirements, present criteria for monitoring changes in the toughness surveillanceproperties of reactor programs. The vessel Systembeltline 80+ materials reactor through vessel lD surveillance program adheres to all of the requirements of ASTM E-185-82 and satisfies the intent of 10 CFR 50, Appendix H. D 5.3.1.6.1 Test Material Selection Materials selected for the surveillance program are those judged most likely to be controlling with regard to radiation g embrittlement according to the recommendations of Regulatory Guide 1.99, Revision 2. Surveillance test materials are prepared from the actual , materials used in fabricating the beltline region of the reactor 1 pressure vessel. The test materials are processed so that thuy l are representative of the materials in the completed reactor l vessel. Specimens are prepared from three metallurgically different materials, including base metal, weld metal and heat-affected zone (HAZ) material. In addition, material is included from a standard heat of ASME D SA-533 Grr.au B Class 1 manganese-molybdenum-nickel steel made available by the USNRC sponsored Heavy Section Steel Technology . (HSST) Program. This standard reference material (SRM) is used as a monitor for Charpy impact tests, permitting comparisons among the irradiation data from operating power reactors and i experimental reactors. Compilation of data generated from post-irradiation tests of these correlation monitors will be carried out by the HSST program. Base metal test material is from a section of the shell course forging selected from the beltline of the reactor vessel. Selection shall be based on an evaluation of initial toughness , Amendment E 5.3-4 December 30, 1988 j
CESSAREnacm. o 3 (characterized by an index temperature such as RT and the estimated enfect of chemical composition and neutrgT)luence f on _i' the toughness during reactor operation. Normally, the forging with the highest adjusted 4 ndex temperature (initial value plus predicted radiation induced increase in index temperature) at end-of-life shall be selected as the surveillance base metal test material. In certain circumstances, however, selection of two separate shell course forgings, from both intermediate and lower ! shell courses, may.be warranted if both should be monitored for their effect on the pressure-temperature operating limits during the design lifetime of the reactor vessel. Weld region test material, representative of the controlling E reactor pressure vessel weld, is produced by welding together sections of forgings from the beltline of the reactor vessel. The HAZ test material is manufactured from a section of the srme forgings used for base metal surveillance test material. The weld metal test material is produced from the same heat of weld-wire or rod and lot of flux used in the beltline of the reactor: vessel. Welding parameters duplicate those used for the beltline welds. Representative stock (archive material) to provide two additional ( sets of test specimens for each material shall be provided with full documentation and identification. 5.3.1.6.2 Test Specimens 5.3.1.6.2.1 Type and Quantity Drop weight, standard and procracked Charpy impact, tensile test, E and compact tension fracture toughness specimens are provided for unirradiated tests. Drop weight tests will be conducted in , accordance with ASTM E-208. Charpy impact tests will be j conducted in accordance with ASTM E-23. Tensile tests are I conducted in accordance with ASTM E-8 and E-21. Correlation of j drop weight and Charpy impact tests to establish RT will be made in accordance with NB2300 of the ASME Code, S N ion III. Charpy impact and tensile- test specimens are provided for post-irradiation tests. The total quantity of specimens furnished for carrying out the , overall requirements of this program is presented in Table 5.3-1. ] For surveillance programs including two base metal forgings, the ) total number of specimens shown in Table 5.3-1 will be distributed between the two forgings, but there will be a sufficient quantity of specimens from each forging to meet the requirements of ASTM E-185-82. B Amendment E 5.3-5 December 30, 1988 i
CESSAR HInce 4 For baseline testing, the base metal specimens shown in Table , 5.3-2, with the exception of longitudinal base metal drop ] weights, will be divided equally for both forgings. For )' irradiation encapsulation, the base metal specimens shown in Table 5.3-3 will be divided equally for both forgings. Three capsule assemblies will contain specimens from one base metal q forging and three capsule assemblies will contain specimens from ' the other base metal forging. 5.3.1.6.2.2 Baseline Specimens The type and quantity of test specimens provided for establishing , the properties of the unirradiated reactor vessel naterials are presented in Table 5.3-2. The data from tests of these specimens l' provide the basis for determining the neutron-induced property changes or the reactor vessel materials. Twelve drop weight test specimens each of base metal weld metal and HAZ material are provided for establishing the NDTT of the unirradiated surveillance materials. These data form the basis for P determination. RT is the reference temperature fro. Nch subsequent neutro N nduced changes are determined. Twenty-four standard Charpy test specimens each of base metal lE (longitudinal and transverse) , weld metal, and HAZ material are provided. This quantity exceeds the minimum number of test I specimens recommended by ASTM E-185 for developing a Charpy impact energy transition curve and is intended to provide a sufficient number of data points for establishing accurate Charpy impact energy transition temperatures for these materials. These , data, together with the drop weight NDTT, are used to establish l an RT NDT f r each maMal. Twelve precracked Charpy impact test specimens each of base metal (longitudinal and transverse) and weld metal material are provided in addition to the standard Charpy impact specimens. E This quantity is sufficient to determine fracture toughness properties (critical stress intensity factors under dynamic i loading) over the range extending from linear clastic to clastic-plastic fracture. Twelve tensile test specimens each of base metal (longitudinal lD ' and transverse), weld metal and HAZ materials are provided. This l quantity also exceeds the minimum number of test specimens recommended by ASTM E-185 and is intended to permit a sufficient number of tests for accurately establishing the tensile properties for these materials at a minimum of three test temperatures (e.g., ambient, operating and design). Amendment E 5.3-6 December 30, 1988
CESSAREE&b- j l O I V ! Eight compact tension lt test specimens and four 1/2t compact , tension test specimens each of base metal (transverse) and weld metal material are provided to augment the fracture toughness ; data determined from the precracked Charpy tests. This quantity of specimens is sufficient to determine fracture toughness ) properties over the range extending from linear elastic to elastic-plastic fracture behavior. ) 5.3.1.6.2.3 Irradiated Specimens I l Tensile test, standard and precracked Charpy impact and 1/2t' l l compact tension test specimens are used for determining changes E in the strength and static and dynamic toughness properties of the materials due to neutron irradiation. A total of 333 standard Charpy impact, 81 precracked Charpy impact, 60 1/2t compact tension and 36 tensile test specimens are provided. The I type and quantity of tests specimens provided for establishing the properties of irradiated materials over the lifetime of the vessel are presented in Table 5.3-3. Compact tension specimens provided for capsule irradiation are precracked prior to insertion to reduce the time required for post irradiation testing. The types and quantity of specimens provided exceed the p minimum requirements of ASTM E-185-82. I 1 Q 5.3.1.6.3 Surveillance Capsules The surveillance test specimens are placed in corrosion resistant capsule assemblies for protection from the primary coolant during irradiation. The capsules also serve to physically locate the test specimens in selected positions within the reactor vessel ; and to facilitate the removal of a desired quantity of test ' specimens when a specified radiation exposure has been attained. Six surveillance capsule assemblies are provided for the reactor vessel. Three of the capsule assemblies contain precracked Charpy test specimens (PC Capsule) and three capsule assemblies contain 1/2t compact tendion specimens (CT capsule). The type and quantity of specimens contained in each type of capsule E assembly is presented in Table 5.3-4. A, typical capsule assembly, illustrated in Figure 5.3-1, consists of a series of three specimen compartments, connected by wedge couplings, and a lock assembly. Each compartment enclosure of the capsule assembly is internally supported by the surveillance specimens and is externally pressure tested to 3125 psi during final fabrication. The wedge couplings also serve as end caps for the specimen compartments and position the compartments w3'.nin the capsule holders which are attached to the reactor vossel cladding. The lock assemblies fix the locations of the Amendment E 5.3-7 December 30, 1988
CESSAR nair"lCATION O capsules within the holders by exerting axial forces on the wedge coupling assemblies; this causes the wedges to exert horizontal ; forces against the sides of the holders preventing relative motion. The lock assemblies also serve as a point of attachment c for the tooling used to remove the capsules from the reactor. Each capsule assembly is made up of two Charpy, flux, compact tension compartment assemblies and one temperature, flux, E tensile, Charpy compartment assembly. Each capsule compartment is assigned a unique identification so that a complete record of test specimen and flux / temperature monitor (tensile-monitor) compartment. Each capsule compartment is assigned a unique identification so that a complete record of test specimen location within each compartment can be maintained. 5.3.1.6.3.1 Charpy, Flux and Compact Tension lE Compartment Assembly I This assembly (Figure 5.3-2) contains 15 impact test specimens and a set of five flux spectrum monitors in the top section. The . bottom section contains either 18 precracked Charpy test ! specimens or ten 1/2t compact tension specimens. The Charpy test p I specimens are arranged vertically in 1 x 3 arrays and are oriented with the notch toward the reactor core. The 1/2t 4 compact tension specimens are oriented so that the opening of the crack starter notch is facing the top of the compartment. This orientation will result in a neutron flux gradient parallel to the crack front. The temperature differential between the specimens and the reactor coolant is minimized by using spacers between the specimens and the compartment and by sealing both sections of the assembly in an atmosphere of helium. This quantity of specimens provides an adequate number of data points for establishing a Charpy impact energy transition curve for a given irradiated material. Comparison of the unirradiated and j irradiated Charpy impact energy transition curves permits i determination of the RT NDT changes due to irradiation for the I various materials. { 5.3.1.6.3.2 Temperature, Flux, Tensile and Charpy lE l Compartment Assembly l This assembly (Figure 5.3-3) contains 3 base metal (transverse) tensile test specimens and 12 impact test specimens in the top section. The tension specimens are placed in a housing machined I to fit the compartment. Split spacers are placed around the gage D length of the specimens to minimize the temperature differential between the specimen gage length and the coolant. The impact specimens are arranged vertically in 1 x 3 arrays and are oriented with the notch toward the reactor core. Spacers are utilized between the test specimens and the compartment. The l Amendment E 5.3-8 December 30, 1988
CESSAR HE"icarie. O bottom section contains a set of nine flux spectrum monitors, a l set of temperature monitors, 9 SRM (Standard Reference Material) J D impact test specimens and 3 weld metal tension test specimens. Both compartment sections are sealed within an atmosphere of helium. 5.3.1.6.4 Neutron Irradiation and Temperature Exposure The changes in the RT of the reactor vessel materials are derived from specimens T radiated to various fluence levels and 1 in different neutron energy spectra. In order to permit accurate i predictions of the RT of the vessel materials, complete i information on the n E ron flux energy spectra, and the irradiation temperature of the encapsulated specimens must be available. 5.3.1.6.4.1 Flux Measurements Fast neutron flux measurements are obtained by insertion of . threshold detectors into each of the six irradiation capsules. Such detectors are particularly suited for the proposed application, because their effective threshold energies lie - in the range of interest (0.5 to 15 MeV). V These neutron threshold detectors and the. thermal neutron ! detectors, presented in Table 5.3-5, can be used to monitor the i thermal and fast neutron spectra incident on the test specimens. These detectors possess reasonably long half-lives and activation cross sections covering the desired neutron energy range. Three sets ofEach assembly. flux spectrum monitors are includedain each capsule lD detector is placed j inside sheath which I ; identifies the material and facilitates handling. Cadmium covers are used for those materials (e.g., uranium, nickel, copper, cobalt and neptunium) which have competing neutron capture 3 activities. The flux monitors are placed in holes drilled in stainless steel housings as shown in Figure 5.3-3 at three axial locations in each capsule assembly (Figure 5.3-1) to provide an axial profile of the level of fluence which the specimens attain. In addition to these detectors, the program also includes correlation monitors (Charpy impact test specimens made from a ; reference heat of ASME SA-533 Grade .B Class 1, manganese-molybdenum-nickel steel) which are irradiated alonglD with the specimens made from reactor vessel materials. The changes in impact properties of the reference material O V provide a cross-check on the dosimetry in any given surveillance (SRM) f 1
+
Amendment E 5.3-9 December 30, 1988 L__________________-__
CESSAR nSincari:n l I O program. These changes also provide data for correlating the , I results from this surveillance program with the results from experimental irradiations and other reactor surveillance programs using specimens of the same reference material. 5.3.1.6.4.2 Temperature Estimates Because the changes in mechanical and impact properties of irradiated specimens are highly dependent on the irradiation temperature, it is necessary to have knowledge of specimens as well as the pressure vessel. During irradiation, instrumented capsules are not practical for a surveillance program extending over the design lifetime of a power reactor. The maximum temperature of the irradiated specimens can be estimated with reasonable accuracy by including in the capsule assemblies small pieces of low melting point alloys or pure metals. The compositions of candidate materials with melting points in the operating range of power reactors are listed in Table 5.3-6. The monitors are selected to bracket the operating temperature of the reactor vessel. The temperature monitors concist of a helix of low melting alloy wire inside a sealed quartz tube. A stainless steel weight is provided to destroy the integrity of the wire when the melting point of the alloy is reached. The compositions and therefore the melting temperatures of the temperature monitors are differentiated by the physical lengths of the quartz tubes which contain the alloy wires. A set of temperature monitors is included in each capsule D assembly. The temperature monitors are placed in holes drilled in stainless steel housings as shown in Figure 5.3-3 and are also placed in three axial locations in each capsule assembly (Figure 5.3-1) to provide an axial profile of the maximum temperature to which the specimens are exposed. 5.3.1.6.5 Irradiation Locations The test specimens are enclosed within six capsule assemblies the axial positions of which are bisected by the midplane of the core. A summary of the specimens contained in each of these ; capsule assemblies is presented in Table 5.3-4. j i The test specimens contained in the capsule assemblies are used for monitoring the neutron-induced property changes of the reactor vessel materials. These capsules, therefore, are positioned near the inside wall of the reactor vessel so that the , irradiation conditions (fluence, flux spectrum, temperature) of the test irradiation specimens resemble as closely as possible the conditions of the reactor vessel. The neutron 8l W 3 Amendment D i 5.3-10 September 30, 1988
v CESSAR ninne- l J (* i fluence of the test specimens is expected to be 1.15 times higher than that seen by the' adjacent vesselapproximately wall, so lE , the measured changes in properties of the surveillance materials 1 will be able to predict the radiation induced changes in the D reactor vessel beltline materials. The capsule assemblies are placed in capsule holders positioned l circumferentially about the core at locations which include the ( regions of maximum flux. Figure 5.3-4 presents the typical exposure locations for the capsule assemblies. l All capsule assemblies are inserted into their respective capsule I holders during the final reactor assembly operation. l 5.3.1.6.6 Withdrawal Schedule The capsule assemblies remain within their holders until the specimens contained therein have been exposed to predetermined D , levels of fast neutron fluence. At that time, the capsule ' assembly is removed and the surveillance materials are evaluated. The capsule assembly removal schedule and the associated target fluence are presented in Table 5.3-7. . D ( Q The target fluence levels for the surveillance capsules are determined at the azimuthal locations for the recommended E ; withdrawal schedule of ASTM E-185-82, extended to a design life I of 60 years (48 EFPY). The fluence values in Table 5.3-7 are f accurate within +30%, -30%. The uncertainty is composed of q errors in the calculational method and errors in the combined i radial and axial power distribution. l Withdrawal schedules may be modified to coincide with those refuelingoutagesorplantshutdownsmostcloselyapproachingthelE withdrawal schedule. The two standby capsules are provided in the event they are needed to monitor the effect of a major core change or annealing of the vessel, or to provide supplemental l toughness data for evaluating a flaw in the beltline. l 5.'3.1.6.7 Irradiation Effects Prediction Basis Irradiation induced RT T shift and reduction of upper shelf energy are predicted bated on Regulatory Guide 1.99 Revision 2, p
" Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." Predicted changes in RT and upper shelf energy are used to select the surveillarbeT materials (Section 5.3.1.6.1) and to formulate the initial heatup and cooldown limit curves for plant operation. Once actual post-irradiation surveillance data become available for each reactor vessel, these data will be used to adjust plant operating
) J limit curves. l l Amendment E 5.3-11 December 30, 1988 I L________________
4 CESSARnnince I l O, ! 5.3.1.7 Reactor Vessel Fasteners The bolting material for the reactor vessel closure head is fabricated from SA 540, B23 or B24, Class III material. This material conforms to the requirements of 10 CFR 50, Appendix G l and Regulatory Guide 1.65, " Materials and Inspections for Reactor Vessel Closure Studs." C-E specifies the use of a manganese phosphate coating on threads of studs, nuts and washers to improve anti-galling properties and resistance to corrosion. In addition, Super Moly lubricant (containing molybdenum disulfide) is specified to be added' to threads and bearing surfaces at installation to further enhance ; anti-galling properties. Laboratory testing and field experience to date have shown no evidence of deleterious breakdown of either phosphate coating or lubricant. 5.3.2 PRESSURE-TEMPERATURE LIMITS All components in the Reactor Coolant System (RCS) are designed to withstand the effects of cyclic loads due to RCS temperature and pressure changes. These cyclic loads are introduced by normal unit load transients, reactor trips, and startup and shutdown operation. During unit startup and shutdown, the rates of temperature and pressure changes are limited by the Technical Specifications. The design number of cycles for heatup and cooldown is based on a rate of 100"F/hr. l The maximum allowable RCS pressure at the corresponding minimum allowabic temperature is based upon the stress limitations for brittle fracture. These limitations are derived using linear E l clastic fracture mechanics (LEFM) principles, the procedures i prescribed by Appendix G to Section III of the ASME Code, l
" Protection Against Nonductile Fracture," Appendix G to 10 CFR 50, " Fracture Toughness Requirements," NRC Regulatory Guide 1.99,
" Radiation Embrittlement of Reactor Vessel Materials," and the procedures recommended by WRC Bulletin 175, "PVRC Recommendations on Toughness Requirements for Ferritic Materials." Compliance i with Appendix H to 10 CFR 50, " Reactor Vessel Material 1 Surveillance Program Requirements," is discusomt in Section 5.3.1.6. l l
)
)
E Amendment E ! l 5.3-12 December 30, 1988 l l l
CESSARn h mw n v 5.3.2.1 P-T Limit Curves 5.3.2.1.1 Material Properties Pressure-temperature limitations (P-T Limits) are determined, in l part, using material property test data for Reactor Coolant j Pressure Boundary materials, as required by Appendix G to Section ] III of the ASME Code. Based on considerations of existing l material property test data, an initial RTN fr the reactor l' vessel beltline material of -20*F is assumed.DTThe initial RT for the remaining material of the Reactor Coolant System of +1 b is also assumed. l As a- result of fast neutron irradiation in the region of the core, the RT of irradiated material will increase with l t operation. N techniques used to analytically and 1 I experimentally predict the integrated fast neutron (E 2 1 Mev) ! fluxes of the reactor vessel are described in Sections 5.3.1.4 I and 5.3.1.6. Analytical predictions of the shift in RT l reactor vessel beltline materials are based on the procedubI described in NRC Regulatory Guide 1.99. Based on these procedures, the maximum calculated shift (without any margin) in (g
'j the RT at 1/4t over a 60-year design life is 49'F. The total predicbd shift (calculated with a conservative margin of 50*F for the beltline forging) is 99'F. The corresponding RT shift value for the beltline girth seam weld is 88'T (38'F cbbulated shift plus a conservative margin of 50*F).
l The measured shift in RT for a sample is applied to the adjacent section of the reNNor vessel for later stages. in plant E life because the measured neutron spectra and flux at the samples and reactor vessel inside radius should be nearly identical. The measured shift in RT for a sample is adjusted for the difference in calculatNE' lux magnitudes between the sample and the point of interest In the reactor vessel wall. The maximum exposure to the reactor ve. .ael will be obtained from the measured l sample exposure by application of the calculated azimuthal neutron flux variation. The neutron fluence and the actual shift in RT will be established periodically during plant operation by te b[ng the reactor vessel surveillance material samples which are irradiated in holders secured to the inside wall of the reactor vessel as described in Section 5.3.1.6 and shown in Figure 5.3-4. Since the surveillance materials are irradiated l Very close to the vessel inside surface, the samples when tested will demonstrate the accuracy of the RT shift prediction as wellastheeffectofanyadditionallonkermagingphenomenon. l The pressure-temperature limits will then be adjusted n periodically, if necessary, to stay within allowable stress ( i limits during normal operation.
-g Amendment E 5.3-13 December 30, 1988
CESSAR MSinCATIEN O! ! Figures 5.3-Sa and 5.3-5b are typical examples of the intended l pressure-temperature limitations determined in accordance with ( Appendix G to 10 CFR 50 for normal operation of the Reactor j Coolant System. The P-T Limit curves are based on following: A. Minimum Boltup Temperature The minimum boltup temperature is principally governed by Appendix G to Section III of the ASME Code. The ASME Code 3 requires that when the flange and adjacent shell region are l stressed by the full bolt preload and by pressure not exceeding 20% of the preoperational system hydrostatic test - pressure, the minimum metal temperature in the stressed region must be at least the initial RT t plus any effects of irradiation. From the assumption thaF the initial RT for all regions of the RCS other than the beltline is +10 Y and that the flange and adjacent shell regions are not subjected to significant irradiation, the minimum boltup ! temperature per this requirement is +10*F. Based on considerations of typical refueling water temperatures, ; however, the minimum boltup temperature is taken to be 60*F. B. Maximum Pressure Below Lowest Service Temperature et d The maximum allowable pressure below the lowest service temperature is defined by paragraph NB-2332, Section III of E the ASME Code, to be 20% of the preoperational hydrostatic test pressure. This test pressure is 125% of the design pressure (2500 psia), or 3125 psia, and 20% of this value is i 625 psia. Therefore, the maximum pressure below the lowest service temperature is 625 psia (actual fluid pressure at the inside surface of the beltline) . C. Lowest Service Temperature The lowest service temperature (LST) is defined by paragraph NB-2332, Section III of the ASME Code, to be the minimum , allowable temperature at pressures above 20% of the l preoperational hydrostatic test pressure. This value is l defined to be no lower than RTI + 100*F, where the RT is considered to be for the mos@ limiting component in b$ RCS other than the beltline. From the assumption that the initial RT for all regions of the RCS other than the beltline ibN10*F, the LST per this requirement is 10*F + 100*F = +110*F. , I Amendment E I 5.3-14 December 30, 1988 [ J l
CESSAR 8!ninco,, n V D. Operation, Heatup and Cooldown Curves
- 1. Reactor Vessel ~ Beltline P-T Limits for the reactor vessel beltline are exanined for heatup and cooldown conditions. For the heatup analysis, both the 1/4t and 3/4t locations are examined. For heatup, the thermal stresses, o , are ->
compressive at the inside surface and tensile bt the outside surface. The membrane stresses due to pressure, o are always tensile, but more so at the inside than*,outside surface. As a result, the total stress will always be greater at the outside than the , inside surface. However, the . maximum allowable I stresses, taking into account irradiation effects, will decrease more at the inside than outside surface because the effects of irradiation are more pronounced there. Which surface stresses will. approach this maximum first is not always clear and is a function of several variables. Therefore, both ' locations are examined for the heatup transient. For the cooldown analysis, only the 1/4t location needs (' to be examined. During cooldown, the thermal stresses are tensile at the inside surface and compressive at i the outside. Again, membrane stresses due to pressure ! are always positive. As a result, the total stress l will always be greater at the inside than the outside surface. Since the maximum allowable stresses, taking into account irradiation effects, will decrease more at , the inside than outside surface, it is clear that the ! total stress at the inside surface will approach the i maximum allowable before those at the outside surface. { Therefore, only the 1/4t location needs to be examined for the cooldown transient. E Heatup and cooldown rates from O'F/hr (" Isothermal" pressurization) to the design limit of 100'F/hr at 10
'F/hr intervals are examined in determining the allowable heatup and cooldown rates as a function of temperature to meet low-temperature overpressure protection (LTOP) requirements for the Reactor Coolant system. LTOP considerations are discussed in Section 5.2.2.10.
For the beltline analysis, during normal operations, the following condition is maintained: O Amendment E 5.3-15 December 30, 1988
CESSAREnace - O IR " Im + It where, K = Reference stress intensity factor IR specified by Figure G-2210-1 in Appendix G to Section III of the ASME Code. intensity factor for membrane KI " = Stress stress due to pressure.
=a M where M = Membrane correction f$ct$r, defined i$ Figure G-2214-1 in Appendix G to Section III of the ASME Code, Pr o = , where P = internal reactor vessel m t pressure, psia, r = inside reactor vessel radius, in.,
t = reactor vessel wall thickness, in. K = Stress intensity factor for thermal It stress.
= AT M , where M = Thermal correction fa$ tor defined bn Figure G-2214-2 in E Appendix G to Section III of the ASME Code, AT y = Maximum steady-state temperature differential across the reactor vessel wall, 'F.
The right side of the above equation (2KI + K is . calculated for various pressures and then Eet eq$)1 to l the left side (Km) . The 2 represents a safety factor j required by Appedd'ix G to Section III of the ASME Code. The minimum allowable temperature corresponding to a given pressure can then be calculated from this equation. The resulting combination represents a (maximum) pressure-(minimum) temperature limit which must not be exceeded to ensure nonbrittle material behavior. A set of such pressure-temperature coordinates define a P-T limit curve for a specific heatup or cooldown rate, at a particular point in plant life. Figures 5.3-Sa and 5.3-5b illustrate representative P-T Limits for the end-of-(plant) life-60 years, or 48 EFPY's. Figure 5.3-6 will illustrate the corresponding allowable baatup and cooldown rates. Amendment E 5.3-16 December 30, 1988
C E S S A R ! anne m :u V The pressure used in computing K and K and the corresponding computed temperature
- are foht,the crack tip. Correction factors for pressure and temperature are taken into account when expressing P-T Limits !
graphically so as to show the limits in terms of indicated pressurizer pressure and indicated RCS temperature. This is done for P-T Limit Curve presentation in Technical Specifications. The correction factors take into account the effects of instrument error, pressure differentials due to flow in ' the Reactor Coolant System, static pressure differentials due to elevation differences, and temperature differentials due to thermal gradients in the reactor vessel wall. The P-T Limits shown in Figures 5.3-5a and 5.3-5b, however, are in terms of actual fluid conditions at the reactor vessel inside wall without any correction factors applied. K is a function of the temperature, T, and the ab$ustedreferencetemperature, ART, of the material at the crack-tip. The analytical expression for K i8 IR K = 26.78 + 1.233 e[0.145(T - ART + 160)] as specified by Appendix G to Section III of the ASME Code. The calculation of the ART is in accordance with the procedure described in NRC Regulatory Guide 1.99. The ART is a function of the initial RT NDT of the E , material, (Initial)RT DT, the shift in RT due to I irradiation over a pe[ igd of time, ART NDT~' " " "" "W margin. The equation for the ART is ART = (Initial) RTNDT + 0 NDT + Margin. The shift, ART NDT, s calculated from LRT yg7 = (CF) f(0.28 - 0.10logf) where CF is a chemistry factor and f is the neutron fluence at the point of interest in the reactor vessel i wall. The fluence f is calculated from the surface l fluence, f 8 surface' ! f=fsurface(e where x is the depth into the reactor vessel wall. 1 _ The margin is calculated based on
'V Margin = 2 /g y+y2 2 Amendnent E 5.3-17 December 30, 1988
CESSAREnnnce O i where o is the standard deviation characteristic of the (Iditial)RT deviation for AggT data, and o is the standard i data as desbibed in Regulatory l Position C.1.1 of S C Regulatory Guide 1.99. o is I assigned a value of zero because measured propedies l will be used to establish RT not estimated values. i o for forgings !s17*FperNku,latoryGuide1.99, but ~ tbe full margin (2o) for prediction purposes is assigned the conservative value of 50*F to cover unanticipated long-term aging phenomena. Once confirmatory data are available from the reactor vessel 1 surveillance program, the margin will be reduced in subsequent shift calculations consistent with Regulatory Guidelines.
- 2. Reactor Vessel Flange PT Limits for the reactor vessel flange are examined for heatup and cooldown in accordance with tha procedures specified by G-2220 of Appendix G to Section III of the ASME Code. For flange analysis, during normal operations, the follwing condition is maintained:
E K = 2K Iprimary + K secondary IR t "boltup primary " pressure boltup K + b t secondary pressure
= Stress intensity for membrane and K mpressure stress due to pressure, '
mpressure *' K = Stress intensity factor for "boltup membrane stress due to boltup,
* "mboltup "'
K = Stress intensity factor for bending l b boltup stress due to boltup, j bboltup ' o, = Membrane (hoop) stress due to pressure pressure, psi, o = Membrane stress due to boltup, psi, o = Bending stress due to boltup, psi, 1 Amendment E l S.3-18 December 30, 1988 i
CESSAR inEncimw j l4 I V .
= Membrane correction factor, [ili, !
M"
= 1.1 M g 6 /a/Q I M = Bending correction factor, /in, i b =M B f /"/0 where, M g and M B = correction factors defined by Figure A3-1 and A3-2 in l Appendix 3 to WRC - :
Bullentin 175. a = crack depth, in. ! Q = the flaw shape : factor defined in f Appendix 3 to WRC ' Bulletin 175. ; K = Stress intensity factor for bending b ] j pressure stress due to pressure, i
" "bpressure o
b
= Secondary bending stress due to p
- pressure pressure, ps1.
t K = Stress intensity factor for thermal t i stress,
=
)
(AT) E where, a = Coefficient of thermal expansion, in/in, ! E = Modulus of Elasticity, psi, , p = Poisson's ratio, AT = Temperature differential causing the stress,
- F .,
l The left side of the above equation (K7p) is calculated ! for various temperatures. The right side (2K + K is calculated as a function of IhfbE$6Ee. Tb5*E8 bmbE) allowable pressure corresponding to a given l temperature can then be calculated from this equation. i The resulting (maximum) pressure-(minimum) temperature coordinates define the flange PT Limit for a specific heatup or cooldown, at a particular point in plant j life. In no case is the flange minimum temperature limit allowed to be less than the flange RT 120 *F during normal operation (when the pressure Ng{ce+eds 20% of the preoperational system hydrostatic pressure) , by 1 l-Amendment E 5.3-19 December 30, 1988
l 1 CESSAR MEnCATl:N e RT + 90 'F during hydrost&.ic pressure tests and leb tests, or by RT + -
F when the core is critical, in accordance Y th App mdix G to 10 CFR 50.
E. In-service Inspection and Hydrostatic Test Curves P-T Limits for in-service inspection and hydrostatic test are developed in the same manner as for the reactor vessel beltline described in D.1 above. The exception is that a safety factor of 1.5 is applied to the stress intensity factor for membrane stress due to pressure, K rather than 2, as allowed by G-2400 in Appendix G to Secd1"o,n III of the E ASME Code. i F. Core Critical Curves The core critical curve is intended to provide additional s margins of safety during core operation. The limit is defined as 40*F above the minimum allowable temperature for heatup or cooldown nor lower than the minimum temperature allowable for inservice hydrostatic pressure test, in accordance with paragraph IV.A.3 in Appendix G to 10 CFR 50. < 5.3.2.2 O_pe ratinct Procedures Pressure-temperature limitations and additional information are I described in the Technical Specifications. The pressure-temperature limit curves provided have been prepared in accordance with Appendix G, ASME Code, Section III. Maintenance of Reactor Coolant System pressure and temperature within these prescribed limits ensures that the integrity of the reactor coolant pressure boundary is maintained. 1 1 Il D i e l l Amendment E j 5.3-20 December 30, 1988
CESSAREnnnem O 5.3.3 REACTOR VESSEL INTEGRITY C-E designs reactor pressure vessels and specifies corresponding E fabrication methods. C-E has been involved in reactor vessel design and fabrication since the late 1950s, and this proven ! expertise is reflected in the System 80+ reactor vessel and the D satisfactory performance of large numbers of reactor vessels in operating plants. j Vessel integrity is ensured because proven fabrication techniques are employed and because weil characterized steels, which exhibit uniform properties and consistent behavior, are used. The characterization of these materials was established through industrial and governmental studies which examined the prefabrication material properties through to irradiated service operation. Inservice inspection and material surveillance programs are also conducted during the service life of the vessel, which further ensure that vessel integrity is maintained. 5.3.3.1 DesiQn Applicable design codes are found in Table 5.2-1. A schematic of ; the reactor vessel is shown in Figure 5.3-5. information on design may be found in Section 5.3.1.2. Additional lD ( The design permits all required inspections to be performed, and does not preclude access to areas requiring inservice inspection. 5.3.3.2 Materials of Construction The materials used in the construction of the reactor vessel, as listed in Table 5.2-2, are in accordance with Section III of the ASME Boiler and Pressure Vessel Code. 5.3.3.3 Fabrication Methods Fabrication of the reactor vessel is described in Section
- 5. 3 .1. 2 . Fabrication processes used in construction of the reactor vessel comply with Sectionn III and IX of the ASME B&PV Code.
5.3.3.4 Inspection Requirements i Inspection requirements of ASME Code, Section III are discussed in Section 5.3.1.3. 5.3.3.5 shipment and Installation [ The requirements of Regulatory Guide 1.38 are followed in the
\~- packaging and shipment of the reactor vessel. Regulatory Guides ;
1.37 and 1.39 and are addressed addressrequirementsduringtheconstructionphaselB in the site-specific SAR. Amendment E 5.3-21 December 30, 1988 L_----_---
1 CESSARnn% . O The reactor vessels are prepared to be shipped by barge or rail to the site, while mounted on the shipping skid used for installation. The vessels are protected by closing all openings (including the top of the vessel) with wooden shipping covers. The closure heads are shipped with separate skids and covers. Vessel surfaces and covers are sprayed with a strippable coating for protection against corrosien during shipping and installation. Prior to the welding of inter-connecting piping, and installation of insulation; the temporary protective coating is removed by peeling. 5.3.3.6 Operating Conditions See site-specific SAR. B 5.3.3.7 In-service Surveillance The reactor vessel surveillance program is described in detail in Section 5.3.1.6. It is designed on the basis of 10 CFR 50, Appendix H and ASTM E-185-82. Standard reference material to corroborate the post-irradiation surveillance data and precracked l B Charpy impact specimens to enable determination of fracture toughness properties before and after irradiation are included in the program. Standard Charpy specimens in excess of the number required by E-185-82 are included for key materials to increase lB the accuracy in defining post-irradiation index temperatures. When combined with the use of highly radiation resistant materials in the beltline of the reactor vessel, this surveillance program provides maximu. tasurance, consistent with commercial requirements, of the integr,ty of the reactor pressure vessel in terms of strength and fracture resistance. E l O Amendment E 5.3-22 December 30, 1988
CESSAR E!ninem:n n U TABLE 5.3-1
]
10TAL QUANTITY OF SPECIMENS I Quantity Type of Base Weld SDecimen Orientation Metal tietal HAZ SRM* Totals Drop Weight Transverse 12 12 12 - 36 Standard Charpy Longitudinal 51 - - 69 120 Transverse 114 114 96 - 324 Precracked Charpy Longitudinal' 39 - - - 39 D Transverse 39 39 - - 78 ! Compact Tension It Transverse 8 8 - - 16 l l 1/2t Transverse 34 34 - - 68 I Tensile Test Longitudinal 12 - - - 12 lE Transverse 30 30 - - 60 (O Total 339 237 108 69 753 l 1 l 1 l
- Standard Reference Material characterized by Heavy Section Steel Technology (HSST) l Program 1
1 l O Amendment E December 30, 1988
CESSAR !annem:,. O TABLE 5.3-2 TYPE AND OUANTITY OF SPECIMENS FOR BASELINE TESTS Ouantity Type of Base Weld Specimen Orientation Metal Metal HAZ SRM* Totals Drop Weight Transverse 12 12 12 - 36 Standard Charpy Longitudinal 24 - - 15 39 D Transverse 24 24 24 - 72 Precracked Charpy Longitudinal 12 - - - 12 Transverse 12 12 - - 24 Tensile Test Longitudinal 12 - - - 12 lE Transverse 12 12 - - 24 Compact Tension It Transverse 8 8 - - 16 1/2t Transverse 4 4 - - 8 g O Total 120 72 36 15 243 l 4 l l l
- Standard Reference Material i
I I l l l O Amendment E l December 30, 1988 l
CESSAR ENHncamw o V TABLE 5.3-3 TYPE AND OUANTITY OF SPECIMENS FOR IRRADIATION EXPOSURE AND IRRADIATED TESTS Ouantity Type of Base Weld SDecimen Orientation Metal Metal HA7. SRM* Totals - l Standard Charpy Longitudinal '27 - - 54 81 Transverse 90 90 72 - 252 D Precracked Charpy Longitudinal 27 - - - 27 Transverse 27 27 - - 54 Tensile Test Transverse 18 18 - - 36_ lE Compact Tension 1/2t Transverse 30 30 - - 60 D Total 219 165 72 54 460 O O 1
- Standard Reference Material l
lO Amendment E December 30, 1988
~ CESSAR 8lnincum l C ( TABLE 5.3-4 TYPE AND OUANTITY OF SPECIMENS CONTAINED IN EACH IRRADIATION CAPSULE ASSEMBLY PCy Capsule Assembly CT Capsule Assembly lE Standard Precracked Standard 1/2t Compact Material Charpy. Charpy Tension Charpy Tension Tension Base Metal 9 9 - - - - D (Longitudinal) Base Metal 15 9 3 15 10 3 (Transverse) Weld Metal 15 9 3 15 10 3 Heat-Affected 12 - - 12 - - (HAZ) FT Standard Reference U# Material (SRM) 9 - - 9 - - 60 27 6 51 20 6 O ,V Amendment E December 30, 1988
C E S S A R EEnne. m . O TABLE 5.3-5 CANDIDATE MATERIALS FOR NEUTRON THRESHOLD DETiCTORS Material Reaction Threshold Enercy (MeV) Hal f-Life 93 Niobium Nb (n,n) Nb 93m 0.03 16 years B Neptunium Np237 (n,f) Cs137 0.5 30.2 years 238 Uranium U (n,f)'Cs 137 0.7 30.2 years Iron Fe54 (n,p) Mn54 4.0 314 days Nickel NiS8 (n,p) CoS8 5.0 71 days Copper Cu63 (n,a) Co60 7.0 5.3 years Titanium Ti46 (n,p) Sc46 8.0 84 days Cobalt CoS9 (n,y) Co60 Thermal 5.3 years O Amendment D September 30, 1988
l CESSAR EnEncmeu O TABLE 5.3-6 COMPOSITION AND MELTING POINTS OF CANDIDATE MATERIALS FOR TEMPERATURE MONITORS Composition Melting Temperature (wt%) (*F) 80 Au, 20 Sn 536 90.0 Pb, 5.0 Sn, 5.0'Ag 558 97.5 Pb, 2.5 Ag 580 97.5 Pb, 0.75 Sn, 1.75 Ag 590 O l O l l
i CESSAR Ennneui ,, 1 O I i TABLE 5.3-7 CAPSULE ASSEMBLY REMOVAL SCHEDULE l 1 i Target ] Azimuthal Removal Fluenge ! Capsule Location Time (n/cm ) j 38' 1 E0L ------- 2 43' Standby ------- 18 3 137' 5-8 EFPY 9.0x10 4 142' Standby ------- I 5 230' 22-26 EFPY 3.6x10 l9 19 E 6 310' 38-42 EFPY 6.0x10 O NOTE: Schedule may be modified to coincide with those refueling outages or schedule shutdowns most closely approximating the withdrawal schedule. l O . V L Amendment E December 30, 1988
v 0 5 g = LOCK ASSEMBLY I
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NOZZLE SCHEDULE nk gy:% u xg s SERVICE ) COOLANT INLET 4 COOLANT OUTLET 2 CEDM NOZZLES 97 1 j INSTRUMENTATION 61 Tt i T.P'- C LOSU D RE VENT 1 ja 14'-7" rfq 1 l SEAL LEAK MONITOR 1 10" DVI NOZZLE (8.5" 1.D.) 30"ID \ 42"ID INLET I I OUTLET i NOZZLE -W _ NOZZLE 91/16" + e ,, ,, CLADDING 1/8" MIN
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'lNSTRUM ENT NOZZLES Amendment E December 30,1988 ggg REACTOR VESSEL 53-7
CESSAR MEncamn 1 fh 5.4 COMPONENT AND SUBSYSTEM DESIGN 1 5.4.1 REACTOR COOLANT PUMPS The reactor coolant pumps provide sufficient forced circulation flow through the Reactor Coolant System to assure adequate heat removal from the reactor core during power operation. A low limit on reactor coolant pump flow rate (i.e., design flow) is established to assure that Specified Acceptable Fuel Design Limits (SAFDLs) are not exceeded. Design flow is derived on the basis of the thermal-hydraulic considerations presented in Section 5.2. The reactor coolant pump and motor assembly in conjunction with the flywheel, provide sufficient coastdown flow following loss of . power to the pumps to assure adequate core cooling. 1 The reactor coolant pump pressure boundary is designed for the transients given in Section 3.9 so that the ASME Code Section III allowable stress limits are not exceeded for the specified number of cycles. Stress criteria concerning earthquake and pipe rupture conditions are presented in Section 3.9.3. m (L The design overspeed of the reactor coolant pump is 125 percent of normal speed. 5.4.1.1 Pump Flywheel Intecrity A. The material used to manufacture the flywheel of the reactor coolant pump motor will be produced by a commercially acceptable process that minimizes flaws, such as the vacuum melt and degassing process. This provides adequate fracture toughness properties under reactor operating conditions. The acceptance criteria for flywheel design will be compatible with the safety philosophy of the Pressure Vessel Research Committee (PVRC) of the Welding Research Council D (WRC) primary coolant pressure boundary criteria as appropriate considering the inherent design and functional requirement differences between the pressure boundary and the flywheel. 1
- 1. The reference nil-ductility transition temperature I (RT ) of the material, as determined per ASME Code D NB- N 1(a), will be no greater than 10'F.
- 2. The Charpy V-notch (Cv) upper shelf energy level, in the " weak" (Wr) direction, as obtained per ASTM-A-370 will be no less than 50 ft-lb. A minimum of three Cv specimens will be tested from each plate or forging.
Amendment D 5.4-1 September 30, 1988
CESSAR 8EnCATl"N O
- 3. The minimum fracture toughness of the material at the normal operating temperature of the flywheel will be equivalent to a dynamic stress intensity factor (K dynamic) of at least 100 ksi/in. Compliance will N demonstrated by either of the following:
- a. Testing of the actual material of the flywhed to establish the K d value at the normal operating temperhur(e,ynamic) or !
- b. Use of a lower bound fracture toughness curve obtained from tests on the same type of material.
The curve will be translated along the temperature coordinate until the KI (dynamic) value of 45 ! ksi/in. is indicated at khe NDT of the material, as obtained from drop-weight tests.
- 4. Each finished flywheel will be subjected to a 100 percent volumetric ultrasonic inspection from the flat surface per ASME BPVC Section III.
This inspection will be performed on the flywheel after final machining and the overspeed test.
- 5. If the flywheel is flame cut, at least 1/2 inch of '
stock will be left on the outer and bore radii, for machining to final dimensions. 4
- 6. 'Ihe flywheel will be subjected to a magnetic particle or liquid-penetrant examination per "Section III" before final assembly. The inspection will be !
performed on finished machined bores, keyways, and on both flat surfaces to a radial distance of 8 inches minimum beyond the final largest machined bore diameter but not including small drilled holes. There will be i no stress concentrations such as stamp marks, center punch marks, or drilled or tapped holes within 8 inches of the edge of the largest flywheel bore. t B. The flywheels will be designed to withstand normal operating conditions, anticipated transients, and the largest mechanistic pipe break size remaining after application of D leak before break as described in Section 3.6, combined with i the Safe Shutdown Earthquake. The following criteria will be satisfied: j
- 1. The combined stress, both centrifugal and interference, !
at normal operating speed will not exceed one-third of the minimum specified yield strength for the material selected in the direction of maximum stress. Amendment D 5.4-2 September 30, 1988 i
f C E S S A R !annca m i
- 2. The design speed of the flywheel will be 125 percent of normal operating speed.
The lowest of the critical speeds of the flywheel will ~ be at least 10% above the highest anticipated overspeed-of the pump. .The highest: anticipated . overspeed is predicted for the largest break size remaining .after application. of leak before break as described in D Section 3.6. 3.' The combined centrifugal and interference stresses at the design speed will be . limited to two-thirds of the minimum specified yield strength. Design' speed' is defined as 125 percent.of normal operating' speed.
- 4. The motor and pump shaft or bearings and coupling will withstand any combination of normal operating loads or.
anticipated transients, and the largest remaining pipe break after application of leak before. break as D described in Section 3.6, combined with the Safe Earthquake Shutdown. Each flywheel will be tested'at design speed, 125 percent of normal operating speed, as defined in B.2 above. The flywheel will be accessible for 100 percent in-place-volumetric ultrasonic inspection. The flywheel-motor assembly is designed to allow such inspection with a minimum j of motor disassemb f. 1 5.4.1.2 Description i Table 5.4.1-1 lists the principal parameters of the reactor ) coolant pumps and Figure 5.4.1-l ' depicts the arrangement of the - pump and motor. Reactor coolant pump supports are discussed in Section 5.4.14. The pump piping and instrument diagram is given in Figure 5.1,2-2. The four reactor coolant pumps are vertical, single' stage bottom suction, horizontal discharge, motor-driven centrifugal pumps. i The pump impeller is keyed and locked'to.its shaft. Pump shaft alignment is maintained by a water lubricated radial bearing within the pump and by radial and thrust bearings located in the ! motor stand. The pump and motor shafts are directly. connected by ! a coupling. L The shaft eeal assembly consists of two face-type, mechanical f seals in eeries, with controlled leakage bypass-to. provide the i same pressure differential
- across each. seal. The seal assembly' l is designed for 2500 psi differential and to reduce the leakage pressure from Reactor Coolant System pressure - to the volume Amendment D 5.4-3 September 30, 1988
CESSAR MMICATION O control tank pressure. A third, face-type, low-pressure vapor seal at the top is designed to withstand system operating pressure when the pumps are not operating. The leakage past the second pressure seal and the controlled leakage is piped to the volume control tank in the Chemical and Volume Control System. Leakage past the low pressure vapor seal is collected and piped to the reactor drain tank. The temperature of the water in the seal assembly is maintained within acceptable limits by a water-cooled heat exchanger. Water is also injected into the seal area from an external seal injection system. The performance of the shaft seal system is monitored by pressure and temperature sensing devices in the seal system. The seal assembly can be replaced without draining the pump casing or removing the shaft. The seal assemblies are designed to limit seal leakage plus controlled bypass flow to approximately the values given below: Seal leakage plus controlled bypass flow, per pump: All seals functioning (normal) 3.9 gpm one seal functioning (abnormal) 5.4 gpm The motor is sized for continuous operation at the flows resulting from four-pump or one-pump operation with 1.0 to 0.74 specific gravity water. The motors are designed to start and accelerate to speed under full load with a drop to 80 percent of normal rated voltage at the motor terminals. Each motor is provided with an anti-reverse rotation device. The , device is designed to prevent impeller rotation in the reverse j direction due to each of the following conditions: motor' I starting torque, if the motor was incorrectly wired for reverse rotation; and, reactor coolant flow through the pump in the reverse direction due to the largest remaining pipe break after application of leak before break as described in Section 3.6, D which could result in reverse flow through the pump. 5.4.1.3 Evaluation The reactor coolant pumps are sized to deliver flow that equals or exceeds the design flow rate utilized in the thermal hydraulic analysis of the Reactor Coolant System. Analysis of steady-state and anticipated transients is performed assuming the minimum design flow rate. Tests are performed to evaluate reactor coolant pump performance during the post-core load hot functional testing to verify adequate flow. Leakage from the pump via the pump shaft is controlled by the shaft seal assembly. Reactor coolant entering the seal chambers Amendment D i 5.4-4 September 30, 1988
CESSAR H%nCATION 77 (v) 19 cooled and collected in closed systems to prevent reactor coolant leakage to containment. Instrumentation is provided to monitor seal cperation. The design speed of the flywheel is 125 percent of normal speed. l An overspeed test of each flywheel at the design speed is performed prior to assembly. Refer to pump flywheel integrity Section 5.4.1.1. In the event of a break which is not eliminated by leak before break which would result in reverse flow thrcugh the pump, the D anti-reverse rotation device prevents impeller rotation in the i reverse direction. In the event of a break which is not eliminated by laak before break and which could result in D increased flow through the pump tending to accelerate the pump impeller,the highest predicted pipe break induced overspeed is lD less than the lowest critical speed of the flywheel. l The pump and motor oil lubricated bearings are lubricated by j internal oil systems. Each bearing assembly has its own internal { oil system consisting of either an oil bath or force-feed type j system. During normal operation, no external pumps will be required because pumping action is accomplished by internal f_} pumping devices. Lubricating oil is cooled by cooling coils () submerged in the oil sumps. Both sumps and cooling coils are internal to the motor otructural frame and are designed for , Seismic Category I operation, and use the intent of the ASME j Boiler and Pressure Vessel Code, Section III, Class 3 as a guide for design and construction. This is established within the ! Combustion Engineering Topical Report, CENPD-201-A, which demonstrates the reactor coolant pump performance during a loss of component cooling water incident. Although the pump-motor , assembly operation is not considered necessary for plant safety, J this design minimizes the direct effects of seismic events on the l reactor coolant pump and motor assembly oil lubricating systems so that adequate coast down characteristics are not detrimentally affected. Bearing metal temperatures, oil flow and/or pressure, oil levels, l cooling water flow and/or pressure are continuously monitored and i alarmed in the control room. . In the unlikely event that component cooling water to the reactor coolant pump and motor oil lubricating systems is not available or that an oil leak occurs during operation, the operator is alerted as soon as cooling water to the oil system is lost and has a time period of at least 30 minutes in which to reduce power, if necessary, and isolate cooling water and shut the reactor coolant pump motor assembly down to prevent bearing (A} v seizure. Amendment D 5.4-5 September 30, 1988
CESSAR 8! Enc-0 1 This time limit is established in the Topical Report CENPD-201-A. 1 I Combustion Engineering will perform a test to verify the analysis in Appendix A. The component cooling water will be secured to the pump lubrication oil sumps and data taken so as to demonstrate the heat up rate of the oil sump up to a maximum sump temperature of 200*F. A test securing cooling water to the pump seals will also be performed to demonstrate the seals operability as detailed in Combustion Engineering Topical Report CENPD-201-A. In the remote possibility of a simultaneous loss of component cooling water to all reactor coolant pump motor assemblies, 30 minutes is adequate to secure the plant and maintain the normal coast down capabilities of the reactor coolant pump motor assemblies. During a loss of component cooling water event, it is unlikely that a shaft seizure due to bearing failure will occur for the following reasons: A. The design is such that the heat generated in the bearing normally carried away by the cooling water, is transferred by alternate paths. The lube oil sump baths surrounding the bearings, the stagnant cooling water remaining in the heat exchanger coils, and the bearing and sump assembly metal masses, all act as heat sinks. Also conduction down the pump shaft and outer sump shell radiation help to reduce the temperature rise. B. The rotation of the bearing assemblies insures adequate oil flow and mixing of heated oil to insure the heat transfer as described in A. C. In the event that the oil temperature rises such that the viscosity degrades significantly, the design of the thrust bearing continues to produce a hydrodynamic film so as to preclude metal to metal contact. D. Operation and test experience has demonstrated that the reactor coolant pump motor assembly will operate without cooling water to the lubricating oil system for at least the calculated 30 minute time period. Should an oil leak occur, redundant instrumentation will alert the operator to shut down the reactor coolant pump motor assembly and thereby avoid bearing damage. In the event of an oil leak, the separation of lubrication systems would limit the problem to a single reactor coolant pump. j 1 0 5.4-6 4
CESSAR Ennricarieu n V The loss of the oil in the bearing oil reservoir would not result in bearing seizure for the following reasons: A. Temperature and oil level monitors will provide appropriate indication of an abnormal condition. B. The vibration monitoring device furnished on the pump will respond to bearing degradation and allow the operator to shutdown the pump. , I C. If the above protective measures fail, the high torque 1 produced by the motor will cause a slow breakdown of the bearings and not a rapid shaft seizure. Industry experience indicates that the babbitt bearing surfaces . wear away and the bearing pads and sleeves will be badly worn but the , shaft will continue to rotate. 1 1 If the extremely remote possibility of bearing seizure occurs l while the reactor coolant pump motor assembly is in operation, adequate flow to the core is available from the other reactor coolant pump motor assemblies as demonstrated by the one pump loss of flow study. ['N Figure 5.1.2-2 shows a separate oil lift system which is required (
'v) for start-up of the pump assembly. The oil lift system furnishes high pressure oil to the pump assembly thrust bearings, thereby lifting the rotor and reducing bearing friction during pump start-up. Interlocking devices are furnished which prevent pump start-up until oil lift flow is established. The oli lift system is automatically shutdown when the pump reaches full speed.
Since oil lift is not required during normal operation, an oil ; leak in this system will not cause a bearing failure. 5.4.1.4 Tests and Inspections The reactor coolant pump pressure boundary is nondestructively inspected as required by ASME Section III for Class 1 components. The pump casing inspections include complete radiography and l liquid penetrant or ultrasonic testing. The pump receives a hydrostatic pressure test in the vendor's shop and with the Reactor Coolant System. Inservice inspection of the pump pressure boundary will be performed during plant life in accordance with ASME Section XI. The pump assembly is performance tested in the vendor's shop over at least the normal operating range in accordance with the Standards of the Hydraulic Institute. The tests also demonstrate ability of the pumps to function under the various operating conditions specified. Tests commonly performed are hot and cold 5.4-7
i CESSAR8HFinc-Oj i performance and stop-start cycling. Special testing will also be j performed on one pump. Such testing will include losc of cooling l and/or neal injection water. Vibrations are monitored at several places on the pump during shop testing. In addition to meeting an absolute criterion for vibration amplitude, the test results are examined for evidence of critical speed problems. The pump motors undergo a " routine" test in accordance with NEMA MG-1. This test also confirms that the motors are within their vibration limits. At least one motor is tested by being used as the driver for the pump assemblies, during the pump manufacturer's shop testing. The following testing may also be performed where significant seal experience is lacking to develop confidence in the sealing system; o Seal materials testing for suitability in reactor coolant environment. o Long term testing of an entire seal assembly. To the greatest extent practicable, all conditions of operation 9,f within the reactor coolant pump will be duplicated. Reactor coolant pump flywheel inspections and testing are described in Section 5.4.1.1. I B 1 O Amendment B 5.4-8 March 31, 1988 l l k
CESSAR !!nificui:,i O 1 TABLE 5.4.1-1 REACTOR COOLANT PUMP PARAMETERS l 1 Number of Units 4 Type Vertical, single stage centrifugal I Design Total Dynamic Head, ft* 365 Design Flow, gpm 111,400 i Design Pressure, psia 2500 Design Temperature, F 650 ; Normal Operating Pressure, psia 2250 Normal Operating Temperature, F* 558 lE NPSH Required (at design flow), ft* 220 Suction Temperature, F* 557 E 3 Water Volume, each, ft 134 Weight (including motor), dry, lbs. 279,000 O Shaft Seals Mechanical Face Seals Pump Speed, rpm
- 1190 Motor Synchronous Speed, rpm 1200 Motor Type AC Induction Horsepower, hot
- 9000 cold 12,000 Rated Brake Horsepower 12,000 Voltage 13,200 1
Phase 3 Frequency 60 Hz Insulation Class F . 1 Starting Current, at 100% Voltage, amps 3,000 j i I
- Parameters are related to four-pump, full power operating conditions.
O 1 Amendment E December 30, 1988
O a r a COOLER y OTOR M
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Amendment B
!! arch 31,1988 jygyg[/ h REACTOR COOLANT PUMP 5.4.1-1 1
CESSARin&%. , V(D 5.4.2 STEAM GENERATORS 5.4.2.1 Desian Bases The two steam generators are designed to transfer 3817 MWt from thg RCS to the secondary system, producing approximately 17.12 x 10 lb/hr of 1000. psia saturated steam, when provided with 450*F lB feedwater. Moisture separators and steam driers in the shell side of the steam generator limit the moisture content of the steam to 0.25 wt% during normal operation at full power. The steam generator design parameters are listed in Table 5.4.2-1. l The steam generators, including the tubes, are designed for the l RCS transients listed in Section 3.9.1 so that the code allowable i stress limits are not exceeded for the specified number of j cycles. All transients have been established based on conservative assumptions of operating conditions in consideration ; of supportive system design capabilities. The steam generators ! will be capable of sustaining the following additional design transients without exceeding code allowable stress limits: A. Fif teen secondary side hydrostatic tests with secondary side lB pressurized to 1-1/4 times the design pressure and the J g primary side pressurized so that the tube differential ) ! t < pressure does not exceed 820 psid (test condition) ; l V - l B. Three hundred secondary side leak tests with the secondary lB ) side pressurized from 820 psia to design pressure, with the primary side pressurized so that tube differential pressure (secondary to primary) does not exceed 820 psid (test condition); C. Less than ten thousand cycles of adding 40*F feedwater at 800 gpm to the steam generator through the downcomerl0 feedwater nozzle when at hot standby conditions (normal condition); D. Seven hundred and fifty cycles of adding 40*F feedwater at 800 gpm to the steam generator through the .i downcomer lD feedwater nozzle during loading conditions (normal l condition); ) E. Seven hundred and fifty cycles of adding 100*F feedwater at 800 gpm to the steam generator through the. downcomer lD feedwater nozzle during loading conditions (normal condition); F. Seven cycles of adding 40*F feedwater at 800 gpm to thelD steam generator through the downcomer feedwater nozzles B Amendment D 5.4-9 September 30, 1988
CESSAREHWnc-O\ during a steam line break. This provides for one steam line break incident with the emergency feedwater cycled a maximum of seven times (faulted condition); G. Four hundred and twenty cycles of adding 40*F feedwater at lB 850 gpm to the steam generator through the downcomer lE feedwater nozzles with the flow initiated 30 seconds after a loss of normal feedwater. This provides for 60 loss of normal feedwater incidento with the emergency feedwater lB cycled a maximum of seven times (upset condition) ; H. Six thousand pressure transients of 85 psi across the lB primary divider plate in either direction caused by starting and stopping reactor coolant pumps (normal condition). The steam generator was designed to ensure that critical vibration frequencies are well out of the range expected during normal operation and during abnormal conditions. The tubing and tubing supports are designed and fabricated with considerations given to both secondary side flow induced vibration and reactor coolant pump induced vibrations. In addition, the steam generator assemblies are designed to withstand the blowdown forces resulting from the severance of a steam nozzle. The steam generator assemblies are also designed to withstand the severance of any one of the feedwater nozzles. The two accidents are not considered simultaneously. The steam generator tubes are Ni-Cr-Fe alloy 690, 3/4-inch OD, with 0.042-inch nominal wall thickness. A steam generator tube rupture incident is a penetration of the barrier between the reactor coolant system and the main steam system. The integrity of this barrier is significant from the standpoint of radiological safety in that a leaking steam I generator tube allows the transfer of reactor coolant into the main steam system. Radioactivity contained in the reactor coolant would mix with water in the shell side of the affected steam generator. This radioactivity would be transported by steam to the turbine and then to the condenser, or directly to i the condenser via the Turbine Bypass System. Noncondencible I radioactive gases in the condenser are removed by the main condenser's evacuation system and discharged to the plant ventilation system. Experience with nuclear steam generators indicates that the probability of complete severance of a tube is remote. A double-ended rupture has never occurred in a steam generator of l this design. The more probable modes of failure, which result in smaller penetrations, are those involving the occurrence of Amendment E 5.4-10 December 30, 1988
CESSAReMecmu O pinholes or small cracks in the tubes, and of cracks in the seal welds between the tubes and tube sheet. Detection and control of steam generator tube leakage is described in Section 5.2. The concentration of radioactivity in the secondary side of the steam generators is dependent upon the concentration of radionuclides in the reactor coolant, the' primary-to-secondary leak rate, and the rate of steam generator blowdown. The expected specific activities in the secondary side of the steam generators during periods of normal operation are given in Section 11.2. The recirculation water within the steam generators will contain volatile additives.necessary for proper chemistry control. These and other chemistry considerations of the main steam system are discussed in Section 10.3.5. 5.4.2.2 Description The steam generator is illustrated in Figure 5.4.2-1. Moisture-separating equipment in the shell side of the steam generators limits moisture content of the exit steam. Manways 3 and handholes are provided for access to the steam generator internals. Reactor coolant enters at the bottom of each steam
) generator through the single inlet nozzle, flows through the U-tubes, and leaves through the two outlet nozzles. A vertical divider plate separates the inlet and outlet plenums in the lower head.
The steam generator with integral economizer is in most respects ! similar to earlier U-tube recirculating steam generators. The basic difference is that instead of introducing feedwater .only through a sparger ring to mix with the recirculating water flow in the downcomer channel, feedwater is also introduced ' into a separate, but integral section of the steam generator. A semi-cylindrical section of the tube bundle, at the cold leg or exit end of the U-tubes, is separated from the remainder of the tube bundle by vertical divider plates. Feedwater is introduced directly into this section and pre-heated before discharge into the evaporator section. l The lower portion of the evaporator section and the downcomer l channel occupy only one-half of the steam generator l cross-section. The effect of this non-symmetry is considered in calculation of recirculation ratio, internal flow considerations, and in design of tube support structures. 1 5.4-11
CESSAR n%ncarcu The steam-water mixture leaving the vertical U-tube heat transfer O surface enters the separators which impart a centrifugal motion to the mixture and separate the water particles from the steam. The water exits from the perforated separator housing and recirculates through the downcomer channel to repeat the cycle. Final drying of the steam is accomplished by passage of the steam through corrugated plate dryers. A recirculation system allows the circulation of water through the steem generator during wet lay-up and the addition of chemical cleaning agents. The recirculation system consists of a distribution ring locat6d above the tube bundic below the normal water level with connecting piping to the blowdown system. This piping requires a nozzle penetration through the pressure D boundary (shell). Suction is taken at the blowdown nozzle and recirculated through the distribution ring. The recirculating header can effect a rapid changeover of the steam generator inventory should a chemical intrusion occur requiring rapid removal of impurities. Primary head draining capability is provided by channel head drains. Tubesheet drains allows secondary side draining. Such drain capability enhance access for inspection and maintenance. The pressure drop from the steam generator feedwater nozzles to the steam outlet nozzle including the economizer is approximately 40 psi. The steam generator supports are described in Section 5.4.14. 5.4.2.3 Economizer Intecrity The economizer section is designed in full consideration of operating transients, startup and standby operation, and accident conditions such as loss of feedwater flow and feedwater line 1 break. The structural design of the various parts is adequate to withstand the thermal and pressure loadings from these various conditions, consistent with the appropriate load classifications and design rules in the ASME Code, Section III, see Appendix G. The components of the steam generator economizer section have , been designed for the primary stresses which occur due to the l blowdown associated with a feedline break. The divider plates, ' which separate the economizer region from the evaporator region of the secondary side, are supported from the vessel shell and the central cylindrical support welded to the tubesheet. This divider cylinder becomes an extension of the primary tubesheet stay cylinder, though less massive, and extends the full height of the economizer. The tube support / flow baffle plates are supported from the vessel shell, the divider cylinder and the Amendment D 5.4-12 September 30, 1988
CESSAR ME"icarian _ l I tubesheet via an array of support rods. The support rods, which l also serve as support plate spacers are solid and designed for either tensile or buckling loads. An effort has been made to avoid the use of thin plates which may collapse when subjected to differential pressure. 5.4.2.4 Steam Generator Materials I l The pressure boundary materials used in the construction of the l l steam generator are listed in Table 5.2-2. These materials are I in accordance with the ASME Boiler and Pressure Vessel Code, I Section III. Code cases used in the fabrication of the steam generator are discussed in Section 5.2.1. The Class 1 components of the steam generator will meet the fracture toughness requirements of the ASME code. An additional discussion of fracture toughness testing is included in Section 5.2.3. Discussion of the techniques used to maintain cleanliness during final assembly and shipment are discussed in Section 5.2.3. Onsite cleaning and cleanliness control for the steam generator p is discussed in the site-specific SAR. lD l 5.4.2.4.1 Steam Generator Tubes l The method of fastening tubes to the tube sheet conforms with the requirements of Section III and IX of the ASME Code. Tube expansion into the tube sheet is total with no voids or crevices occurring along the length of the tube in the tube sheet. Localized corrosion of tubing material has led to steam generator tube leakage in some operating reactor plants. Examination of , tube defects that have resulted in leakage has shown that two l mechanisms are primarily responsible. These localized corrosion i mechanisms are referred to as (1) stress assisted caustic cracking, and (2) wastage or beavering. Both of these types of l corrosion have been related to steam generators that have , operated on phosphate chemistry. The caustic stress corrosion i type of failure is precluded by controlling bulk water chemistry to the specification limits shown in Section 10.3.4. Removal of solids from the secondary side of the steam generator is discussed in Section 10.4.8. Localized wastage or beavering has been eliminated by removing phosphates from the chemistry control program. Volatile chemistry (discussed in Section 10.3.4) has been O successfully used in all C-E steam generators that have gone into operation since 1972. Amendment D 5.4-13 September 30, 1988
CESSAR !!Einem:u O 5.4.2.5 Tests and Inspections Prior to, during and after fabrication of the steam generator, nondestructive tests based upon Section III of the ASME Code are performed. Initial hydrostatic tests of the primary and secondary sides of the steam generator were conducted in accordance with ASME Code, Section III. Leak tests were also performed. Following satisfactory performance of the hydrostatic tests, magnetic-particle inspections are made on all accessible welds. Inservice ir.spection of the steam generator is described in Section 5.2.4. O t' \ l r < .* .
, l l
1 k O 1 Amendment D 5.4-13a September 30, 1988
C E S S A R ENMnc m ,. O TABLE 5.4.2-1 STEAM GENERATOR PARAMETERS (a) Parameter Value Number of units 2 Heat transfer rate g SG, Btu /hr g 6.517 x 10 9 s w/0.2% Blowdown (a) w/1.0% Blowdown 6.530 x 10 Primary Side Design pressure / temperature (psig/*F) 2485/650 i Coolant inlet temperature, 615 3 Coolant outlet temperature, 'F 557 Coolant flow rate, each, lb/hr 82.9 x 10 6 lg Coolant volume at 68F each, ft 3 2800 j B Tube size, OD, in. 0.75 Tube thickness, nominal, in. 0.042 Secondary Side Design pressure / temperature (psia /*F) 1200/570 B Steam pressure, psia 1000 Steam flowrate (at 0.25% moisture) per SG, lb/hr 8.56 x 10 6
- Feedwater temperature at full power, 'F 450 Moisture carryover, weight maximum, % 0.25 Primary inlet nozzle, No./ID, in. 1/42 Primary outlet nozzle, No./10, in. 2/30 Steam nozzle, No./ID, in. 2/28 l
l Feedwater nozzles, No./ size / schedule (Economizer) 2/14/80 feedwater nozzles, No./ size / schedule (Downcomer) 1/6/80 l Overall heat transfer coe{ficient Evaporator, Btu /hr-ft 'F 1371 y Economirer, Btu /hr-ft 2 *F 593-(r.) Blowdown w/12% quality. Amendment E December 30, 1988 l
m' i NO. i NO. SERVICC R E Q'D 1 PRIMARY INLET 1 0 2 3 4 5 PRIMARY OUTLET STEAM OUTLET DOWNCOMER BLOWDOWN FEEDWATEF. l 2 2 1 f 2 I I 6 PRIMARY MANWAY 2 ) 7 SECONDARY MANWAY 2 I B HANDHOLE 2 400 l 9 ECONOMlZEh FEEDWATER 2 l 10 RECIRCULATION NOZZLE 1 \ / STEAM OUTLET . NOZZLES (2)
@% p PRESSURE TEST a
'\ NOZZLE PEERLEOS DRYERS 194 STEAM 1
IPI s( '! i p'tl SEPARATORS WITHOUT i ! U '
- k- -
20'-5" O.D. U RECIRCULATION l 18" MANWAYS (2)" / NOZZLE l - 7 nn a b\ DOWNCOMER FEEDWATER I > I~ NOZZLE 180 RING t p . . c, Jr t HEADER 3 , . 1
-(
75' PRIMARY i - l 18" MANWAYS O 5 L) , -@ > ' i
/0 f ,; 4 50 f.p- g INCONE L 690 TT
( .._ } \ s 7e < TUBING 0 ,\/ ., 4 o, ., , J' I ll s\
>- d = 16'6"
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- 450 '; I O.D.
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g'u ' , c , o (3) HALF EGGCRATES : a { INTEGRAL AXIAL FLOW ECONOMlZER (8") HANDHOLES . 1
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l ' HIGH CAPACITY BLOWDOWN F.CONDARY 5 TUBESHEET DRAIN PERMANENTLY MARKED
/N TUBESHEET TONGUE AND GROOVE
, 2 DIVIDER PLATE
/ \\ \ V Amendment D-PRIMARY SIDE DRAIN \ NOZZLES PREPARED FOR g
, September 30, 1988 NOZZLE DAM INSTALLATION ,
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m Figure STEAM GENERATOR 5.4.2.-1 l l
7 O THIS PAGE INIDfTIONA11Y BIRK O ! O
2 CESSAR 25nCATCN O' 5.4.3 REACTOR COOLANT PIPING 5.4.3.1 Desian Basis Applicable design codes are found in Table 5.2-1. The reactor : coolant loop piping is designed and analyzed for all transients specified in Section 3.9.1. In addition, those nozzles subjected to local thermal transients, caused by fluid entering the Reactor Coolant System from an auxiliary system, are analyzed to ensure that the nozzles can accommodate the additional transients. In addition to being specified as Seismic category I, the piping is designed to ensure that critical vibration frequencies are well out of the range expected during normal operation and during abnormal conditions. Additional presentations relating to seismic and dynamic analysis and criteria for the reactor coolant , piping is contained in Sections 3.7.2 and 3.9.2, respectively. l 5.4.3.2 p_e_gg. fiction Each of the two heat transfer loops contains five sections of ' pipe: one 42-in. internal diameter pipe between the reactor vessel outlet nozzle and steam generator inlet nozzle, two 20-in. internal diameter pipes from the steam generator's two outlet nozzles to the reactor coolant pumps suction nozzle, and two 4 30-in. internal diameter pipes from the pumps discharge nozzle to i the reactor vessel inlet nozzles. These pipes are referred to as the hot leg, the sur tion legs, and the cold legs, respectively. The other major pieces of reactor coolant piping are the surge line, a 12-in. pipe between the pressurizer and the hot leg, and i the spray line, a 4-inch pipe at the pressurizer end reduced to a i 3-inch pipe and connected to two (2) cold legs. The 42-in. and 30-in. pipe diameter are selected to obtain coolant velocities which provide a reasonable balance between erosion-corrosion, pressure drop, and system volume. The surge line is sized to limit the fr4 tional pressure loss through it during the maximum in-surge s; that the pressure differential between the pressurizer and tb heat transfer loops is no more than 5 percent of the system .4 sign pressure. The spray line sizing is discussed in Section 5.4.10. , To reduce the amount of field welding during plant fabrication, the 42-in, and 30-in. pipes are supplied in maj Ur pieces, complete with shop-installed instrumentation nozzles and connecting nozzles to the auxiliary systems. Where required, the nozzles are supplied with safe ends to facilitate field welding ) of the connecting piping. O, 5.4-14 i
l i l CESSARE!ELma I l /O k )
%/
Flow restricting orifices (7/32" dia. x 1" long) are provided in ) the nozzles for the RCS instrumentation and sampling lines to l I limit flow in the event of a break downstream of a nozzle. 1 5.4.3.3 Etterials j l The materials used in the fabrication of the piping are listed in l Table 5.2-2. These materials are in accordance with the ASME Code, Section III. The proviisions taken to control those factors that contribute to stress corrosion cracking are discussed in 4 Section 5.2.3. 1 1 Fracture toughness of the reactor coolant piping is discussed in i Section 5.2.3.- j l 5.4.3.4 Tests and Inspections { l Prior to, during and after fabrication of the reactor coolant 1 piping, nondestructive tests based upon Section III of the ASME Code were performed. In addition, the fully assembled reactor coolant system is hydrostatically tested in accordance with the Code. Inservice inspection of the reactor coolant system piping is discussed in Section 5.2.4. 5.4.4 MAIN STEAM LINE RESTRICTIONS I I The steam generator outlet nozzles are one piece forgings with an integral venturi type flow restrictor. The venturi section of the nozzle is designed to reduce the flow area by 70%. 5.4.5 MAIN STEAM LINE ISOLATION SYSTEM l The Main Steam Line Isolation System is composed of portions of the Main Steam System and the Engineered Safety Features Actuation System. Discussed here are those portions of these l systems that respond to a Main Steam Isolation Signal, as defined I' in Section 7.3. A discussion of radiological considerations is provided in Section 12.3. 5.4.5.1 Desian Bases 1 A. The Main Steam Line Isolation Valves are designed to isolate the steam generators and the main stearn lines in the event of a main steam line rupture. ( V 5.4-15 u____________-.____
4
)
CESSAR HL"icmu ! I I B. The Main Steam Line Isolation Valves are designed to perform )i containment isolation functions for the main steam lines in ' the event of a design basis accident, as discussed in Section 6.2.4. In the event of a steam line break outside the containment, the isolation function serves to reduce the potential leakage of radioactivity to the environment. C. The Main Steam Isolation Valves are designed to isolate the i main steam lines and the steam generators as required for maintenance. C-E interface requirements for the Main Steam Isolation Valves are listed in Section 5.1.4. 5.4.5.2 System Desian 5.4.5.2.1 General Description Each of the four main steam lines is provided with a power-actuated Main Steam Isolation Valve designed to stop flow from either direction when it is tripped closed. Each valve is located outside containment and is provided with means of actuation from the Engineered Safety Features Actuation System, meeting the requirements of IEEE Standard 279. The logic circuitry required to isolate the main steam lines is discussed in Section 7.3. The main steam system valves and arrangement are discussed in Section 10.3.2. lD 5.4.5.2.2 Component Description i The main steam isolation system consists of the Main Steam Isolation Valves and their associated controls and instrumentation. The Main Steam Isolation Valves are remotely operated valves designed to either fail closed or be guaranteed to close upon receipt of a Main Steam Isolation Signal. The Main Steam Isolation Valves can be monitored and controlled locally and in the control room. 5.4.5.2.3 System Operation The Main Steam Isolation Valves are designed to isolate the main steam lines and the steam generators as required during operation and under accident conditions. A steam line break inside containment would result in a pressure ! rise in the containment. Reverse flow protection is also achieved through the Main Steam Isolation Valves. To achieve : reverse flow protection in the case of the main steam pipe ! rupture, the valve is fully closed within 5.0 seconds from receipt of the initiating signal. Amendment D 5.4-16 September 30, 1988 l l
CESSARn h mw o The main steam line isolation system components are qualified to serve in the environment specified in Section 3.11. 5.4.5.3 Desian Evaluation Design evaluations are listed to corresoond with the design bases listing. A. The Main Steam Isolation Valves are capable of isolating the steam generators within 5.0 seconds after receiving a signal lB from the Engineered Safety Features Actuation System. In the event of a steam line break, this action prevents continuous uncontrolled steam release from more than . one steam generator. Protection is offered for breaks inside or outside the containment. B. The Main Steam Isolation Valves, their operatori, and associated circuitry are Seismic Category I, and are protected against missiles and the effect of high-energy line breaks. 5.4.5.4 Tests and Inspections ( ('- All main steam isolation valves are designed, fabricated, tested, and installed in accordance with the codes and standards identified in the interface requirements described in Section 5.1.4. Assurance of operability is discussed in Section 3.9.3 of the site-specific SAR. lD I 5.4.6 REACTOR CORE ISOLATION COOLING SYSTEM This system is not applicable to a Pressurized Water Reactor, i O Amendment. D 5.4-17 September-30, 1988 _ _ _ _ _ _ _ _ l
l l l l CESSARH?Re- - 1 5.4.7 SHUTDOWN COOLING SYSTEM lC , 5.4.7.1 Desian Bases ! l 5.4.7.1.1 Summary Description The Shutdown Cooling System (SCS) is used in conjunction with the Main Steam and Main or Emergcacy 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 temperature to the refueling temperature and to lC maintain this temperature during refueling. lC The SCS is used in addition to the SG atmospheric steam celease capability and the Emcrgency Feedwater System to cooldown d e RCS following a small break LOCA (see Section 6.3). The SCS would also be used subsequent to steam and feedline breaks, steam generator tube ruptures, and is used prior to RCP start to maintain flow through the core during plant startup. 5.4.7.1.2 Functional 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. ! J B. The design bases defined in Section 5.4.7.1.1 are met assuming the failure of a single active component during shutdown cooling or a single active or limited leakage , passive failure of a component during long-term operations C ' (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). C. The SCS is designed such that the SCS pumps and containment spray pumps are functionally interchangeable. D. The SCS shall be designed for a nominal pressure of 900 psig and a temperature of 400*F. Amendment C 5.4-18 June 30, 1988
~
CESSAREP4%ma I t V E. No single failure allows the SCS to be overpressurized by the RCS. SCS components whose design pressure is less than the RCS design pressure are provided with overpressure protection (see section 5.4.7.2.3). C 4 i F. The SCS reduces the Reactor Coolant System temperature as I' follows-l
- 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 shutdown 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 and 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 exchangere 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-flow 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 MWicaritu i I criteria considered to be relevant, and are not imposed as ! interface requirements unless specifically called out as such in a particular interface requirement. 1 Relevant GDC - 1, 2, 3, 4, 10, 34, 35, 36, 37, 38, 39, 40, l l 50, 54, 56, 57 1 Relevant Reg. I Guides - 1.1, 1.4, '
. 2 6, 1.28, 1.29, 1.31, 1.34, j 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 l 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, 673, 670, 658 on Figure 6.3. 2-1B) shall receive electt3 cal power such that no fault to a single power supply could open the valves to l 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. I
- d. Two independent instrument power supplies shall be I 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 4
Amendment C 5.4-20 June 30, 1988 l 1
)
CESSAR1!'41 nema . O t and seiches. for applicable sites) or- the effects thereof per the requirements of' Criterion 2 of 10CFR50 i J will not' prevent then from . performing -their safety functions. .f d
- 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 such that they will withstand the.
effects 'of earthquakes per the requirements . ' of ' Criterion 2 of 10CFR50 without loss of the capability ,
- to perform their safety functions.
- 4. Failure'of.non-seismic systems and structures.shall not I 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 be protected from the effects c" postulated 'high and i moderate energy pipe ruptures.
- 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 NPSH) in.a
[ connecting system- (e.g., Safety Injection System or. l- Containtaent Spray . System) shall preclude the' l availability of minimum acceptable shutdown- cooling l- capability.- Minimum acceptable. shutdown capability is defined as that provided by one SCS ' cooling pump l C and its associated heat exchanger train. l
- 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: Os . Amenciment C 5.4-21 , l June 30, 1988
CESSARsnec- i i J
- a. Protection of the SCS from the cr anIuences of a pestulated 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.
i D. Missiles
- 1. For the portion of the SCS located inside containment, appropriate design procedures shall be used to insure lC 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 I equipment from carrying out its specified safety functions. ) i I
- 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 credible dynamic effects originating within the containment under the conditions of accidents the SCS is required to mitigate. Separation via physical plant layout, pipe restraints, isclation capability, or other 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
CESSAR EMi?iCATl!N l p . \. 0 . 1
- 2. Containment isolation valves, operators, and associated !
power and control systems located outside the l containment that are part of the SCS shall be protected i from dynamic effects and loss of function resulting { from equipment failures and pipe ruptures originating ' in adjacent areas. Protection from such failure and rupture e f f ec'.s shall be by separation, enclosure, restraint, viter-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 laakage passive failure. ]
F. Independence
- 1. Electrical - See 5.4.7.1.3 (A.2.b.)
- 2. Environmental - See 5.4.7.1.3 (Q) l
- 3. Mechanical - See 5.4.7.1.3 (C, D, E) 1 l
G. Thermal Limitations Q Q 1. Component Cooling Water - See 5.4.7.1.3 (P.2)
\C l
- 2. Environmental - See 5.4.7.1.3 (Q)
H. Monitoring
- 1. The safety related instrumentation of the SCS is identified in Table 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, Sectic n III components shall be arranged to provide adequate clearances to permit inservice inspection.
- 2. Manually operated valves which contain reactor coolant p or other potentially radioactive liquids during normal ig plant operations shall be provided with handwheel extensions and shielding,.to allow periodic actuation.
Amendment C 5.4-23 June 30, 1988
CESSAR EnglCATl N I
- 3. SCS components which contain reactor coolant or other potentially radioactive liquids during normal plant operations, and whicn require access for periodic pressure tests and nondestructive examination, shall be j capable of being flushed prior to testing.
C
- 4. System and component arrangement shall allow adequate clearances for performance of inspections identified in Cliapter 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. j
?
- 5. 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 l 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 shs11 be compatible with the radiation dose as well as the chemistry of the system fluid. Amendment C 5.4-24 June 30, 1988
CESSAR nnince,, I lO
- 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 10 CFR 50.
r
- 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 steel to the SCS, and (2) the field welds of the'SCS.
- 4. Controls shall be exercised during plant construction to assure 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 as 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 p effect of boron dilution during shutdown cooling initiation. C
- 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 shutdown cooling is 20 feet.
lE
- b. The elevation of these pumps when used for j containment spray functions, shall be low enough C
)
such that adequate NPSH is available when all I 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 flow measurement device orifices are provided of sufficient length to comply with: ASME Fluid Meters; Their Theory and Application, Parts 1 & 2.
- 4. The SCS suction lines shall be arranged such that no portion is physically above the lowest point of the RCS hot leg piping.
I Amendment E l 5.4-25 December 30, 1988
i CESSAR HSL"lCATION
- 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 compensata. The C elevation of SCS relief valves (SD-769, SD-769) 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.
- 7. In the event of a limited leakage passive failure in Ic 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) C shall be designed to accept relief valve discharge from the shutdown cooling suction line overpressure relief valves. l
- 0. 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 rainimum, the following features:
- a. Facilities for fire detection and alarming.
- b. Facilities or methods to minimize the probability of fire and its associated effects.
- c. Facilities for fire extinguishment.
- d. Methods of fire prevention such as use of fire resistant and non-combustible materials whenever practical, and minimizing exposure of combustible materials to fire hazards.
Amendment C 5.4-26 June 30, 1988
CESSAR22ece l
/n\ ,
N) j i
- e. Assurance that fire protection systems do not l adversely 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 l inadvertent operation does not significantly ' impair the capability of safety related i structures, systems, and components.
- 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 l l pipe whip, jet impingement, and environmental l ef fects resulting from postulated piping ruptures I l in the fire protection system.
- 2. Cooling Water System Requirements
- a. The cooling water system design shall be such that
,es cooling water is available to supply the SCS heat )
exchangers when an irradiated core is present in C (~~ ) the reactor vessel or the spent fuel pool.
- b. For all conditions, cooling water shall be supplied as follows:
Required Value Parameter Per Heat Exchancer 3 i Normal Allowable Delivery Pressure 100 psig l 1 Maximum Allowable Delivery Pressure 150 psig i 2 Required Flow rate 11,000 gpm l 1 Maximum Allowable Flow rate 13,000 gpm 3
- c. Cooling water piping supplying the shutdown ;
cooling heat exchangers shall be designed and j fabricated in accordance with ASME B&PVC, Section l III, Class 3, as a minimum, and shall be designed 1 as Seismic Category I, Safety Class 3, as a l minimum.
]
l I l (%) i i ) J \ i l l l l Amendment C j l S.4-27 June 30, 1988 j l 1 i __- J
CESSAR RE"lCATION O
- 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.
- c. 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 j accordance with the requirements of Section 3.11. j l
5.4.7.2 System Desian j 5.4.7.2.1 System Schematic I The SCS is shown on the RCS P&ID (Figure 5.1. 2-1) and on the SIS E and SCS P& ids (Figures 6.3.2-1A, 6.3.2-1B and 5.4.7-3. The ; i pressure and temperature of the RCS system vary from 400 psia and 350*F at initiation of shutdown cooling to atmospheric pressure and 120'F at refueling conditions. SCS design parameters are lC 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 equal 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 pumps are required to meet normal cooldown design criteria. C Amendment E ' 5.4-28 December 30, 1988
CESSAREnnncma o V 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 Shutdown cooling flow is measured by orifice meters installed in each train of the SCS discharge piping. The information provided l by these flow elements is used by the operator for flow control l during SCS operation. l The cooldown rate is controlled by adjusting flow through the heat exchangers with throttle valves on the discharge of each l heat exchanger. The operator maintains a constant total SCS-flow to the core by adjusting the heat 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 l C during cold shutdown. The units are sized to maintain a refueling water temperature of 120*F with the service water C temperature 95'F at 96 hours after shutdown following an assumed reactor core average burnup of two years. A l conservative 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 The design temperature is based upon the temperature of the reactor coolant at the initiation of shutdown cooling plus a design tolerance. l 1 i B. Instrumentation l The operation of the SCS is controlled nd monitored through l the use of installed instrumentation. "he instrumentation provides the capability to monitor cooldown rate and shutdown cooling flow to detect degradation of flow or-SCS C l heat removal capabilities. The instrumentation provided for the SCS is summarized in Section 7.5. lE C. Piping All SCS piping is austenitic stainless steel. All piping- [m
.q)
\ joints and connections are welded, except for a minimum i
l Amendment E 5.4-29 December 30, 1988
CESSARnaince Ol 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 pressure. Relief valves will be provided as required lC 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 C override 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. E. Shutdown Cooling System Pumps C The function of the SCS pumps is to provide flow through the reactor core and SCS heat exchangers for normal plant shutdown operation or as required for long term core cooling. ( 9 Amendment C 5.4-30 June 30, 1988 I
CESSARininc-t 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 supply the CSS. l The design temperature for the SCS pumps is based upon the temperature of the reactor coolant at the initiation of shutdown cooling - (350
- F 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 I 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 operation with a pumped fluid temperature of 400*F. The pump motors are specified to have the s capability of starting and accelerating the driven Os equipment, under 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 C 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 l to facilitate reduction of radiation levels before , maintenance. The pressure containing parts are fabricated I from stainless steel; the internals are selected for compatibility with boric acid solutions. The pumps 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 Overpressure Prevention A. Overpressurization of the SCS by the RCS is prevented in the following ways:
- 1. The shutdown cooling suction isolation valves (SD-673, 672, 671, 670) are powered by four independent power C Amendment C 5.4-31 June 30, 1988
i CESSARUnh o l OiI supplies such that a fault in one power supply or valve i will neither line up the RCS to either of the two SCS ! trains inadvertently nor prevent the initiation shutdown cooling with at least one SCS train. oflC
- 2. Interlocks associated with the shutdown cooling suction ;
isolation valves prevent the valves from being opened if RCS pressure exceeds 608 psia, and close these lE valves automatically if RCS pressure rise above the accumulation pressure of the SCS suction line relief lc valves. The instrumentation and controls which 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.
I Alarms on SD-673, 672, 671 and 670 annunciate when the lC 4. SCS suction isolation valves are not fully open. Also, l if SD-673 and 671 or SD-672 and 670 valves are open and i RCS pressure exceeds the maximum pressure for SCS l operation, an alarm will notify the operator that a 1 pressurization transient is occurring during low temperature conditions. , 1
- 5. Relief valves are provided as discussed in Section 5.4.7.2.2.
- 6. Conservative system piping design and maximum E utilization of welded connections.
B. Inadvertent overpressurization of the SCS is also precluded by the use of Safety Injection Tank isolation valves (see C j Sections 7.6). l 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 including SD-671 and 670 are designed to ASME B&PVC Section III, Class C 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. D. The component cooling water side of the SCS heat exchanger is designed to ASME B&PVC Section III, Class 3. Amendment E l 5.4-32 December 30, 1988
CESSAR inacamu ! i th E. The power operated valves are designed to the applicable ! IEEE Standards. F. The SCS is a Seismic Category I System. i 5.4.7.2.5 System Reliability Considerations The SCS is designed to perform its design function assuming a single failure, as described in Section 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 lC time period. The SCS does not utilize any pneumatically operated valves. lC ! l The instrumentation, control, and electric equipment pertaining to the SCS is designed to applicable portions of IEEE ! 279, 308 and 603. Standards lC In addition to normal offsite power sources, physically and , electrically independent and redundant emergency _ power supply I systems are provided to power safety-related components. See j ' Chapter 8 for further information. Since the SCS is essential 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 I used. l Environmental envelopes are specified for system components to l ensure acceptable performance in normal and applicable accident 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 be identified via appropriate leak detection provisions. (see Section 5.4.7.1.3.C.2). Make-up of the leakage is provided by C the manual alignment of the SIS to the IRWST or by opening the safety Injection Tank isolation valves. The affected SCS train can then be isolated and core ' cooling continued with the other train. Amendment C 5.4-33 June 30, 3988
)
CESSAR n%ncamu I I O' 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 Thisleakagetothepumpcompartmentwilldraintotheroomsump.lC From there it is pumped to the waste management system. The sump 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. When the RCS pressure falls below the normal operating range, thelC Safety Injection Actuation Signal (SIAS) setpoint can be manually decreased as discussed in Section 7.2.1.1.1.6. When RCS pressure reaches approximately 625 psig, the Safety Injection Tank pressure can be reduced to acceptable values. E When RCS pressure reaches approximately 400 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 lC 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 the 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 overpressurization of the SCS. See Section 7.6 and 5.4.7.2.3 for details. Also, if SD-673 and 671 or SD-672 C and 670 SCS suction isolation valves are open and RCS f pressure exceeds the maximum pressure for SCS operation, an 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 C is warmed up and placed in operation as follows (refer to Figures 6.3.2-1A, 6.3-2-1B and 5.4.7-3): l l Amendment E 3 5.4-34 December 30, 1988
I CESSAR 'casicuc,,
,v I ) %j
- 1. The SCS suction line isolation valves (SD-673*, 672, 671*, 670, 659*, 658) are opened.
- 2. The SCS throttle valves (SD-652, 653*) are cracked C open.
- 3. The SCS warmup line isolation valves (SD-657*, 656) are opened and the SCS pumps are started to recirculation flow through the SCS. induce l E
- 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 C 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. When complete, the system is in its E normal operational mode.
- 6. The SCS throttle valves (SD-652, 653*) and the SCS C
q bypass flow control valves (SD-651*, 652) are adjusted i i as necessary to maintain the RCS cooldown rate at
- 75'F/ hour or less, at the total SCS flow through each lE 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 SCS 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 l 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 C exchangers is increased to maintain the maximum permissible cooldown rate. 1 l g
- Odd numbered valves are located in SCS Train 1 and even numbered Lj
' G'j valves are located in SCS Train 2. Amendment E 5.4-35 December 30, 1988
)
I
CESSAR nutricariou O 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 after shutdown for a normal design basic cooldown is presented in Figure 5.4.7-1. lC Shutdown cooling is continued throughout the entire period of plant shutdown to maintain a of 120*F or less. Whenever refuelingwatertemperaturelC shutdown cooling is in 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 duringcoldshutdowntocontrolreactorcoolanttemperature.lC Prior to plant heatup, the SCS heat exchangers are bypassed to maintain flow through the core without the heat removal effect of the heat exchangers. Flow can be initiated to the heat exchangers if necessary to control the heatup rate. When the reactor coolant pumps 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 normal i shutdown cooling lineup for use of SCS pumps, exceptlC 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 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
CESSAR !!!acmu I l I l O 1 D. Design Bases Event Operations Following certain Design Bases Events (feedwater line break, small break LOCA, steam line break, or loss of offsite power), shutdown cooling can be initiated with RCS hot leg conditions which exceed the normal shutdown cooling lC initiation temperature of 350*F. However, shutdown cooling .) will never be initiated at conditions which exceed ; design temperature of the SCS components. the l C ' 5.4.7.3 _P__erformance Evaluation The design point of the SCS is taken at 96 hours after plant At this point, the design basis is to maintain a 120*F C shutdown. refueling temperature with a service water temperature of 95*F. Two SCS heat exchangers and two SCS pumps are assumed to be ir operation at the design flow. The SCS heat exchanger size is determined at this point, since it requires the greatest heat lE 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 O 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. , E At each time interval in the cooldown, an iterative process is utilized to analyze trana. ' ut performance, whereby the heat removal is established by 4 m ting the available heat load with the SCS heat exchanger heat moval capability. The cooldown l rate is limited to a maximum of 75*F/ hour throughout the i cooldown. The normal two train cooldown curve is shown in Figure 5.4.7-1. i 'I l With the most limiting single active failure in the SCS, RCS l l 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. k l Amendment E 5.4-37 December 30, 1988
~
CESSAR EnWicarios O The SCS is designed utilizing a philosophy of total physical separation of redundant trains such that the syster 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 C periods >24 hr) after event initiation. Total train separation 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 Preoperational Testinq 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 SCS 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 accordance with Section III of the ASME Boiler and Pressure Vessel Code. O Amendment C 5.4-38 June 30, 1988
.CESSAR !!=nco,.
4 O 1 THIS PAGE INTENTIONALLY BLANK O O . Amendment c 5.4-39 June 30, 1988
CESSAR 8lnhuo. O THIS PAGE INTENTIONALLY BLANK O l O Amendment C 5.4-40 June 30, 1988
CESSAR 8!niflCAT10N O THIS PAGE INTENTIONALLY BLANK O O Amendment C 5.4-41 June 30, 1988
CESSAR E!aincuion y A l THIS PAGE INTENTIONALLY BLANK l b
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e l O Amendment C 5.4-42 June 30, 1988
CESSAR EMWicari:n j O TABLE 5.4.7-1 (Sheet 1 of 2) SHUTDOWN COOLING DESIGN PARAMETERS SYSTEM DESIGN PARAMETERS l Shutdown cooling system startup Approximately 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 120 C Nominal shutdown cooling flow, gpm/HX 5000 !
. COMPONENT DESIGN PARAMETERS Shutdown Coolina Heat Exchanaer Data Quantity 2 1 Type 2
Shell and tube, horizontal U-tube { Service transfer rate,28tu/hr *F-ft 350 lC l Heat Transfer area, ft /HX 6010 lE Tube Side Fluid Reactor coolant Design pressure, psig 900 lE Design temperature, 'F 400 lC Material Austenitic stainless steel Code ASME Section III, Class 2 1 Shell Side lE Fluid Component cooling water Design pressure, psig 150 Design temperature, 'F 250 Material Carbon Steel Code ASME Section III, Class 3 lE At 96 hours after shutdown: 1 Tube Site C i Flow, million 1b/hr. 2.47 i Inlet temperature, *F 120 j Shell Side E Flow, million 1b/hr 5.47 Inlet temperature, 'F 98 Amendment E December 30, 1988 l i
l CESSAR H5Gcam, O ! IMM_5.4a l-1 (Cont'd) (Sheet 2 of 2) ; SHUTDOWN COOLING DESIGN PARAMETERS I Shutdown Coolina Pumo Data Quantity 2 Type Single Stage, Vertical, Centrifugal l Safety Classification 2 Code ASME III, Class 2 ! Design Pressure, psig 900 Maximum Operating Suction Pressure, psig 385 E Design Temperature, 'F 400 , Design Flow Rate, gpm 5,000 l Design Head, ft 400 l Materials Stainless Steel Type 304, 316 or approved alternate C Seals Mechanical E I
'l L
l 1 4 l O Amendment E December 30, 1988
CESSAR E*Ricamn (" TABLE 5.4.7-2 SHUTDOWN COOLING SYSTEM INTERFACE REQUIREMENTS FOR COMPONENT COOLING WATER Shutdown Shutdown Cooling Cooling Mode (3.5 hrs.) _.{96 hrs) Supply Temp, 'F (Max)(I) 120 (LATER) C Outlet Temp, 'F 142 (LATER) Minimum Flow per SDCHX, gpm(2) 11,000 11,000 Total Heat Load 6 242 51 for Both SDCHX, 10 Btu /hr 1 0 NOTES:(1) For maximum supply temperatures lower than those l listed, the minimum flow listed may be reduced provided that the heat removal capability of the Shutdown Cooling System , is not adversely affected. Conversely, for Component Cooling Water supply temperature:: lower than those listed, it may be necessary to reduce flow, so that the heat capacity of the ultimate heat sink is not exceeded. C (2) Maximum allowable component cooling water flow through each shutdown cooling heat exchanger is 13,000 gpm. .O i Amendment C June 30, 1988
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CESSAR HEnce,, i O 5.4.8 ' REACTOR COOLANT CLEANUP SYSTEM One function of the Chemical and Volume Control System (CVCS) is - to providt radiological and chemical cleanup of the Reactor ; Coolant System. A description of the CVCS is given in Section 9.3.4. Radiological considerations are described in Chapters 11 and 12. 5.4.9 MAIN STEAM LINE AND FEEDWATER PIPING j See Chapter 10. l lB 5.4.10 PRESSURIZER 5.4.10.1 Desian Bases The pressurizer is designed to: A. Maintain RCS operating pressure such that the minimum pressure during operating transients is above the setpoint 3 for the Safety Injection Actuation Signal and low pressure reactor trip, and such that the maximum pressure is below the high pressure reactor trip setpoint. B. Meet the design transients specified in Section 3.9.1 except that the maximum allowable rate of change in pressurizer temperature during plant heatup and cooldown is 200*F/hr. C. Provide sufficient water volume in the pressurizer to prevent uncovering the heaters as a result of a reactor B trip. D. Provide sufficient water volume to prevent pressurizer l heaters from being uncovered by the outsurge fcllowing step load decreases of 10% starting from 100% to 25% or a 5% per B l minute ramp decrease from 100% to 15%. l l E. Provide sufficient steam volume to avoid lifting the primary safety valves as a result of a loss of condenser vacuum (LOCV) event (normal control systems are operational). E F. Provide sufficient steam volume to allow acceptance of the insurge resulting from any loss of load transient without 3 liquid or two-phase flow reaching the primary safety valve nozzles. G. Minimize the total reactor coolant mass change and associated charging and letdown flow rates in order to reduce the quantity of wastes generated by load follow 1 operations. ! Amendment E i l 5.4-43 December 30, 1988
CESSARE h w O
- 11. Provide sufficient pressurizer heater capacity to heat up the pressurizer, filled with water at the zero power level, at a rate that ensures a pressurizer temperature (and thus pressure) which will maintain an adequate degree of subcooling of the water in the reactor coolant loop as it is heated by core decay heat and/or pump work from the reactor coolant pumps.
I. Contain a total water volume that does not adversely affect the total mass and energy released to the containment during the maximum hypothetical accident. J. Ensure that, in addition to being specified as Seismic Category I, the pressurizer vessel, including heaters, baffles, and supports shall be designed such that no damage to the equipment is caused by the frequency ranges of 19-20 cps and 95-100 cps. The lower frequency is defined as for the reactor vessel. The design basis for the higher frequency consists of a pressure pulse of 5 psi which diminishes internally within the vessel. K. The combination of maximum heat loss from the pressurizer and the pressurizer heater capacity shall be such as to maintain the pressurizer at normal operating pressure during hot standby conditions. This capability shall be provided by redundant trains of heaters powered from off-site power and Class lE emergency power. L. The total spray flows shall be sufficient to keep the pressure below the reactor trip setpoint during an insurge of water during the " Maneuvering and Load Follow" and " Loss of Load" transients. E M. The pressurizer size and spray capacity shall be adequate so the pressurizer safety valves are not actuated by overpressure events initiated by normal operation transients. E 5.4.10.2 Depeript. ion the pressurizer, as shown in Figure 5.4.10-1, is a vertically mounted, bottom supported, cylindrical pressure vessel. Replaceable direct immersion electric heaters are vertically mounted in the bottom head. The pressurizer is furnished with nozzles for the spray, surge, and safety lines, and pressure and level instrumentation. A canway is provided in the top head for access for inspection of the pressurizer internals. The pressurizer surge line is connected to one of the reactor coolant hot legs and the spray lines are connected to two of the cold a legs at the reactor coolant pump discharge. Heaters are Amendment E 5.4-44 December 30, 1988
~ CESSAR E%ncari:n l supported inside the pressurizer to preclude damage from B vibration and seismic loadings. Principal design parameters are listed in Table 5.4.10-1. The pressurizer is designed and fabricated in accordance with the ASME Code listed in Table 5.2-1. The interior surface is clad with weld deposited stainless steel. j The total volume of the pressurizer is established by consideration of the factors given in Section 5.4.10.1. To ; account for these factors and to provide adequate margin at all ! power levels, the water level in the pressurizer is programmed as a function of average coolart . temperature as shown in Figure 5.4.10-2, in conjunction with Figure 5.4.10-3. High or low water level error signals result in the control actions shown in Figure 5.4.10-4. The pressurizer surge line is sized to accommodate the flow rates associated with the RCS expansion and contraction due ! to the transients specified in Section 3.9.1. j The pressurizer maintains Reactor Coolant System operating pressure and, in conjunction with the Chemical and Volume Control ] System (CVCS), Section 9.3.4, compensates for changes in reactor 1 coolant volume during load changes, heatup, and cooldown. During O full-power operation, the pressurizer is about one-half full of aaturated steam. Reactor Coolant System pressure may be controlled automatically or manually by maintaining the . temperature of the pressurizer fluid at the saturation i temperature corresponding to the desired system pressure. A ) small continuous spray flow is maintained to the pressurizer to I avoid stratification of pressurizer boron concentration and to maintain the temperature in the surge and spray lines, thereby reducing thermal shock as the spray control valves open. An auxiliary spray line is provided from the charging pumps to permit pressurizer spray during plant heatup, or to allow cooling if the reactor coolant pumps are shut down. During load changes, the pressurizer limits pressure variations caused by expansion or contraction of the reactor coolant. The average reactor coolant temperature is programmed to vary as a function of load as shown in Figure 5.4.10-3. A reduction in load is followed by a decrease in the average reactor coolant temperature to the programmed value for the lower power level. l The resulting contraction of the coolant lowers the pressurizer l water level, causing the reactor system pressure to decrease. Titis pressure reduction is partially compensated by flashing of pressurizer water into steam. All pressurizer heaters are j automatically energized on low system pressure, generating steam I and further limiting pressure decrease. Should the water level I in the pressurizer drop sufficiently below its setpoint, the letdown control valves close to a minimum value, and the charging fB Amendment B i 5.4-45 March 31, 1988 1
CESSAR ES's"icari:n ! l Oi flow control valve (s) open in the chemical and volume control B system are automatically controlled to add coolant to the system and restore pressurizer level. When steam demand is increased, the average reactor coolant temperature is raised in accordance with the coolant temperature program. The expanding coolant from the reactor coolant piping hot leg enters the bottom of the pressurizer through the surge line, compressing the steam and raising system pressure. The increase in pressure is moderated by the condensation of steam 1 during compression and by the decrease in bulk temperature in the ; liquid phase. Should the pressure increase be large enough, the pressurizer spray valves open, spraying coolant from the reactor coolant pump discharge (cold leg) into the pressurizer steam j space. The relatively cold spray water condenses some of the steam in the steam space, limiting the system pressure increase. The programmed pressurizer water level is a temperature dependent function. A high level error signal, produced by an in-surge, causes the letdown control valves to modulate open, releasing coolant to the chemical and volume control system, and the B charging flow control valve is closed to a minimum value, thus I restoring the pressurizer to the programmed level. Small I pressure and primary coolant volume variations are accommodated i J by the steam volume that absorbs flow into the pressurizer and by the water volume that allows flow out of the pressurizer. The pressurizer heaters are single unit, direct immersion heaters that protrude vertically into the pressurizer through sleeves welded in the lower head. Each heater is internally restrained from high amplitude vibrations and can be individually removed l for maintenance during plant shutdown. A number of the heaters are connected to proportional , controllers, which adjust the heat input to account for ! steady-state losses and to maintain the desired steam pressure in l the pressurizer. The remaining heaters are connected to on-off l controllers. These heaters are normally deenergized but are automatically turned on by a low pressurizer pressure signal or a j high level error signal. This latter feature is provided since i load increases result in an in-surge of relatively cold coolant i into the pressurizer, thereby decreasing the bulk water temperature. The CVCS acts to restore level, resulting in a transient pressure below normal operating pressure. To minimize i the extent of this transient, the backup heaters are energized, { contributing more heat to thu water. Backup heaters are 6 deenergized in the event of concurrent high-level error and high-pressurizer pressure signals. A low-low pressurizer water level signal deenergizes all heaters before they are uncovered to prevent heater damage. The pressure control program is shown in Figure 5.4.10-5. Amendment B j 5.4-46 March 31, 1988 j l
CESSAR EHHncamn O 5.4.10.3 Evaluation i It is demonstrated by analysis in accordance with requirements for ASME Code, Section III, Class 1 vessels that the pressurizer is adequate for all normal operating and transient conditions expected during the life of the facility. Following completion of fabrication, the pressurizer is subjected to the required ASME i Code, Section III hydrostatic test and post-hydrostatic test non-destructive testing. l During hot functional testing, the transient performance of the pressurizer is checked by determining its normal heat losses and maximum depressurization rate. This information is used in setting the pressure controllers. Further assurance of the structural integrity of the pressurizer during plant life will be obtained from the inservice inspections performed in accordance with ASME Code, Section XI, and described in Section 5.2. Overpressure protection of the Reactor Coolant System is provided by four ASME Code spring-loaded safety valves. Refer to Section 5.4.12 and 5.4.13. 5.4.10.4 Tests and Inspections Prior to and during fabrication' of the pressurizer, non-destructive testing is performed in accordance with 'the requirements 'of Section III of the ASME Boiler and Pressure Vessel Code. Table 5.4.10-2 summarizes the pressurizer l inspection program, which also includes tests not required by the Code. Refer to Section 5.2.1 for inservice inspections of the pressurizer. 1 1 l l O S.4-47
CESSAR nutriCATION O TABLE 5.4.10-1 PRESSURIZER PARAMETERS Property Parameter Design pressure, psia 2500 Design temperature, 'F 700 flormal operating pressure, psia 2EoD flormal operating temperature, *F 652.7 Internal free volume, ft 2400 flormal (full power) operating water volume, ft 3 1200 B flormal (full power) steam volume, ft 3 1200 Installed heater capacity, kW 2400 lE Heater type Immersion Spray flow, minimum design capacity, gpm 375 lD Spray flow, (maximum) continuous, gpm 3 lE f10ZZles Surge, in. (nominal) 12, schedule 160 Spray, in. (nominal) 4, schedule 160 Safety valves, in. (nominal) 6, schedule 160 lE Instrument Level, in. (nominal) 3/4, schedule 160 Temperature, in. (nominal) 1, schedule 160 Pressure, in. (nominal) 3/4, schedule 160 Heater, 0.D., in. 1-1/4 O Amendment E December 30, 1988
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C E S S A R n nincuia O TABLE 5.4.10-2 PRESSURIZER TESTS Component Tests (a) Heads Plates UT, MT Cladding Shell Plates UT, MT Cladding UT, PT q Heaters Tubing UT, PT Centering of elements RT End Plug UT, PT Nozzle (Forgings) UT, MT Studs UT, MT Welds l Shell, longitudinal RT, MT Shell, circumferential RT, MT l l Cladding UT, PT Nozzles RT, MT 1 Nozzle safe ends RT, PT ! l Instrument connections PT Support Skirt MT, RT Temporary attachment after removal MT ! All welds after hydrostatic test MT or PT Heater assembly, end plug weld PT (a) Key: UT = ultrasonic testing MT - magnetic particle testing PT - dye-penetrant testing RT - radiographic testing O
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CESSARnahmu O 5.4.11 PRESSURIZER RELIEF TANK J l The in-containment refueling water storage tank (IRWST) is used h, I as the pressurizer relief tank. The design and description of this tank are given in Section 5.4.7. 5.4.12 VALVES 5.4.12.1 p_e s_i_gn Ba s i s The safety-related functions of valves within the reactor coolant pressure boundary are to act as pressure retaining vessels and leaktight barriers during normal plant operation, accidents and seismic disturbances. These valves are designed in accordance with ASME Code, Section III, Class I requirements and must withstand the effects of the system design transients (see Section 3.9.1) plus other transients associated with each valves location or service requirements. The valves meet Sciumic Category 1 requirements. Valve stem leakage control for reactor coolant pressure boundary valves is accomplished through the use of double packings with a leak-off connection at the lantern ring to provide for collection of leakage. Backseats are specified on manual and motor-operated gate and globe valves to further minimize potential leakage. Materials of construction are specified to assure compatibility with the environment and contained fluids. 5.4.12.2 Design Description All valves in the reactor coolant system are constructed primarily of stair.less steel. Other materials in contact with the coolant, such as hard facing and packing, are compatible materials. Fasteners, packing, gland aFsemblics, and yoke fanteners are constructed of stainless steel to eliminate corrosion problems. Valves not constructed with diaphragms, or bellows, or not normally backscated, are provided with two independent sets of packirg, separated by a lantern ring and leak-off connections piped to a collection system to eliminate uncontrolled leakage. 5.4.12.3 Design Evaluation All valves within the reactor coolant pressure boundary are stress analyzed in accordance with ASME Code, Section 11I, Cle.ss I, taking into consideration cyclic loadings. O Amendment E 5.4-48 December 30, 1988
CESSARnnac-O v 5.4.12.4 Tests and Inspections The valves are hydrostatically tested and leak tested across the seats and across the packing in accordance with the ASME B&PV Code as specified in Table 5.2-1. 5.4.13 SAFETY AND RELIEF VALVES 5.4.13.1 Desjon Basis l l The safety valves on the pressurizer are designed to protect the primary system, as required by the ASME D&PV Code, Section III. l0 The design basis for establishing the relieving capacity of the pressurizer safety valves is presented in Appendix 5A. For the postulated transients presented in Chapter 15, the results indicate that relieving capacity of the safety valves is sufficient to provide overpressure protection in accordance with Section III of the ASME Code. Safety valves on the steam side of each steam generator are designed to protect the steam system, as required by the ASME o Code, Section III. They are conservatively sized to pass a ( steady flow equivalent to the maximum expected power level at the design pressure of the steam system. 5.4.13.2 Description The RCS has four safety valves to provide overpresenre protection. A typical safety valve is illustrated in Figure 5.4.13-1. The design paraneters are given in Table 5.4.13-1. These valves are connected by piping to the top of the pressurizer. They are direct acting, spring-loaded safety valves meeting ASME Code requirements. They have an enclosed bonnet and have a balanced bellows to compensate for backpressure. The s'afety valves pass sufficient pressurizer steam to limit the Reactor Coolant System pressure to 110% of design pressure (2750 psia) following a complete loss of turbine generator load without simultaneous reactor trip. A delayed reactor trip is assumed on a high-pressurizer pressure signal. To determine maximum steam flow through the pressurizer safety valves, the main steam safety valves are assumed to be operational. Values for the system parameters, delay times, and core moderator coefficient are given in Chapter 15. Overpressure protection for the shell side of the steam generators and the main steam line up to the inlet of the turbine stop valve is provided by the secondary safety valves.
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, Amendment D 5.4-49 September 30, 1988
CESSARE5a m i O These valves are each sized to pass a steam flow of 945,292 lb/hr ; at 1308 psia. This limits steam generator pressure to less than ! 110% of steam generator design pressure during worst case I transients. The secondary safety valves consist of two banks of j 10 valves with staggered set pressures. The valves are spring-loaded safety valves procured in accordar.ce with ASME Boiler and Pressure Vessel Code, Section III (see Table 5.2-1). l Parameters for the secondary safety valves are given in Table 4 5.4.13-2. 5.4.13.3 RyAluation overpressure protection is discussed in Section 5.2.2. The ASME Code report on Overpressure Protection is included as Appendix SA. 5.4.13.4 Tests and Inspections The valves are inspected during fabrication in accordance with ASME III Code requirements. I i 5.4.13.4.1 Pressurizer Safety Valves I The inlet and outlet portions of the valves are hydrostatically tested with water at the appropriate pressures required by the applicable section of the ASME Code. Set pressure and seat l leakage tests can be performed with steam using a pro-rated I spring. Final set pressure tests are performed with the final springs using either high pressure steam or low pressure steam , with an assist device. Final seat leakage tests are performed l prior to shipment with the final springs using either hot air or hot nitrogen. Valve adjustment shall be made to a valve ring settingcombinationselectedtoprovidestablevalveopegionon 1 the basis of the EPRI Safety Valve Test Program results. 5.4.13.4.2 Main Steam Safety Valves The inlet portion of the valve is hydrostatically tested with water in accordance with the ASME Code. Set pressure and set , leakage tests are performed using steam. Adjustment is made to l provide a valve blowdown meeting the requirement specified in Table 5.4.13-2. l (1) CEN-227 " Summary Report on the Operability of Pressurizer Safety Relief Valve in C-E Designed Plants", December 1982. 5.4-50
1 CESSAR E!nificui . J i l I A TABLE 5.4.13-1 PRESSURIZER SAFETY VALVE PARAMETERS Property Parameter Design pressure, psia 2500 Design temperature, 'F 700 Fluid Saturated Steam, 4400 ppm boron, pH = 4.5 to 10.6 Set pressure, psia 2500 1% Hin. capacity, lb/hr at, accumulation 460,000 pressure, each Type Spring loaded safety-balanced bellows, D enclosed bonnet O l Orifice area, in.2 4,34 Accumulation, % 3 ' Backpressure B Max. buildup / max superimposed, psig 700 /340 Blowdown, % 18.5 Min. blowdown pressure, psia 2040 i Typical materials 1 Body ASME SA 182, GR. F316 Disc ASTM B637, UNS N07750 0 Nozzle ASME SA 182, GR. 347 Maximum allowable back pressure. Amendment D September 30, 1988
CESSAR En'iincamn O TABLE 5.4.13-2 MAIN STEAM SAFETY VALVE PARAMETERS l l ProDerty Parameter j Design pressure, lb/in.2g 1375 l Design temperature, 'F 575 l Fluid Saturated Steam j Set pressure, 1b/in.29 1255, 1290, 1315 6 Min. capacity, lb/h at accumulation pressure 19x10 Total (20 Valves) Type Spring loaded Orifice area, in.2 16 Accumulation, % 3 Backpressure ! Max. buildup / max superimposed, lb/in.29 125/0 Approx. dry weight, lbs. 1545 Minimum blowdown pressure, psig 1175 Typical materials Body ASME SA 105 Disc ASTM A565, GR. 616 Nozzle ASME SA 182, GR. F316 O m a am-mm
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CESSAR 85L,.r 5.4.14 COMPONENT SUPPORTS 5.4.14.1 Desien Basis The criteria applied in the design of the Reactor Coolant System supports are that the specific function of the supported equipment be achieved during all normal, earthquake, safety valve ' actuation and Branch Line Pipe Break (BLPB) conditions. (BLPB includes feedwater line breaks and all loss-of-coolant-accident conditions resulting from breaks not eliminated by B leak-before-break analysis in piping to branch nozzles of the reactor coolant system.) Specifically, the supports are designed to support and restrain the Reactor Coolant System components under the combined Safe Shutdown Earthquake and Branch Line Pipe B Break loadings in accordance with the stress and deflection limits of Section III, ASME Code. 5.4.14.2 DescriDtien Figure 5.4.14-1 illustrates the Reactor Coolant System support points. A description of the supports for each supported component follows: A. Reactor Vessel Supports The reactor vessel is supported by four vertical columns located under the vessel inlet nozzles. These columns are designed to flex in the direction of horizontal thermal expansion and thus allow unrestrained heatup and cooldown. They also act as holddown devices for the vessel. Horizontal keyways located alongside the upper portion of the column guide the vessel during thermal expansion and contraction of the Reactor Coolant System and maintain the vessel centerline. Four horizontal keys are welded to the bottom vessel head. The column base plate ac'~ as a keyway for these keys to restrain the bottom of the vessel. The supports are designed to accept normal, seismic, and Branch Line Pipe Break loads. Irradiation effects are eddressed in the fracture mechanics analysis of columns, in combination with the design basis loads, D to ensure that structural integrity will be maintained. Reactor vessel supports are shown in Figure 5.4.14-2. Amendment D 5.4-51 September 30, 1988
CESSAR Enecm2 O B. Steam Generator Supports The steam generator is supported at the bottom by a sliding base bolted to an integrally attached conical skirt. The sliding base rests on low friction bearings which allow < unrestrained thermal expansion of the Reactor Coolant System. Two keyways within the sliding base mate with embedded keys to guide the movement of the steam generator during expansion and contraction of the Reactor Coolant ! System and limit movement of the bottom of the steam D generator during seismic events and Branch Line Pipe Breaks. A system of keys and snubbers located on the steam drum guide the top of the steam generator during expansion and contraction of the Reactor Coolant System and provide support during seismic events and following Branch Line Pipe D Breaks. Typical steam generator supports are shown in Figure 5.4.14-3. C. Reactor Coolant Dump Supports Each reactor coolant pump is provided with four vertical support columns, four horizontal support columns, and two horizontal snubbers. The rigid structural columns provide support for the pumps during normal operation, earthquake conditions, and Branch Line Pipe Breaks. An illustration of D the pump supports is shown in Figure 5.4.14-4. D. Pressurizer Supports The pressurizer is supported by a cylindrical skirt welded to the pressurizer and bolted to the support structure. The skirt is designed to withstand deadweight and normal operating loads as well as the loads due to earthquakes, safety valve actuation, and Branch Line Pipe Breaks. Four keys welded to the upper shell provide additional restraint for earthquake, safety valve actuation, and Branch Line Pipe D Break conditions. 5.4.14.3 Evaluation The structural integrity of the reactor coolant system support components is ensured by quality assurance inspections in accordance with Section III of the ASME Code during fabrication. The non-integral supports are procured by individual equipment specifications which impose appropriate quality assurance , requirements commensurate with the respective component's I functions. Amendment D 5.4-52 September 30, 1988 '
5'2" CESSAR "ERTIFICAT13N C V During pre-operational testing of the Reactor Coolant System, the support displacements will be monitored for concurrence with calculated displacements and/or clearances. Subsequent inspections of supports which are integral with Reactor Coolant System components will be in accordance with Section XI of the ASME Code. w l l 5.4-53
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