DCL-17-038, Diablo Canyon Power Plant, Units 1 & 2, Revised Updated Final Safety Analysis Report, Rev. 23, Chapter 5, Reactor Coolant System
Text
DCPP UNITS 1 &
2 FSAR UPDATE Chapter 5 REACTOR COOLANT SYSTEM CONTENTS Section Title Page i Revision 23 December 2016 5.1
SUMMARY
DESCRIPTION 5.1-1 5.1.1 DESIGN BASES 5.1-3 5.1.1.1 General Design Criterion 2, 1967 - Performance Standards 5.1-3 5.1.1.2 General Design Criterion 3, 1971 - Fire Protection 5.1-3 5.1.1.3 General Design Criterion 4, 1967 - Sharing of Systems 5.1-3 5.1.1.4 General Design Criterion 4, 1987 - Environmental and
Dynamic Effects Design Bases 5.1-3 5.1.1.5 General Design Criterion 6, 1967 - Reactor Core Design 5.1-3 5.1.1.6 General Design Criterion 9, 1967 - Reactor Coolant Pressure
Boundary 5.1-4 5.1.1.7 General Design Criterion 11, 1967 - Control Room 5.1-4 5.1.1.8 General Design Criterion 12, 1967 - Instrumentation
and Controls 5.1-4 5.1.1.9 General Design Criterion 13, 1967 - Fission Process Monitors
and Controls 5.1-4 5.1.1.10 General Design Criterion 15, 1967 - Engineered Safety
Features Protection Systems 5.1-4 5.1.1.11 General Design Criterion 21, 1967 - Single Failure Definition 5.1-4 5.1.1.12 General Design Criterion 40, 1967 - Missile Protection 5.1-4 5.1.1.13 General Design Criterion 49, 1967 - Containment Design Basis 5.1-5 5.1.1.14 General Design Criterion 54, 1971 - Piping Systems
Penetrating Containment 5.1-5 5.1.1.15 General Design Criterion 55, 1971 - Reactor Coolant Pressure
Boundary Penetrating Containment 5.1-5 5.1.1.16 General Design Criterion 56, 1971 - Primary Containment
Isolation 5.1-5 5.1.1.17 Reactor Coolant System Safety Function Requirements 5.1-5 5.1.1.18 10 CFR 50.49 - Environmental Qualification of Electrical
Equipment Important to Safety for Nuclear Power Plants 5.1-6 5.1.1.19 10 CFR 50.55a(f) - Inservice Testing Requirements 5.1-6 5.1.1.20 10 CFR 50.55a(g) - Inservice Inspection Requirements 5.1-6 5.1.1.21 10 CFR 50.63 - Loss of All Alternating Current Power 5.1-6 5.1.1.22 10 CFR Part 50 Appendix R (Sections III.G, III.J, II I.L, and III.O)
- Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979 5.1-7 5.1.1.23 Regulatory Guide 1.89, November 1974 - Environmental Qualification of Class 1E Equipment for Nuclear Power Plants 5.1-7 5.1.1.24 Regulatory Guide 1.97, Revision 3, May 1983 - Instrumentation DCPP UNITS 1 &
2 FSAR UPDATE Chapter 5 REACTOR COOLANT SYSTEM CONTENTS Section Title Page ii Revision 23 December 2016 for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident 5.1-7 5.1.1.25 Regulatory Guide 1.121, August 1976 - Bases for Plugging Degraded PWR Steam Generator Tubes 5.1-8 5.1.1.26 NUREG-0737 (Items II.B1, II.D.1, IIE.3.1, II.F.2, II.
G.1, II.K.3.5, and II.K.3.25), November 1980 - Clarification of TMI Action Plan Requirements 5.1-8 5.1.1.27 Generic Letter 83-37, November 1983 - NUREG-0737 Technical Specifications 5.1-9 5.1.1.28 Generic Letter 88-05, March 1988 - Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants 5.1-9 5.1.1.29 Generic Letter 90-06, June 1990 - Resolution of Generic Issue 70, "Power-Operated Relief Valve and Block Valve Reliability" and Generic Issue 94, "Additional Low-Temperature Over Pressure Protection for Light-Water Reactors" Pursuant to 10 CFR 50.54(f) 5.1-9 5.1.1.30 Generic Letter 95-07, August 1995 - Pressure Locking and Thermal Binding of Safety-Related Power-Operated Valves 5.1-9 5.1.1.31 NRC Bulletin 88-09, July 1988 - Thimble Tube Thinning in Westinghouse Reactors 5.1-9 5.1.1.32 NRC Bulletin 88-11, December 1988 - Pressurizer Surge Line Thermal Stratification 5.1-9 5.1.1.33 Branch Technical Position ASB 10-2, March 1978 - Design Guidelines for Avoiding Water Hammers in Steam Generators 5.1-9
5.1.2 SCHEMATIC FLOW DIAGRAMS 5.1-10
5.1.3 PIPING AND INSTRUMENTATION DIAGRAMS 5.1-10
5.1.4 ELEVATION DRAWINGS 5.1-10
5.1.5 REACTOR COOLANT SYSTEM COMPONENTS 5.1-11
5.1.6 REACTOR COOLANT SYSTEM PERFORMANCE AND SAFETY FUNCTIONS 5.1-11 5.1.6.1 Reactor Coolant System Flow Determination and Safety Analyses 5.1-11 5.1.6.2 Reactor Coolant Flow 5.1-12 DCPP UNITS 1 &
2 FSAR UPDATE Chapter 5 REACTOR COOLANT SYSTEM CONTENTS Section Title Page iii Revision 23 December 2016 5.1.6.3 Best Estimate Flow 5.1-12 5.1.6.4 Thermal Design Flow 5.1-13 5.1.6.5 Mechanical Design Flow 5.1-13 5.1.6.6 Minimum Measured Flow 5.1-13 5.1.6.7 Minimum Required Reactor Coolant System Flow Rate 5.1-14
5.1.7 SYSTEM OPERATION 5.1-14 5.1.7.1 Plant Startup 5.1-14 5.1.7.2 Power Generation and Hot Standby 5.1-15 5.1.7.3 Plant Shutdown 5.1-16 5.1.7.4 Refueling 5.1-17 5.1.7.5 Mid-Loop Operation 5.1-17 5.1.8 SAFETY EVALUATION 5.1-17 5.1.8.1 General Design Criterion 2, 1967 - Performance Standards 5.1-17 5.1.8.2 General Design Criterion 3, 1971 - Fire Protection 5.1-17 5.1.8.3 General Design Criterion 4, 1967 - Sharing of Systems 5.1-18 5.1.8.4 General Design Criterion 4, 1987 - Environmental and
Dynamic Effects Design Bases 5.1-18 5.1.8.5 General Design Criterion 6, 1967 - Reactor Core Design 5.1-18
5.1.8.6 General Design Criterion 9, 1967 - Reactor Coolant Pressure Boundary 5.1-18 5.1.8.7 General Design Criterion 11, 1967 - Control Room 5.1-19 5.1.8.8 General Design Criterion 12, 1967 - Instrumentation and Controls 5.1-19 5.1.8.9 General Design Criterion 13, 1967 - Fission Process Monitors and Controls 5.1-20 5.1.8.10 General Design Criterion 15, 1967 - Engineered Safety Features Protection System 5.1-20
5.1.8.11 General Design Criterion 21, 1967 - Single Failure Definition 5.1-20 5.1.8.12 General Design Criterion 40, 1967 - Missile Protection 5.1-21 5.1.8.13 General Design Criterion 49, 1967 - Containment Design Basis 5.1-21 5.1.8.14 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment 5.1-21 5.1.8.15 General Design Criterion 55, 1971 - Reactor Coolant Pressure Boundary Penetrating Containment 5.1-21 5.1.8.16 General Design Criterion 56, 1971 - Primary Containment Isolation 5.1-22 5.1.8.17 Reactor Coolant System Safety Function Requirements 5.1-22 DCPP UNITS 1 &
2 FSAR UPDATE Chapter 5 REACTOR COOLANT SYSTEM CONTENTS Section Title Page iv Revision 23 December 2016 5.1.8.18 10 CFR 50.49 - Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants 5.1-23 5.1.8.19 10 CFR 50.55a(f) - Inservice Testing Requirements 5.1-23 5.1.8.20 10 CFR 50.55a(g) - Inservice Inspection Requirements 5.1-24 5.1.8.21 10 CFR 50.63 - Loss of All Alternating Current Power 5.1-24 5.1.8.22 10 CFR Part 50 Appendix R (Se ctions III.G, III.J, III.L, and III.O) - Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979 5.1-24 5.1.8.23 Regulatory Guide 1.89, November 1974 - Environmental Qualification of Class 1E Equipment for Nuclear Power Plants 5.1-25
5.1.8.24 Regulatory Guide 1.97, Revision 3, May 1983 - Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident 5.1-25 5.1.8.25 Regulatory Guide 1.121, August 1976 - Bases for Plugging Degraded PWR Steam Generator Tubes 5.1-25 5.1.8.26 NUREG-0737 (Items II.B.1, II.D.1, II.E.3.1, II.F.2, II.G.1, II.K.3.5, and II.K.3.25), November 1980 - Clarification of TMI Action Plan Requirements 5.1-25 5.1.8.27 Generic Letter 83-37, November 1983 - NUREG-0737 Technical Specifications 5.1-27 5.1.8.28 Generic Letter 88-05, March 1988 - Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants 5.1-27 5.1.8.29 Generic Letter 90-06, June 1990 - Resolution of Generic Issue 70, "Power-Operated Relief Valve and Block Valve Reliability" and Generic Issue 94, "Additional Low-Temperature Over Pressure Protection for Light-Water Reactors" Pursuant to 10 CFR 50.54(f) 5.1-27 5.1.8.30 Generic Letter 95-07, August 1995 - Pressure Locking and Thermal Binding of Safety-Related Power-Operated Valves 5.1-28 5.1.8.31 NRC Bulletin 88-09, July 1988 - Thimble Tube Thinning in Westinghouse Reactors 5.1-28 5.1.8.32 NRC Bulletin 88-11, December 1988 - Pressurizer Surge Line Thermal Stratification 5.1-28 5.1.8.33 Branch Technical Position ASB 10-2, March 1978 - Design Guidelines for Avoiding Water Hammers in Steam Generators 5.1.28
5.
1.9 REFERENCES
5.1-28
DCPP UNITS 1 &
2 FSAR UPDATE Chapter 5 REACTOR COOLANT SYSTEM CONTENTS Section Title Page v Revision 23 December 2016 5.1.10 REFERENCE DRAWINGS 5.1-29 DCPP UNITS 1 &
2 FSAR UPDATE Chapter 5 CONTENTS (Continued)
Section Title Page vi Revision 23 December 2016 5.2 INTEGRITY OF THE REACTOR COOLANT PRESSURE BOUNDARY 5.2-1 5.2.1 DESIGN BASES 5.2-1 5.2.1.1 General Design Criterion 2, 1967 - Performance Standards 5.2-1 5.2.1.2 General Design Criterion 4, 1987 - Environmental and
Dynamic Effects Design Bases 5.2-1 5.2.1.3 General Design Criterion 9, 1967 - Reactor Coolant
Pressure Boundary 5.2-1 5.2.1.4 General Design Criterion 11, 1967 - Control Room 5.2-1 5.2.1.5 General Design Criterion 12, 1967 - Instrumentation
and Controls 5.2-1 5.2.1.6 General Design Criterion 16, 1967 - Monitoring Reactor
Coolant Pressure Boundary 5.2-2 5.2.1.7 General Design Criterion 33, 1967 - Reactor Coolant Pressure
Boundary Capability 5.2-2 5.2.1.8 General Design Criterion 34, 1967 - Reactor Coolant Pressure
Boundary Rapid Propagation Failure Prevention 5.2-2 5.2.1.9 General Design Criterion 35, 1967 - Reactor Coolant Pressure
Boundary Brittle Fracture Prevention 5.2-2 5.2.1.10 General Design Criterion 36, 1967 - Reactor Coolant Pressure
Boundary Surveillance 5.2-2 5.2.1.11 General Design Criterion 51, 1967 - Reactor Coolant Pressure
Boundary Outside Containment 5.2-3 5.2.1.12 Reactor Coolant Pressure Boundary Safety Function
Requirement 5.2-3 5.2.1.13 10 CFR 50.55a- Codes and Standards 5.2-3 5.2.1.14 10 CFR 50.55a(f) - Inservice Testing Requirements 5.2-3 5.2.1.15 10 CFR 50.55a(g) - Inservice Inspection Requirements 5.2-3 5.2.1.16 10 CFR 50.60 - Acceptance Criteria for Fracture Prevention
Measures for Lightwater Nuclear Power Reactors for Normal Operation 5.2-3 5.2.1.17 10 CFR 50.61- Fracture Toughness Requirements for
Protection against Thermal Shock Events 5.2-3 5.2.1.18 10 CFR Part 50 Appendix G- Fracture Toughness
Requirements 5.2-4 5.2.1.19 10 CFR Part 50 Appendix H- Reactor Vessel Material
Surveillance Program Requirements 5.2-4 5.2.1.20 Safety Guide 14, October 1971 - Reactor Coolant Pump
Flywheel Integrity 5.2-4 5.2.1.21 Regulatory Guide 1.14, Revision 1, August 1975 -
DCPP UNITS 1 &
2 FSAR UPDATE Chapter 5 CONTENTS (Continued)
Section Title Page vii Revision 23 December 2016 Reactor Coolant Pump Flywheel Integrity 5.2-4 5.2.1.22 Regulatory Guide 1.44, May 1973 - Control of the Use
of Sensitized Stainless Steel 5.2-4 5.2.1.23 Regulatory Guide 1.45, May 1973 - Reactor Coolant Pressure
Boundary Leakage Detection Systems 5.2-5 5.2.1.24 Regulatory Guide 1.97, Revision 3, May 1983 - Criteria for
Accident Monitoring Instrumentation for Nuclear Power Plants 5.2-5 5.2.1.25 Regulatory Guide 1.99, Revision 2, May 1988 - Radiation
Embrittlement of Reactor Vessel Materials 5.2-5 5.2.1.26 NUREG-0737 (Items II.B.1, II.D.
1, II.D.3, II.K.2.13, and
III.D.1.1), November 1980 - Clarification of TMI Action Plan Requirements 5.2-5 5.2.1.27 Generic Letter 1989-10, June 1989 - Safety-Related Motor-
Operated Valve Testing and Surveillance 5.2-6 5.2.1.28 Generic Letter 1990-06, June 1990 - Enclosure B, Resolution
of Generic Issue 94 - Additional Low-Temperature Overpressure Protection For Light-Water Reactors 5.2-6 5.2.2 SYSTEM DESCRIPTION 5.2-6 5.2.2.1 Design of Reactor Coolant Pressure Boundary Components 5.2-6 5.2.2.2 Overpressurization Protection 5.2-37 5.2.2.3 General Material Considerations 5.2-40 5.2.2.4 Fracture Toughness 5.2-43 5.2.2.5 Austenitic Stainless Steel 5.2-58 5.2.3 SAFETY EVALUATION 5.2-63 5.2.3.1 General Design Criterion 2, 1967 - Performance Standards 5.2-63 5.2.3.2 General Design Criterion 4, 1987 - Environmental and
Dynamic Effects Design Bases 5.2-63 5.2.3.3 General Design Criterion 9, 1967 - Reactor Coolant Pressure
Boundary 5.2-64 5.2.3.4 General Design Criterion 11, 1967 - Control Room 5.2-64 5.2.3.5 General Design Criterion 12, 1967 - Instrumentation and
Controls 5.2-65 5.2.3.6 General Design Criterion 16, 1967 - Monitoring Reactor
Coolant Pressure Boundary 5.2-65 5.2.3.7 General Design Criterion 33, 1967 - Reactor Coolant
Pressure Boundary Capability 5.2-66 5.2.3.8 General Design Criterion 34, 1967 - Reactor Coolant
Pressure Boundary Rapid Propagation Failure Prevention 5.2-66 5.2.3.9 General Design Criterion 35, 1967 - Reactor Coolant DCPP UNITS 1 &
2 FSAR UPDATE Chapter 5 CONTENTS (Continued)
Section Title Page viii Revision 23 December 2016 Pressure Boundary Brittle Fracture Prevention 5.2-67 5.2.3.10 General Design Criterion 36, 1967 - Reactor Coolant
Pressure Boundary Surveillance 5.2-68 5.2.3.11 General Design Criterion 51, 1967 - Reactor Coolant
Pressure Boundary Outside Containment 5.2-68 5.2.3.12 Reactor Coolant Pressure Boundary Safety Function
Requirement 5.2-69 5.2.3.13 10 CFR 50.55a- Codes and Standards 5.2-69 5.2.3.14 10 CFR 50.55a(f) - Inservice Testing Requirements 5.2-69 5.2.3.15 10 CFR 50.55a(g) - Inservice Inspection Requirements 5.2-69 5.2.3.16 10 CFR 50.60 - Acceptance Criteria for Fracture Prevention
Measures for Lightwater Nuclear Power Reactors for Normal Operation 5.2-70 5.2.3.17 10 CFR 50.61 - Fracture Toughness Requirements for
Protection against Pressurized Thermal Shock Events 5.2-71 5.2.3.18 10 CFR Part 50 Appendix G - Fracture Toughness
Requirements 5.2-71 5.2.3.19 10 CFR Part 50 Appendix H - Reactor Vessel Material Surveillance Program Requirements 5.2-71 5.2.3.20 Safety Guide 14, October 1971 - Reactor Coolant Pump Flywheel Integrity 5.2-71 5.2.3.21 Regulatory Guide 1.14, Revision 1, August 1975 -
Reactor Coolant Pump Flywheel Integrity 5.2-73 5.2.3.22 Regulatory Guide 1.44, May 1973 - Control of the Use
of Sensitized Stainless Steel 5.2-74 5.2.3.23 Regulatory Guide 1.45, May 1973 - Reactor Coolant
Pressure Boundary Leakage Detection Systems 5.2-74 5.2.3.24 Regulatory Guide 1.97, Revision 3, May 1983 - Criteria for
Accident Monitoring Instrumentation for Nuclear Power Plants 5.2-82 5.2.3.25 Regulatory Guide 1.99, Revision 2, May 1988 - Radiation
Embrittlement of Reactor Vessel Materials 5.2-82 5.2.3.26 NUREG-0737 (Items II.B.1, II.D.
1, II.D.3, II.K.2.13, and
III.D.1.1), November 1980 - Clarification of TMI Action Plan Requirements 5.2-83 5.2.3.27 Generic Letter 1989-10, June 1989 - Safety-Related
Motor-Operated Valve Testing and Surveillance 5.2-84 5.2.3.28 Generic Letter 1990-06, June 1990 - Enclosure B, Resolution
of Generic Issue 94 - Additional Low-Temperature Overpressure Protection For Light-Water Reactors 5.2-84
DCPP UNITS 1 &
2 FSAR UPDATE Chapter 5 CONTENTS (Continued)
Section Title Page ix Revision 23 December 2016 5.
2.4 REFERENCES
5.2-85 5.2.5 REFERENCE DRAWINGS 5.2-88 5.3 THERMAL HYDRAULIC SYSTEM DESIGN 5.3-1 5.3.1 ANALYTICAL METHODS AN D DATA 5.3-1 5.3.2 OPERATING RESTRICTIONS ON REACTOR COOLANT PUMPS 5.3-1
5.3.3 TEMPERATURE-POWER OPERATING MAP 5.3-1 5.3.4 LOAD FOLLOWING CHARACTERISTICS 5.3-1 5.3.5 TRANSIENT EFFECTS 5.3-2 5.3.6 THERMAL AND HYDRAULIC C HARACTERISTICS
SUMMARY
TABLE 5.3-2 5.4 REACTOR PRESSURE VESSEL AND APPURTENANCES 5.4-1
5.4.1 REACTOR PRESSURE VESSEL DESCRIPTION 5.4-1 5.4.1.1 Design Bases 5.4-1 5.4.1.2 Design Transients 5.4-1 5.4.1.3 Codes and Standards 5.4-1 5.4.1.4 Reactor Pressure Vessel Description 5.4-2 5.4.1.5 Inspection Provisions 5.4-3 5.4.2 FEATURES FOR IMPROVED RELIABILITY 5.4-4 5.4.3 PROTECTION OF CLOSURE STUDS 5.4-4 5.4.4 MATERIALS AND INSPECTIONS 5.4-4 5.4.5 SPECIAL PROCESSES FOR FABRICATION AND INSPECTION 5.4-5 5.4.5.1 Fabrication Processes 5.4-5 5.4.5.2 Tests and Inspections 5.4-5 5.4.6 QUALITY ASSURANCE SURVEILLANCE 5.4-6 5.4.7 REACTOR PRESSURE VESSEL DESIGN DATA 5.4-7 DCPP UNITS 1 &
2 FSAR UPDATE Chapter 5 CONTENTS (Continued)
Section Title Page x Revision 23 December 2016
5.4.8 REACTOR PRESSURE VESSEL EVALUATION 5.4-7
5.5 COMPONENT AND SUBSYSTEM DESIGN 5.5-1
5.5.1 REACTOR COOLANT PUMPS 5.5-1 5.5.1.1 Design Bases 5.5-1 5.5.1.2 Design Description 5.5-1 5.5.1.3 Design Evaluation 5.5-3 5.5.1.4 Tests and Inspections 5.5-8
5.5.2 STEAM GENERATORS 5.5-10 5.5.2.1 Design Bases 5.5-10 5.5.2.2 Design Description 5.5-11 5.5.2.3 Design Evaluation 5.5-12 5.5.2.4 Tests and Inspections 5.5-15 5.5.2.5 Steam Generator Tube Surveillance Program 5.4-16
5.5.3 REACTOR COOLANT PIPING 5.5-17 5.5.3.1 Design Bases 5.5-17 5.5.3.2 Design Description 5.5-17 5.5.3.3 Design Evaluation 5.5-20 5.5.3.4 Tests and Inspections 5.5-20
5.5.4 MAIN STEAM LINE FLOW RESTRICTORS 5.5-21
5.5.5 MAIN STEAM LINE ISOLATION SYSTEM 5.5-21
5.5.6 RESIDUAL HEAT REMOVAL SYSTEM 5.5-21 5.5.6.1 Design Bases 5.5-22 5.5.6.2 System Description 5.5-26 5.5.6.3 Design Evaluation 5.5-31 5.5.6.4 Safety Evaluation 5.5-32 5.5.6.5 Tests and Inspections 5.5-39 5.5.6.6 Instrumentation Applications 5.5-39 5.5.7 REACTOR COOLANT CLEAN UP SYSTEM 5.5-39 5.5.8 MAIN STEAM LINE AND FEEDWATER PIPING 5.5-39 5.5.9 PRESSURIZER 5.5-39 DCPP UNITS 1 &
2 FSAR UPDATE Chapter 5 CONTENTS (Continued)
Section Title Page xi Revision 23 December 2016 5.5.9.1 Design Bases 5.5-40 5.5.9.2 Design Description 5.5-41 5.5.9.3 Design Evaluation 5.5-42 5.5.9.4 Tests and Inspections 5.5-45 5.5.10 PRESSURIZER RELIEF TANK 5.5-45 5.5.10.1 Design Bases 5.5-45 5.5.10.2 Design Description 5.5-45 5.5.10.3 Design Evaluation 5.5-46
5.5.11 VALVES 5.5-46 5.5.11.1 Design Bases 5.5-46 5.5.11.2 Design Description 5.5-47 5.5.11.3 Design Evaluation 5.5-48 5.5.11.4 Tests and Inspections 5.5-48 5.5.12 SAFETY AND RELIE F VALVES 5.5-48 5.5.12.1 Design Bases 5.5-48 5.5.12.2 Design Description 5.5-48 5.5.12.3 Design Evaluation 5.5-49 5.5.12.4 Tests and Inspections 5.5-49 5.5.13 COMPONENT SUPPORTS 5.5-50 5.5.13.1 Design Bases 5.5-50 5.5.13.2 Design Description 5.5-50 5.5.13.3 Design Evaluation 5.5-52 5.5.14 REACTOR VESSEL HEA D VENT SYSTEM 5.5-53 5.5.14.1 Design Bases 5.5-53 5.5.14.2 Design Description 5.5-53 5.5.14.3 Supports 5.5-54 5.5.15 REFERENCES 5.5-54 5.5.16 REFERENCE DRAWINGS 5.5-55 5.6 INSTRUMENTATION REQUIREMENTS 5.6-1 5.6.1 REACTOR COOLANT SYSTEM 5.6-1 5.6.1.1 Inadequate Core Cooling Instrumentation 5.6-2 DCPP UNITS 1 &
2 FSAR UPDATE Chapter 5 CONTENTS (Continued)
Section Title Page xii Revision 23 December 2016 5.6.1.2 Loose Parts Monitoring 5.6-3 5.6.2 RESIDUAL HEAT REMOVAL SYSTEM 5.6-3
5.
6.3 REFERENCES
5.6-4
DCPP UNITS 1 &
2 FSAR UPDATE Chapter 5 TABLES Table Title xiii Revision 23 December 2016 5.0-1 Applicable Design Basis Criteria
5.1-1 System Design and Operating Parameters
5.2-1 ASME Code Cases for Westinghouse PWR Class A Components (Historical)
5.2-2 Equipment Code and Classification List
5.2-3 Procurement Information Components Within Reactor Coolant System Boundary 5.2-4 Summary of Reactor Coolant System Design Transients
5.2-5 Stress Limits for PG&E Quality/Code Class I Loop Piping and Valves
5.2-6 Load Combinations and Stress Criteria for Primary Equipment
5.2-6a Load Combinations and Acceptance Criteria for Replacement Primary Equipment 5.2-7 Faulted Condition Stress Limits for PG&E Quality/Code Class I Components
5.2-8 Loading Combinations and Accept ance Criteria for Primary Equipment Supports 5.2-8a Load Combinations and Accepta nce Criteria for Integrated Head Assembly (IHA)
5.2-9 Active and Inactive Valves in the Reactor Coolant Pressure Boundary
5.2-10 Reactor Coolant System Nominal Pressure Setpoints (psig)
5.2-11 Reactor Vessel Materials
5.2-12 Pressurizer, Pressurizer Relief Tanks, and Surge Line Materials
5.2-13 Reactor Coolant Pump Materials
5.2-14 Steam Generator Materials DCPP UNITS 1 &
2 FSAR UPDATE Chapter 5 TABLES Table Title xiv Revision 23 December 2016
5.2-15 Reactor Coolant Water Chemistry Specification
5.2-16 Reactor Coolant Boundary Leakage Detection Systems
5.2-17 Deleted in Revision 2
5.2-17A DCPP Unit 1 Reactor Vessel Toughness Data
5.2-17B DCPP Unit 2 Reactor Vessel Toughness Data
5.2-18 Deleted in Revision 2
5.2-18A Identification of Unit 1 Reactor Vessel Beltl ine Region Base Material
5.2-18B Identification of Unit 2 Reactor Vessel Beltl ine Region Base Material
5.2-19 Deleted in Revision 2 5.2-19A Fracture Toughness Properties of Unit 1 Reactor Vessel Beltline Region Base Material
5.2-19B Fracture Toughness Properties of Unit 2 Reactor Vessel Beltline Region Base Material
5.2-20 Deleted in Revision 2
5.2-20A Identification of Unit 1 Reactor Vessel Beltline Region Weld Metal
5.2-20B Identification of Unit 2 Reactor Vessel Beltline Region Weld Metal
5.2-21 Deleted in Revision 2
5.2-21A Fracture Toughness Properties of Unit 1 Reactor Vessel Beltline Region Weld Metal
5.2-21B Fracture Toughness Properties of Unit 2 Reactor Vessel Beltline Region Weld Metal
5.2-22 Reactor Vessel Material Surveillance Program Withdrawal Schedule DCPP UNITS 1 &
2 FSAR UPDATE Chapter 5 TABLES Table Title xv Revision 23 December 2016
5.2-23 Reactor Coolant System Pressure Boundary Isolation Valves
5.4-1 Reactor Vessel Design Parameters (Both Units)
5.4-2 Reactor Vessel Construction Quality Assurance Program (Historical)
5.5-1 Reactor Coolant Pump Design Parameters (Both Units)
5.5-2 Reactor Coolant Pump Quality Assurance Program (Historical)
5.5-3 Steam Generator Design Data
5.5-4 Deleted in Revision 4
5.5-5 Steam Generator Quality Assurance Program (Both Units) (Historical)
5.5-6 Reactor Coolant Piping Design Parameters (Both Units)
5.5-7 Reactor Coolant Piping Quality Assurance Program (Both Units) (Historical)
5.5-8 Design Bases for Residual Heat Removal System Operation (Both Units)
5.5-9 Residual Heat Removal System Codes and Classifications (Both Unit 1 and Unit 2)
5.5-10 Residual Heat Removal System Component Data (Both Units)
5.5-11 Recirculation Loop Leakage
5.5-12 Pressurizer Design Data
5.5-13 Pressurizer Quality Assurance Program (Both Units) (Historical)
5.5-14 Pressurizer Relief Tank Design Data
5.5-15 Reactor Coolant System Boundary Valve Design Parameters DCPP UNITS 1 &
2 FSAR UPDATE Chapter 5 TABLES Table Title xvi Revision 23 December 2016
5.5-16 Pressurizer Valves Design Parameters
5.5-17 Reactor Vessel Head Vent System Equipment Design Parameters
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 5 FIGURES Figure Title xvii Revision 23 December 2016 5.1-1 Deleted in Revision 1
5.1-2 Pump Head-Flow Characteristics
5.1-2A Safety Analysis - RCS Flow Parameters
5.2-1 Identification and Location of Beltline Region Material for the Reactor Vessel (Unit 1)
5.2-2 Reactor Coolant Loop Model FOR STATIC AND LOCA
5.2-2A Reactor Coolant 4-Loop Model
5.2-3 THRUST RCL Model Showing Hydraulic Force Location
5.2-3A Deleted in Revision 19
5.2-4 Identification and Location of Beltline Region Material for the Reactor Vessel (Unit 2)
5.2-5 Deleted in Revision 2
5.2-6 Deleted in Revision 9
5.2-7 Lower Bound Fracture Toughness A533, Grade B, Class 1
5.2-8 Transition Temperature Correlation Between K ld (Dynamic) and C v for a Series of Unirradiated Steels
5.2-9 Containment Monitor Response Time Versus Primary Leak Rate
5.2-10 Air Ejector Radiogas Monitor Response Time Versus Primary Leak Rate 5.2-11 Blowdown Liquid Monitor Response Time Versus Primary Leak Rate
5.2-12 Containment Cooling Water Liquid Monitor Response Time Versus Primary Leak Rate
5.2-13 Containment Area Monitor Response Time Versus Primary Leak Rate DCPP UNITS 1 & 2 FSAR UPDATE Chapter 5 FIGURES Figure Title xviii Revision 23 December 2016 5.2-14 Containment Radiogas Monitor Count Rate Versus Primary Leak Rate after Equilibrium
5.2-15 Containment Particulate Monitor Count Rate Versus Primary Leak Rate after Equilibrium
5.2-16 Surveillance Capsule Elevation View (Unit 1)
5.2-17 Surveillance Capsule Plan View (Unit 1)
5.2-18 Surveillance Capsule Elevation View (Unit 2)
5.2-19 Surveillance Capsule Plan View (Unit 2)
5.3-1 Hot Leg, Cold Leg, and Average Reactor Coolant Loop Temperature as a Function of Percent Full Power
5.4-1 Reactor Vessel (Unit 1)
5.4-2 Reactor Vessel (Unit 2)
5.4-3 Integrated Head Assembly Seis mic Support Structure Assembly
5.5-1 Reactor Coolant Controlled Leakage Pump
5.5-2 Reactor Coolant Pump Estimated Performance Characteristics
5.5-3 Reactor Coolant Pump Spool Piece and Motor Support Stand
5.5-4 Westinghouse Delta 54 Steam Generator
5.5-4A Deleted in Revision 19
5.5-5 Deleted in Revision 19
5.5-6 Deleted in Revision 19
5.5-7 Deleted in Revision 1
5.5-8 Pressurizer
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 5 FIGURES Figure Title xix Revision 23 December 2016 5.5-9 Reactor Support 5.5-10 Steam Generator and Reactor Coolant Pump Supports
5.5-11 Component Supports 5.5-12 Pressurizer Support 5.5-13 (a) U1: Function Diagram, Reactor-Turbine Generator Protection
5.5-14 Schematic Flow Diagram of the Reactor Vessel Head Vent System
5.5-14A Deleted in Revision 20
5.5-15 Deleted in Revision 20
5.5-16 Deleted in Revision 8
5.5-17 (a) U2: Function Diagram, Reactor-Turbine Generator Protection
5.5-18 Seven Nozzle RSG Outlet Flow Restrictor
NOTE:
(a) This figure corresponds to a controlled engineering drawing that is incorporated by reference into the FSAR Update. See Table 1.6-1 for the correlation between the
FSAR Update figure number and the corresp onding controlled engineering drawing number.
DCPP UNITS 1 &
2 FSAR UPDATE xx Revision 23 December 2016 Chapter 5 APPENDICES Title Title___
_______________
5.5A Deleted in Revision 22 DCPP UNITS 1 &
2 FSAR UPDATE 5.1-1 Revision 23 December 2016 Chapter 5 REACTOR COOLANT SYSTEM 5.1
SUMMARY
DESCRIPTION The reactor coolant system (RCS) consists of four similar heat transfer loops connected
in parallel to the reactor pressure vessel (RPV), which are located inside the containment. Each loop contains a reactor coolant pump (RCP), steam generator (SG), and associated piping and valves. The system also includes a pressurizer, a
pressurizer relief tank (PRT), interconnecting piping, and instrumentation necessary for operation.
During operation, the RCS, using coolant flow provided by the RCPs, transfers heat generated in the core to the SGs where the steam that drives the turbine-generator is produced. Borated pressurized water circulates in the RCS at a flowrate and
temperature consistent with the reactor core thermal-hydraulic performance
requirements. The water also acts as a neutron moderator and reflector, and as a solvent for the boric acid neutron absorber used as chemical shim control.
The reactor coolant pressure boundary (RCPB) provides a barrier against the release of radioactivity generated within the reactor, and is designed to ensure a high degree of
integrity throughout plant life.
RCS pressure is controlled by the pressurizer in which water and steam are maintained
in equilibrium by electrical heaters or water sprays. Steam can be formed (by the heaters) or condensed (by the pressurizer spray) to minimize reactor coolant pressure
variations. Spring-loaded pressurizer safety valves (PSVs) and power-operated relief valves (PORVs) are mounted on the pressurizer, and discharge to the PRT where the steam is condensed and cooled by mixin g with water. Noncondensable gases (primarily) or steam can be removed from th e reactor vessel closure head (RVCH) by the reactor vessel head vent system (RVHVS).
The chemical and volume control system (CVCS) is designed to avoid uncontrolled reductions in boric acid concentration or reactor coolant temperature. The reactor coolant is the core moderator, reflector, and solvent for the chemical shim. As a result, changes in coolant temperature or boric acid concentration affect the reactivity level in the core.
Whenever the RCS boron concentration is varied, good mixing is provided to ensure uniform boron concentration throughout the RCS. Coolant flow is provided by either an RCP or a residual heat removal (RHR) pump to ensure uniform mixing whenever the boron concentration is varied. Although pressurizer mixing is not achieved to the same degree, the fraction of the total RCS volume, which is in the pressurizer is small.
Pressurizer spray provides homogenization of boron concentration.
DCPP UNITS 1 &
2 FSAR UPDATE 5.1-2 Revision 23 December 2016 Also, the distribution of flow around the system is not subject to the degree of variation that would be required to produce non-homogeneities in coolant temperature or boron concentration as a result of areas of low coolant flowrate.
The RCS design arrangement eliminates dead-ended sections and other areas of low coolant flow in which non-homogeneities in coolant temperature or boron concentration could develop.
The RCS is designed to operate within the coolant temperature change limitations.
Refer to Tables 5.5-1, 5.5-3, 5.5-6, 5.5-12, and 5.5-14 through 5.5-17 for system design pressures and temperatures.
The design basis and safety evaluation of the RCS and its associated structures, systems, and components (SSCs), with the exception of the RCPB are discussed in this section. The RCPB is discussed in Sectio n 5.2. The following are interfacing functions and SSCs that are discussed in the identified section:
(1) Reactor and reactor core design, nuclear design, and thermal-hydraulic design are discussed in Chapter 4; (2) The RHR system is discussed in Section 5.5.6.
(3) Emergency core cooling is discussed in Section 6.3; (4) RCS instrumentation associated with the reactor trip system (RTS) are discussed in this section, in conjunction with Section 7.2, and other RCS instrumentation, including the reactor vessel level instrumentation system (RVLIS), are discussed in this section, in conjunction with Sections 7.3, 7.4 and 7.5; (5) RCP seal cooling is provided by the component cooling water (CCW) system and seal injection is provided by the CVCS (refer to Sections 9.2.2 and 9.3.4, respectively);
(6) RCS inventory and volume control, in conjunction with the CVCS, is discussed in this section and, with respect to CVCS, in Section 9.3.4; (7) RCS coolant as a solvent, in conjunction with CVCS, functions as a neutron moderator and reflector and is discussed in this section. Refer to Section 9.3.4.2.8.2.4 for discussion on chemical shim and reactor coolant makeup; (8) Sampling of the RCS using the nuclear steam supply system (NSSS) sampling system is described in Section 9.3.2.1.
DCPP UNITS 1 &
2 FSAR UPDATE 5.1-3 Revision 23 December 2016 (9) Main steam flow restriction, in conjunction with the main steam system, is discussed in this section with respect to the integral flow restrictor in the SGs (refer to Sections 5.5, 10.3, and 15.4 for further discussion of the main steam flow restrictors);
(10) Main steam isolation is discussed in Section 10.3; (11) Decay heat removal, in conjunction with Sections 10.3 and 10.4.8, is discussed in this section; (12) The main feedwater system is discussed in Section 10.4.7, with exception of the main feedwater ring in the SGs, which is covered in this section.
5.1.1 DESIGN BASES 5.1.1.1 General Design Criterion 2, 1967 - Performance Standards The PG&E Design Class I portion of the RCS is designed to withstand the effects of, or is protected against, natural phenomena, such as earthquakes, tornadoes, flooding, winds, tsunamis, and other local site effects.
5.1.1.2 General Design Criterion 3, 1971 - Fire Protection The RCS is designed and located to minimize, consistent with other safety requirements, the probability and effects of fires and explosions.
5.1.1.3 General Design Criterion 4, 1967 - Sharing of Systems The RCS and components are not shared by the Diablo Canyon Power Plant (DCPP) units unless it is shown safety is not impaired by the sharing.
5.1.1.4 General Design Criterion 4, 1987 - Environmental and Dynamic Effects Design Bases Consideration of the dynamic effects associated with main reactor coolant loop (RCL) piping postulated pipe ruptures are excluded from the DCPP design basis with the approval of leak-before-break (LBB) methodology by demonstrating that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.
5.1.1.5 General Design Criterion 6, 1967 - Reactor Core Design The RCS is designed to provide decay heat removal so that fuel damage limits are not exceeded under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the DCPP UNITS 1 &
2 FSAR UPDATE 5.1-4 Revision 23 December 2016 effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all offsite power.
5.1.1.6 General Design Criterion 9, 1967 - Reactor Coolant Pressure Boundary The RCS design includes provisions for the control of RCS chemistry such that the materials of construction of the pressure-retaining boundary of the RCS are protected from corrosion that might otherwise reduce t he system structural integrity during its service lifetime.
5.1.1.7 General Design Criterion 11, 1967 - Control Room The RCS is designed to or contains instrumentation and controls that support actions to maintain and control the safe operational status of the plant from the control room or from an alternate location if control room access is lost due to fire or other causes.
5.1.1.8 General Design Criterion 12, 1 967 - Instrumentation and Controls Instrumentation and controls are provided, as required, to monitor and maintain RCS variables within prescribed operating ranges.
5.1.1.9 General Design Criterion 13, 1967 - Fission Process Monitors and Controls The RCS design includes means for monitoring and maintaining control over the fission process throughout core life and for all conditions that can reasonably be anticipated to cause variations in reactivity of the core, such as indication of position of control rods and concentration of soluble reactivity control poisons.
5.1.1.10 General Design Criterion 15, 1967 - Engineered Safety Features Protection Systems The RCS is provided with instrumentation for sensing accident situations and initiating the operation of necessary engineered safety features (ESFs).
5.1.1.11 General Design Criterion 21, 1967 - Single Failure Definition Portions of the RCS are designed to perform their function after sustaining a single failure. Multiple failures resulting from a single event are treated as a single failure.
5.1.1.12 General Design Criterion 40, 1967 - Missile Protection The RCS is designed to be protected against dynamic effects and missiles that might result from plant equipment failures.
DCPP UNITS 1 &
2 FSAR UPDATE 5.1-5 Revision 23 December 2016 5.1.1.13 General Design Criterion 49, 1967 - Containment Design Basis The RCS is designed so that the containment structure can accommodate, without exceeding the design leakage rate, the pressures and temperatures resulting from the largest credible energy release following a loss-of-coolant accident (LOCA), including a considerable margin for effects from metal-water or other chemical reactions that could occur as a consequence of failure of emergency core cooling systems (ECCSs).
5.1.1.14 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment The RCS piping that penetrates containment is provided with leak detection, isolation, redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating this system. The piping is designed with a capability to test periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits.
5.1.1.15 General Design Criterion 55, 1971 - Reactor Coolant Pressure Boundary Penetrating Containment Each RCS line that penetrates the containment is provided with containment isolation valves (CIVs).
5.1.1.16 General Design Criterion 56, 1971 - Primary Containment Isolation The RCS contains piping that penetrates containment and that is connected directly to the containment atmosphere. Normally closed isolation valves are provided outside containment and automatic (check) valves are provided inside containment to ensure containment integrity is maintained.
5.1.1.17 Reactor Coolant System Safety Function Requirements (1) Protection from Missiles and Dynamic Effects PG&E Design Class I RCS SSCs are designed to be protected against the effects of missiles and dynamic effects which may result from plant equipment failure.
(2) Reactor Heat Removal The RCS provides sufficient heat transfer capability to transfer the heat produced during power operation, plant cooldown, cold shutdown, operational transients, and accidents.
(3) RCS Thermal-Hydraulic Requirements The RCS thermal-hydraulic design provides appropriate limits on RCS pressure and ensures adequate RCP net positive suction head (NPSH).
DCPP UNITS 1 &
2 FSAR UPDATE 5.1-6 Revision 23 December 2016 (4) RCS Coolant Functional Properties The RCS contains the water used as a core neutron moderator and reflector and as a solvent for chemical shim control.
(5) RCS Pressure and Volume Control The pressurizer maintains system pressure and volume and limits pressure transients using the surge line, pressurizer (via free volume), heaters, spray, and the PORVs.
(6) Steam Flow Restriction The RCS is designed with flow restrictors that limit the steam flow in the event of a main steam line break (MSLB) at any location on the main steam line.
(7) RCP Coastdown The RCP is designed to mitigate a loss of RCS flow by coasting down upon a loss of motive power.
(8) Pressurizer Relief Tank The PRT is designed to prevent collapse under a full vacuum.
5.1.1.18 10 CFR 50.49 - Environmental Qualif ication of Electrical Equipment Important to Safety for Nuclear Power Plants PG&E Design Class I RCS components that require environmental qualification (EQ) are qualified to the requirements of 10 CFR 50.49.
5.1.1.19 10 CFR 50.55a(f) - Inservice Testing Requirements RCS American Society of Mechanical Engineers (ASME) Code components are tested to the requirements of 10 CFR 50.55a(f)(4) and 10 CFR 50.55a(f)(5) to the extent practical.
5.1.1.20 10 CFR 50.55a(g) - Inservice Inspection Requirements RCS ASME Code components are inspected to the requirements of 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(g)(5) to the extent practical.
5.1.1.21 10 CFR 50.63 - Loss of All Alternating Current Power The RCS is designed to provide: (1) cooling of the core by natural circulation of reactor coolant through the core and SGs; (2) RCS pressure control; and (3) system monitoring DCPP UNITS 1 &
2 FSAR UPDATE 5.1-7 Revision 23 December 2016 in the event of a station blackout (SBO), including RCS temperature, pressurizer pressure, pressurizer level, and source range monitors.
The RCPs are capable of withstanding an SBO event without a loss of seal integrity.
5.1.1.22 10 CFR Part 50 Appendix R (Sections III.G, III.J, III.L, and III.O) - Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979 Section III.G - Fire Protection of Safe Shutd own Capability: RCS SSCs are designed with fire protection features that are capable of limiting fire damage so that the RCS SSCs necessary to achieve and maintain hot shutdown conditions from either the control room or hot shutdown panel (HSP) are free of fire damage. Fire Protection of the RCS SSCs is provided by a combination of physical separation, fire-rated barriers, and/or automatic suppression and detection.
Section III.J - Emergency Lighting: Emergency lighting or battery operated lights (BOLs) are provided in areas where indication of RCS parameters may be required to safely shut down the unit following a fire.
Section III.L - Alternative and Dedicated Shut down Capability: Alternate shutdown capability is provided to achieve and maintain hot standby conditions, achieve cold shutdown conditions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and maintain cold shutdown conditions thereafter.
Section III.O - Oil Collection System for Reactor Coolant Pumps: The RCPs are equipped with an oil collection system that is designed, engineered , and installed such that failure will not lead to fire during normal or design basis accident conditions and that there is reasonable assurance that the system will withstand earthquakes.
5.1.1.23 Regulatory Guide 1.89, November 1974 - Environmental Qualification of Class 1E Equipment for Nuclear Power Plants The subcooled margin monitors (SCMMs) are designed to be environmentally qualified in accordance with the requirements of Regulatory Guide 1.89, November 1974 (refer to Section 3.11).
5.1.1.24 Regulatory Guide 1.97, Revision 3, May 1983 - Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident The RCS provides instrumentation to monitor system variables during and following an accident.
DCPP UNITS 1 &
2 FSAR UPDATE 5.1-8 Revision 23 December 2016 5.1.1.25 Regulatory Guide 1.121, August 1976 - Bases for Plugging Degraded PWR Steam Generator Tubes DCPP has established criteria defining the limiting safe conditions of tube degradation of SG tubing beyond which defective tubes, as established by inservice inspection (ISI), are removed from service by installing a tube plug at each end of the tube.
5.1.1.26 NUREG-0737 (Items II.B.1, II.D.1, II.E.3.1, II.F.2, II.G.1, II.K.3.5, and II.K.3.25), November 1980 - Clarificat ion of TMI Action Plan Requirements Item II.B.1 - Reactor Coolant System Vents: RVCH high point vents are capable of being remotely operated from the control room.
Item II.D.1 - Performance Testing of Boiling-Water Reactors and Pressurized-Water Reactor Relief and Safety Valves (originally Recommendation 2.1.2 of NUREG-0578, July 1979): The pressurizer PORVs and PSVs are capable of operating under expected operating conditions for design-basis transients and accidents.
Item II.E.3.1 - Emergency Power Supply for Pressurizer Heaters: All four pressurizer heater groups can be supplied with power from the offsite power system when offsite power is available. In addition, power can be provided to two of the four heater groups from the standby power system through the Class 1E buses when offsite power is not available. This arrangement is adequate for establishing and maintaining natural circulation during hot standby conditions. Redundancy is provided by supplying each of the two groups of heaters from a different Class 1E bus.
Item II.F.2 - Instrumentation for Detection of Inadequate Core Cooling: Instrumentation is provided to unambiguously detect inadequ ate core cooling. The instrumentation includes reactor water level indication and provides an advance warning of the approach to inadequate core cooling. The instrumentation covers the full range from normal operation to the complete uncovering of the core.
Item II.G.1 - Emergency Power for Pressurizer Equipment: PORVs, PORV block valves, and pressurizer level instruments are capable of being supplied from either the offsite power system or the onsite distribution system, when offsite power is unavailable, through PG&E Design Class I motive and control components.
Item II.K.3.5 - Automatic Trip of R eactor Coolant Pumps During Loss-Of-Coolant Accident: DCPP provides alternative means to automatically trip the RCPs in response to small-break LOCAs (SBLOCAs).
Item II.K.3.25 - Effect of Loss of Alternating-Current Power on Pump Seals: RCP pump seals are designed to withstand a complete loss of alternating-current power for at least two hours.
DCPP UNITS 1 &
2 FSAR UPDATE 5.1-9 Revision 23 December 2016 5.1.1.27 Generic Letter 83-37, November 1983 - NUREG-0737 Technical Specifications Item II.B.1 - Reactor Coolant System Vents: The RVHVS is designed with testing and surveillance provisions to ensure operability.
5.1.1.28 Generic Letter 88-05, March 1988 - Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants A comprehensive boric acid corrosion control program (BACCP) is established to address boric acid corrosion concerns associated with RCS leakage at less than Technical Specification limits.
5.1.1.29 Generic Letter 90-06, June 1990 - Resolution of Generic Issue 70, "Power-Operated Relief Valve and Block Valve Reliability" and Generic Issue 94, "Additional Low-Temperature Over Pressure Protection for Light-Water Reactors" Pursuant to 10 CFR 50.54(f)
The RCS PORVs and PORV block valves are included in the inservice testing (IST) program. 5.1.1.30 Generic Letter 95-07, August 1995 - Pressure Locking and Thermal Binding of Safety-Related Power-Operated Valves RCS PG&E Design Class I, power-operated gate valves meet the requirements of Generic Letter 95-07, August 1995.
5.1.1.31 NRC Bulletin 88-09, July 1988 - Thimble Tube Thinning in Westinghouse Reactors An inspection program is established to perform periodic, non-destructive examination of the incore neutron monitoring thimble tubes for the purposes of measuring and monitoring thimble tube wear.
5.1.1.32 NRC Bulletin 88-11, December 1988 - Pressurizer Surge Line Thermal Stratification DCPP implemented a program to confirm pressurizer surge line integrity with respect to thermal stratification and striping concerns.
5.1.1.33 Branch Technical Position ASB 10-2, March 1978 - Design Guidelines for Avoiding Water Hammers in Steam Generators The SGs are designed and demonstrated to reduce the possibility and/or consequences of feedwater hammer. The DCPP design prevents or delays water draining from the DCPP UNITS 1 &
2 FSAR UPDATE 5.1-10 Revision 23 December 2016 feedring following a drop in SG level and minimizes the volume of feedwater piping external to the SG which could pocket steam.
5.1.2 SCHEMATIC FLOW DIAGRAMS Figure 3.2-7 is a schematic flow diagram of the RCS. Principal pressures, temperatures, flowrates, and coolant volume under normal full power operating
conditions are listed in Table 5.1-1.
The RCPB (refer to Section 5.2) is defined as:
(1) The RPV, including the RVCH and control rod drive mechanism (CRDM) housings (2) The reactor coolant side of the SGs (3) RCP casings (4) A pressurizer attached to one of the RCLs (5) Pressurizer PSVs and PORVs (6) The interconnecting piping, valves, and fittings between the principal components listed above (7) The piping, fittings, and valves leading to connecting auxiliary or support systems up to and including the second isolation valve (from the
high-pressure side) on each line
Piping and components designated as part of the RCS but that do not contain reactor coolant at design temperature and pressure (such as the PRT and associated piping) are outside the bounds of the RCPB.
5.1.3 PIPING AND INSTRUMENTATION DIAGRAMS
RCS piping and instrumentation are shown schematically in Figure 3.2-7.
5.1.4 ELEVATION DRAWINGS Physical layout of the RCS is sho wn in the following figures:
Figures 1.2-4, 1.2-5, and 1.2-6 (plan views inside containment)
Figures 1.2-22, 1.2-24, and 1.2-28 (section views inside containment)
Figure 5.5-10 (SG and RCP supports)
Figure 5.5-11 (component supports)
Figure 5.5-12 (pressurizer support)
DCPP UNITS 1 &
2 FSAR UPDATE 5.1-11 Revision 23 December 2016 5.1.5 REACTOR COOLANT SYSTEM COMPONENTS The principal RCS components are described in Sections 5.2 and 5.5. Refer to Section 5.2.2.1.15.4 for a description of the RPV, Section 5.5.1 for a description of the RCPs, Section 5.5.2 for a description of the SGs, Section 5.5.3 for a description of the RCS piping, Section 5.5.9 for a description of the pressurizer, Section 5.5.10 for a description of the PRT, and Section 5.5.12 for a description of the RCS PSVs and PORVs.
5.1.6 REACTOR COOLANT SYSTEM PERFOR MANCE AND SAFETY FUNCTIONS
The RCS transfers heat from the reactor to the SGs under conditions of both forced and natural circulation flow. The heat transfer capability of the SGs is sufficient to transfer to the steam and power conversion system (SPCS) the heat generated during normal operation and the initial phase of plant cooldown under natural circulation conditions.
The RCS, in conjunction with the reactor control and protection systems, maintains the reactor coolant at conditions of temperature, pressure, and flow adequate to protect the core from damage. The safety design requirements are to prevent conditions of high power, high reactor coolant temperature, or low reactor coolant pressure, or buildup of noncondensable gases which could interfere with core cooling, or combinations of these which could result in a departure from nucleate boiling ratio (DNBR) smaller than the applicable limit value (refer to Sections 4.4.3.3 and 4.4.4.1).
Design and performance characteristics of the RCS are provided in Table 5.1-1 and
Figures 5.1-2 and 5.1-2A.
5.1.6.1 Reactor Coolant System Flow Determination and Safety Analyses Reactor coolant flow is an important parameter in most of the non-LOCA safety
analyses. Figure 5.1-2 provides a representation of how the Thermal Design Flow (TDF) and Mechanical Design Flow (MDF) w ere established for the DCPP original design. These values were generated based on the best estimate flow expected after start-up. The TDF, a conservatively low flow, and the MDF, a conservatively high flow, are used in various safety analyses, depending on whether low flow or high flow is
conservative for each particular analysis. Figure 5.1-2A provides a representation of
the relationship between the best estimate flow, the MDF, the minimum measured flow (MMF), and the TDF. The values of these parameters are presented in Table 5.1-1.
The total RCS flow assumed in the safety analyses depends on the methodology for
each specific analysis. For departure from nucleate boiling (DNB) analyses that employ the Improved Thermal Design Procedure (ITDP), the MMF value is assumed directly in
the analysis. In the ITDP, a random flow uncertainty of 2.4% of flow is accounted for in a statistical square root of the sum of the squares (SRSS) combination with other appropriate plant input parameter uncertainties to set the DNBR limit. For non-DNB DCPP UNITS 1 &
2 FSAR UPDATE 5.1-12 Revision 23 December 2016 related events or DNB events for which the ITDP is not employed, the TDF value is used.
RCS flow is measured in accordance with the surveillance frequency control program, and is compared directly to the Unit 1 and Unit 2 Technical Specification flow limits in Limiting Condition of Operation (LCO) 3.4.1c and Surveillance Requirements 3.4.1.3 and 3.4.1.4. The RCS minimum flow limits provided by the LCO and surveillance requirements include both the TDF values for each unit explicitly a pproved by the U.S.
Nuclear Regulatory Commission (NRC) and the MMF values provided in the Unit 1 and Unit 2 cycle specific Core Operating Limits Report to ensure continued plant operation consistent with the safety analyses.
5.1.6.2 Reactor Coolant Flow The reactor coolant flow, a major parameter in the design of the system and its
components, is established using a detailed design procedure supported by operating plant performance data, by pump model tests and analysis, and by pressure drop tests
and analyses of the RPV and fuel assemblies.
Evaluation of the RCS flow involves a number of parameters. RCS best estimate flow, TDF, MDF, MMF, and Minimum Required Total RCS Flow Rate are parameters established during original des ign and are evaluated in the safety analyses of record.
Figures 5.1-2 and 5.1-2A provide a representation of these RCS flow parameters relative to original design considerations and the current safety analyses of record.
RCS flow is measured using the cold leg elbow differential pressure taps. The cold leg elbow tap flow methodology was established using Reference 2 and Reference 3.
5.1.6.3 Best Estimate Flow The best estimate flow is the most likely value for the actual plant operating condition.
The best estimate flow is used in developing the TDF and MDF. This flow is calculated based on the best estimate of the RPV, SG and piping flow hydraulic resistance, and on the best estimate of the RCP head-flow performance, with no uncertainties assigned to either the RCS component flow resistance or the pump head. The best estimate flow is
calculated based on hydraulic analyses.
The best estimate flow is also used to confirm the cold leg elbow tap flow measurement
while limiting the elbow flow tap measuremen t to a maximum value corresponding to the best estimate flow plus an allowance for the elbow tap flow repeatability uncertainty.
The hydraulic analysis uncertainty is 2%, while the instrument analysis repeatability
allowance is 0.4%, for a total uncertainty of 2.4%. Application of this acceptance criterion results in definition of a conservative current cycle flow, confirmed by both the
elbow tap flow measurements and the best estimate hydraulic analysis.
DCPP UNITS 1 &
2 FSAR UPDATE 5.1-13 Revision 23 December 2016 In the event that changes are made to the plant primary side hydraulic resistance or RCP characteristics, the best estimate flow must be recalculated.
Although the best estimate flow is the most likely value to be expected in operation, more conservative flowrates are applied in the thermal and mechanical designs, as discussed in Sections 5.1.6.4 and 5.1.6.5, below. The relationship between these parameters is reflected in Figures 5.1-2 and 5.1-2A.
5.1.6.4 Thermal Design Flow TDF is the basis for the reactor core thermal performance, the SG thermal performance, and the design plant parameters used throughout the design. To provide the required
margin in the safety analyses, the TDF accounts for the uncertainties in the RPV, SG, and piping flow resistances, RCP head, and the methods used to measure flowrate.
The combination of these uncertainties, which includes a conservative estimate of the pump discharge weir flow resistance, is equivalent to increasing the initial plant design
best estimate RCS flow resistance by approximately 19 percent. The intersection of this
conservative flow resistance with the initial plant design best estimate pump curve established the TDF. Figures 5.1-2 and 5.1-2A illustrate the relationship of TDF to other design and operating parameters. This procedure provides a flow margin for TDF of approximately 4 percent from the best estimate flow. The TDF is the initial flow assumed for non-DNB related accident and transient analyses and DNB analyses for which the ITDP is not used. The TDF for each unit is maintained during plant operation by satisfying the minimum RCS flow requirements of Technical Specification LCO 3.4.1c. Refer to Section 4.4.4.1 for a discussion of ITDP.
Data from all operating plants have indicated that the actual flow has been well above the flow specified for the thermal design of the plant. Tabulations of important design and performance characteristics of the RCS, as provided in Table 5.1-1, are based on
the TDF, as indicated.
5.1.6.5 Mechanical Design Flow MDF is the flow used in the mechanical design of the RPV internals and fuel assemblies. To ensure that a conservativel y high flow is specified, the MDF was based on a reduced system resistance (90 percent of initial plant design best estimate) and on
increased pump head capability (107 percent of initial plant design best estimate). The intersection of this flow resistance with the higher pump curve established the MDF.
Figures 5.1-2 and 5.1-2A illustrate the relationship of MDF to other design and operating parameters.
5.1.6.6 Minimum Measured Flow The plant MMF, the RCS minimum measured flow, is the flow used in reactor core DNB analyses for the ITDP. The MMF is defined as the TDF plus at least one flow measurement uncertainty. The MMF value included in the Unit 1 and Unit 2 cycle DCPP UNITS 1 &
2 FSAR UPDATE 5.1-14 Revision 23 December 2016 specific Core Operating Limits Report allows for a measurement uncertainty error of 2.4%. The MMF for each unit is also provide d in Tables 4.1-1 and 5.1-1. The MMF for each unit is maintained during plant operation by satisfying the minimum RCS flow requirements of Technical Specification LCO 3.4.1c.
5.1.6.7 Minimum Required Reactor Coolant System Flow Rate The minimum required RCS flow rate is the RCS total flow rate limit provided for each
unit in the Technical Specifications Limiting Condition of Operation (LCO) 3.4.1c and verified under Surveillance Requirements 3.4.1.3 and 3.4.1.4. These RCS total flow rate limits incorporate a measurement error of no more than 2.4%. The RCS flow rate
allowable values and nominal trip setpoints reflected in Technical Specification 3.3.1, Function 10, are based on a percentage of the loop flow measured every 24 months
under Technical Specification Sur veillance Requirement 3.4.1.4. This is determined using the cold leg elbow taps.
The RCS cold leg taps indicated total flow is continuously compared to the Reactor
Coolant Flow-Low nominal trip setpoint (refer to Section 7.2.2.1.4). The best estimate flow (refer to Section 5.1.6.2) may not be used as a substitute for the Technical Specification 3.4.1.4 Surveillance Requirement for flow measurement.
5.1.7 SYSTEM OPERATION
Brief descriptions of normal plant operations covering plant startup, power generation and hot standby, plant shutdown, refueling, and mid-loop operation are provided below.
5.1.7.1 Plant Startup
Plant startup encompasses the operations which bring the reactor plant from cold
shutdown to no-load power operating temperature and pressure.
Before plant startup, the RCLs and pressurizer are filled completely with reactor coolant to eliminate noncondensable gases.
If the vacuum refill method of filling the RCS is performed, the vacuum process will remove noncondensable gases and the pressurizer will not need to be filled completely. The water contains the correct concentration of
boron to maintain shutdown margin (SDM). The secon dary side of the SG is filled with water to normal startup level.
Coolant temperature variation during normal operation is limited and the associated reactivity change is well within the capability of the rod control group movement. For design evaluation, the RCS heatup and cooldown transients are analyzed using a rate of temperature change equal to 100°F per hour. Over certain temperature ranges, fracture prevention criteria will impose a lower limit to heatup and cooldown rates.
The RCS is then pressurized using the low-pressure control valve and either the
centrifugal charging pump (CCP3) (preferentially) or the centrifugal charging pumps DCPP UNITS 1 &
2 FSAR UPDATE 5.1-15 Revision 23 December 2016 (CCP1 and CCP2) to obtain the required pressure drop across the No. 1 seal of the RCPs. The pumps may then be operated intermittently in venting operations. As an alternative, a vacuum process can be used in filling the RCS. If this method is used, operating the RCPs intermittently to aid venting noncondensable gases may not be required.
During RCP operation, a charging pump and the low-pressure letdown path from the RHR system to the CVCS maintain the necessary RCS pressure. RCP operation is initiated after the required pressure differential across the No. 1 seal is achieved. The brittle fracture prevention temperature limitations of the RPV impose an upper pressure limit during low temperature operation. The charging pump supplies seal injection water
for the RCP shaft seals. A nitrogen atmosphere and normal operating temperature, pressure, and water level are established in the PRT.
After venting, the RCS is pressurized, all RCPs are started, and the pressurizer heaters are energized to begin heating the reactor coolant in the pressurizer, which leads to
formation of the steam bubble. If the vacuum refill method of filling the RCS is
performed, a pressurizer steam bubble may be formed prior to starting the RCPs. The pressurizer liquid level is reduced until the no-load power level volume is established.
During the initial heatup phase, hydrazine is added to the reactor coolant to scavenge the oxygen in the system; the heatup is not taken beyond 180
°F until the oxygen level has been reduced to the specified level.
As the reactor coolant temperature increases, the pressurizer heaters are manually
controlled to maintain adequate suction pressure for the RCPs.
5.1.7.2 Power Generation and Hot Standby Power generation includes steady state operation, ramp changes not exceeding the rate
of 5 percent of full power per minute, step changes of 10 percent of full power (not
exceeding full power), and step load decreas es with steam dump not exceeding 50 percent of full power.
During power generation, RCS pressure is maintained by the pressurizer controller at or
near 2235 psig, while the pressurizer liquid level is controlled by the charging-letdown
flow control of the CVCS.
When the reactor power level is less than 15 percent, the reactor power is controlled
manually. At powers above 15 percent, the operator may select the automatic mode of
operation. The rod motion is then controlled by the reactor control system that
automatically maintains an average coolant temperature, which follows a program
based on turbine load.
During hot standby operations, when the reactor is subcritical, the RCS temperature is
normally maintained by steam dump to the main condenser. This is accomplished by valves in the steam line, operating in the pressure control mode, which is set to maintain DCPP UNITS 1 &
2 FSAR UPDATE 5.1-16 Revision 23 December 2016 the SG steam pressure, or manually. Residual heat from the core and/or operation of an RCP provides heat to overcome RCS heat losses.
5.1.7.3 Plant Shutdown Before plant cooldown is initiated, the boron concentration in the RCS is increased to the value required for the corresponding target temperature. Subsequent reactor coolant samples are taken to verify that the RCS boron concentration is correct.
During plant cooldown, minimum SDM is maintained in accordance with requirements of the Technical Specifications. The temperature changes imposed on the RCS during its normal modes of operation do not cause any unacceptable reactivity changes.
Plant shutdown is the operation that brings the reactor plant from no-load power
operating temperature and pressure to cold shutdown. During plant cooldown from hot
standby to hot shutdown conditions, concentrated boric acid solution from the CVCS is added to the RCS to increase the reactor coolant boron concentration to that required
for cold shutdown. If the RCS is to be opened during the shutdown, the hydrogen and
fission gas in the reactor coolant is reduced by degassing the coolant in the volume
control tank (VCT).
Plant shutdown is attained in two phases: first, by the combined use of the RCS and
steam systems, and, second, by the RHR system.
During the first phase of shutdown, residual core and reactor coolant heat are transferred to the main steam system via the SG. Steam from the SG is dumped to the main condenser or to the atmosphere. At least one RCP is kept running to ensure uniform RCS cooldown. Pressurizer heaters and spray flow are manually controlled to cool the pressurizer while maintaining the
required RCP suction pressure. The plant does not permit the pressurizer to go water-solid without the RHR system and low temperature overpressure protection (LTOP) systems in service. As the pressurizer cools, the low-pressure control valve, pressurizer spray, pressurizer heaters, and the charging pumps maintain the required
RCS pressure.
When the reactor coolant temperature is below approximately 350
°F and the nominal pressure is less than or equal to 390 psig, the second phase of shutdown commences
with the operation of the RHR system.
During the second phase of plant cooldown and during cold shutdown and refueling, the heat exchangers of the RHR system are employed. Their capability is discussed in Section 5.5.
At least one RCP (either of those in a loop containing a pressurizer spray line) is kept running until the coolant temperature is reduced in accordance with plant procedures.
Pressurizer cooldown continues by initiating auxiliary spray flow from the CVCS if the RCPs are not available. Plant shutdown continues until the reactor coolant temperature is 140°F or less.
DCPP UNITS 1 &
2 FSAR UPDATE 5.1-17 Revision 23 December 2016 5.1.7.4 Refueling Before removing the RVCH for refueling, the system temperature is reduced to 160
°F or less, and hydrogen and fission product levels are reduced. Water level is monitored to indicate when the water level is below the top of the RVCH. Draining continues until the water level is below the RPV flange. The RVCH is then removed and the refueling cavity is flooded. Upon completion of refueling, the system is refilled for plant startup.
5.1.7.5 Mid-Loop Operation During refueling conditions, SG nozzle dams may be used in accordance with approved plant procedures to isolate the SG U-tubes and channel heads from the RCS for inspection and maintenance. The SGs are discussed further in Section 5.5.2.
Use of SG nozzle dams requires lowering the water level in the RCS to a level below that necessary to remove the RVCH (i.e., partial drain or mid-loop operation). Mid-loop operation, when performed in accordance with approved plant procedures, is
acceptable when core decay heat is less than or equal to 15.3 MWt (Reference 1).
During mid-loop operation, water level is closely monitored to ensure adequate RHR pump suction and decay heat removal by the RHR system.
5.1.8 SAFETY EVALUATION 5.1.8.1 General Design Criterion 2, 1967 - Performance Standards All RCS components are located within the PG&E Design Class I auxiliary and containment buildings. These buildings, or a pplicable portions thereof, are designed to withstand the effects of winds and tornadoes (refe r to Section 3.3), floods and tsunamis (refer to Section 3.4), external missiles (refer to Section 3.5), earthquakes (refer to Section 3.7), and other natural phenomena, to protect RCS SSCs, ensuring their design functions will be performed.
PG&E Design Class I RCS SSCs are design ed to perform their function of providing shutdown capability under Double Design Earthquake (DDE) and Hosgri Earthquake (HE) loading. The seismic requirements are defined in Sections 3.7 and 3.10, and the provisions to protect the system from seismic damage are discussed in Sections 3.7, 3.9, and 3.10.
5.1.8.2 General Design Criterion 3, 1971 - Fire Protection The RCS is designed to the fire protection guidelines of Branch Technical Position APCSB 9.5-1 (refer to Appendix 9.5B, Table B-1).
DCPP UNITS 1 &
2 FSAR UPDATE 5.1-18 Revision 23 December 2016 5.1.8.3 General Design Criterion 4, 1967 - Sharing of Systems RCP vibration monitoring is provided by field equipment mounted in instrument racks in each containment building. Vibration data from the instrument racks is collected and stored on a shared server in the common control room. The RCP vibration monitoring system does not perform a safety function, or provide a direct control function; it only provides indication and alarms in the control room. Refer to Section 5.5.1.2 for additional information on RCP vibration monitoring.
5.1.8.4 General Design Criterion 4, 1987
- Environmental and Dynamic Effects Design Bases Detailed analysis has shown that the primary loops are highly resistant to stress corrosion cracking and high and low cycle fatigue. Based on this analysis, dynamic effects of RCS primary loop pipe breaks need not be considered in the structural design basis. Protection from the dynamic effects of the most limiting breaks of auxiliary branch lines needs to be considered. This includes RCS branch line breaks and other high energy line breaks as described in Sections 5.2.2.1.9, 5.2.2.1.10, 5.2.2.1.11, 5.2.2.1.14, 5.2.2.1.15, and 5.2.2.1.16. Refer to Section 3.6.2.1.1.1 for discussion of the LBB methodology and application to the primary loops of DCPP Unit 1 and Unit 2.
5.1.8.5 General Design Criterion 6, 1967 - Reactor Core Design Each reactor core is designed to function throughout its design lifetime without exceeding acceptable fuel damage limits. The RCS is a reliable process and decay heat removal system that provides for this capabil ity under all expected conditions of normal operation, with appropriate margins for uncertainties and anticipated transient situations.
5.1.8.6 General Design Criterion 9, 1967 - Reactor Coolant Pressure Boundary The provisions for the control of water chemistry to protect the RCS from corrosion are discussed in Sections 5.2.2.3.4 and 5.5.2.3.5, and therefore ensure the RCPB is maintained.
CVCS provides RCP seal injection to ensure RCP seal integrity, and therefore maintaining the RCPB (refer to Section 9.3.4.
3.21). Refer to Section 5.2.2.3 for a discussion of the RCPB materials of construction. Refer to Section 5.2.3.23.2 for leakage limits for the RCS pressure isolation valves (PIVs).
DCPP UNITS 1 &
2 FSAR UPDATE 5.1-19 Revision 23 December 2016 5.1.8.7 General Design Criterion 11, 1967 - Control Room Instrumentation and controls are provided in the control room for operators to maintain the RCS within design parameters. RCS instrumentation and controls in the control room include:
(1) RCS temperature, pressure, and flow indication (2) RCS subcooling margin and RPV level indication (3) Pressurizer pressure, level, and temperature indication, and heater group controls and power indication (4) PRT pressure, level, and temperature indication (5) RCP controls and motor amps indication (6) RCP seal flow, differential pressure, and temperature indication (7) PORVs/PSVs discharge temperature indication (8) PSV acoustic monitor flow indication (9) PORV and PORV block valve, PRT valve, and RVHVS valve controls and position indication In the event control room access is lost, instrumentation and controls required for safe shutdown, are provided outside the control room (refe r to Section 7.4.2.1) at the HSP and the dedicated shutdown panel. Pressurizer heater on-off control is provided on the HSP for two backup heater groups; however, these controls are not required for safe shutdown (refer to Section 7.4.2.1.2.4). Instr umentation requirements for the RCS are discussed in Section 5.6.1.
5.1.8.8 General Design Criterion 12, 1 967 - Instrumentation and Controls RCS Instrumentation and controls are provided, as required, to monitor and maintain the RCS variables within prescribed operating ranges, and to provide post-accident monitoring (refer to Section 5.6.1 for additional information).
Monitored RCS variables include:
(1) RCS temperature (2) RCS pressure (3) Pressurizer pressure and level (4) RCS flow (5) RCP motor amps (6) Subcooling Margin (7) RPV level (8) PORV and PSV position DCPP UNITS 1 &
2 FSAR UPDATE 5.1-20 Revision 23 December 2016
5.1.8.9 General Design Criterion 13, 1967 - Fission Process Monitors and Controls The RCS instrumentation monitors and provides continuous indication of RCS temperature for additional fission process information. Refer to Sections 7.7.3.3 and 9.3.4.3.7 for additional information.
5.1.8.10 General Design Criterion 15, 1967 - Engineered Safety Features Protection System The pressurizer pressure circuit initiates safety injection (SI) when 2-out-of-4 pressurizer pressure channels read below the specified setpoint (refer to Sections 7.3.2.1 and 7.3.3.3).
5.1.8.11 General Design Criterion 21, 1967 - Single Failure Definition The PG&E Design Class I RCS SSCs descri bed below are designed so that a single failure will not prevent the RCS from perform ing its design function. Redundant Class 1E power is provided, as necessary, for PG&E Design Class I SSCs.
Redundant pressurizer PORVs function in the event of an accident (refer to Sections 5.2.2, 15.2.15, and 15.4.3).
RVHVS vent valves provide redundant capability to vent noncondensible gases from the RCS which might inhibit core cooling during natural circulation assuming a single failure (refer to Section 5.5.14.2).
The RVLIS supplements RCS pressure and temperature sensors and the SCMM in detection of inadequat e core cooling (refer to Sections 5.6.1 and 7.5.2.2).
Refer to Sections 5.1.8.15, 5.1.8.16, and 6.2.4 for a discussion of the configuration of the containment isolation system (CIS).
Refer to Section 5.1.8.10 for a discussion of the pressurizer pressure circuit for initiation of SI. Refer to Section 5.1.8.26, Items II.E.3.1 and I I.G.1 for a discussion of the emergency power supplies for pressurizer equipment.
Refer to Sections 3.9 and 5.2 for a discussion of active valves.
DCPP UNITS 1 &
2 FSAR UPDATE 5.1-21 Revision 23 December 2016 5.1.8.12 General Design Criterion 40, 1967 - Missile Protection There are no credible missiles generated by the failure of the RCS components that would prevent the ESFs SSCs inside containment from performing their design functions.
Precautionary measures, taken to preclude missile formation from RCP components, ensure that the pumps will not produce missiles under any anticipated accident condition (refer to Sections 5.2.3.20 and 5.5.1.3.7).
A failure of a CRDM housing sufficient to allow a control rod to be rapidly ejected from the core is not considered credible based on the precautionary measures; however, a missile shield structure is provided over the CRDMs which will block missiles which might be generated in the event of a fractur e of the pressure housing of any mechanism (refer to Sections 3.5.2.1.1, 3.5.2.
2.1, 3.5.2.3.1, 3.5.2.4 and 3.5.2.5).
Missiles generated by smaller components such as valves, temperature and pressure element assemblies, and pressurizer heaters are either not credible or have been shown to not impact safety functions (refer to Sections 3.5.2.1.1, 3.5.2.2.1, 3.5.2.3.1, 3.5.2.4 and 3.5.2.5).
5.1.8.13 General Design Criterion 49, 1967 - Containment Design Basis The RCS piping routed through containment penetrations is designed and analyzed to withstand the pressures and temperatures that could result from a LOCA without exceeding the design leakage rates. Refer to Sections 3.8.2.1.3 and 6.2.4, and Table 6.2-39 for additional details.
5.1.8.14 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment The RCS CIVs required for containment closure are periodically tested for operability and leakage. Test connections are provided in the penetration and in the piping to verify valve leakage and penetration leakage are within prescribed limits. Testing of the components required for the CIS is discussed in Section 6.2.4.
5.1.8.15 General Design Criterion 55, 1971 - Reactor Coolant Pressure Boundary Penetrating Containment The RCS penetrations that are part of the CIS include the PRT makeup and gas analyzer lines, and the RVLIS lines, which comply with the requirements of GDC 55, 1971, as described in Section 6.2.4 and Table 6.2-39. Refer to Section 9.3.6 for the nitrogen line to the PRT.
DCPP UNITS 1 &
2 FSAR UPDATE 5.1-22 Revision 23 December 2016 5.1.8.16 General Design Criterion 56, 1971 - Primary Containment Isolation The RCS penetrations that are part of the CIS include the common inlet line to the PRT that accepts discharge from the various ECCS relief valves, which complies with the requirements of GDC 56, 1971, as described in Section 6.2.4 and Table 6.2-39.
5.1.8.17 Reactor Coolant System Safety Function Requirements (1) Protection from Missiles and Dynamic Effects The PG&E Design Class I RCS SSCs are protected from the effects of missiles and/or dynamic effects as discussed in Sections 3.5.3.1 and 3.6, respectively, and Section 5.2.
(2) Reactor Heat Removal The RCS provides sufficient heat transfer capability, using coolant flow from the RCPs, to transfer the heat produced during power operation and the initial phase of plant
cooldown, when the reactor is subcritical, to the steam system via the SGs.
The system provides sufficient heat transfer capability to transfer the heat produced
during the subsequent phase of plant cooldown and cold shutdown to the RHR system.
The system heat removal capability under power operation and normal operational
transients, including the transition from forced to natural circulation, ensures that no fuel
damage occurs within the operating bounds permitted by the reactor control and
protection systems.
(3) RCS Thermal-Hydraulic Requirements The RCS thermal-hydraulic design provides appropriate limits on RCS pressure (refer to Section 5.3.1) and ensures adequate RCP N PSH (refer to Sections 5.3.2 and 5.5.1).
Refer to Section 7.7.2 for discussion of T avg control.
(4) RCS Coolant Functional Properties The RCS contains the water used as the core neutron moderator and reflector and as a
solvent for chemical shim control. The system, together with the CVCS, maintains the homogeneity of soluble neutron poison concentration and controls the rate of change of
coolant temperature, preventing uncontrolled reactivity changes.
(5) RCS Pressure and Volume Control The pressurizer maintains RCS pressure and volume through the surge line during operation and limits pressure changes during transients. During plant load reduction or increase, reactor coolant volume changes are accommodated in the pressurizer via the
surge line, pressurizer sprays and/or heaters, and the PORVs. Refer to Sections 5.5.9 DCPP UNITS 1 &
2 FSAR UPDATE 5.1-23 Revision 23 December 2016 and 5.5.12 for additional information. For RCS pressure control during a steam generator tube rupture (SGTR), refer to Section 15.4.3.
(6) Steam Flow Restriction Each SG has an integral flow restrictor located in the steam outlet nozzle to limit the steam blowdown from the SGs in the event of a main steam line rupture. The flow restrictor consists of seven 6.03-inch ID venturi nozzles. These flow restrictors are separate from the in-line 16-inch diameter flow restrictors in the MSS described in Section 10.3.3.13 (5). The flow restrictors are discussed in detail in Sections 5.5.4 and 15.4.2. (7) RCP Coastdown Sufficient pump rotation inertia is provided by a flywheel, in conjunction with the impeller and motor assembly, to provide adequate RCS flow during coastdown. The flywheel inertia of the four RCPs sustains reactor coolant flow for a period of time sufficient to assure the minimum heat removal needed to prevent immediate damage to the core.
The assumption of RCP coastdown in relation to the safety analyses is discussed further in Sections 5.5.1.3.2, 15.2.5, 15.2.9, 15.3.1, 15.3.4, and 15.4.2.
(8) Pressurizer Relief Tank The PRT and rupture disks are designed for a vacuum to prevent tank collapse if the contents cool following a discharge without nitrogen being added.
5.1.8.18 10 CFR 50.49 - Environmental Qualif ication of Electrical Equipment Important to Safety for Nuclear Power Plants RCS instrumentation and control equipment required to function in a harsh environment under accident conditions is qualified to the applicable environme ntal conditions to ensure that they will continue to perform their PG&E Design Class I functions. Section 3.11 describes the DCPP EQ Program and the requirements for the environmental design of electrical and related mechanica l equipment. The affected equipment is listed in the EQ Master List and includes junction boxes, switches, solenoid valves, valve motors, acoustic monitors, resistance temperature detectors (RTDs), differential pressure indicating switches, and pressure transmitters.
5.1.8.19 10 CFR 50.55a(f) - Inservice Testing Requirements The PG&E Design Class I RCS components comply with the ASME Code for Operation and Maintenance of Nuclear Power Plants and are tested to the requirements of 10 CFR 50.55a(f)(4) and 10 CFR 50.55a(f)(5) to the extent practical.
DCPP UNITS 1 &
2 FSAR UPDATE 5.1-24 Revision 23 December 2016 5.1.8.20 10 CFR 50.55a(g) - Inservice Inspection Requirements The PG&E Design Class I portion of the RCS ASME BPVC Section XI components are inspected to the requirements of 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(g)(5) to the extent practical. Refer to Section 5.2.3.15 for ISI of the RPV and RVCH. Refer to Section 5.2.3.21 for ISI of the RCP flywheel.
5.1.8.21 10 CFR 50.63 - Loss of All Alternating Current Power For DCPP, safe shutdown for SBO is assumed to be Mode 3. Core cooling in this mode is to be provided by natural circulation of the reactor coolant through the core and SGs, with heat removal from the SGs provided by the atmospheric steam dump valves.
The SBO event will result in RCP trip with the simultaneous loss of seal injection flow and CCW flow to the RCP, which allows hot RCS water to enter the pump bearing and seal areas. The thermal barrier heat exchanger is designed to cool the RCS water upon restoration of CCW flow using the ac standby power supply to prevent seal damage for the duration of the SBO event.
5.1.8.22 10 CFR Part 50 Appendix R (Sections III.G, III.J, III.L, and III.O) - Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979 Section III.G - Fire Protection of Safe Shutd own Capability: Tables 9.5G-1 and 9.5G-2 for DCPP Unit 1 and Unit 2, respectively, list the minimum equipment required to bring the plant to a cold shutdown condition as defined by 10 CFR Part 50, Appendix R, Section III.G. Active RCS components included are the PORVs, the PORV block valves, and RCS temperature, pressure, and pressurizer level instrumentation. These SSCs are provided fire protection features appropriate to the requirements of Section III.G.Section III.J - Emergency Lighting: Emergency lighting or BOLs are provided in areas where indication of RCS parameters may be required to safely shutdown the unit following a fire.
Section III.L - Alternative and Dedicated Shutdown Capa bility: Safe shutdown capabilities are provided in the control room and at alternate locations via the HSP (refer to Section 7.4). The ability to safely shut down the plant following a fire in any fire area is summarized in Section 4.0 of Ap pendix 9.5A and Appendix 9.5E.
Section III.O - Oil Collection System for Reactor Coolant Pumps:
Appendix 9.5C describes compliance of the RCP oil collection system with 10 CFR Part 50, Appendix R, Section III.O requirements.
DCPP UNITS 1 &
2 FSAR UPDATE 5.1-25 Revision 23 December 2016 5.1.8.23 Regulatory Guide 1.89, November 1974 - Environmental Qualification of Class 1E Equipment for Nuclear Power Plants The SCMM is qualified in accordance with the requirements of Regulatory Guide 1.89, November 1974 (refer to Section 3.11).
5.1.8.24 Regulatory Guide 1.97, Revision 3, May 1983 - Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident RCS post-accident variables required to be monitored for meeting Regulatory Guide 1.97, Revision 3, requirements consist of: R CS soluble boron concentration; RCS cold leg water temperature; RCS hot leg water temperature; RCS pressure; core exit temperature; coolant level in the reactor; subcooling margin indication; pressurizer level; SG pressure; RCP status; PSV position; pressurizer heater status; PRT level, temperature and pressure; and CIV position (refer to Table 7.5-6).
5.1.8.25 Regulatory Guide 1.121, August 1976 - Bases for Plugging Degraded PWR Steam Generator Tubes Regulatory Guide 1.121, August 1976 provides guide lines for establishing criteria for SG tube defects, minimum wall thickness and analytical and loading criteria for tubes exhibiting partial or complete thru-wall cracks and wastage. DCPP uses the guidance of Regulatory Guide 1.121, August 1976 to assess the limits of tube degradation criteria.
DCPP procedures ensure that SG tube inspections and tube integrity assessments are conducted on the appropriate frequency as specified in the technical specifications, and that all SG tubes satisfying the tube repair criteria are plugged. Refer to Section 5.5.2.5 for a discussion of the SG tube inspection program.
5.1.8.26 NUREG-0737 (Items II.B.1, II.D.1, II.E.3.1, II.F.2, II.G.1, II.K.3.5, and II.K.3.25), November 1980 - Clarificat ion of TMI Action Plan Requirements Item II.B.1 - Reactor Coolant System Vents: The RVHVS can be used to remove non-condensable gases or steam from the RVCH to support natural circulation cooling by remote-manual operation from the control ro om (refer to Section 5.5.14).
Item II.D.1 - Performance Testing of Boiling-Water Reactors and Pressurized-Water Reactor Relief and Safety Valves (NUREG-0578, July 1979, Section 2.1.2): Refer to Sections 3.9.2.1.7, 5.2.3.26, and 5.5.12.4 for a discussion of testing of RCS PSVs and PORVs. Item II.E.3.1 - Emergency Power Supply for Pressurizer Heaters: All of the four pressurizer heater groups can be supplied with power from off-site power sources when they are available. In addition, power can be provided to two of the four heater groups from the Class 1E power source through the Class 1E buses when off-site power is not DCPP UNITS 1 &
2 FSAR UPDATE 5.1-26 Revision 23 December 2016 available. This arrangement is adequate for establishing and maintaining natural circulation during hot standby conditions. Redundancy is provided by supplying each of the two groups of heaters from a different Class 1E bus.
Plant procedures direct the operators to connect the required pressurizer heaters to the emergency buses. Loading of each Class 1E bus can be accomplished from the main control board. Procedures identify under what conditions selected loads can be shed from the Class 1E bus to prevent overloading when the pressurizer heaters are connected. The procedures also include provisions to reset the safety injection actuation signal to permit the operation of the heaters. Transfer to the power supplies can be accomplished within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after a loss of offsite power.
Devices which supply the pressurizer heaters with motive and control power from the Class 1E buses are PG&E Design Class I.
Item II.F.2 - Instrumentation for Detection of Inadequate Core Cooling: To meet the requirements for supplementing existing instrumentation, the instrumentation for detection of inadequate core cooling includes the SCMM, core exit thermocouple system, and RVLIS, which covers the range from normal operation to complete uncovering of the core. Refer to Sections 5.6.1.1 and 7.5.2.2 for further discussion.
Item II.G.1 - Emergency Power for Pressurizer Equipment: The PORVs are air-to-open, fail-closed valves. They are normally supplied by the plant air compressors. Two of the three valves have a backup supply from the nitrogen system to function on loss of air, including PG&E Design Class I high pressure accumulators which have sufficient capability to operate each valve more than 100 times after the loss of both air and nitrogen. The third PORV is not supplied with a backup motive power supply.
Each PORV is opened by a solenoid valve which is energized-to-open, spring-to-close.
The circuits to the solenoid valves are supplied with redundant interlocks which prevent energization below normal operating pressures. These control circuits are powered from the redundant Class 1E station batteries.
The backup PORV air supply is PG&E Desig n Class I. The piping, accumulators, control power connections, and the solenoid valves are PG&E Design Class I.
The PORV block valves, including the control power connections, are powered from Class 1E buses which are served by either offsite power or the standby power supply.
Each of the three valves is powered from a separate Class 1E 480-V bus.
The pressurizer level indication circuits are PG&E Design Class I. AC power for all Class 1A instrument channels is supplied from inverters which are supplied from the Class 1E buses with automatic backup from the Class 1E 125-Vdc batteries.
Item II.K.3.5 - Auto Trip of RCPs During Loss-of-Coolant Accident: The RCPs do not automatically trip on a SBLOCA as sufficient time is available for manual trip (Reference DCPP UNITS 1 &
2 FSAR UPDATE 5.1-27 Revision 23 December 2016 4). Plant procedures direct the operators to manually trip the RCPs if necessary. The implementation of this item utilized the required Westinghouse Owner's Group RCP trip criteria that had been submitted by the Westinghouse Owners Group in response to Generic Letter 83-10c, February 1983. Generic Letter 85-12, June 1985, also provided guidance concerning implementation of the approved RCP trip criteria.
Item II.K.3.25 - Effect of Loss of Alternating-Current Power on Pump Seals: The CCW system that provides cooling water to the RCP thermal barriers can be supplied from the standby power supply and its operability will not be lost with a loss of ac power.
5.1.8.27 Generic Letter 83-37, November 1983 - NUREG-0737 Technical Specifications Item II.B.1 - Reactor Coolant System Vents: One of the two separate vent paths, consisting of at least two valves in series which are powered from Class 1E buses, is required to be operable by plant procedures. Refer to Section 5.5.14 for a description of the RVHVS.
5.1.8.28 Generic Letter 88-05, March 1988 - Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants The DCPP BACCP monitors and maintains the integrity of the RVCH, the RCS and its supports, and all other borated systems pressure boundary components in accordance with Generic Letter 88-05, March 1988.
The BACCP is established to minimize boric acid induced corrosion by providing for:
(1) Early detection of boric acid leaks.
(2) Thorough inspection of the surrounding areas.
(3) Proper evaluation of areas where leakage has occurred. Of special concern is any impact to ASME Code Class 1 equipment.
(4) Prompt action to mitigate the leak, perfor m repairs, and avoid future damage.
5.1.8.29 Generic Letter 90-06, June 1990 - Resolution of Generic Issue 70, "Power-Operated Relief Valve and Block Valve Reliability" and Generic Issue 94, "Additional Low-Temperature Over Pressure Protection for Light-Water Reactors" Pursuant to 10 CFR 50.54(f)
The pressurizer PORVs and PORV block valves are included in the scope of the IST program. The block valves are also included in the DCPP Generic Letter 89-10, June 1989, motor-operated valves (MOV) program (refer to Section 5.2.3.27). Technical Specifications ensure valves required for LTOP protection will be able to perform their design function.
DCPP UNITS 1 &
2 FSAR UPDATE 5.1-28 Revision 23 December 2016 5.1.8.30 Generic Letter 95-07, August 1995 - Pressure Locking and Thermal Binding of Safety-Related Power-Operated Valves PG&E Design Class I power-operated gate valves in the RCS that were determined to be susceptible to pressure locking have been modified by drilling a hole in the high pressure side of the disk to prevent pressure locking. No power operated gate valves in the RCS were found susceptible to thermal binding.
5.1.8.31 NRC Bulletin 88-09, July 1988 - Thimble Tube Thinning in Westinghouse Reactors The incore neutron monitoring system is inspected during each outage in accordance with plant procedures. The inspection confirms the integrity of the incore neutron monitoring system thimble tube.
5.1.8.32 NRC Bulletin 88-11, December 1988 - Pressurizer Surge Line Thermal Stratification Analyses were performed to evaluate the st ress and fatigue effects due to thermal stratification and thermal striping of the pressurizer surge lines. The fatigue evaluation determined the pressurizer surge lines meet the acceptance criteria of ASME BPVC Section III-1986.
5.1.8.33 Branch Technical Position ASB 10-2, March 1978 - Design Guidelines for Avoiding Water Hammers in Steam Generators The SGs include top-discharge spray nozzles in the feedwater ring which reduce the possibility of steam pockets being trapped in the feedwater ring (refer to Section 5.5.2).
The main feedwater piping has been design ed minimizing the length of horizontal feedwater piping which could be emptied when the SG water level drops below the level of the feedwater ring (refer to Section 10.4.7.3.20).
5.
1.9 REFERENCES
- 1. RCS Pressurization Analysis for Diab lo Canyon Shutdown Scenarios, Westinghouse Technical Report, April 1997.
- 2. RCS Flow Measurement Using Elbow Tap Methodology at Diablo Canyon Units 1 and 2, WCAP-15113, Revision 1, Westinghouse Electric Company LLC, April 2002. 3. RCS Flow Verification Using Elbow Taps at Westinghouse 3-Loop PWRs, WCAP-14750, Revision 1, Westinghouse Electric Company, September 1999.
- 4. Analysis of Delayed Reactor Cool ant Pump Trip during Small Loss of Coolant Accidents for Westinghouse Nuclear Steam Supply Systems, WCAP-9584, Westinghouse Electric Company LLC, August 1979.
DCPP UNITS 1 &
2 FSAR UPDATE 5.1-29 Revision 23 December 2016 5.1.10 REFERENCE DRAWINGS Figures representing controlled engineering drawings a re incorporated by reference and are identified in Table 1.6-1. The contents of the drawings are controlled by DCPP procedures.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-1 Revision 23 December 2016 5.2 INTEGRITY OF THE REACTOR COOLANT PRESSURE BOUNDARY The RCPB is designed to accommodate the system pressures and temperatures attained under all expected modes of plant operation, including all anticipated transients, and to maintain the resulting stresses within allowable values. The system is
protected from overpressure by means of pressure relieving devices as required by
applicable codes and a special system for low temperature operation. Materials of
construction are specified to minimize corrosion and erosion and to provide a structure
and system pressure boundary that will maintain its integrity throughout the life of the
plant. Inspections in accordance with Reference 8, and provisions for surveillance of
critical areas to enable periodic assessment of the boundary integrity, are made.
5.2.1 DESIGN BASES 5.2.1.1 General Design Criterion 2, 1967 - Performance Standards The RCPB is designed to withstand the effects of, or is protected against, natural phenomena such as earthqua kes, tornadoes, flooding, wind s, tsunamis and other local site effects.
5.2.1.2 General Design Criterion 4, 1987 - Environmental and Dynamic Effects Design Bases Consideration of the dynamic effects associated with main RCL piping postulated pipe ruptures are excluded from the DCPP des ign basis with the approval of LBB methodology by demonstrating that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping systems.
5.2.1.3 General Design Criterion 9, 1967 - Reactor Coolant Pressure Boundary The RCPB is designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its lifetime.
5.2.1.4 General Design Criterion 11, 1967 - Control Room The RCPB is designed to or contains instrumentation and controls that support actions to maintain the safe operational status of the plant from the control room or from an alternate location if control room access is lost due to fire or other causes.
5.2.1.5 General Design Criterion 12, 1967 - Instrumentation and Controls Instrumentation and controls are provided as required to monitor and maintain the RCPB variables within prescribed operating ranges.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-2 Revision 23 December 2016 5.2.1.6 General Design Criterion 16, 1967 - Monitoring Reactor Coolant Pressure Boundary Means are provided for monitoring the RCPB to detect leakage.
5.2.1.7 General Design Criterion 33, 1967 - Reactor Coolant Pressure Boundary Capability The RCPB is capable of accommoda ting without rupture, and wi th only limi ted allowance for energy absorption through plastic deformation, th e static and dynamic loads imposed on any boundary components as a result of any inadvertent and sudden release of energy to the coolant.
5.2.1.8 General Design Criterion 34, 1967 - Reactor Coolant Pressure Boundary Rapid Propagation Failure Prevention The RCPB is designed to minimize the probability of rapidly propagating type failures.
Consideration is given (a) to notch-toughness properties of materials extending to the upper shelf of the Charpy transition curve, (b) to the state of the stress of materials under static and transient loadings, (c) to the quality control specified for materials and component fabrication to limit flaw sizes, and (d) to the provisions for control over service temperature and irradiation effects which may require operational restrictions.
5.2.1.9 General Design Criterion 35, 1967 - Reactor Coolant Pressure Boundary Brittle Fracture Prevention Under conditions where RCPB system components constructed of ferritic materials may be subjected to potential loadings, such as a reactivity induced loading, service temperatures are at least 120°F above the nil ductility transition (NDT) temperature of the component material if the resulting energy release is expected to be absorbed by plastic deformation or 60°F above the NDT temperature of the component material if the resulting energy release is expected to be absorbed within the elastic strain energy range.
5.2.1.10 General Design Criterion 36, 1967 - Reactor Coolant Pressure Boundary Surveillance RCPB components have provisions for inspection, testing, and surveillance by appropriate means to access the structural and leaktight integrity of the boundary components during their service lifetime. For the reactor vessel, a material surveillance program conforming to ASTM-E-185-66 is provided.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-3 Revision 23 December 2016 5.2.1.11 General Design Criterion 51, 1967 - Reactor Coolant Pressure Boundary Outside Containment For the portion of the RCPB outside containment, appropriate features as necessary are provided to protect the health and safety of t he public in case of accidental rupture in that part. Determination of the appropriateness of features such as isolation valves and additional containment include consideration of the environmental and population conditions surrounding the site.
5.2.1.12 Reactor Coolant Pressure Boundary Safety Function Requirement (1) Protection from Missiles and Dynamic Effects The RCPB is designed to be protected against missile and dynamic effects which may result from equipment failures.
5.2.1.13 10 CFR 50.55a- Codes and Standards The RCPB is designed in accordance with the requirements of 10 CFR 50.55a to the extent practical.
5.2.1.14 10 CFR 50.55a(f) - Inservice Testing Requirements RCPB ASME Code components are tested to the requirements of 10 CFR 50.55a(f)(4) and 10 CFR 50.55a(f)(5) to the extent practical.
5.2.1.15 10 CFR 50.55a(g) - Inservice Inspection Requirements The RCPB ASME Code components are inspected to the requirements of 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(g)(5) to the extent practical.
5.2.1.16 10 CFR 50.60 - Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation The fracture toughness and material surveillance program requirements are implemented for the RCPB set forth in Appendices G and H to 10 CFR Part 50.
5.2.1.17 10 CFR 50.61- Fracture Toughness Requirements for Protection against Thermal Shock Events 10 CFR 50.61 specifies the calculation of projected values of the reference temperature for reactor vessel material evaluated for the highest neutron fluence expected throughout expiration of the operating license.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-4 Revision 23 December 2016 5.2.1.18 10 CFR Part 50 Appendix G- Fracture Toughness Requirements The fracture toughness requirements for ferritic materials of pressure-retaining components of the RCPB are implemented to provide adequate margins of safety during any condition of normal operation, incl uding anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime.
5.2.1.19 10 CFR Part 50 Appendix H- Reactor Vessel Material Surveillance Program Requirements A surveillance program is implemented to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region which result from exposure of these materials to neutron irradiation and the thermal environment. Under the program, fracture toughness test data ar e obtained from material specimens exposed in surveillance capsules, which are withdrawn periodically from the reactor vessel.
5.2.1.20 Safety Guide 14, October 1971 - Reactor Coolant Pump Flywheel Integrity Missile protection, with regards to the flywheels of the RCP motors, is provided in accordance with Safety Guide 14, October 1971, with exception to the ISI requirement C.4. The ISI requirements are in accordance with Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1.
5.2.1.21 Regulatory Guide 1.14, Revision 1, August 1975 - Reactor Coolant Pump Flywheel Integrity A program provides for the inspection of each RCP flywheel per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, with exception to the examination requirements given by Regulatory Guide 1.14, Revision 1, Positions C.4.b(1) and C.4.b(2).
5.2.1.22 Regulatory Guide 1.44, May 1973 - Control of the Use of Sensitized Stainless Steel Regulatory Guide 1.44, May 1973, describes methods for control of the application and processing of stainless steel to avoid severe sensitization to diminish occurrences of stress corrosion cracking.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-5 Revision 23 December 2016 5.2.1.23 Regulatory Guide 1.45, May 1973 - Reactor Coolant Pressure Boundary Leakage Detection Systems Leakage detection systems are designed with acceptable methods to detect and identify the location of the source of RCPB leakage.
5.2.1.24 Regulatory Guide 1.97, Revision 3, May 1983 - Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants Instrumentation is provided to monitor RCPB integrity following an accident.
5.2.1.25 Regulatory Guide 1.99, Revision 2, May 1988 - Radiation Embrittlement of Reactor Vessel Materials
NDT) values are derived for 1/4T and 3/4T (thickness) in the limiting material by using the method described in Regulatory Guide 1.99, Revision 2 (Reference 27), and the maximum fluence for the applicable service period. Methods acceptable to the U.S. Nuclear Regulatory Commission (NRC) for estimating the embrittlement of reactor vessel beltline materials is provided in Regulatory Guide 1.99, Revision 2, which was endorsed in Generic Letter 88-11, July 1988.
5.2.1.26 NUREG-0737 (Items II.B.1, II.D.1, II.D.3, I I.K.2.13, and III.D.1.1), November 1980 - Clarification of TMI Action Plan Requirements Item II.B.1 - Reactor Vessel Head Vent System: A RVHVS is provided to exhaust non-condensable gases and/or steam from the RCS that could inhibit natural circulation core cooling. The configuration of the RCS vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve power supply or control system does not prevent isolation of the vent path.
Item II.D.1 - Performance Testing of Pressurized-Water Reactor Relief and Safety Valves: PSVs and PORVs and block valves: a program has been implemented for testing to qualify RCS relief and safety valves under expected design transients.
Item II.D.3 - Valve Position Indication for PSVs and PORVs: Positive PSV and PORV position indication is provided in the control room.
Item II.K.2.13 - Thermal Mechanical Report: An analysis has been performed to evaluate the effects of high pressure injection on vessel integrity.
Item III.D.1.1 - Integrity of Systems Outside Containment Likely to Contain Radioactive Material for Pressurized-Water Reactors and Boiling Water Reactors: A program has been implemented for preventative maintenance for leakage testing and reduction of DCPP UNITS 1 &
2 FSAR UPDATE 5.2-6 Revision 23 December 2016 leakage for primary systems that could conta in highly radioactive fluids during a serious accident or transient.
5.2.1.27 Generic Letter 1989-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance The RCPB PG&E Design Class I and position changeable MOVs are included in the MOV Program for Generic Letter 89-10, June 1989, and associated Generic Letter 96-05, September 1996.
5.2.1.28 Generic Letter 1990-06, June 1990 - Enclosure B, Resolution of Generic Issue 94 - Additional Low-Temperature Overpressure Protection For Light-Water Reactors The RCPB is designed such that brittle fracture of the RPV while at low temperature, if combined with a critical crack in the reactor coolant pressure vessel welds or plate material, will not occur.
5.2.2 SYSTEM DESCRIPTION 5.2.2.1 Design of Reactor Coolant Pressure Boundary Components The RCPB is defined as tho se piping systems and compon ents that contain reactor coolant at design pressu re and temperature. RCPB pipin g systems and components are defined as PG&E Quality/Code Class I, with the exception of those RCPB components excluded from PG&E Quality/Code Class I requirements by 10 CFR 50.55a as described in Section 3.2.2.3. With the exception of the reactor coolant sampling lines, the entire RCPB, as defined above, is located entirely within the con tainment structure.
The RCS boundaries are designed to accommoda te the system pressures and temperatures attained under all expected modes of pl ant operation, including all anticipated transients, and to maintain the stresses wi thin applicable stress limits.
5.2.2.1.1 Performance Objectives The performance objectives of the RCS are describe d in Section 5.1. Equipment codes and classification of the components within the RCS boundary are listed in Table 5.2-2.
Procurement information for major RCS components is provided in Table 5.2-3.
The following five operating conditions are considered in the design of the RCS:
(1) Normal Conditions Any condition in the course of startup, operation in the design power
range, hot standby and system shutdown, other than upset, emergency, faulted, or testing conditions.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-7 Revision 23 December 2016 (2) Upset Conditions Any deviations from normal conditions anticipated to occur often enough that the design should include a capability to withstand the conditions
without operational impairment. The upset conditions include those transients that result from any single operator error control malfunction, transients caused by a fault in a system component requiring its isolation
from the system, and transients due to loss of load or power.
Upset conditions include any abnormal incidents not resulting in a forced outage and also forced outages for which the corrective action does not
include any repair of mechanical damage. The estimated duration of an upset condition was included in the design specifications.
(3) Emergency Conditions Emergency conditions are those deviations from normal conditions that
require shutdown for correction of the conditions or repair of damage in
the system. These conditions have a low probability of occurrence but are
included to ensure that no gross loss of structural integrity results as a
concomitant effect of any damage developed in the system. The total
number of postulated occurrences for such events will not cause more
than 25 stress cycles having an Sa value greater than that for 106 cycles
from the applicable ASME BPVC Section III, fatigue design curves.
(4) Faulted Conditions Faulted conditions are those combinations of conditions associated with
extremely low probability postulated events whose consequences are
such that the integrity and operability of the nuclear energy system may be
impaired to the extent that considerations of public health and safety are
involved. Such conditions require compliance with safety criteria as may be specified by jurisdictional authorities.
(5) Testing Conditions Testing conditions are those tests, in addition to the hydrostatic or
pneumatic tests, permitted by the ASME BP VC Section III, including leak tests or subsequent hydrostatic tests.
5.2.2.1.2 Design Parameters The design parameters of the RCS are described in Section 5.1 and Table 5.1-1.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-8 Revision 23 December 2016 5.2.2.1.3 Compliance with 10 CFR 50.55a Codes and standards applicable to RCPB components are specified in 10 CFR 50.55a.
They depend on when the plant was designed and constructed. Construction permits for DCPP Unit 1 and Unit 2 were issued on April 23, 1968, and December 9, 1970, respectively. Therefore, codes and standards specified in 10 CFR 50.55a for
construction permits issued before January 1, 1971, are applicable to the DCPP.
The codes, standards, and component classifications used in the design and
construction of the DCPP RCPB components are shown in Table 5.2-2 and are in
accordance with the applicable pr ovisions of 10 CFR 50.55a. Where text refers to codes in general, the applicable code edition and addenda are as specified in Table 5.2-
- 2. These design codes specify applic able surveillance requirements including allowances for normal degradation.
Although use of the normal, upset, emergency, and faulted condition terminology was
introduced in codes (ASME BPVC,Section III, Summer 1968 Addenda) and standards after the code applicability date for the DCPP, analyses of RCS components in
accordance with the ASME BPVC conditions (normal, upset, and faulted) have been performed for the load combinations and associated stress limits identified in Tables
5.2-5through 5.2-7. Formal analysis of the (LOCA + Hosgri) faulted load combination is in progress. This analysis is being tracked in the DCPP corrective action program.
For piping, the 1967 or 1973 versions of the B31.1 Code do not contain explicit description for the load combinations or allowable stress limits for the faulted loading conditions. As a result, the load combinations and allowable stress limits for the faulted loading conditions are provided in Table 5.2-5. The stresses due to the above conditions are combined using the equations described in two editions of the B31.1 Code (both the 1967 Code with 1971 Addendum, and the 1973 Code with Summer 1973 Addendum). The combined stresses are compared with the allowable stress limits as shown in Table 5.2-5. Valves have been designed in accordance with USAS B16.5, in general, and ASME BPVC Section VIII, for flange connections.
5.2.2.1.4 Applicable Code Cases Application, by Westinghouse or other vendors, of the code cases in Table 5.2-1 is in accordance with ASME Code guidelines. Specific application of any of these code
cases to both DCPP units has not been identified since, at the time of their fabrication, there was neither code, nor NRC requirements to maintain and update a centralized list of these code cases.
5.2.2.1.5 Design Transients The design transients in this section and in Table 5.2-4, in general, apply to the RCPB ASME III Components. Additional, specific transient analysis which applies to DCPP UNITS 1 &
2 FSAR UPDATE 5.2-9 Revision 23 December 2016 individual components may be found in Section 5.2.2.1.5.6, Component Transients, and Section 5.5, Component and Subsystem Design.
To ensure the high degree of integrity of RCS equipment over the design life of the plant, fatigue evaluation is based on conservative estimates of the magnitude and
frequency of temperature and pressure transients resulting from various operating
conditions in the plant. To a large extent, the specific transient operating conditions to
be considered for equipment fatigue analyses were determined by Westinghouse. The
transients selected represent operating conditions that should be prudently anticipated
during plant operation and are sufficiently severe or frequent to be of possible
significance to component cyclic behavior.
The design cycles discussed herein are conservative estimates for equipment design
purposes only and are not intended to be an accurate representation of actual
transients or to reflect operating experience.
As such, the number of occurrences specified in Table 5.2-4 is not an absolute limit, but reflect design bases assumptions.
The design limit requires that the cumulative fatigue usage factor (as calculated per
ASME code guidance) for the equipment or component is less than 1.0. Therefore, a
higher number of occurrences may be allowable based upon evaluation of actual
stresses.
A program has been established and will be maintained which includes tracking the number of cyclic or transient occurrences of Table 5.2-4 to ensure that components are
maintained within their design limit unless the program demonstrates by other means
that the design limit will not be exceeded.
DCPP Unit 1 and Unit 2 are licensed for LBB for the main RCL piping. LBB allows the elimination of the dynamic effects of pipe rupture from the design basis. Dynamic effects of pipe rupture are defined as missile generation, pipe whip, pipe break reaction forces, jet impingement, decompression waves within the ruptured pipe, and local pressurizations. Although the dynamic effects of pipe rupture have been eliminated from the design basis, LBB cannot be applied to: containment design, ECCS performance, and EQ of electrical and mechan ical equipment. For these applic ations, the main RCL pipe breaks must be used. The LOCA transient included in Section 5.2.2.1.5, Design Transients, for the plant and for each component provides limiting pressure and temperature blowdown curves which were o riginally generated for the main loop pipe breaks. Since PG&E has applied LBB to the Reactor Coolant (main loop) piping, the LOCA transient presented herein bounds the transients which would be generated by the large branch line breaks, i.e., the Pressurizer Surge Line, the RHR Suction Line, and the Accumulator Lines.
5.2.2.1.5.1 Normal Conditions The following five transients are considered normal conditions:
(1) Heatup and Cooldown DCPP UNITS 1 &
2 FSAR UPDATE 5.2-10 Revision 23 December 2016 For design evaluation, the heatup and cooldown cases are conservatively represented by continuous heatup or cooldown at a rate of 100°F per
hour, which corresponds conceivably to a heatup or cooldown rate that
could only occur under upset or emergency conditions. Heatup brings the
RCS from ambient to the no-load temperature and pressure conditions.
Cooldown represents the reverse situation.
The limitations on heatup reflect:
(a) Criteria for prevention of nonductile failures that establish maximum permissible temperature change rates, as a function of plant
pressure and temperature.
(b) Slower initial heatup rates when using pumping energy only.
(c) Interruptions in the heatup and cooldown cycles due to such factors as drawing a pressurizer steam bubble, rod withdrawal, sampling, water chemistry, and gas adjustments.
(2) Unit Loading and Unloading The unit loading and unloading c ases under automatic reactor control are conservatively represented by a continuous and uniform ramp power
change of 5 percent per minute between 15 percent load and full load.
This load swing is the maximum possible consistent with operation under automatic reactor control. The reactor temperature varies with load as prescribed by the temperature control system.
(3) Step Load Increase and Decrease of 10 percent of Full Power The +/-10 percent step change in load demand is a control transient that is assumed to be a change in turbine control valve opening that might be caused by disturbances in the outside electrical network. The reactor
control system is designed to restore plant equilibrium without reactor trip following a
+/-10 percent step change in turbine load demand initiated from nuclear plant equilibrium conditions in the range between 15 and 100 percent of full load, the power range for automatic reactor control.
During load change conditions, the reactor control system attempts to match turbine and reactor outputs in such a manner that peak reactor
coolant temperature is minimized and reactor coolant temperature is
restored to its programmed setpoint at a sufficiently slow rate to prevent an excessive change in pressurizer pressure.
Following a step load decrease in turbine load, the secondary side steam
pressure and temperature initially increase since the decrease in nuclear DCPP UNITS 1 &
2 FSAR UPDATE 5.2-11 Revision 23 December 2016 power lags behind the step decrease in turbine load. During the same time increment, the RCS average temperature and pressurizer pressure
also increase initially. Because of the power mismatch between the
turbine and reactor and the increase in reactor coolant temperature, the
control system automatically inserts the control rods to reduce core power.
The reactor coolant temperature is ultimately reduced from its peak value
to a value below its initial equilibrium value at the beginning of the
The reactor coolant average temperature setpoint changes as a function of turbine-generator load, as determined by first-stage turbine pressure
measurement. Pressurizer spray causes the pressurizer pressure to decrease from its peak pressure value. At some point during the decreasing-pressure transient, the saturated water in the pressurizer
begins to flash, reducing the rate of pressure decrease. Subsequently, the pressurizer heaters come on to restore the plant pressure to its normal value. Following a step load increase in turbine load, the reverse situation
occurs; i.e., the secondary side steam pressure and temperature initially
decrease and the reactor coolant average temperature and pressure
initially decrease. The control system automatically withdraws the control
rods to increase core power. The decreasing pressure transient is
reversed by actuation of the pressurizer heaters, and eventually the
system pressure is restored to its normal value. The reactor coolant average temperature is raised to a value above its initial equilibrium value at the beginning of the transient.
(4) Large Step Decrease in Load This transient applies to a step decrease in turbine load from full power of such magnitude that the resultant rapid increase in reactor coolant
average temperature and secondary side steam pressure and
temperature automatically initiates a secondary side steam dump system
response that prevents a reactor shutdown or lifting of SG safety valves.
DCPP was originally designed to accept step load reductions from 0 to 95 percent without a reactor trip. Industry experience and operational analysis have shown a 95 percent step load decrease to be very difficult to recover from without the occurrence of a reactor trip. Therefore, the design basis load reduction transient for DCPP has been revised to a 50
percent step load reduction. However, for equipment fatigue and design purposes, the large step decrease in load transient continues to be based on a 95 percent step decrease since it results in more severe pressure and temperature changes.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-12 Revision 23 December 2016 (5) Steady State Fluctuations The reactor coolant average temperature, for purposes of design, is assumed to increase or decrease at a maximum rate of 6°F in 1 minute.
The temperature changes are assumed to be around the programmed value of Tavg (Tavg
+/-3°F). The corresponding reactor coolant pressure is assumed to vary accordingly, and thus be within 2250
+/- 50 psia. It is assumed that an infinite number of these fluctuations occur during the design life of the plant.
5.2.2.1.5.2 Upset Conditions The following seven transients are considered upset conditions:
(1) Loss of Load Without Immediate Turbine or Reactor Trip This transient applies to a step decrease in turbine load from full power
occasioned by the loss of turbine load without immediately initiating a reactor trip and represents the most severe transient on the RCS. The
reactor and turbine eventually trip as a consequence of a high pressurizer
level trip initiated by the reactor protection system (RPS). Since redundant means of tripping the reactor are provided as a part of the RPS, transients of this nature are not expected but are included to ensure a
conservative design.
(2) Loss of Power This transient involves the loss of outside electrical power to the station with a reactor and turbine trip. Under these circumstances, the RCPs are de-energized and, following their coastdown, natural circulation is established in the system to some equilibrium value. This condition permits removal of core residual heat through the SGs that are being fed by the auxiliary feedwater system (AFWS) powered either by a diesel generator or main steam. Steam is initially removed for reactor cooldown
through atmospheric dump valves provided for this purpose.
(3) Partial Loss of Flow This transient applies to a partial loss of flow accident from full power in
which a RCP is tripped as a result of a loss of power to the pump. The consequences of such an accident are a reactor and turbine trip on low reactor coolant flow followed by automatic opening of the steam dump system and flow reversal in the affected loop. The flow reversal results in
a reactor coolant at cold leg temperature, being passed through the SG and cooled still further. This cooler water then passes through the hot leg
piping and enters the reactor vessel outlet nozzles. The net result of the DCPP UNITS 1 &
2 FSAR UPDATE 5.2-13 Revision 23 December 2016 flow reversal is a sizable reduction in the hot leg coolant temperature of the affected loop.
(4) Reactor Trip from Full Power A reactor trip from full power may occur for a variety of causes resulting in
RCS and SG secondary side temperature and pressure transients. It results from continued heat transfer from the reactor coolant to the SG.
The transient continues until the reactor coolant and SG secondary side temperatures are in equilibrium at zero power conditions. A continued
supply of feedwater and controlled dumping of secondary steam remove
the core residual heat and prevent the SG safety valves from lifting. The reactor coolant temperatures and pressures undergo a rapid decrease
from full power values as the RPS causes the control rods to move into the core.
(5) Inadvertent Auxiliary Spray The inadvertent pressurizer auxiliary spray transient will occur if the auxiliary spray valve is opened inadvertently during normal operation.
This will introduce cold water into the pressurizer causing a very sharp
pressure decrease.
Auxiliary spray water temperature depends on regenerative heat
exchanger performance. The most conservative case occurs when the
letdown stream is shut off and unheated charging fluid enters the pressurizer.
The design assumes a spray water temperature of 100°F and a flowrate of 200 gpm. It is also assumed that, if activated, the auxiliary spray will
continue for 5 minutes until shut off.
The pressure decreases rapidly to the low-pressure reactor trip point and
the pressurizer low-pressure reactor trip is assumed to be actuated. This
accentuates the pressure decrease until the pressure is finally limited to
the hot leg saturation pressure. After 5 minutes the spray is stopped and
the pressurizer heaters return the pressure to 2250 psia.
For design purposes, it is assumed that RCS temperature changes do not occur as a result of auxiliary spray initiation except in the pressurizer.
(6) Design Earthquake (DE)
The DE loads are a part of the mechanical loading conditions specified in equipment specifications. The origin of their determination is separate
and distinct from those transient loads resulting from fluid pressure and DCPP UNITS 1 &
2 FSAR UPDATE 5.2-14 Revision 23 December 2016 temperature. Their magnitude is considered in the fatigue design analysis for comparison with appropriate stress limits.
(7) RCS Cold Overpressurization RCS cold overpressurization may occur during startup and shutdown conditions at low temperature, with or without the existence of a steam bubble in the pressurizer. The event is inadvertent, and can potentially occur by any one of a variety of malfunctions or operator errors. The function of the cold overpressure mitigation system (COMS), also known as the LTOP system, is twofold: 1. To provide RCS pressure relief capability to maintain RCS pressure below the limit based on the more limiting of; the fracture toughness requirements of Appendix G of 10 CFR Part 50 for the reactor vessel at low RCS temperatures, or the maximum RCS pressure requirements as dictated by the PORV discharge piping limits. 2. To comply with the minimum RCS pressure constraint consistent with RCP No. 1 seal integrity. This limit is of concern after COMS is actuated (i.e., a PORV opens) and the transient pressure decreases to its minimum value.
All LTOP events can be categorized as belonging to either of the two following transient mechanisms: 1. Events resulting in the addition of mass (mass input transient), or 2. Events resulting in the input of heat (heat input transient).
Umbrella cases of the temperature and pressure transients that can occur from each mechanism are provided for use in the component design.
5.2.2.1.5.3 Emergency Conditions No transient is classified as an emergency condition.
5.2.2.1.5.4 Faulted Conditions The following transients are considered faulted conditions:
(1) RCPB Pipe Break
This accident involves the postulated rupture of a pipe within the RCPB. It is conservatively assumed that system pressure is reduced rapidly and the
ECCS is initiated to introduce water into the RCS. The SI signal will also initiate a turbine and reactor trip.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-15 Revision 23 December 2016 (2) Steam Line Break For RCS component evaluation, the following conservative conditions are considered:
(a) The reactor is initially in hot, zero power subcritical condition assuming all rods in, except the most reactive rod, which is
assumed to be stuck in its fully withdrawn position.
(b) A steam line break occurs inside the containment.
(c) Subsequent to the break, there is no return to power and the reactor coolant temperature cools down to 212°F.
(d) The ECCS pumps restore the reactor coolant pressure.
The above conditions result in the most severe temperature and pressure
variations that the component will encounter during a steam line break accident.
(3) Double Design Earthquake The mechanical stress resulting from the DDE is considered for each component. The seismic analysis is described in Section 3.7.
(4) Hosgri Earthquake The mechanical stress resulting from the HE is considered for each component. The seismic analysis is described in Section 3.7.
The design transients and the number of occurrences of each are shown in Table 5.2-4.
5.2.2.1.5.5 Preoperational Tests and Condition Transients The following hydrostatic tests and leak test conditions were considered in RCS component fatigue evaluations. In some instances, these tests were conducted prior to plant startup.
(1) Turbine Roll Test This test was imposed upon the plant during the hot functional test period
for turbine cycle checkout. RCP power heats the reactor coolant to operating temperature and the steam generated is used to perform a
turbine roll test. Plant cooldown during the test exceeds, however, the
100°F per hour maximum rate.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-16 Revision 23 December 2016 (2) Hydrostatic and Leak Test Conditions Each of the major NSSS components (SG, RCPs, reactor vessel, CRDMs, loop piping and pressurizer) may be subjected to a maximum of 10 hydrostatic tests without exceeding ASME BPVC criteria.
The pressure tests are:
(a) Primary Side Hydrostatic Test Before Initial Startup Pressure tests include both shop and field hydrostatic tests that
occur as a result of component or system testing. This hydrostatic
test was performed prior to initial fuel loading at a water
temperature of at least 168°F (calculated using the methods
presented in Paragraph NB2300 of ASME BPVC Section III-1971, Summer 1972 Addenda), which is compatible with reactor vessel
fracture prevention criteria requirements, and a maximum test
pressure. In this test, the primary side of the SG is pressurized to 1.25 times design pressure (3107 psig) coincident with no pressurization of the secondary side.
(b) Secondary Side Hydrostatic Test Before Initial Startup The secondary side of the SG is pressurized to 1356 psig (1.25 times the design pressure of the secondary side) coincident
with the primary side at zero psig.
(c) Primary Side Leak Test Each time the primary system is opened, a leak test will be performed. During this test the primary system pressure is assumed, for design purposes, to be raised to 2500 psia, with the
system temperature above design transition temperature, while the
system is checked for leaks.
In actual practice, the primary system is pressurized to less than
2500 psia to prevent the PSVs from lifting during the leak test. The secondary side of the SG is pressurized by closing off the steam lines, so that the pressure differential across the tubesheet does
not exceed 1600 psi.
(d) Secondary Side Leak Test During the life of the plant it may be necessary to check the SG secondary side, particularly the manway closure, for leakage. For design purposes, the secondary side is assumed to be pressurized DCPP UNITS 1 &
2 FSAR UPDATE 5.2-17 Revision 23 December 2016 just below 1085 psig (the design pressure of the secondary side of the SG) to prevent the main steam safety valves from lifting. The primary side will also be pressurized so as to not exceed a differential pressure of 670 psi.
(e) Tube Leakage Test During the life of the plant it may be necessary to check the SG for tube leakage and tube-to-tube sheet leakage. This is done by visual inspection of the undersid e (channel head side) of the tube sheet for water leakage, with the secondary side pressurized. Tube
leakage tests are performed during plant cold shutdown.
For these tests, the secondary side of the SGs is pressurized with water, initially at a very low pressure, and the primary system
remains depressurized (i.e., 0 psig). The underside of the tube
sheet is examined visually for leaks. If any leaks are observed, the
secondary side is depressurized and repairs made by tube
plugging. The secondary side is then repressurized (to a higher
pressure) and the underside of the tube sheet is again checked for
leaks. The process is repeated until all the leaks are repaired. The
maximum (final) secondary-side test pressure reached is 840 psig.
The total number of tube leakage tests considered as part of the SG design is 800 during the life of the component. The following is a breakdown of the anticipated number of occurrences at each secondary side pressure.
Case Test Pressure, psig No. of Occurrences Case 1 200 400 Case 2 400 200 Case 3 600 120 Case 4 840 80
Both the primary and secondary sides of the SGs will be at ambient temperature during these tests.
Since the tests outlined under items (a) and (b) occur prior to plant startup, the number of cycles is independent of other operating plant conditions.
5.2.2.1.5.6 Component Transients The following transients apply to the components listed, and are provided to clarify the specific transient applicable to the component.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-18 Revision 23 December 2016 (1) Steam Generator Evaluation Hot standby operation / feedwater cycling is a normal transient which occurs when the plant is being maintained at hot standby or no load conditions. It is assumed that the low steam generation rate is made up by intermittent slug feeding of 32°F feedwater into the SG. Feedwater additions required during plant heatup and cooldown are also assumed to be covered by the feedwater cycling transient, but with no increase in the total number of cycles.
The fatigue analysis for the SG design also considers a one-time upset event where 32 o F feedwater is introduced to a hot, dried-out secondary side of the SG.
(2) Pressurizer Evaluation Normal heatup cases, as mentioned in Section 5.2.2.1.5.1, are conservatively represented by continuous operation at a uniform temperature rate of 100ºF per hour. In actual practice, the rate of temperature change is lower because of other limitations such as material considerations, use of RCP for heatup, or interruptions during normal startup operations to draw a steam bubble, etc. For the pressurizer, the design cooldown rate is 200ºF per hour.
Following any large change in boron concentration in the RCS, the pressurizer spray is operated to equalize concentration between the loops and the pressurizer. This can be done by manually operating the pressurizer backup heaters, thus causing a pressure increase and initiation of spray flow. The pressurizer pressure increases initially before being returned to nominal pressure by the proportional spray. The pressure is then maintained at nominal pressure by spray operation, matching the heat input from the backup heaters until the concentration is equalized. The only effects of these operations on the primary system are as follows:
- 1. The reactor coolant pressure varies in step with the pressurizer pressure.
- 2. The pressurizer surge line nozzle at the hot leg will experience the thermal shocks associated with outflow from the pressurizer 5.2.2.1.6 Identification of Active Pumps and Valves Pumps and valves are classified as either active or inactive components for faulted
conditions. Active components are those whose operability is relied upon to perform a
PG&E Design Class I function as described in Section 3.2.2.1. Inactive components are those whose operability (e.g., valve opening or closure, pump operation or trip) is not
relied upon to perform a PG&E Design Clas s I function. The RCPs are the only pumps in the RCS boundary and are classified as "inactive" in the event of a RCL pipe rupture.
Valves in sample lines are not considered to be part of the RCS boundary because the nozzles where these lines connect to the RCS are orificed to a 3/8-inch hole. This hole
restricts the flow such that loss through a severance of one of these lines is sufficiently DCPP UNITS 1 &
2 FSAR UPDATE 5.2-19 Revision 23 December 2016 small to allow operators to execute an orderly plant shutdown (refer to Section 9.3.2 for description of the portion of the sample lines that are part of the RCPB).
Table 5.2-9 lists the active and inactive valves between major components in the main
process lines of the RCPB, along with the actuation type, valve types, and location. The
listed valves are those that are within the RCPB. Check valves are also included in Table 5.2-9. Check valves are a credited means of pressure boundary isolation for the
original design. Vents, drains, test and instrument root valves are excluded from the table as they meet the isolation requirements and are not between major components of
the RCPB. Manual valves are passive components and are not considered either active
or inactive, therefore they are not included on Table 5.2-9.
5.2.2.1.7 Design of Active Pumps and Valves The design criteria for active PG&E Design Class I pumps outside the RCS boundary are discussed in Section 3.9.2. All these PG&E Design Class I pumps are designated either PG&E Quality/Code Class II or III.
The valves were designed to function at normal operating conditions, maximum design
conditions, and DDE/Hosgri conditions. Active valves that are used for accident
mitigation only, and do not serve to support safe shutdown following a HE, were qualified for active function for a HE to provide increased conservatism in accordance with Reference 30. The design meets the requirements of the ANSI B31.1, ANSI B16.5, and MSS-SP-66 codes (refer to Table 5.2-2).
The stress limits for the valves in the RCS pressure boundary are indicated in Table 5.2-5.
In addition, all valves 1 inch and larger withi n the RCPB were checked for wall thickness to ANSI B16.5, MSS-SP-66, or ASME BPVC Section III-1968 (some 1974) requirements, as applicable, and subjected to nondestructive tests in accordance with
ASME and American Society for Testing and Materials (ASTM) codes.
The valves were designed to the requirements of ANSI B16.5 or MSS-SP-66 pertaining
to minimum wall thickness for pressure containing components. Analyses were
performed to qualify active valves.
HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED These valves were subjected to a series of stringent tests prior to service and during the plant life. Prior to installation, the following tests were performed: shell hydrostatic tests to MSS-SP-61 requirements, backseat and main seat leakage tests. Cold hydrostatic
tests, hot functional qualification tests, periodic ISIs and operability tests have been and are performed to verify and assure the funct ional ability of the valves. These tests assure reliability of the valves for the design life of the plant.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-20 Revision 23 December 2016 On all active valves, an analysis of the extended structure was performed for static equivalent seismic loads applied at the center of gravit y (CG) of the extended structure.
The minimum stress limits allowed in these analyses will assure that no significant
permanent damage occurs in the extended structures during the earthquake.
Motor operators and other electrical appurtenances necessary for operation were
qualified.
The natural frequencies of all active valves were determined by test or by analysis. If
the natural frequencies of the valves were shown to be less than 33 Hz, one of the
following options was employed:
(1) The valve was qualified by dynamic testing.
(2) The valve was modified to increase the minimum frequency to greater than 33 Hz.
(3) The valve was qualified conservatively using static accelerations that are sufficiently in excess of accelerations it might experience in the plant to
take into account any effect due to both multifrequency excitation and
multi-mode response (a factor of 1.5 times peak acceleration is generally
accepted, although lower coefficients can be used when shown to yield
conservative results).
(4) A dynamic analysis of the valve was performed to determine the equivalent acceleration to be applied during the static analysis. The analysis provided the amplification of the input acceleration considering the natural frequency of the valve and the frequency content of the
applicable plant floor response spectra. The adjusted accelerations were then used in the static analysis and the valve operability was assured by
the methods outlined above, using the modified acceleration input.
Swing check valves are characteristically simple in design and their operation is not affected by seismic accelerations or applied nozzle loads. The check valve design is
compact and there are no extended structures or masses whose motion could cause
distortions which could restrict operation of the valve. The nozzle loads due to seismic
excitation do not affect the functional ability of the valve since the valve disc is typically
designed to be isolated from the casing wall. The clearance available around the disc prevents the disc from becoming bound or restricted due to any casing distortions
caused by nozzle loads. Therefore, the design of these valves is such that once the
structural integrity of the valve is assured using standard design or analysis methods, the ability of the valve to operate is assured by the design features. For the faulted
condition evaluations, since piping stresses are shown to be acceptable, the check valves are qualified.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-21 Revision 23 December 2016 The valves have undergone the following tests: (a) in-shop hydrostatic test, (b) in-shop seat leakage test, and (c) periodic in-plant exercising and inspection to assure
functional ability.
By the above methods, all active valves are qualified for operability for the faulted
condition seismic loads. These methods simulate the seismic event and assure that the active valves will perform their PG&E Desig n Class I functions when necessary.
5.2.2.1.8 Inadvertent Operation of Valves The inactive valves within the RCPB listed in Table 5.2-9 are not relied upon to function after an accident. They meet redundancy requirements and will not increase the
severity of any of the transients discussed in Section 5.2.2.1.5, if operated inadvertently during any such transient.
5.2.2.1.9 Stress and Pressure Limits System hydraulic and thermal design parameters are the basis for the analysis of
equipment, coolant piping, and equipment support structures for normal and upset
loading conditions. The analysis uses a static model to predict deformation and
stresses in the system. The analysis gives six components, three moments, and three forces. These moments and forces are resolved into pipe stresses in accordance with
applicable codes. Stresses in the structural supports are determined by the material and section properties based on linear elastic small deformation theory.
In addition to the loads imposed on the system under normal and upset conditions, the design of mechanical equipment and equipment supports requires that consideration also be given to faulted loading conditio ns such as those experienced during seismic and pipe rupture events.
Analysis of the RCLs and support systems for seismic loads is based on a three dimensional, multi-mass elastic dynamic model. The floor response spectra are used
as input to the detailed dynamic model, which includ es the effects of the supports and the supported equipment. The loads developed from the dynamic model are
incorporated into a detailed loop and support model to determine the support member
stresses.
The dynamic analysis employs the displacement method, lumped parameter, stiffness
matrix formulations, and assumptions that all components behave in a linearly elastic
manner. Seismic analyses are covered in detail in Section 3.7.
5.2.2.1.10 Stress Analysis for Structural Adequacy Methods and models used to determine the structural adequacy of components under
the normal and upset conditions are described herewith.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-22 Revision 23 December 2016 5.2.2.1.10.1 Analysis Method for Reactor Coolant System Loading combinations and all owable stresses for RCS components are provided in Tables 5.2-5through 5.2-7.
The load combinations considered in the design of structural steel members of
component supports are summarized in Tables 5.2-8 and 5.2-8a. Formal analysis of the (LOCA + Hosgri) faulted load combination is in progress. This analysis is being tracked in the DCPP corrective action program. The RCS design is described in Section 5.5.
The following paragraphs define the loads applied in the analysis:
(1) Deadweight The deadweight loading imposed by piping on the supports consists of the dry weight of the coolant piping and weight of the water contained in the
piping during normal operation. In addition, the total weight of the primary
equipment components, including water, forms a deadweight loading on
the individual component supports.
(2) Thermal Expansion The free vertical thermal growth of the reactor vessel nozzle centerlines is
considered to be an external anchor movement transmitted to the RCL.
The weight of the water in the SG and the RCP is applied as an external force in the thermal analysis to account for equipment nozzle
displacement as an external movement.
For the SGs, the RCL piping was reanalyzed for thermal expansion. The thermal expansion reanalysis was performed in a similar manner to the original analysis except water weights were included as a part of the deadweight analysis.
(3) Earthquake Loads (DE, DDE and Hosgri)
The earthquake (DE, DDE and Hosgri) acceleration, which produces transient vibration of the equipment mounted within the containment
building, is specified in terms of the floor response spectrum curves at various elevations within the containment building.
These floor response spectrum curves for earthquake motions are
described in detail in Section 3.7.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-23 Revision 23 December 2016 (4) Pressure The steady state hydraulic forces based on the system's initial pressure are applied as internal loads to the RCL model for determination of the
RCL support system deflections and support forces.
(5) LOCA The RCL piping was analyzed for dynamic effects of pipe rupture events.
Since the dynamic effects of pipe rupture events in the main RCL piping no longer have to be considered in the design basis analyses (refer to Section 3.6.2.1.1.1), pipe rupture loads for the analysis are defined for RCL branch line breaks. The pipe rupture load analysis considered
double-ended circumferential breaks in the RHR, SI (accumulator line), and pressurizer surge line connections to the RCL piping.
(6) Other Pipe Rupture Loads The analysis also considered the effects of main steamline breaks on the
RCL piping and the RCS equipment supports. The feedline break was not explicitly analyzed as the main st eamline break is more limiting.
5.2.2.1.10.2 Analytical Models The static and dynamic structural analyses assume linear elastic behavior and employ
the displacement (stiffness) matrix method and the normal mode theory for lumped-parameter, multimass structural repre sentation to formulate the solution.
(1) Reactor Coolant Loop Model The RCL model is constructed for the WESTDYN (Reference 18) computer program. This is a special purpose program designed for the
static and dynamic analysis of redundant piping systems with arbitrary
loads and boundary conditions.
The RCL and support model used for static and LOCA analysis is a one-loop model that describes the spatial geometry, lumped-mass locations, and the node points as shown in Figure 5.2-2. Stiffness characteristics of the equipment support structures are incorporated as linear elastic restraints in the RCL model. The WESTDYN program computes internal member forces, support structure reactions, nodal point deflections, and stresses and also determines system natural frequencies.
The RCL seismic model for DE, DDE and Hosgri is a four-loop model that describes the spatial geometry, lumped-mass locations, and other node points as shown in Figure 5.2-2a. Stiffness characteristics of the DCPP UNITS 1 &
2 FSAR UPDATE 5.2-24 Revision 23 December 2016 equipment support structures are incorporated as linear elastic restraints in the RCL model. Two lumped masses represent the vessel shell. One is located at the vessel flange and the other is located at the lower radial restraints. Three lumped masses represent the core barrel and internals.
One mass is located at each of the upper and lower core plates. The third mass is lumped at the lower radial restraints. The SG is represented by a three-mass, lumped model. The lower mass position is located at the intersection of the inlet and outlet nozzles of the SG. The middle mass position is located at the SG upper support elevation. The upper mass position is located at the top of the SG.
In the DE and DDE analyses, the RCP is represented by a two-mass, lumped model for three of the loops and by a five-mass, lumped model for one of the loops. For the Hosgri model the RCP is represented by a two-mass, lumped model for all four loops. For the RCP two-mass, lumped model, the lower mass position is located at the intersection of the pump suction and discharge nozzles. The upper mass position is located at the CG of the pump motor. For the RCP five-mass, lumped model, the lowest mass position is located at the intersection of the pump suction and discharge nozzles. The middle mass position is located at the CG of the pump motor. Three lumped-masses were used to represent the RCP motor and casing.
(2) Support Structure Models The equipment support structure models have dual purposes since they are required:
(a) To quantitatively represent, in terms of 6 x 6 or 3 x 3 stiffness matrices, or stiffness values, the elastic restraints which the supports impose upon the loop (b) To evaluate the individual support member stresses due to the forces imposed upon the support by the loop.
The loadings on the component supports are obtained from the analysis of an integrated RCL support system's dynamic structural
model, as shown in Figures 5.2-2 and 5.2-2a.
Figures 5.2-2 and 5.2-2a show the RCL model and component supports included in the RCL piping analysis . The analysis considered the pipe rupture restraints on the main RCL piping to be
inactive. The pipe rupture restraints on the main RCL piping were
made inactive by either removing shims or by removing the
support.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-25 Revision 23 December 2016 The primary equipment supports were evaluated using ANSYS, GTSTRUDL and WESPLAT finite element analysis computer programs.
(3) Hydraulic Models The hydraulic model is constructed to quantitatively represent the behavior of the coolant fluid within the RCLs in terms of the concentrated time-
dependent loads imposed upon the loops.
In the original analysis, in evaluating the hydraulic forcing functions during
a LOCA, the pressure and the momentum flux terms are dominant. Inertia and gravitational terms were neglected although they were taken into
account when evaluating the local fluid conditions.
Thrust forces resulting from a LOCA were calculated in a two-step
process. First, the MULTIFLEX 3.0 (Reference 6) code calculated
transient pressure, flowrates, and other coolant properties as a function of
time. Second, the THRUST (Reference 18) code used the results
obtained from MULTIFLEX 3.0 and calculated time-history of forces at locations where there is a change in either direction or area of flow within
the RCL. These locations for the broken loop are shown in Figure 5.2-3.
For the RCL piping analysis , thrust forces and blowdown loads were determined for RCS branch line and main steamline breaks identified in
Section 5.2.2.1.10.1.
5.2.2.1.10.3 Analysis and Solutions (1) Static Load Solutions The static solutions for deadweight, thermal expansion, and pressure load conditions are obtained by using the WESTDYN computer program.
(2) Normal Mode Response Spectral Seismic Load Solution The stiffness matrices representing various supports for dynamic behavior are incorporated into the RCL model. The response spectra for the DE, DDE and HE are applied along the horizontal and vertical axes simultaneously. From the input data, the overall stiffness matrix of the
three-dimensional RCL is generated and the natural frequencies and
normal modes are obtained by the modified Jacobi method.
The forces, moments, deflections, rotations, support structure reactions
and stresses are then calculated for each significant mode. The total DCPP UNITS 1 &
2 FSAR UPDATE 5.2-26 Revision 23 December 2016 seismic response is computed by combining the contributions of the significant modes by the SRSS method.
5.2.2.1.10.4 Reactor Coolant Loop Stress Analysis Results The stress for the normal and upset conditions shows that the stresses in the piping are
below the code-allowable values.
(1) Normal Conditions Stresses due to primary loading of pressure and deadweight are
combined and compared with the stress value for the applicable material property. Refer to Section 5.2.2.1.3. for the applicable code edition. The thermal expansion stress is a secondary stress. The magnitude of the thermal stress is compared with the B31.1 Piping Code allowable
expansion stress limit.
The stress evaluation for the normal condition shows that the stresses in
all RCL members are within the allowable stress values.
(2) Upset Conditions The DE stresses are added to the stresses due to primary loadings of
pressure and deadweight. The st ress evaluation for the upset condition
shows that stresses in all RCL members are within the allowable stress values. 5.2.2.1.10.5 Component Supports Stress Analysis Results (1) Normal Conditions Thermal, weight, and pressure forces (obtained from the RCL analysis) acting on the support structures are combined algebraically.
(2) Upset Conditions DE support forces are added alge braically to normal condition forces. The interaction and stress equations are compared to the allowable limits specified by AISC-1969, which includes a 1/3 allowable stress increase for seismic.
The stress evaluation for the normal and upset conditions shows that the stresses in all
members are within the allowable values.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-27 Revision 23 December 2016 5.2.2.1.11 Analysis Method for Faulted Condition The analysis of the RCLs and support systems for blowdown loads resulting from a LOCA is based on the time-history response of simultaneously applied blowdown
forcing functions on a broken and unbroken loop dynamic model. The forcing functions
are defined at points in the system loop where changes in cross-section or direction of flow occur such that differential loads are generated during the blowdown transient.
Stresses and loads are checked and compared to the corresponding allowable stress.
The stresses in components resulting from normal sustained loads and the worst case
blowdown analysis are combined with the results of seismic faulted condition analyses, using absolute sum or SRSS methodology to determine the maximum stress for the combined loading case. Combining LOCA and seismic loads is considered very
conservative since it is highly improbable that both maxima will occur at the same instant. These stresses are combined to ensure that the main reactor coolant piping
loops and connected primary equipment support system will not lose their intended
functions under this highly improbable situation.
Combining seismic faulted condition and LOCA dynamic loads using SRSS methodology is subject to the conditions and limitations of NUREG-0484, May 1980, Methodology for Combining Dynamic Responses:
- The SRSS technique is acceptable contingent upon performance of a linear, elastic, dynamic analysis to meet the appropriate ASME BPVC Section III, Service Limit for faulted load condition.
For components not designed to ASME Section III, a code reconciliation to ASME
BPVC Section III is required to apply the above.
For faulted conditions, the limits are provided in Table 5.2-7.
Further details of the stress analysis for faulted conditions are presented in
Sections 5.2.2.1.14 and 5.2.2.1.15.
Protection criteria against dynamic effects associated with pipe breaks are covered in Section 3.6. For the RCL analysis, thrust forces and b lowdown loads were determined for RCS branch line b reaks identified in Section 5.2.2.1.10.1.
5.2.2.1.12 Protection Against Environmental Factors
Protection provided for the RCS against environmental factors is discussed in
Sections 3.3, 3.4, 3.5 and 3.6. Fire protection is discussed in Section 9.5.1.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-28 Revision 23 December 2016 5.2.2.1.13 Compliance with Code Requirements In the PG&E classification of DCPP fluid systems and fluid system components, the vessels, piping, valves, pumps and their supports of the RCS pressure boundary are designated PG&E Design Class I, PG&E Quality/Code Class I. The comparison of DCPP system Design/Quality/Code classifications to non-licensing basis regulations
and codes is discussed in Section 3.2 and delineated in Table 3.2-4.
For conservative fatigue evaluations of the reactor vessel, SG, RCP, and pressurizer in accordance with the ASME BPVC per Table 5.2-2, maximum stress intensity ranges are derived from combining the normal and upset condition transients discussed in Section
5.2.2.1.5. The stress ranges and number of occurrences are then used in conjunction with the fatigue curves in the ASME BPVC per Table 5.2-2 to get the associated cumulative usage factors.
The criterion presented in the ASME BPVC per Table 5.2-2 is used for fatigue analysis.
The cumulative usage factor is less than 1, hence, the fatigue design is adequate.
The reactor vessel stress reports include a summary of the critical stress locations analyzed in the vessel, a discussion of the results including a comparison with the
corresponding code limits, descriptions of the methods of analysis and computer programs used, a presentation of some of the actual hand calculations performed, and a tabulation of the references cited in the report.
The content of the stress report is in accordance with the requirements of the ASME BPVC per Table 5.2-2.
For the RVCH, the content of the stress report is in accordance with the requirements of the ASME BPVC per Table 5.2-2.
5.2.2.1.14 Stress Analysis for Faulted Condition Loadings (Double Design Earthquake, Hosgri and Loss-of-Coolant Accident)
Stress analyses of the RCS for faulted conditions employ the displacement (stiffness)
matrix method and lumped-parameter, multimass representation of the system. The analyses are based on adequate and accurate representation of the system using an
idealized, mathematical model. This section discusse s the RCL and support structures analysis. Refer to Section 5.2.2.1.15 for component analysis.
5.2.2.1.14.1 Analysis Method (1) Reactor Coolant Loop The procedure for evaluation of the piping stresses due to combined loadings of weight, pressure, DDE or Hosgri, and LOCA is as follows:
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-29 Revision 23 December 2016 (a) The LOCA stress analysis yields the time-history of stresses at various crosssections in the RCL piping. Axial stress due to pressure is included.
(b) The DDE and Hosgri analysis of the RCL piping was performed using the response spectra method. The RCL seismic model was constructed for the WESTDYN computer program.
(c) Containment internal concrete structure horizontal response spectra at elevations corresponding to the SG upper supports, RCP supports, and the reactor vessel supports were used in the analysis. For the Hosgri seismic analysis, a vertical response spectra envelope from 114 foot elevation to the base slab 87 foot elevation was used in the analys is. For the DDE seismic analysis, the vertical response spectrum is two-thirds of the ground horizontal response spectrum.
(d) With mode, the results due to the vertical shock were combined by direct addition with the results of the horizontal shock directions.
The modal contributions were then added by the SRSS method.
(e) For the Hosgri seismic analyses, two seismic cases were considered; north-south plus vertical and east-west plus vertical.
Each horizontal shock was combined with the vertical shock and
the worst combined response was used in the evaluation of the
system. (f) For the DDE seismic analyses, six horizontal shock directions were performed and the six directions were made up of three pairs of perpendicular shock directions. The shock directions correspond to the follow: 1) parallel with the north-south axis, 2) parallel with the east-west axis, 3) 22 degrees counterclockwise off the north-south axis, 4) 22 degrees counterclockwise off the east-west axis, 5) 22 degrees clockwise off the north-south axis, and 6) 22 degrees clockwise off the east-west axis.
(g) The results of the analysis are as follows: The results of the seismic evaluation were combined with the pressure and
deadweight stresses. The revise d piping stresses were all under the allowable stress limits in Table 5.2-5.
(h) Since the DDE results are obtained by the response spectra method, the six components of a state vector for deflection at a
point or for internal member force cannot be assigned absolute
and/or relative algebraic signs.
Consequently, the maximum values of the DDE axial and shear stresses at a pipe cross-section DCPP UNITS 1 &
2 FSAR UPDATE 5.2-30 Revision 23 December 2016 are calculated from the internal force state vector at that cross-section by considering all possible permutations of signs of the six components of the state vector.
The DDE axial and shear stresses are combined with the time-history of LOCA axial and shear stresses.
(i) Dynamic LOCA loads resulting from pipe rupture events in the RCL branch lines were considered in the design basis stress analyses and were included in the loading combinations.
(j) The resultant axial and shear stresses are combined with the hoop and radial stresses (due to pressure) to determine the principal stresses, 1, 2, and 3. (k) The previous steps are performe d for various cross-sections in the RCL piping. It should be emphasized that, for a given location of
the pipe cross-section, the stress intensity calculation is performed at every step computed from the time-history analysis.
(l) Maximum resultant deadweight, DDE or Hosgri, and LOCA moments were determined at locations located along the RCL
piping, elbows, and connections to equipment. At each location, the maximum resultant moment for DDE is the largest resultant
moment from the various shock cases that were performed for the
seismic analysis. The largest resultant moment for LOCA is the largest moment from the pipe rupture analyses for RCL branch line
breaks and the main steamline break.
(m) As explained in Section 5.2.2.1.3, B31.1 Code pipe stress equations were used with the resultant moments to determine
deadweight, DDE or Hosgri, and LOCA pipe stresses at locations along the RCL corresponding to the maximum resultant moment
locations. At each location, the stresses were combined by
absolute sum and were added to the pressure stress to determine
the maximum stress at that location. This maximum stress was
then verified to be within the stress limit provided in Table 5.2-5. It should be emphasized that the above analysis method is very
conservative since the peak DDE or Hosgri, and LOCA pipe stresses are considered to occur at the exact same instant in time and that the resultant moments for each load type are considered
to be aligned such that the maximum pipe stress occurs at the
same location around the pipe circumference for each load type.
Formal analysis of the (LOCA + Hosgri) faulted load combination is in progress. This analysis is being tracked in the DCPP corrective action program.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-31 Revision 23 December 2016 (2) Evaluation of Support Structures The support loads are computed by multiplying the support stiffness values by the displacement values at the support point. The support loads are used for support member evaluation.
For the support qualification, the following inputs were entered into a GTSTRUDL or ANSYS finite element analysis computer program:
(a) Loads acting on the supports obtained from the RCL analysis (including time-history LOCA forces) (b) Support seismic self-weight excitation loads. (c) Attached platform loads. (d) Attached pipe support loads. (e) Asymmetric compartment pressurization loads. (f) Jet impingement loads. (g) Support structure member properties. (h) Support Geometry. (i) Material properties and code parameters.
The resulting member and component stresses were compared to the acceptance criteria allowable stresses as specified in Table 5.2-8.
RCS component supports were shown ad equate by evaluating the supports for the loads determined in the integrated RCLs seismic analysis.
Stress analyses for structural qualification of the primary equipment supports were performed using the load combinations described in Table 5.2-8. ANSYS, GTSTRUDL and WESPLAT finite element analysis computer programs were used to perform the analysis. The independent loadings included deadweight (DW), thermal expansion (TH), system pressure (P), design earthquake (DE), double design earthquake (DDE), Hosgri earthquake (HE), loss-of-coolant accident (LOCA), other pipe ruptures (OPR), jet impingement (JI) and asymmetric compartment pressurization loads.
Input loads applied to the primary equipment supports were taken from the results of an integrated reactor coolant loop/support model. Loads from pipe supports and platforms attached to the primary equipment supports were also applied. For seismic self-weight excitation (SWE) of the support structure, the containment internal building structure seismic data (peak accelerations, zero period accelerations (ZPAs) and response spectra) for the DE, DDE and HE earthquakes were applied. The use of ZPAs in lieu of peak accelerations was supported by modal analysis to show that supports behave as rigid structures.
ANSYS and GTSTRUDL finite element analysis computer programs were also used to obtain support stiffness values for the equipment supports.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-32 Revision 23 December 2016 In summary, stresses in all RCS component support members are within the acceptance criteria limits specified in Table 5.2-8 for the DE, DDE and Hosgri seismic events combined as specified in Table 5.2-8.
(3) Integrated Head Assembly (IHA)
The ANSYS general purpose finite element program was used to perform structural analysis of the IHA. The IHA was evaluated for stresses due to
combined loadings of weight (dead load), pressure, thermal, maintenance, missile impact, seismic (DDE or Hosgri) and LOCA. The seismic loading
associated with the DDE and Hosgri was developed as described in
Section 3.7.3.15.4. LOCA loads were applied where the IHA is attached
to the reactor head. Seismic, LOCA, and other loads were combined as
shown in Table 5.2-8a. The resulting loads and stresses for the various components of the IHA were evaluated using the requirements of ASME
Section III, Division I, 2001 Editio n through 2003 Addenda, Subsection NF and Appendix F as shown in Table 5.2-8a.
5.2.2.1.14.2 Time-history Dynamic Solution for Loss-of-Coolant Accident Loading The initial displacement configuration of the mass points is defined by applying the initial steady state hydraulic forces to the RCL model. These initial displacement conditions, natural frequencies, normal modes, the time-history hydraulic forcing functions and
reactor vessel nozzle displacements are used by the WESTDYN program to calculate the time-history dynamic response for the RCL model. The time-history response is used to determine pipe moments, support forces, and pipe deflections.
5.2.2.1.14.3 Analysis Results
All support system elements were evaluated to verify that the supported equipment and piping remain within their respective faulted condition stress limits. Stresses in the
support system elements for faulted conditions are below the limits provided in
Table 5.2-8. Stresses in the piping for faulted conditions are below code-allowable values. Stresses in the PG&E Design Class I members and connections of the IHA for
the specified load conditions meet the acceptance criteria provided in Table 5.2-8a.
5.2.2.1.15 Component Stress Analysis for Faulted Condition Loadings (Double Design Earthquake, Hosgri, and Loss-of-Coolant Accident)
5.2.2.1.15.1 Integrated Reactor Coolant Loop Analysis Stress analysis for faulted condition loadings (DDE, Hosgri, and LOCA) is discussed in Section 5.2.2.1.14.1 for the RCL. Formal analysis of the (LOCA + Hosgri) faulted load combination is in progress. This analysis is being tracked in the DCPP corrective action program.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-33 Revision 23 December 2016 5.2.2.1.15.2 Steam Generator Evaluation The SGs are designed and analyzed in accordance with the ASME Boiler and Pressure Vessel Code of the 1998 Edition through the 2000 Addenda of the ASME Code,Section III. The SG primary side is classified as ASME Code Class 1 (PG&E Quality/Code Class I); the SG secondary side is classified as ASME Code Class 2 (PG&E Quality/Code Class II). The evaluation of the LOCA and seismic (Faulted) conditions are based on meeting the acceptance criteria of ASME III, Appendix F. The design load combinations are identified in Table 5.2-6 and the stress criteria are identified in Table 5.2-7.
Loss-of-coolant primary pipe break hydraulic forcing functions time history were obtained from the LOCA hydraulic forces analysis for the SG. Seismic response spectra, pipe rupture loadings, and pipe nozzle loadings were obtained from applicable design specifications and the RCL analysis for the SG. The Design Limits for Level D (Faulted) Conditions were obtained from Subsection NB and Appendix F, respectively, of the ASME Code. Service Limits were also obtained from the ASME Code.
A finite element based model of the DCPP SGs is used for the dynamic analysis. The model consists of a system of pipe and beam elements, lumped mass elements, and general matrix elements (with both mass and stiffness options). The primary piping stiffness, secondary piping stiffness, external supports stiffness, and some internals stiffness are represented using general matrix elements. The SG dynamic model is coded for use with the ANSYS computer pro gram. ANSYS is a general purpose finite element program with a wide variety of capabilities and an extensive library of finite elements.
A seismic faulted analysis has been performed to predict the response of the SG and its internals to DDE and HE loadings. A linear response spectrum dynamic analysis is used to predict the seismic response of the SG. The seismic analysis of the SGs has been performed using the plant response spectra and external support and piping stiffness for the DCPP Unit 1 and Unit 2 site configuration.
5.2.2.1.15.3 Reactor Coolant Pump Evaluation
The RCPs are designed and analyzed in accordance with the ASME Boiler and Pressure Vessel Code of 1968 Edition through Winter of 1970 Addenda of the ASME Code,Section III. The evaluation of the LOCA and seismic (Faulted) conditions are based on meeting the stress limits in Table 5.2-7. The design load combinations are identified in Table 5.2-6.
The seismic analyses of the RCP were performed using dynamic modal methods with a finite element computer program. The seismic response spectra corresponding to the
elevation of the RCP support structure were used.
The RCP and motor were modeled as a system of nodes and elements (pipes, beams, mass with rotary inertia, springs, fluid elements and stiffness matrices). A modal DCPP UNITS 1 &
2 FSAR UPDATE 5.2-34 Revision 23 December 2016 analysis was performed to determine mode frequencies, mode shapes, and mode participation factors of the pump and motor. The seismic response spectra analyses were performed using the ANSYS finite element computer program. Seismic response spectra, pipe rupture loadings, and pipe nozzle loadings were determined and the RCP evaluated for the faulted conditions.
The LOCA analysis was performed by a nonlinear transient dynamic method using the WECAN computer program. Time histories of the blowdown forces and moments acting on the RCP and the severed cross-over leg were applied to the finite element model of the RCP, the RCP support, and the piping geometry for the SG outlet nozzle break.
The nozzles and support feet of the RCP were analyzed by static stress analysis methods with externally applied design loads.
For the Hosgri spectra the external loads applied to the inlet and outlet nozzles of the RCP by the RCL piping are all below the load for which the nozzles previously were shown acceptable. No further analysis was necessary for the nozzles.
The loads resulting from piping reactions for the Hosgri spectra were lower than the
DDE loads for which the RCP support feet were analyzed. No further analysis was necessary for the support feet.
5.2.2.1.15.4 Reactor Vessel Evaluation The reactor vessel is designed and analyzed in accordance with the ASME BPVC 1965 (Unit 1) through Winter 1966 Addenda and 1968 (Unit 2) Editions of the ASME BPVC Section III. The reactor vessel is classified as ASME Code Class A. The evaluation of the LOCA and seismic (Faulted) conditions are based on meeting the stress limits in Table 5.2-7. The design load combinations are identified in Table 5.2-6.
Loss-of-coolant primary pipe break hydraulic forcing functions were obtained from the LOCA hydraulic forces analysis for the reactor vessel. Seismic response spectra, pipe rupture loadings, and pipe nozzle loadings were determined and evaluated for the faulted condition.
A seismic faulted analysis has been performed to predict the response of the reactor vessel and its internals to DDE and HE loadings. A linear response spectrum dynamic analysis is used to predict the seismic response of the reactor vessel. The seismic analysis of the reactor vessel has been performed using the plant response spectra and external support and piping stiffness for the DCPP Unit 1 and Unit 2 site configuration.
Several portions of the reactor vessel were evaluated using static stress analysis
methods with externally applied design loads.
The CRDM and core exit thermocouple head adapter, closure head flange, vessel flange, closure studs, inlet nozzle, outlet
nozzle, vessel support, vessel wall transition, core barrel support pads, bottom head
shell juncture and bottom head instrumentation penetrations were analyzed by this DCPP UNITS 1 &
2 FSAR UPDATE 5.2-35 Revision 23 December 2016 method. The design loads for all areas evaluated are based on the actual plant loads.
All stresses and fatigue usage factors were found to be acceptable.
5.2.2.1.15.5 Reactor Vessel Internals Evaluation The reactor vessel internals evaluation is presented in Sections 3.7.3.15 and 3.9.2.3.
5.2.2.1.15.6 Fuel Assembly Evaluation The fuel assembly evaluation is presented in Sections 3.7.3.15 and 3.9.2.
5.2.2.1.15.7 Control Rod Drive Mechanism The SYSTUS finite element computer code was used to perform structural analysis of
the replacement CRDM pressure housings. The CRDM pressure housings were
evaluated for stresses due to combined loadings of deadweight, pressure, thermal, seismic, LOCA, and other pipe ruptures as shown in Table 5.2-6a. The seismic
loadings were developed as described in Section 3.7.3.15.3. The combined stresses were determined at each critical location along the length of the CRDM assembly
including locations along the rod t ravel housing, latch housing, and CRDM penetration into the RVCH. The resulting loads and stresses for the CRDM pressure housings were
evaluated using the requirements of ASME Section III, Division I, 2001 Edition through 2003 Addenda, Subsection NB a nd Appendix F as shown in Table 5.2-6a. The results demonstrated the ASME code limits are met for the CRDM pressure housings.
5.2.2.1.15.8 Primary Equipment Support Evaluation
The evaluation of primary equipment supports is presented in Section 5.2.2.1.14.1(2).
5.2.2.1.15.9 Pressurizer Evaluation The pressurizer is designed and analyzed in accordance with the ASME Boiler and Pressure Vessel Code of the 1965 Edition through Summer 1966 Addenda of the ASME Code,Section III. The evaluation of the LOCA and seismic (Faulted) conditions are based on meeting the stress limits in Table 5.2-7. The design load combinations are identified in Table 5.2-6.
A dynamic modal analysis was performed to evaluate the response of the pressurizer and its internals to DDE and Hosgri loadings. The seismic dynamic analysis modeled the heater rods, pressurizer vessel, and vessel support with beam elements and lumped mass elements. Seismic response spectra, pipe rupture loadings, and pipe nozzle loadings were determined and the pressurizer evaluated for the faulted conditions.
Refer to Section 3.7.1.4 for the applicable DDE and Hosgri percent of critical damping values.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-36 Revision 23 December 2016 The Hosgri response spectra for 4 percent damping at the 140 feet elevation has a peak of 5.1 g horizontally, well below the value used to qualify the pressurizer.
A dynamic RCL analysis, which included a surge line model and was performed with the DDE and Hosgri response spectra, produced loads (forces and moments) on the support skirt, surge nozzle, and upper seismic lugs which were evaluated and shown to be acceptable.
5.2.2.1.15.10 Reactor Vessel Closure Head
The RVCH was designed and analyzed in accordance with the ASME Boiler and Pressure Vessel Code Section III, Division I, 2001 Edition through 2003 Addenda, Subsection NB and Appendix F. The RVCH is classified as ASME Code Class 1.
A finite element model of the reactor vessel closure region was used to perform the
dynamic analysis of the RVCH. To adequately analyze the effects of the important
structural items, a 46.67 degree circumferential segment of the reactor vessel closure region was modeled. The ANSYS general purpose finite element program was used to
perform the dynamic analysis of the RVCH model. The RVCH model was evaluated for
stresses due to combined loadings of dead weight, pressure, thermal, seismic (DE, DDE or Hosgri), LOCA, and other pipe ruptures as shown in Table 5.2-6a. The seismic
and LOCA loadings were developed as described in Section 3.9.2.1.3. The resulting loads and stresses for the RVCH were evaluated using the requirements of ASME
Section III, Division I, 2001 Edition through 2003 Adden da, Subsection NB and Appendix F as shown in Table 5.2-6a. The results demonstrated the ASME code limits
are met for the replacement RVCH.
5.2.2.1.15.11 Core Exit Thermocouple Nozzle Assembly The ANSYS finite element computer code was used to perform structural analyses of the core exit thermocouple nozzle assembly (CETNA) pressure boundary components.
The CETNA pressure-retaining components were evaluated for stresses due to combined loadings of deadw eight, pressure, thermal, seismic, and LOCA, as shown in Table 5.2-6a.
The seismic and LOCA loadings were developed using the response spectrum modal superposition method described in Section 3.7.3.15.3. Horizontal and vertical response spectra at the RVCH CG elevation were input to the model. The two orthogonal horizontal direction response spectra were combined using the SRSS method.
Horizontal and vertical seismic responses were combined using SRSS method. Seismic (DDE and HE) and LOCA stresses were combined using the SRSS method. Response spectra critical damping values were analyzed according to welded steel structures in Regulatory Guide 1.61, October 1973.
The combined stresses were determined at each critical location and were evaluated using the requirements of ASME BPVC Section III, Division I, 1989 Edition, Subsection DCPP UNITS 1 &
2 FSAR UPDATE 5.2-37 Revision 23 December 2016 NB. A code reconciliation is performed for the requirements of ASME BPVC Section III, Division I, 2001 Edition through 2003 Addenda. The results demonstrated that the ASME code limits are met fo r the CETNA pressure-retaining components.
5.2.2.1.16 Stress Levels in Reactor Coolant Pressure Boundary Components Sections 5.2.2.1.14 and 5.2.2.1.15 discuss RCS PG&E Quality/Code Class I components and the resulting stress levels under normal, upset and faulted conditions.
5.2.2.1.17 Analytical Methods for Stresses in Pumps and Valves
The design and analysis to ensure structural integrity and operability of the RCPs and valves used the load combinations and stress limits as defined in Table 5.2-5 for stress limits for Class A loop piping and valves, and Tables 5.2-6 and 5.2-7 for the RCPs. As a result, the design and analyses of these components are based on the requirements of various codes and procedures that were in effect when the equipment was
purchased.
These codes and procedures have been widely used by the nuclear industry and were, to a large extent, incorporated or referenced in ASME BPVC Section III-1971 (refer to Section 3.9.2). Every valve and pump is hydrostatically tested to the applicable ASME BPVC requirements, as listed in Table 5.2-2, ensures the integrity of the pressure boundary parts.
5.2.2.1.18 Analytical Methods for Evaluation of Pump Speed and Bearing Integrity RCP overspeed evaluation is covered in Section 5.5.1.
5.2.2.1.19 Operation of Active Valves Under Transient Loadings Operation of active valves under transient loadings is discussed in Sections 3.9.2 and
3.10. Refer to Table 5.2-9 for a listing of active and inactive valves within the RCPB.
HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED During plant startup testing, the preoperational piping dynamics effe cts test program described in Section 3.9.1 will note and correct excessive piping deflections and vibrations. Since all valves are supported as part of adjoining piping, this testing and any required corrective action, will ensure that the deflections by the pipe (and valve) supports will not impair the operability of active PG&E Design Class I valves, including those in the RCS pressure boundary.
5.2.2.2 Overpressurization Protection The pressurizer is designed to accommodate pressure increases (as well as decreases)
caused by load transients. The spray system condenses steam to prevent the DCPP UNITS 1 &
2 FSAR UPDATE 5.2-38 Revision 23 December 2016 pressurizer pressure from reaching the setpoint of the PORVs during a step reduction in power level of 10 percent of load. Flashing of water to steam and generation of steam by automatic actuation of the heaters keeps the pressure above the low-pressure
reactor trip setpoint.
The spray nozzles are located on the top of the pressurizer. Spray is initiated when the
pressure controlled spray demand signal is above a given setpoint. The spray flow increases proportionally with increasing pressure and pressure error until it reaches a
maximum value.
Overpressure protection is discussed in the Sections 5.2.2.2.1 through 5.2.2.2.3.
Protection against overpressurization during low temperature operation is provided by
the LTOP system, which is described in Section 5.2.3.28.
5.2.2.2.1 Location of Pressure-Relief Devices The pressurizer is equipped with three PORVs that limit system pressure for a large
power mismatch and thus prevent actuation of the fixed high-pressure reactor trip. The
relief valves are operated automatically or by remote-manual control. The operation of
these valves also limits the undesirable opening of the spring-loaded safety valves.
Remotely operated block valves are provided to isolate the PORVs if excessive leakage
occurs. The relief valves are designed to limit the pressurizer pressure to a value below
the high-pressure trip setpoint for all design transients, up to and including, the design
percentage step load decrease with steam dump but without reactor trip.
Isolated output signals from the pressurizer pressure protection channels are used for pressure control. These are used to control pressurizer spray and heaters, and PORVs.
In the event of a complete loss of heat sink, protection of the RCS against overpressure (Reference 1) is afforded by pressurizer and SG safety valves along with any of the following reactor trip functions:
(1) Reactor trip on turbine trip (2) Pressurizer high-pressure reactor trip (3) Overtemperature T reactor trip (4) SG low-low water level reactor trip
A detailed functional description of the process equipment associated with the
high-pressure trip is provided in Reference 2.
The overpressure protection upper limit is based on the positive surge of the reactor
coolant produced as a result of turbine trip under full load, assuming the core continues
to produce full power and normal feedwater is maintained. The self-actuated safety DCPP UNITS 1 &
2 FSAR UPDATE 5.2-39 Revision 23 December 2016 valves are sized on the basis of steam flow from the pressurizer to accommodate this surge at a setpoint of 2500 psia and a total accumulation of 3 percent. Each of the
safety valves is rated to carry 420,000 lb/hr, which is greater than one-third of the total
rated capacity of the system. Note that no credit is taken for the relief capability provided by the PORVs during this surge.
The RCS design and operating pressures, together with the safety, power-relief, and
pressurizer spray valve setpoints, and the protection system setpoint pressures are
listed in Table 5.2-10. A schematic representation of the RCS showing the location of
pressure-relieving devices is shown in Figure 3.2-7.
System components whose design pressure and temperature are less than the RCS design limits are provided with overpressure protection devices and redundant isolation
means. System discharge from overpressure protection devices is collected in the PRT in the RCS. Isolation valves are provided at all connections to the RCS. Figures 3.2-8 through 3.2-10 show those systems that communicate directly with the RCS, and all
pressure-relieving devices to prevent reactor coolant pressure from causing overpressure in auxiliary emergency systems in the event of leakage into those
systems.
All pressurizer relief piping was manufactured, installed, and tested in accordance with ANSI B31.7-1969 with 1970 Addenda. The piping from the pressurizer to the relief valves is designed to ANSI B31.1-1967. The valve discharge piping to the PRT is designed to ANSI B31.7-1967 with 1970 Addenda -
Class III piping. Refer to Section 3.9.2.1.7 for piping analysis.
5.2.2.2.2 Mounting of Pressure-Relief Devices The pressurizer safety and relief valve piping system has undergone extensive analysis considering combined loads due to internal pressure, pipe and valve deadweight, thermal growth of the pressurizer, seismic accelerations due to earthquakes, and
hydraulic hammer forces due to operation of the valve and the volume of water in the
water seal at the inlet to the valve.
A vertical loop in the pipe between the pressurizer and the safety valve is provided to
allow for differential thermal growth between the safety valves and the pressurizer.
Previously, the loop provided a water seal against the valve seat to prevent gas and
steam leakage through the valve from damaging the seat. The safety valves have been
modified from a water-seated to a steam-seated design and water in the loop is
continuously drained. The hydraulic hammer analysis was a dynamic time-history type
of analysis taking into account the water seal volume, the valve opening time, the
location and number of bends in the downstream piping, and the lengths of each piece
of straight pipe on the discharge of the valves. Analyses consider combinations of all
three valves open or shut to determine the most highly stressed condition. The
analyses have not been revised to reflect the absence of the water seal volume, DCPP UNITS 1 &
2 FSAR UPDATE 5.2-40 Revision 23 December 2016 resulting in a conservative design since the loads are less severe without the water seal volume.
5.2.2.2.3 Report on Overpressurization Protection The design bases for overpressurization protection of the RCS are discussed in
Section 5.5.9. Additional information is also provided in Section 5.2.3.28.
5.2.2.3 General Material Considerations This section discusses the materials used in the RCS.
5.2.2.3.1 Material Specifications The reactor vessels for Unit 1 and Unit 2 were fabricated to the 1965 Edition through Winter 1966 Addenda for Unit 1 and the 1968 Edition for Unit 2, of the ASME BPVC,Section III.
Materials of construction for the RVCHs meet the requirements of the ASME BPVC,Section III, 2001 Edition with Addenda through 2003.
Materials of construction for the SGs meet the requirements of the 1998 Edition of the ASME BPVC,Section III, with addenda through the 2000 Addenda. SG pressure boundary ferritic material is procured with reference temperature at nil ductility transition (RT NDT) of 0°F.
Materials of construction for the pressurizers for Unit 1 and Unit 2 meet the requirements of the 1965 Edition of the ASME BPVC,Section III, and addenda through the 1966 Addenda. Charpy tests in the major working or rolling direction were performed at 10°F to ensure that the required toughnes s levels were obtained. The fracture toughness of these materials is considered sufficient to ensure a margin of safe
operation.
Pipe is seamless forged stainless steel conforming to ASTM A376, Type 316 with weld
repair limited to 3 percent of nominal wall thickness. Fittings in the main RCLs for both Unit 1 and Unit 2 are cast stainless steel conforming to ASTM A351, Gr. CF8M. The
90-degree elbows are cast in sections and joined by electroslag welds. The cobalt
content is limited to 0.20 percent.
The minimum wall thickness of the pipe and fittings is not less than that calculated using
ASA B31.1, Section 1, formula of paragraph 122 with an appropriate allowable stress
value provided in Nuclear ASA Code Cases N-7 (for piping) and N-10 (for fittings).
The pressurizer surge line pipe conforms to ASTM A-376, Type 316, with
supplementary requirements S2 (transverse tension tests) and S6 (ultrasonic test). The S2 requirements apply to each length of pipe. The S6 requirements apply to 100 DCPP UNITS 1 &
2 FSAR UPDATE 5.2-41 Revision 23 December 2016 percent of the piping wall volume.
The pipe wall thickness for the pressurizer surge line is Schedule 140 for Unit 1 and Schedule 160 for Unit 2. There are two 90-degree elbow fittings in the pressurizer surge line for both Unit 1 and Unit 2. The Unit 1 and Unit 2 surge line 14-inch elbows are wrought stainless steel conforming to ASTM A-403, WP316. The Unit 1 and Unit 2 surge line nozzles are forged stainless steel conforming to ASTM A-182, F316.
Branch nozzles conform to SA-182, Grade F316. Thermal sleeves for Unit 1 and Unit 2 conform to SA-312 or SA-240, Type 316 or 304. The sample scoop conforms to SA-182, Type 316. The pressurizer spray scoop conforms to SA-403, Grade WP 316.
Stainless steel pipe conforms to ANSI B36.19 for sizes 1/2 through 12 inches and wall thickness schedules 40S through 80S. Stainless steel pipe outside of the scope of
ANSI B36.19 conforms to ANSI B36.10, exclusive of the RCL piping of special sizes 27-1/2, 29, and 31 inches I.D. Flanges conform to ANSI B16.5. Socket weld fittings and socket joints conform to ANSI B16.11.
Radiographic or ultrasonic examination was performed throughout 100 percent of the wall volume of each pipe and fitting. Acceptance standards for ultrasonic testing are in
accordance with ASME BPVC Section III, except that the defect standard for acceptance is a Charpy-V notch not exceeding 1 inch in length and 3 percent of wall
thickness in depth. Acceptance standards for radiographic examination are in
accordance with ASTM E-186 Severity Level 2, except that defect Categories D and E are not acceptable.
A liquid penetrant examination was perform ed on both the entire outside and inside surfaces of each finished hot, cold, and crossover loop fitting and pipe in accordance
with the procedure of ASME BPVC Section VIII, Appendix VIII, and the acceptance standards of ASA B31.1, Code Cases N-9 or N-10.
All unacceptable defects were eliminated in accordance with the requirements of
ASME BPVC Section III. All butt welds an d nozzle welds are of a full penetration design; welds 2 inches and smaller are socket-welded joints. The mechanical properties of representative material heats in the final heat treat condition were no less
than 1.20 times the allowable stress tabulated in ASA Code Case N-7 corresponding to
650°F.
Type 308 weld filler material was used for all welding applications to avoid microfissuring. As an option, Type 308L weld filler metal analysis was substituted for
consumable inserts when this technique was used for the weld root closure. All welding was performed in accordance with the ASME BPVC S ection IX. In all welding, except for the RVCH cladding operations, the interpass temperature was limited to 350°F maximum. The methodology used for the RVCH cladding was qua lified using the guidance in Regulatory Guide 1.43, May 1973.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-42 Revision 23 December 2016 5.2.2.3.2 Compatibility with Reactor Coolant The materials of construction of the RCPB were specified to minimize corrosion and
erosion. To avoid the possibility of accelerated erosion, the internal coolant velocity is
limited to about 50 fps.
The reactor vessel is constructed of carbon steel with a 0.125 inch minimum of stainless
steel or Inconel cladding on all internal surfaces that are in contact with the reactor
coolant. The pressurizer is also constructed of carbon steel with austenitic stainless
steel or Inconel (Unit 2 only) cladding on all surfaces exposed to the reactor coolant. All parts of the RCP in contact with the reactor coolant are austenitic stainless steel except for seals, bearings, and secondary seals (O-rings made of elastomer material).
The portions of the SG in contact with the reactor coolant water are clad with austenitic stainless steel. The SG tubesheet is weld clad with Inconel and the heat transfer tubes are made of Inconel. Tables 5.2-11 through 5.2-14 summarize the materials of
construction of these RCS components.
The reactor coolant piping and fittings that make up the loops are austenitic stainless
steel. All smaller piping that comprises part of the RCS boundary, such as the
pressurizer surge line, spray and relief line, loop drains, and connecting lines to other
systems, is also made of austenitic stainless steel. All valves in the RCS that are in
contact with the coolant are constructed primarily of stainless steel. Other materials in contact with the coolant, such as materials for hard surfacing and packing, are special
materials.
5.2.2.3.3 Compatibility with External Ins ulation and Environmental Atmosphere
The materials of construction of the RCPB were specified to ensure compatibility with
the containment-operating environment. All insulation u sed on the RCPB, as defined by the ASME BPVC Section XI, is of the reflective stainless steel type or as described in Section 6.3.3.35. Additional information on the compatibility of RCPB materials with the containment environment to which they are exposed is provided in Section 3.11.
5.2.2.3.4 Chemistry of Reactor Coolant The RCS water chemistry is selected to minimize corrosion. A periodic analysis of the
coolant chemical composition is performed to verify that the reactor coolant quality meets the specifications for coolant chemistry, activity level, and boron concentration.
The CVCS provides a means for adding chemicals to the RCS that control the pH of the coolant during initial startup and subsequent operation, scavenge oxygen from the coolant during startup, and control the oxygen level of the coolant due to radiolysis
during all power operations subsequent to startup.
To ensure thorough mixing, at least one RCP or RHR pump is always in service when chemicals are being added to the system or when changing the boron concentration.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-43 Revision 23 December 2016 The chemical used for pH control is lithium hydroxide. This chemical is chosen for its
compatibility with the materials and water chemistry of the borated water/stainless
steel/zirconium/Inconel system. In addition, lit hium is present in solution from the neutron irradiation of dissolved boron in the coolant. The lithium hydroxide is introduced
into the RCS via the charging flow. The solution is prepared in the plant and poured
into the chemical mixing tank. Reactor makeup water is then used to flush the solution
to the suction manifold of the charging pumps.
The concentration of lithium hydroxide in the RCS is maintained in the range specified
for pH control. If the concentration excee ds this range, a demineralizer is valved in to remove the excess lithium. Since the amount of lithium to be removed is small and its
buildup can be readily calculated and determined by analysis, the flow through the cation bed demineralizer is not required to be full letdown flow.
During reactor startup from the cold condition, hydrazine is employed as an oxygen
scavenging agent. The hydrazine solution is introduced into the RCS using the same
injection flow path as the pH control agent, as described above.
Dissolved hydrogen is employed to control and scavenge oxygen produced due to radiolysis of water in the core region. Sufficient partial pressure of hydrogen is
maintained in the VCT such that the specifie d equilibrium concentration of hydrogen is maintained in the reactor coolant. A self-contained pressure control valve maintains a
minimum pressure in the vapor space of the VCT. This can be adjusted to provide the correct equilibrium hydrogen concentration. The RCS water chemistry specifications
are provided in Table 5.2-15.
5.2.2.4 Fracture Toughness This section addresses fracture toughness in the RCPB. The RCS component upon
which operating limitations are based is the reactor vessel.
5.2.2.4.1 Compliance with Code Requirements Assurance of adequate fracture toughness of the RPV is established using methods to estimate the RT NDT (Reference 5). The fracture toughness properties of the reactor vessel wall material surrounding the irradiated core region are the limiting properties.
The stringent fracture toughness requirements were determined based on the ASME BPVC Section III-1971 Edition, and the 1972 Summer Addenda. The estimated RTNDT uses as a guide the fracture toughness requirements of NB2300 of the Summer 1972
Addenda, which meet the intent of 10 CFR Part 50, Appendix G. For materials not in the beltline region, RT NDT was estimated using methods identifie d in Section 5.3.2 of the NRC Standard Review Plan. The upper-shelf energy level of the material is established
using methods (Reference 5), which are responsive to 10 CFR Part 50, Appendix G.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-44 Revision 23 December 2016 The DCPP Unit 1 and Unit 2 reactor vessels were fabricated to the 1965 Edition through the Winter 1966 Addenda for Unit 1, and the 1968 Edition for Unit 2, of the ASME BPVC Section III. Thus, Charpy impact test orientation was parallel to the working or rolling direction of the base materials. Additional impact tests were performed, however, on the intermediate and lower shell course plates of both vessels. These plates surround
the effective height of the fuel assemblies. Full Charpy test curves were obtained on
these plates from specimens oriented normal to the principal rolling direction. Full
Charpy curves for all the base material in the vessels have been obtained by the
fabricator on impact specimens oriented parallel to the principal working or rolling
direction. Reactor vessel fracture toughness data are provided in Tables 5.2-17A and
5.2-18A, and Tables 5.2-17B and 5.2-18B for Unit 1 and Unit 2, respectively.
The RVCH was manufactured to the requirements of the ASME BPVC Section III, 2001 Edition with Addenda through 2003. Fracture toughness data is provided in Table 5.2-
17B.
Reactor vessel beltline region weld test specimens were taken from weldments
prepared from excess production plate, weld wire, and flux materials. After completion of welding, the weldments were subjected to heat treatment to obtain the metallurgical
effects equivalent to those produced during fabrication of the reactor vessel. The
significant properties (e.g., weld wire chemical composition and weld flux type) of the
weld materials in the beltline region were representative of the actual beltline materials and their fracture toughness. The use of test specimens prepared from excess
production plate, weld wire, and flux materials and subjected to heat treatment satisfies
the intent of the specific requirement of 10 CFR Part 50, Appendix G, Section III.C.2 and ensures an adequate margin of safety.
Two hundred forty bolting material specimens were impact tested at 10°F. The average
of all the impact energy values was 50.5 ft-lb. The lateral expansion was measured on
24 specimens, and an average value of 35 mils was recorded. Fracture energy values
obtained on 90 percent of the 240 specimens tested at 10°F either met or exceeded the
fracture toughness requirements of 10 CFR Part 50, Appendix G. The lowest value of 40 ft-lb. exceeded the special mechanical property requirements of paragraph N-330 of
the 1965 Edition of the ASME BPVC, which states that an average of 35 ft-lb. fracture energy is considered adequate for pressure vessel materials to be pressurized at
ambient temperature (70°F).
5.2.2.4.2 Acceptable Fracture Energy Levels The identification and location of reactor vessel beltline region materials for Unit 1 and Unit 2 are shown in Figures 5.2-1 and 5.2-4, respectively. Chemical composition, NDT, and upper-shelf energy at the end-of-license fluence at the vessel wall 1/4 thickness location for materials in the beltline region are provided in Tables 5.2-18A through 5.2-21B for
Unit 1 and Unit 2.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-45 Revision 23 December 2016 The stresses due to gamma heating in the vessel walls were also calculated and combined with the other design stresses. They were compared with the code-allowable
limit for mechanical plus thermal stress intensities to verify that they are acceptable.
The gamma stresses are low and thus have a negligible effect on the stress intensity in
the vessel.
5.2.2.4.3 Operating Limitations During Startup and Shutdown Allowable pressures as a function of the rate of temperature change and the actual temperature relative to the reactor vessel RT NDT are established according to the methods in the 1972 NDT Summer Addenda of the ASME BPVC Section III, Appendix G. Heatup and cooldown curves are provided in the DCPP Pressure Temperature
Limits Report. The heatup and cooldown curves are based on the estimated RTNDT fracture toughness properties of the reactor vessel materials. Toughness data for the reactor vessel base materials are provided in Tables 5.2-17A and 5.2-17B for Unit 1 and Unit 2, respectively.
Predicted reference temperature at pressurized thermal shock (RT PTS) values and upper shelf energy (USE) are derived in the limiting materials by using the method described in Reference 27, 10 CFR 50.61, 10 CFR Part 50, Appendix G, and the maximum fluence for the applicable service period.
For the Unit 1 end of operating license (EOL) at approximately 35.2 effective full-power years (EFPY) on 11/2/2024, the limiting RTPTS values calculated and their respective 10 CFR 50.61 screening limits are:
- RT PTS (weld 9-442) = 198.7 °F, which is <300 °F circumferential weld limit For the Unit 2 EOL at approximately 35.8 EFPY on 8/26/2025, the limiting RTPTS values calculated and their respective 10 CFR 50.61 screening limits are:
throughout the life of the vessel at 1/4T. For Unit 1, the limiting (minimum) 1/4T USE at EOL is 61.1 ft-lbs. This is predicted to occur for axial weld 3-442C. Similarly, for Unit 2, the limiting (minimum) 1/4T USE at EOL is 56.2 ft-lbs.
This is predicted to occur for axial weld 3-201C.
5.2.2.4.4 Compliance with Reactor Vessel Material Surveillance Program Requirements The toughness properties of the reactor vessel beltline material will be monitored
throughout the service life with a material surveillance program that meets the DCPP UNITS 1 &
2 FSAR UPDATE 5.2-46 Revision 23 December 2016 requirements of 10 CFR Part 50, Appendix H. The original surveillance test program (Reference 11) for DCPP Unit 1 complies with ASTM E 185-70, the standard in effect when the vessel was manufactured. With three exceptions, the program also complies
with ASTM E 185-73. The exceptions are the number of capsules in the program
containing the limiting material, the number of Charpy specimens in each capsule, and
the orientation of the base metal specimens.
A supplemental surveillance program was implemented at the Unit 1 fifth refueling outage to improve the existing program by b ringing the overall surveillance program in better compliance with ASTM E 185-82, provide data for the period beyond which the
original surveillance program was designed, and to provide the necessary data to
demonstrate the effectiveness of reactor vessel thermal annealing. Capsule D from
Unit 1, which was meant to be annealed and reinserted into the reactor vessel, was
removed during 1R12 and is stored in the spent fuel pool. There are currently no
industry plans to anneal reactor vessels. The Unit 1 supplemental surveillance program
is described in References 28 and 29. For Unit 2, the specimen orientation, number, selection procedure, and removal schedule conform to ASTM E 185-73. The
surveillance capsule program for Unit 2 is described in Reference 26.
5.2.2.4.4.1 Program Description
The evaluation of radiation damage is based on pre-irradiation testing of Charpy V-notch and tensile specimens, and post-irradiation testing of Charpy V-notch, and tensile specimens; plus wedge opening loadin g (WOL) fracture mechanics test specimens for Unit 1 and compact tension and bend bar fracture mechanics test specimens for Unit 2. These programs are based on transition temperature and fracture mechanics approaches, and conform with ASTM E-185, Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels and 10 CFR Part 50, Appendix H.
Thermal control specimens are not required since the surveillance specimens will be
exposed to the combined neutron irradiation and temperature effects, and the test
results will provide the maximum transition temperature shift. The surveillance program
for Unit 2 does not include correlation monitors, but the program for Unit 1 does.
Neutron dosimeters included in the capsules can be used to measure exposure
throughout the life of the reactor vessel.
5.2.2.4.4.2 Surveillance Capsules The Unit 1 original reactor vessel surveillanc e program included eight specimen capsules and the supplemental surveillance program consists of four additional specimen capsules. The Unit 2 surveillance program consists of six specimen
capsules. The Type II capsules in Unit 1 and all of the Unit 2 capsules utilize fissionable
materials (uranium-238 and neptunium-237) as neutron dosimeters. The fissionable
materials, in the form of U 3 0 8 and Np0 2 powder, are encapsulated in metal (brass or stainless steel) capsules, which are sealed in steel blocks. The capsules are located in
guide baskets welded to the outside of the thermal shield and neutron shield pads for
Unit 1 and Unit 2, respectively, and are positioned directly opposite the center portion of DCPP UNITS 1 &
2 FSAR UPDATE 5.2-47 Revision 23 December 2016 the core. Sketches showing the location and spacing of the capsules for Unit 1 relative to the core, thermal shield, vessel, and weld seams are shown in Figures 5.2-16 and
5.2-17. Sketches showing the location and spacings of the capsules for Unit 2 are
shown in Figures 5.2-18 and 5.2-19. The capsules can be removed when the vessel
head and upper internals are removed and can be replaced when the lower internals
are removed.
The eight capsules in the Unit 1 original surveillance program contain reactor vessel
steel specimens from the intermediate shell plate or plates located in the core region of the reactor. The three Type II capsules also contain weld metal and heat affected zone
specimens. All of the base material specimens are oriented parallel to the principal
rolling direction. In addition, correlation monitors made from fully documented
specimens of SA-533, Grade B, Class 1 material obtained through Subcommittee II of
ASTM Committee E10, Radioisotopes and Radiation Effects, are inserted in the
capsules of Unit 1 only. The eight capsules contain 27 tensile specimens, 256 Charpy
V-notch specimens (which include weld metal and heat affected zone material), and
42 WOL specimens.
The four supplemental surveillance capsules for Unit 1 contain Charpy impact and
tensile specimens machined from intermediate shell plate 4107-1, and oriented such
that the specimen longitudinal axis is normal (transverse) to the plate principal rolling direction. Shell plate 4107 is the limiting base metal at 48 EFPY. These four capsules
also contain surrogate weld metal specimens obtained from ABB Combustion
Engineering. These surrogate weld specimens were made with the same weld wire
heat (27204) and flux type (Linde 1092) as the Unit 1 reactor vessel limiting weld metal, and are representative of the Unit 1 limiting weld.
The four capsules will also contain various Charpy specimens supplied by Electric Power Research Institute (EPRI) which
will be used to obtain data on the effects of a reactor vessel thermal anneal. Two of the capsules will also contain previously irradiated test material from surveillance capsule S.
This material consists of heat-affected zone and limiting weld metal broken Charpy specimens (which can be reconstituted into testable specimens), and weld metal WOL
specimens. The 4 capsules contain 266 Charpy specimens, 24 tensile specimens, 17 reconstitution blanks from surveillance capsule S tested Charpy specimens, and 2 WOL
specimens.
The six capsules for Unit 2 contain reactor vessel steel specimens oriented both parallel
and normal (longitudinal and transverse) to the principal rolling direction of the limiting
shell plate located in the core region of the reactor and associated weld metal and heat
affected zone metal. The six capsules contain 54 tensile specimens, 360 Charpy V-notch specimens (which include weld metal and heat affected zone material), 72
compact tension specimens, and six bend bar specimens.
Dosimeters including Ni, Co, Fe (Unit 2 only), Co-Al, Cd shielded Co-Al, Cd shielded Np-237, and Cd shielded U-238 are placed in filler blocks drilled to contain the
dosimeters. The dosimeters permit evaluation of the flux seen by the specimens and
the vessel walls. In addition, thermal monitors made of low melting alloys are included DCPP UNITS 1 &
2 FSAR UPDATE 5.2-48 Revision 23 December 2016 to monitor the temperature of the specimens.
The specimens are enclosed in a tight fitting stainless steel sheath to prevent corrosion and ensure good thermal conductivity.
The complete capsule is helium leak tested. Vessel base material sufficient for at least
two capsules will be kept in storage should the need arise for additional replacement test capsules in the program. Sufficient weld metal and heat affected zone material
from Unit 2 for two additional capsules will al so be stored. No additional weld metal or heat affected zone material is available for Unit 1.
As part of the surveillance program, a report of the residual elements in weight percent to the nearest 0.01 percent will be made for surveillance material and as deposited weld
metal. Each of five Type I (base metal only) capsules (T, U, W, X and Z) for Unit 1 contains the following specimens:
Material No. of Charpys No. of Tensiles No. of WOL
Plate No. B4106-1 8 1 2 Plate No. B4106-2 8 1 2 Plate No. B4106-3 8 1 2 ASTM Reference 8 - -
The following dosimeters and thermal monitors are included in each of the five
capsules:
Dosimeters Copper Nickel Cobalt-aluminum (0.15% Co.)
Cobalt-aluminum (cadmium shielded)
Thermal Monitors 97.5% Pb, 2.5% Ag (579°F melting point)
97.5% Pb, 1.75% Ag, 0.75% Sn (590°F melting point)
Each of the three Type II capsules (S, V and Y) for Unit 1 contains the following
specimens:
Material No. of Charpys No. of Tensiles No. of WOL Plate No. B4106-3 8 2 2
Weld Metal (a) 8 2 2 Heat Affected Zone Metal 8 - - (Plate B4106-3)
ASTM Reference 8 - -
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-49 Revision 23 December 2016 (a) Weld fabricated from weld wire heat number 27204 using Linde 1092 Flux Lot No. 3714.
The following dosimeters and thermal monitors are included in each of the three Type II
capsules:
Dosimeters Copper Nickel Cobalt-aluminum (0.15% Co.)
Cobalt-aluminum (cadmium shielded)
Np-237 (cadmium shielded)
Thermal Monitors 97.5% Pb, 2.5% Ag (579°F melting point)
97.5% Pb, 1.75% Ag, 0.75% Sn (590°F melting point)
The four supplemental capsules for Unit 1 contain the following specimens, dosimeters, and thermal monitors:
CAPSULE A (N o t e i) CAPSULE B (N o t e i) CAPSULE C (N o t e i) CAPSULE D (N o t e ii) Charpy Tension Charpy Tension WOL Charpy Tension Charpy Tension Weld Metal (Surrogate 27204) 15 3 15 3 30 3 15 3 Base Metal (Plate 4107-1) 15 3 15 3 15 3 15 3 Correlation Monitor (HSST-02 Plate) 12 8 Capsule S Weld Metal (Original 27204) 9 (Note iii) 2 8 (Note iii)
EPRI Specimens 30 35 46 Notes:
(i) Dosimeter wires: copper, iron, nickel and aluminum-0.15% cobalt (cadmium shielded and unshielded) Fission dosimeters: neptunium-237 (cadmium oxide shielded), and uranium 238 (cadmium oxide shielded) Thermal monitors: 97.5% Pb, 2.5% Ag (579°F melt point), 97.5% Pb, 1.75% Ag, 0.75% Sn (590°F melt point) (ii) Capsule D will contain the following dosimeters:
Dosimeter wires: copper, iron, nickel and aluminum-0.15% cobalt (gadolinium shielded and unshielded)
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-50 Revision 23 December 2016 Fission dosimeters: neptunium-237 (gadol inium shielded) and uranium 238 (gadolinium shielded) Thermal monitors: will not be provided because annealing temperature will exceed the melting point of thermal monitors (iii) Broken weld metal and heat-affected zone Charpy specimens from capsule S, suitable for reconstitution Each of the six capsules for Unit 2 will contain the following specimens:
No. of No. of No. of No. of Material Charpys Tensiles Cts Bend Bars Plate B5454-1 (a) 15 3 4 Plate B5454-1 (b) 15 3 4 1 Weld Metal (c) 15 3 4 Heat Affected Zone Metal (Plate B5454-1) 15
(a) Specimens oriented parallel to the principal rolling direction (longitudinal). (b) Specimens oriented normal to the principal rolling direction (transverse). (c) Weld fabricated from weld wire heat numbers 21935 and 12008 using Linde 1092 Flux Lot No. 3869.
The following dosimeters and thermal monitors are included in each of the six capsules:
Dosimeters Iron Copper Nickel Cobalt-aluminum (0.15% Co)
Cobalt-aluminum (cadmium shielded)
NP-237 (cadmium shielded)
Thermal Monitors 97.5% Pb, 2.5% Ag (579°F melting point)
97.5% Pb, 1.75% Ag, 0.75% Sn (590°F melting point)
The fast neutron exposure of the specimens occurs at a faster rate than that experienced by the vessel wall with the specimens being located between the core and the vessel.
Since these specimens experience accelerated exposure and are actual samples from the materials used in the vessel, the changes in material properties are representative of
the vessel at a later time in life.
Data from the fracture toughness specimens (WOL, compact tension, and bend bar) are expected to provide additional information for use in determining fracture toughness for irradiated material.
The reactor vessel surveillance capsules for Unit 1 are shown in Figure 5.2-16 and in
Figure 5.2-18 for Unit 2.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-51 Revision 23 December 2016 Correlation between calculations and measurements on the irradiated samples in the capsules, assuming the same neutron spectrum at the samples and the vessel inner wall, is described in Section 5.2.2.4.4.5 and has indicated good agreement. The degree to which the specimens perturb the fast neutron flux and energy distribution is
considered in the evaluation of the surveillance specimen data. The integrated flux
calculations at the vessel wall are adjusted u sing the surveillance data to provide best estimate fluence values. The calculated maximum EOL fast neutron exposure at the vessel wall is 1.43x10 19 n/cm 2 and 1.68x10 19 n/cm 2 (E > 1 MeV) for Unit 1 and Unit 2, respectively.
5.2.2.4.4.3 Capsule Removal For Unit 1 and Unit 2, the removal schedule conforms to 10 CFR Part 50, Appendix H.
The schedule for removal of the Unit 1 and Unit 2 capsules is provided in Table 5.2-22.
5.2.2.4.4.4 Measurement of Integrated Fast Neutron (E
> 1.0 MeV) Flux at the Irradiation Samples
The use of passive neutron sensors such as those included in the internal surveillance
capsule dosimetry sets does not yield a direct measure of the energy-dependent neutron flux level at the measurement location. Rather, the activation or fission process
is a measure of the integrated effect that the time-and energy-dependent neutron flux
has on the target material over the course of the irradiation period. An accurate
assessment of the average flux level and, hence, time integrated exposure (fluence) experienced by the sensors may be developed from the measurements only if the
sensor characteristics and the parameters of the irradiation are well known.
In particular, the following variables are of interest:
(1) the measured specific activity of each sensor (2) the physical characteristics of each sensor (3) the operating history of the reactor (4) the energy response of each sensor (5) the neutron energy spectrum at the sensor location
In this section, the procedures used to determine sensor-specific activities, to develop
reaction rates for individual sensors from the measured specific activities and the
operating history of the reactor, and to derive key fast neutron exposure parameters
from the measured reaction rates are described.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-52 Revision 23 December 2016 5.2.2.4.4.4.1 Determination of Sensor Reaction Rates The specific activity of each of the radiometric sensors is determined using established
ASTM procedures. Following sample preparation and weighing, the specific activity of
each sensor is determined by means of a lithium-drifted germanium, Ge(Li), gamma
spectrometer. In the case of the surveill ance capsule multiple foil sensor sets, these analyses are performed by direct counting of each of the individual wires or, as in the
case of U-238 and Np-237 fission monitors, by direct counting preceded by dissolution
and chemical separation of cesium from the sensor.
The irradiation history of the reactor over its operating lifetime is obtained from
NUREG-0020, "Licensed Operating Reactors Status Summary Report," or from other
plant records. In particular, operating data are extracted on a monthly basis from
reactor startup to the end of the capsule irradiation period. For the sensor sets utilized in the surveillance capsule irradiations, the half-lives of the product isotopes are long
enough that a monthly histogram describing reactor operation has proven to be an
adequate representation for use in radioactive decay corrections for the reactions of
interest in the exposure evaluations.
Having the measured specific activities, the operating history of the reactor, and the
physical characteristics of the sensors, reaction rates referenced to full power operation
are determined from the following equation:
R ANFYCle o j d=j jref j P P t e t (5.2-1) where: A = measured specific activity (dps/gm)
R = reaction rate averaged over the irradiation period and referenced to operation at a core power level of P ref (rps/nucleus)
N o = number of target element atoms per gram of sensor F = weight fraction of the target isotope in the sensor material Y = number of product atoms produced per reaction
P j = average core power level during irradiation period j (MW)
P ref = maximum or reference core power level of the reactor (MW)
C j = calculated ratio of (E > 1.0 MeV) during irradiation period j to the time weighted averaged (E > 1.0 MeV) over the entire irradiation period = decay constant of the product isotope (sec
-1) t j = length of irradiation period j (sec) t d = decay time following irradiation period j (sec)
and the summation is carried out over the total number of monthly intervals comprising
the total irradiation period.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-53 Revision 23 December 2016 In the above equation, the ratio P j/P ref accounts for month-by-month variation of power level within a given fuel cycle. The ratio C j is calculated for each fuel cycle and accounts for the change in sensor reaction rates caused by variations in flux level due to changes in core power spatial distributions from fuel cycle to fuel cycle. For a single
cycle irradiation C j = 1.0. However, for multiple cycle irradiations, particularly those employing low leakage fuel management, the additional C j correction must be utilized.
5.2.2.4.4.4.2 Corrections to Reaction Rate Data Prior to using the measured reaction rates in the least squares adjustment procedure
discussed in Section 5.2.2.4.4.4.3, additional corrections are made to the U-238 measurements to account for the presence of U-235 impurities in the sensors as well as
to adjust for the build-in of plutonium isotopes over the course of the irradiation.
In addition to the corrections made for the presence of U-235 in the U-238 fission
sensors, corrections are also made to both the U-238 and Np-237 sensor reaction rates
to account for gamma ray induced fission reactions occurring over the course of the
irradiation.
5.2.2.4.4.4.3 Least Squares Adjustment Procedure Values of key fast neutron exposure parameters are derived from the measured
reaction rates using the FERRET least squares adjustment code (Reference 12). The
FERRET approach uses the measured reaction rate data, sensor reaction cross-
sections, and a calculated trial spectrum as input and proceeds to adjust the group
fluxes from the trial spectrum to produce a best fit (in a least squares sense) to the measured reaction rate data. The "measured" exposure parameters along with the associated uncertainties are then obtained from the adjusted spectrum.
In the FERRET evaluations, a log-normal least squares algorithm weights both the trial
values and the measured data in accordance with the assigned uncertainties and correlations. In general, the measured values f are linearly related to the flux by some response matrix A:
i s g(,)()f=A ig(s)g (5.2-2) where i indexes the measured values belonging to a single data set s, g designates the energy group, and delineates spectra that may be simultaneously adjusted. For example, iigg R s= (5.2-3)
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-54 Revision 23 December 2016 relates a set of measured reaction rates R i to a single spectrum g by the multigroup reaction cross-section ig. The log-normal approach automatically accounts for the physical constraint of positive fluxes, even with large assigned uncertainties.
In the least squares adjustment, the continuous quantities (i.e., neutron spectra and cross-sections) are approximated in a multigroup format consisting of 53 energy groups.
The trial input spectrum is converted to the FERRET 53 group structure using the
SAND-II code (Reference 13). This procedure is carried out by first expanding the 47
group calculated spectrum into the SAND-II 620 group structure using a SPLINE
interpolation procedure in regions where group boundaries do not coincide. The
620-point spectrum is then re-collapsed into the group structure used in FERRET.
The sensor set reaction cross-sections, obtained from the ENDF/B-VI dosimetry file (Reference 14), are also collapsed into the 53 energy group structure using the SAND-II
code. In this instance, the trial spectrum, as expanded to 620 groups, is employed as a
weighting function in the cross-section collapsing procedure. Reaction cross-section
uncertainties in the form of a 53 x 53 covariance matrix for each sensor reaction are
also constructed from the information contained on the ENDF/B-VI data files. These
matrices include energy group-to-energy group uncertainty correlations for each of the
individual reactions.
Due to the importance of providing a trial spectrum that exhibits a relative energy distribution close to the actual spectrum at the sensor set locations, the neutron
spectrum input to the FERRET evaluatio n is obtained from plant-specific calculations for each dosimetry location. While the 53 x 53 group covariance matrices applicable to the
sensor reaction cross-sections are developed from the cross-section data files, the
covariance matrix for the input trial spectrum is constructed from the following relation:
gg'n 2gg'gg'MRRRP=+ (5.2-4) where R n specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the set of values. The fractional uncertainties, R g , specify additional random uncertainties for group g that are correlated with a correlation matrix given by:
gg'gg'P 1 e[]H=+ (5.2-5) where:
H=(')gg 2 2 2 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes short-range correlations over a group range DCPP UNITS 1 &
2 FSAR UPDATE 5.2-55 Revision 23 December 2016
( specifies the strength of the latter term). The value of is 1 when g = g' and 0 otherwise.
5.2.2.4.4.5 Calculation of Integrated Fast Neutron (E > 1.0 MeV) Flux at the Irradiation Samples Fast neutron exposure calculations for the reactor geometry are carried out using both
forward and adjoint discrete ordinates transport techniques. A single forward
calculation provides the relative energy distribution of neutrons for use as input to
neutron dosimetry evaluations as well as for use in relating measurement results to the
actual exposure at key locations in the pressure vessel wall. A series of adjoint
calculations, on the other hand, establishes the means to compute absolute exposure
rate values using fuel cycle-specific core power distributions, thus providing a direct
comparison with all dosimetry results obtained over the operating history of the reactor.
In combination, the absolute cycle-specific data from the adjoint evaluations together
with relative neutron energy spectra distributions from the forward calculation provided
the means to:
(1) Evaluate neutron dosimetry from surveillance capsule locations.
(2) Enable a direct comparison of analytical prediction with measurement.
(3) Determine plant-specific bias factors to be used in the evaluation of the best estimate exposure of the RPV.
(4) Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves.
5.2.2.4.4.5.1 Reference Forward Calculation The forward transport calculation for the reactor is carried out in r, geometry using the DORT two-dimensional discrete ordinates code (Reference 15) and the BUGLE-93 cross-section library (Reference 16). The BUGLE-93 library is a 47 neutron group, ENDFB-VI based, data set produced specifically for light water reactor applications. In
these analyses, anisotropic scattering is treated with a P 3 expansion of the scattering cross-sections and the angular discretization is modeled with an S 8 order of angular quadrature. The reference forward calculation is normalized to a core midplane power
density characteristic of operation at the s tretch rating for the reactor.
The spatial core power distribution utilized in the reference forward calculation is
derived from statistical studies of long-term operation of Westinghouse 4-loop plants.
Inherent in the development of this reference core power distribution is the use of an
out-in fuel management strategy, i.e., fresh fuel on the core periph ery. Furthermore, for the peripheral fuel assemblies, a 2 uncertainty derived from the statistical evaluation of plant-to-plant and cycle-to-cycle variations in peripheral power is used.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-56 Revision 23 December 2016 Due to the use of this bounding spatial po wer distribution, the results from the reference forward calculation establish conservative ex posure projections for reactors of this design operating at the stretch rating.
Since it is unlikely that actual reactor operation would result in the implementation of a power distribution at the nominal +2 level for a large number of fuel cycles and, further, because of the widespread implementation of
low leakage fuel management strategies, the fuel cycle-specific calculations for this
reactor will result in exposure rates well below these conservative predictions.
5.2.2.4.4.5.2 Cycle Specific Adjoint Calculations All adjoint analyses are also carried out using an S 8 order of angular quadrature and the P 3 cross-section approximation from the BUGLE-93 library. Adjoint source locations are chosen at several key azimuths on the pressure vessel inner radius. In addition, adjoint
calculations were carried out for sources positioned at the geometric center of all surveillance capsules. Again, these calculations are run in r, geometry to provide neutron source distribution importance functions for the exposure parameter of interest;
in this case, (E > 1.0 MeV).
The importance functions generated from these individual adjoint analyses provide the
basis for all absolute exposure projections and comparison with measurement. These
importance functions, when combined with cycle-specific neutron source distributions, yield absolute predictions of neutron exposure at the locations of interest for each of the operating fuel cycles, and establish the means to perform similar predictions and
dosimetry evaluations for all subsequent fuel cycles.
Having the importance functions and appropriate core source distributions, the
response of interest can be calculated as:
(R 0 , 0) = r E I (r, , E) S(r, , E) r dr d dE (5.2-6) where:
(R 0 , 0) = Neutron flux (E > 1.0 MeV) at radius R 0 and azimuthal angle 0 I (r, , E) = Adjoint importance function at radius r, azimuthal angle , and neutron source energy E
S (r, , E) = Neutron source strength at core location r, and energy E
It is important to note that the cycle-specific neutron source distributions, S(r,,E), utilized with the adjoint importance functions, I(r,,E), permit the use not only of fuel cycle-specific spatial variations of fission rates within the reactor core, but also allow for the inclusion of the effects of the differing neutron yield per fission and the variation in
fission spectrum introduced by the build-in of plutonium isotopes as the burnup of
individual fuel assemblies increases.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-57 Revision 23 December 2016 5.2.2.4.5 Reactor Vessel Annealing HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED There are no special design features that would prohibit the onsite annealing of the vessel. In the event that an annealing opera tion should be required to restore the properties of the vessel material opposite the reactor core because of neutron
irradiation damage, a metal te mperature of approximately 850°F maximum for a period of 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> ma ximum would be applied.
A plan for conducting the thermal annealing must be submitted in accordance with 10 CFR 50.4 at least three years prior to the date at which the limiting fracture toughness criteria in 10 CFR 50.61 or 10 CFR Part 50, Appendix G would be exceeded.
5.2.2.4.6 Loss-of-Coolant Accident Thermal Transient HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED In the event of a large LOCA, the RCS rapidly depressurizes and the loss of coolant
may empty the reactor vessel. If t he reactor is at normal operating conditions before the accident, the reactor vessel temperature is approximately 550°F, and, if the plant has
been in operation for some time, part of the reactor vessel is irradiated. At an early stage in the depressurization transient, the ECCS rapi dly injects cold coolant into the reactor vessel. This produces a thermal stress in the vessel wall. To evaluate the
effect of the stress, three possible modes of failure are considered; ductile yielding, brittle fracture, and fatigue.
(1) Ductile Mode The failure criterion used for this evaluation is that there shall be no gross
yielding across the vessel wall using the material yield stress specified in the ASME BPVC Section III. The combined pressure and thermal stresses during injection through the vessel thickness as a function of time
have been compared to the material yield stress during the SI transient.
The results of the analyses showed that local yielding may occur only in
approximately the inner 18 percent of the base metal and in the vessel
cladding, complying with the above criterion.
(2) Brittle Mode The possibility of brittle fracture of the irradiated reactor vessel core region
has been considered utilizing fracture mech anics concepts. This analysis takes into account the effects of water temperature, heat transfer
coefficients, and fracture toughness as a function of time, temperature, and irradiation. Both a local crack effect and a continuous crack effect DCPP UNITS 1 &
2 FSAR UPDATE 5.2-58 Revision 23 December 2016 have been considered, with the latter requiring the use of a rigorous finite element axisymmetric code. On t he weight of this evidence, the thermal shock resulting from the LOCA will not produce instability in the vessel
wall even at the end of plant life.
(3) Fatigue Mode The failure criterion used for fatigue analysis was based on the ASME
BPVC Section III. In this method the piece is assumed to fail once the combined usage factor at the m ost critical location for all transients applied to the vessel exceeds the code-allowable usage factor of 1.
The location in the vessel below the nozzle level, which will see the emergency core cooling water and have the highest usage factor will be
the incore instrumentation tube attachment welds to the vessel bottom head. As a worst case assumption, the incore instrumentation tubes and
attachment penetration welds are considered to be quenched to the
cooling water temperature while the vessel wall maintains its initial temperature before the start of the transient.
The maximum possible pressure stress during the transient is also taken
into account. This method of analysis is quite conservative and yields
calculated stresses greater than would actually be experienced. The
resulting usage factor for the instru ment tube welds considering all
operating transients and including the SI transient occurring at the end of the plant life is below 0.2, which compares favorably with the code-allowable usage factor of 1.
It is concluded from the results of these analyses that the delivery of cold emergency
core cooling water to the reactor vessel follo wing a LOCA does not cause any loss of integrity of the vessel.
5.2.2.5 Austenitic Stainless Steel The unstabilized austenitic stainless steel materials used in the RCPB, in systems
required for reactor shutdown, and for emergency core cooling, are processed and
fabricated using established methods and techniques to avoid partial or local
sensitization. The measures taken to avoid sensitization are in general conformance
with the recommendations of Regulatory Guide 1.44, May 1973 (Reference 22).
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-59 Revision 23 December 2016 5.2.2.5.1 Cleaning and Contamination Protection Procedures All materials are cleansed and protected by procedures that guard against contaminants capable of causing stress corrosion cracking during storage, fabrication, shipment, erection, testing and operation. Contaminant concentration limits are implemented per
plant approved procedures.
5.2.2.5.2 Solution Heat Treatment Requirements Whenever applicable, solution heat treatment of materials prior to fabrication or assembly into components or systems is discussed in Section 5.2.2.5.5 below. In such cases, solution heat treatment conformed to the requirements of Regulatory Guide 1.44, May 1973.
5.2.2.5.3 Material Inspection Program Austenitic stainless steel materials are procured from raw material produced in the final
heat-treated condition as required by the respective ASTM or ASME material
specification for the particular type or grade of alloy.
Westinghouse-furnished wrought austenitic stainless steel alloy materials are corrosion
tested in the final heat-treated condition. These tests are performed in accordance with
ASTM A262.
5.2.2.5.4 Unstabilized Austenitic Stainless Steel Unstabilized austenitic stainless steel used in componen ts of the RCPB are as follows:
(1) Reactor Vessel (a) (Unit 1) Primary noz zle safe-ends - Type 316 stainless steel forgings.
(Unit 2) Primary nozzle safe-ends - Type 316 stainless steel forgings overlaid with weld metal after final post-weld heat
treatment.
(2) Steam Generators Primary nozzle safe-ends - Grade F316LN forging.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-60 Revision 23 December 2016 (3) Pressurizers Unit 1 Unit 2 (a) Surge nozzle safe-end Type 316 forging Type 316L forging (b) Spray nozzle safe-end Type 316 forging Type 316L forging (c) Relief nozzle safe-end Type 316 forging Type 316L forging (d) Safety valve (3) nozzle Type 316 forging Type 316L forging safe-end 5.2.2.5.5 Avoidance of Sensitization Methods and material techniques used to avoid partial or local severe sensitization are
as follows:
(1) Core Structural Components In all cases where austenitic stainless steel must be given a stress-
relieving treatment above 800°F, a high-temperature stabilizing procedure
was used. This is performed in the temperature range of 1600-1900°F, with holding time sufficient to achieve chromium diffusion to the grain
boundary regions. Proof that such stabilization is achieved is based on
ASTM A393.
(2) Stainless Welding (a) Nozzle safe-ends
- 1. Weld deposit with Ni-Cr-Fe Weld Metal F-Number 43 and attach austenitic stainless steel safe-end after final post-weld
heat treatment.
- 2. Use of a stainless steel weld metal analysis A-7 containing less than 0.02 percent carbon or more than 5 percent ferrite, or both. (b) All welding is conducted using procedures that are in accordance with the ASME BPVC Section IX.
(c) All welding procedures and welders have been qualified to the ASME BPVC rules of Section IX.
When these welding procedure tests are performed on test welds
made from base metal and weld metal materials that are from the same lot(s) of materials used in the fabrication of components, additional testing is frequently required to determine the
metallurgical, chemical, physical, corrosion, etc., characteristics of
the weldment. The additional tests conducted on a technical case DCPP UNITS 1 &
2 FSAR UPDATE 5.2-61 Revision 23 December 2016 basis are as follows: light and electron microscopy, elevated temperature mechanical properties, chemical check analysis, fatigue tests, intergranular corrosion tests or static and dynamic
corrosion tests within reactor water chemistry limitations.
(d) The interpass temperature of all welding methods is limited to 350°F maximum, with the exce ption of the RVCH cladding operations. The methodology used for the RVCH cladding weld operations was qualified using the guidance in Regulatory Guide 1.43, May 1973.
(e) Travel speed, voltage, amperage, as well as thickness of weld metal layers, and degree of weaving (two electrode diameters or ID
of gas cup maximum) are carefully controlled on all welding
processes to minimize sensitization in the completed welds.
(f) All welds are nondestructively examined in accordance with code requirements.
(g) Code-authorized inspectors are required to review and sign off on all welding done both in the shop and field.
(h) For the SGs, ferrite level is 5-18 percent, calculated by WRC sketch. (3) Hard Facing All hard facing procedures on austenitic stainless steel use low (less than
800°F) preheat temperatures to preclude sensitization of the base metal.
Processes approved are limited to those proven by tests not to cause
sensitization.
(4) Bent Pipe Sections Bent pipe sections are solution he at-treated to produce nonsensitized conditions in the material after bending; this is done by controlling
handling temperatures and water quenching time to ensure that all
carbides are in solution.
5.2.2.5.6 Retesting Unstabilized Austenitic Stainless Steel Exposed to Sensitizing Temperatures
It is not normal Westinghouse practice to expose unstabilized austenitic stainless steels to the sensitization range of 800 to 1500°F during fabrication into components except as
described in Section 5.2.2.5.5. If, during the course of fabrication, the steel is inadvertently exposed to the sensitization temperature range, 800 to 1500°F, the DCPP UNITS 1 &
2 FSAR UPDATE 5.2-62 Revision 23 December 2016 material may be tested in accordance with A262 to verify that it is not susceptible to intergranular attack. Testing is not required for:
(1) Cast metal or weld metal with a ferrite content of 5 percent or more.
(2) Material with a carbon content of 0.03 percent or less that is subjected to temperatures in the range of 800 to 1500°F for less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
(3) Material exposed to special processing provided the processing is properly controlled to develop a uniform product and provided that
adequate documentation exists of service experience and/or test data to
demonstrate that the processing will not result in increased susceptibility to intergranular stress corrosion.
If it was verified that such material was susceptible to intergranular attack, the material
would have been solution anneale d again and water quenched or rejected.
5.2.2.5.7 Control of Delta Ferrite Welding of austenitic stainless steel was controlled to mitigate the occurrence of
microfissuring or hot cracking in the weld. Although published data and experience
have not confirmed that fissuring is detrimental to the quality of the weld, it is recognized
that such fissuring is undesirable in a general sense. Also, the presence of delta ferrite
is one of the mechanisms for reducing the susceptibility of stainless steel welds to hot
cracking.
The scope of these controls encompassed welding processes used to join stainless steel parts in components designed, fabricated, or stamped in accordance with the
ASME BPVC Section III, Classes 1, 2, and core support components. Delta ferrite control was appropriate for the above welding requirements except where no filler metal was used if for other reasons such control was not applicable. These exceptions included electron beam welding, autogenous gas shielded tungsten arc welding, explosive welding, and welding using fully austenitic welding materials.
In accordance with Section III, f abrication and installation specifications required welding procedure and welder qualification and included delta ferrite determinations for the austenitic stainless steel welding materials used for welding qualification testing and for production processing. Specifically, the undiluted weld deposits of the "starting" welding materials were required to contain a minimum of 5 percent delta ferrite as
determined by chemical analysis and calcu lation using the appropriate weld metal constitution diagrams in Section III. New welding procedure qualification tests were evaluated for these applications in accordance with the requirements of Sections III
and IX.
The results of all the destructive and nondestructive tests were reported in the
procedure qualification record in additio n to the information required by Section III.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-63 Revision 23 December 2016 The "starting" welding materials used for fabr ication and installation welds of austenitic stainless steel materials and components meet the requirements of Section III. Welding
materials were tested using the welding ener gy inputs to be employed in production welding.
Combinations of approved heats and lots of starting welding materials were used for all
welding processes. The welding q uality assurance program included identification and control of welding material by lots and heats as appropriate. All of the weld processing
was monitored according to approved inspection programs, including review of starting
materials, qualification records, and weldin g parameters. Welding systems are also subject to quality assurance audit including calibration of gauges and instruments;
identification of starting and completed materials; welder and procedure qualifications;
availability and use of approved welding and heat treating procedures; and
documentary evidence of compliance with materials, welding parameters, and
inspection requirements. Fabrication and inst allation welds were inspected using nondestructive examination methods according to Section III rules.
5.2.3 SAFETY EVALUATION 5.2.3.1 General Design Criterion 2, 1967 - Performance Standards Protection provided for the RCS against environmental factors is discussed in Sections 3.3, 3.4, 3.5 and 3.6. In the PG&E quality group classification of DCPP fluid systems and fluid system components, the vessels, piping, valves, pumps and their supports of the RCPB are designated as PG&E Design Class I, PG&E Quality/Code Class I. The RCPB is designed to the requirements of DE, DDE and Hosgri as described in Sections 5.2.2.1.14 and 5.2.2.1.15.
As discussed in Sections 3.9.2.1, 3.9.2.2, a nd 3.9.2.3, the PG&E Design Class I mechanical systems and components are designed to withstand the effects of earthquakes. PG&E Design Class I mechanical systems and components are protected from the effect of winds and tornadoes (refer to Section 3.3), floods and tsunamis (refer to Section 3.4), and external missiles (refer to Section 3.5), ensuring their design function can be performed.
5.2.3.2 General Design Criterion 4, 1987 - Environmental and Dynamic Effects Design Bases The LBB methodology was ap plied to the primary loops of DCPP Unit 1 and Unit 2. The following postulated breaks were eliminated: th e six terminal ends in the cold, hot, and crossover legs; a split in the SG inlet elbow; and the loop closure weld in the crossover leg. Protection from the dynamic effects of the most limiting breaks of auxiliary branch lines needs to be considered. This includes RCS branch line breaks and other high energy line breaks as described in Sections 5.2.2.1.9, 5.2.2.1.10, 5.2.2.1.11, 5.2.2.1.14, 5.2.2.1.15, and 5.2.2.1.16.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-64 Revision 23 December 2016 RCS leakage detection and monitoring is discussed in Section 5.2.3.6 for General Design Criterion 16, 1967 and Section 5.2.3.23 for Regulatory Guide 1.45, May 1973.
5.2.3.3 General Design Criterion 9, 1967 - Reactor Coolant Pressure Boundary The RCPB is designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its lifetime.
The RCPB is designed to accommodate the syst em pressures and temperatures attained u nder all expected modes of plant operation, including all anticipated transients, and to maintain the stresses within applicable stress limits.
Refer to Section 5.2.2.1, Design of Reactor Coolant Pressure Boundary Components.
The RCPB is protected from ove rpressure by means of pre ssure-relieving devices as required by applicable codes. Refer to Section 5.2.2.2, Overpressurization Protection.
The materials of construction of the pressure-retaining boundary of the RCS are protected by control of coolant chemistry from corrosion that mi ght otherwise re duce the system structural integrity during its s ervice lifetime. Refer to Sec tion 5.2.2.3, General Material Considerations.
Generic Letter 87-06, March 1987, applies to the RCPB. PIVs are defined for each interface as any two valves in series within the RCPB which separate the high pressure RCS from an attached low pressure system. These valves are normally closed during power operation. The PIVs are tested for leak tight integrity per Technical Specification 3.4.14. Based on NRC Bulletin 88-11, December 1998, thermal stratification phenomenon could occur in the surge line and may invalidate the analyses supporting the integrity of the surge line with respect to unexpected bending and thermal striping (rapid oscillation of the thermal boundary interface along the piping inside surface) as they affect the overall integrity of the surge line for its design life (e.g., the increase of fatigue).
Consistent with assumptions used and results obtained in the analysis, operating restrictions limit the pressurizer/hot leg differential temperature to 300°F.
5.2.3.4 General Design Criterion 11, 1967 - Control Room Instrumentation and controls necessary to ensure the integrity of the RCPB are provided in the control room. This instrumentation and controls consist of RCPB leakage detection, pressure boundary valve position indication, and post-accident RCS pressure indication.
Refer to Section 5.2.3.23 for further discussion on operational conditions which may indicate changes in RCPB leakage rates.
The RCPB leakage detection instrumentation provided in the control room is listed on Table 5.2-16.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-65 Revision 23 December 2016 The PORV and PSV position indication and controls are provided in the control room.
The PSV position indication system provides the necessary information in the control room to determine the position (open/close) of each of the three PSVs. Refer to Section 7.5.2.8 for details. Also the PORV block valves can be controlled from the control room to isolate the PORV if leaking. Temperature indication in the discharge piping of the PSV and PORVs is provided to identify leaka ge. RHR isolation valve control, valve position indication, and annunciation are provided in the control room as described in Section 7.6.2.1.
The instrumentation required for post-accident monitoring of the RCPB is discussed in Section 5.2.3.24.
Emergency close controls for the PORVs are provided on the HSP in addition to control from the control room (refer to Figure 7.3-21). Indication of RCS pressure is provided by the pressurizer pressure indication located at the HSP.
5.2.3.5 General Design Criterion 12, 1967 - Instrumentation and Controls Instrumentation and controls are provided to monitor and maintain the RCPB.
Valve position indication is provided for RCPB remotely operated valves listed on Table 5.2-9. RHR isolation valve control, valve position indication, and annunciation are provided. PSV and PORV control and ind ication are provided as discussed in Section 5.2.3.4. As described in Section 5.2.3.28, the LTOP function is completely automatic after being manually enabled. Whenever the system is enabled and reactor coolant temperature is below the low temperature setpoint (i.e., PORV arming temperature), a high-pressure signal will trip it automatically and open the PORV until the pressure drops below the reset value.
Regulatory Guide 1.45, May 1973, describes requirements for instruments available for implementing leakage detection systems for the RCPB. Refer to the discussion of reactor coolant leakage requirements in Section 5.2.3.23 and Table 5.2-16.
Instrumentation is provided to monitor RCS integrity following an accident.
Instrumentation related to the RCPB which is required to meet Regulatory Guide 1.97, Revision 3, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, is discussed in Section 5.2.3.24.
5.2.3.6 General Design Criterion 16, 1967 - Monitoring Reactor Coolant Pressure Boundary RCPB components are designed, fabricated, inspected, and tested to the ASME codes and conditions as summarized in Tables 5.2-2 through 5.2-8. Leakage is detected by an DCPP UNITS 1 &
2 FSAR UPDATE 5.2-66 Revision 23 December 2016 increase in the amount of makeup water required to maint ain a normal level in the pressurizer. The reactor vessel closure joi nt is provided with a temperature monitored leakoff between double gasket
- s. Leakage into the reactor containment is drained to the reactor building sump where the level is monitored. Leakage is also detected by measuring the airborne activ ity and quantity of the conden sate drained from each reactor containment fan cooler unit (CFCU). Allowable total leakage rates for the DCPP units are presented in the Technical Specifications.
RCS identified leakage is limited to 10 gpm by Technical Specification 3.4.13. As prescribed by the Technical Specifications, a RCS water inventory balance shall be performed at least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, with exceptions as noted in the Technical Specifications. Tracking the RCS inventory in a consistent manner provides an effective means of quantifying overall system leakages.
Regulatory Guide 1.45, May 1973, describes acceptable methods of implementing this requirement with regard to the selection of leakage detection systems for the RCPB.
Data on leak detection capabilities are provided in Section 5.2.3.23 and in Table 5.2-16.
5.2.3.7 General Design Criterion 33, 1967 - Reactor Coolant Pressure Boundary Capability The RCPB is designed to withstand the static and dynamic loads imposed on boundary components as a result of an inadvertent and sudden release of energy to the coolant.
Design transients are discussed in Section 5.2.2.1.5. Stress and pressure limits are discussed in Section 5.2.2.1.9. The stress analysis for structural integrity is discussed in Section 5.2.2.1.10. The static and dynamic load analyses are described in Sections 5.2.2.1.11, 5.2.2.1.14, 5.2.2.1.15, and 5.2.2.1.16.
5.2.3.8 General Design Criterion 34, 1967 - Reactor Coolant Pressure Boundary Rapid Propagation Failure Prevention The RCPB is designed to minimize the probabil ity of rapidly propagating type failures.
RCS materials exposed to the coolant are corrosio n-resistant stainless steel or Inconel.
The NDT temperature of t he reactor vessel mate rial samples are es tablished by Charpy V-notch and drop weig ht tests. The m aterials testing is consistent with 10 CFR Part 50, Appendices G and H. These te sts also ensure that only materials with adequate toughness proper ties are used.
As part of the reactor vessel specification, certain addi tional tests are performed:
(1) Ultrasonic Testing In addition to code requireme nts, the performance of a 100 percent volumetric ultrasonic test of reactor ves sel plate for shear wave an d a post-hydrote st ultrasonic map of all welds in the pressure vessel are requi red. Cladding bond ultrasonic DCPP UNITS 1 &
2 FSAR UPDATE 5.2-67 Revision 23 December 2016 inspection to more restrict ive requirements than code is also required to preclude interpretation probl ems during ISI.
(2) Radiation Surveillance Program In the surveillance programs, th e evaluation of the radia tion damage is based on pre-irradiation and post-irrad iation testing of Charpy V-notch and tensile specimens.
These programs are directed toward eval uation of the effect of radiation on the fracture toughness of reactor vessel steels based on th e transition temperature approach and the f racture mechani cs approach, and are in accord with ASTM-E-185, Recommended Practic e for Surveillance Tests f or Nuclear Reactor Vessels.
The inspections of reactor ves sel, pressurizer, pi ping, pumps, and SGs are governed by ASME code requirements.
The allowable heatup and cooldown rates as well as the static loading stre sses during plant life are predicted, using conservative values for the c hange in ductili ty transition temperature due to irradiation.
Refer to Sections 5.2.2.4.1 through 5.2.2.4.3 for discussion of tests that ensure only materials with adequate toughness properties are used per 10 CFR Part 50, Appendix G. Refer to Section 5.2.2.4.4 for discussion of 10 CFR Part 50, Appendix H requirements.
5.2.3.9 General Design Criterion 35, 1967 - Reactor Coolant Pressure Boundary Brittle Fracture Prevention Testing and analysis of materi als employed in RCS components is performed to ensure that the required NDT tem perature limits specifie d in the criterion are met. Removable test capsules are installed in the reactor vessel and remov ed and tested at various times in the plant lifetime to determine the effects of operation on syst em materials. Details of the testing and analysis programs are discussed in Section 5.2.2.4, Fracture Toughness.
Close control is main tained over materi al selection and fabrication for the RCS. Materials exposed to the coolant ar e corrosion-resistant stainless st eel or Inconel. Materials testing consistent with 10 CFR Part 50 assures that only mate rials with adequa te toughness properties are used.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-68 Revision 23 December 2016 The fabrication and quality cont rol techniques used in the fabricati on of the RCS are equivalent to those used for the reactor vessel.
The inspections of reactor vessel, SGs, pressurizer, pumps, and piping are go verned by ASME code requirements.
5.2.3.10 General Design Criterion 36, 1967 - Reactor Coolant Pressure Boundary Surveillance The design of the RCPB provides for accessibility during service life to the entire internal surface of the reactor vessel and certain external zones of the vessel, including the nozzle to reactor coolant piping welds and the top and bottom h eads, except where control rod drive or instrument penetrations prevent access. The re actor arrangement within each containment provides sufficient space for inspection of the ex ternal surfaces of the reactor coolant piping, except for the a rea of pipe within the pr imary shielding concrete. The inspection capabili ty complements the lea kage detection system s in assessing the pressure boundary components' integrity.
Monitoring of the NDT temper ature properties of each core region plate, forging, weldment, and associated he at-treated zones are performed in accordance with ASTM-E-185, Recommended Pr actice for Surveillance Tests on Structural Materials in Nuclear Reactors. Samples of reactor vessel plate materials are retai ned and cataloged in case future engineering development sho ws the need for further testing.
The material properties surveillance program includes not only t he conventional tensile and impact tests, but al so fracture mechanics specimens.
The observed shifts in NDT temperature of th e core region mate rials with irradiation are used to confirm the calculated limits to startup and shutdown transients.
To define permissible operatin g conditions below NDT temperature, a pressure range is established that is bounded by a lower limit for pump operati on and an upper limit that satisfies reactor vessel stre ss criteria. To allow for thermal stresses during heatup or cooldown of the re actor vessel, an equivale nt pressure limit is define d to compensate for thermal stress as a function of rate of change of coolant temperature. Since the normal operating temperature of the reactor vessel is well above the maximum expected NDT temperature britt le fracture during normal operation, it is not consid ered to be a credible mode of failure.
5.2.3.11 General Design Criterion 51, 1967 - Reactor Coolant Pressure Boundary Outside Containment The RCPB is defined as tho se piping systems and compon ents that contain reactor coolant at design pressu re and temperature. With the exception of the reactor coolant sampling lines, the enti re RCPB, as defined above, is located entirely within the containment structure. All sampli ng lines are provided with remotely operated valves for isolation in the event of a fai lure. These valves also close automatical ly on a containment DCPP UNITS 1 &
2 FSAR UPDATE 5.2-69 Revision 23 December 2016 isolation signal. Samplin g lines are only used during infrequent sampling and can readily be isolated.
All other piping and components th at may contain reactor c oolant are low-p ressure, low temperature system s which would yield minimal environmental doses in the event of failure. The sampling system and low-pressure systems ar e described in Chap ter 9. An analysis of malfunctions in these system s is included in Chapter 15.
5.2.3.12 Reactor Coolant Pressure Boundary Safety Function Requirement (1) Protection from Missiles and Dynamic Effects The RCPB is protected against postulated missiles sources generated within containment as described in Section 3.5. The design and fabrication of the RCPs ensure that a missile will not be generated under any anticipated accident as described in Section 5.2.3.20, Safety Guide 14, October 1971 and is, therefore, not a credible missile source as described in Section 3.5.
The RCPB is PG&E Design Class I equipment and therefore is designed to be protected against dynamic effects which may result from equipment failures as described in Section 3.6.
5.2.3.13 10 CFR 50.55a- Codes and Standards For codes and standards applicability to the RCPB, refer to Section 5.2.2.1.3, Compliance with 10 CFR 50.55a; and Section 5.2.2.3, General Material Considerations.
For description of the RCPB PG&E Quality/Code Class requirements refer to Section 5.2.2.1, Design of Reactor Coolan t Pressure Boundary Components.
5.2.3.14 10 CFR 50.55a(f) - Inservice Testing Requirements The IST requirements for the RCPB are contained in the DCPP IST Program Plan.
5.2.3.15 10 CFR 50.55a(g) - Inservice Inspection Requirements The ISI program complies, except where relief is granted by the NRC, with the requirements of 10 CFR 50.55a(b)(2), in effect on January 1, 2005, and uses the ASME BPVC Section XI, 2001 Edition with 2002 and 2003 Addenda, as the basis for the inservice examinations and tests conducted during the third 120-month inspection interval. Components that are designated ASME BPVC Class 1, 2, and 3 for ISIs are included in the ISI Program Plan (Reference 8).
The ISI Program Plan also describes the pressure test program for pressure-retaining Code Class 1, 2, and 3 components; examination techniques; Code Cases; and compliance with ASME BPVC Section XI.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-70 Revision 23 December 2016 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED The second interval Containment Inservice Inspection Program Plan implements the ASME Code Section XI, Subsections IWE and IWL, 2001 Edition with 2003 Addenda, within the limits and modifications of 10CFR50.55a. IWE exams of the metallic liner are performed on a 40-month frequency within the 10 year interval starting September 9th, 2008. Concrete shell exams occur on a 5-year frequency as specified by IWL 2410(a) with the initial examinations performed on November 2000 and August 2001, for Unit 1
and Unit 2 respectively.
As part of the inspection effort for Unit 1, a preservice inspection (PSI) program for Class 1, 2, and 3 systems was conducted in co mpliance with the requirements of ASME BPVC Section XI, 1974 Edition including the Summer 1975 Addenda, except where relief was granted by the NRC. For PSI piping exa minations in Unit 1, the examination technique of Appendix III and the acceptance criteria of IWB-3514, both from the Winter 1975 Addenda of the ASME BPVC Section XI, were used. For Unit 2, a PSI program for Class 1, 2, and 3 systems was conducted in compliance with the requirements of
ASME BPVC Section XI, 1977 Edition inclu ding the Summer 1978 Addenda, except where relief was granted by the NRC.
The ISI program for the first inspection interval for Unit 1 and Unit 2 met the requirements of the ASME BPVC Section XI, 1977 Edition including the Summer 1978 Addenda, except where relief was granted by the NRC. The ISI program for the second
inspection interval for Unit 1 and Unit 2 met the requirements of the ASMS BPVC Section XI, 1989 Edition without addenda, except where relief was granted by the NRC.
Where examination techniques differed due to code changes between the PSI and the ISI examinations, or between subsequent ISI examinations, the latest inservice examination data will be used as the new baseline.
Design provisions for access to the reactor vessel are described in Section 5.4.1.5.
Remote access and data acquisition methods have been developed to facilitate
inspection of reactor vessel areas that are not readily accessible for direct examination.
Areas that are inaccessible for the remote e xamination equipment are detailed in PG&E
requests for relief that have been submitted to the NRC.
5.2.3.16 10 CFR 50.60 - Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation Refer to Section 5.2.2.4, Fracture Toughness, for discussion of 10 CFR Part 50, Appendices G and H requirements.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-71 Revision 23 December 2016
5.2.3.17 10 CFR 50.61 - Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events 10 CFR 50.61 provide s requirements for pr otection against pressu rized thermal shock events involving rapid cooldown and high reactor vessel pressure.
10 CFR 50.61 requi res projected values of RTpts for each reactor vess el beltline material using a fluence valu e, f, which is the EOL fluence for the material. DCPP Unit 1 and Unit 2 are currently licensed for 40 y ears of operation, which correspon ds to 35.2 EFPY for Unit 1 and 35.8 EFPY for Unit
- 2. The project ed EOL vessel fluence at the clad/base metal interface (OT) has been sho wn not to exceed the PTS scre ening criteria, i.e., an ART of 270°F for plates and ax ial welds, and an ART of 300 °F for c ircumferential welds, as required by 10 CFR 50.61. Refer to Section 5.2.2.4, Fracture Toughness, for discussion of Appendix G requirements.
5.2.3.18 10 CFR Part 50 Appendix G - Fracture Toughness Requirements 10 CFR Part 50, Appendix G specifies frac ture toughness require ments for ferritic materials of pressur e-retaining components of the RC PB of light water nuclear power reactors to provide adequate margins of safety during any con dition of normal operation, including anticipated ope rational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected ov er its service lifetime.
Refer to Section 5.2.2.4, Fracture Toughness, for discussion of Appendix G requirements.
5.2.3.19 10 CFR Part 50 Appendix H - Reactor Vessel Material Surveillance Program Requirements 10 CFR Part 50, Appendix H implements a sur veillance program to monitor changes in the fracture toughness properti es of ferritic materials in the reactor vessel beltline region of light water nuclear power reac tors which result from exposu re of these materials to neutron irradiation and th e thermal environment. Under the program, fracture toughness test data are obtained from material specimens exposed in surveillance capsules, which are withdrawn periodically from the reactor vessel. Refer to Section 5.2.2.4.4, Materials Surveillance, for discussion of 10 CFR Part 50, Appendix H requirements.
5.2.3.20 Safety Guide 14, October 1971 - Reactor Coolant Pump Flywheel Integrity The flywheel consists of two thick plates bolted together. The flywheel material is produced by a process that minimizes flaws in the material and improves its fracture
toughness properties; i.e., an electric furnace with vacuum degassing. Each plate is
fabricated from SA-533, Grade B, Class 1 steel. Supplier certification reports are DCPP UNITS 1 &
2 FSAR UPDATE 5.2-72 Revision 23 December 2016 available for all plates and demonstrate the acceptability of the flywheel material on the basis of the requirements of Safety Guide 14, October 1971 (Reference 23).
Flywheel blanks are flame cut from SA-533, Grade B, Class 1 plates with at least 1/2 inch
of stock left on the outer and bore surfaces for machining to final dimensions. The
finished machined flywheels, including bores, keyways, and drilled holes, are subjected to magnetic particle or liquid penetrant examinations in accordance with the
requirements of ASME BPVC Section III.
The finished flywheels, as well as the flywheel material (rolled plate), are subjected to 100 percent volumetric ultrasonic inspection
using procedures and acceptance standards specified in ASME BPVC Section III.
The RCP motors are designed such that, by removing the cover to provide access, the
flywheel is available to allow an ISI program in accordance with the Technical Specifications.
Determining acceptability of the flywheel mat erial involves two steps as follows:
(1) Establish a reference curve describing the lower bound fracture toughness behavior for the material in question.
(2) Use Charpy (CV) impact energy values obtained in certification tests at 10°F to fix position of the heat in question on the reference curve.
A lower bound K ld reference curve (refer to Figure 5.2-7) has been constructed from dynamic fracture toughness data generated by Westinghouse (Reference 3) on A-533, Grade B, Class 1 steel. All data points are plotted on the temperature scale relative to the RT NDT. The construction of the lower-bound curve below which no single test point falls, combined with the use of dynamic data when flywheel loading is essentially static, together represent a large degree of conservatism.
The applicability of a 30 ft-lb Charpy energy reference value has been derived from sections on Special Mechanical Property Requirements and Tests in Article 3,Section III, of the ASME BPVC. The implication is that the low test temperature of
+10°F, and the 30 ft-lb. requirement at that temperature provide assurance that RT NDT is less than +10°F. Flywheel plates exhibit an average value of 30 ft-lb or greater in the
weak direction and, therefore, meet the specific Safety Guide 14, October 1971 requirement that RT NDT must be no higher than 10°F. Making the conservative assumption that all materials in compliance with the code requirements are
characterized by an RT NDT of 10°F, it is possible to reassign the reference temperature position RT NDT in Figure 5.2-7 to a value of 10°F.
Flywheel operating temperature at the surface is 120°F. The lower bound toughness
curve indicates a value of 116 ksi-in 1/2 at the (NDT + 110) position corresponding to operating temperature. Thus, the Safety Gui de 14, October 1971 requirement that the operating temperature be at least 100°F above RT NDT is fulfilled.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-73 Revision 23 December 2016 At the time the flywheels were ordered, Charpy V-notch tests were required only at 10°F. However, by assuming a minimum toughness at operating temperature in excess
of 100 ksi-in 1/2 , it can be seen by examination of the correlation in Figure 5.2-8 that the C V upper-shelf energy must be in excess of 50 ft-lb. Therefore, the requirement "b", that the upper-shelf energy must be at least 50 ft-lb, is satisfied.
It is concluded that flywheel plate materials are suitable for use and meet the Safety
Guide 14, October 1971 acceptance criteria on the bases of suppliers' certification data.
The calculated stresses at operating speed are based on stresses due to centrifugal
forces. The stress resulting from the interference fit of the flywheel on the shaft is less
than 2000 psi at zero speed and becomes zero at approximately 600 rpm because of
radial hub expansion.
The RCPs run at approximately 1190 rpm and may operate briefly at overspeeds of up
to 109 percent (at 1295 rpm). Fo r conservatism, however, 125 percent of operating speed was selected as the design speed for the RCPs. The flywheels are given a
preoperational test prior to shipment at 125 percent of the operating speed.
Precautionary measures, taken to preclude missile formation from primary coolant
pump components, ensure that the pumps will not produce missiles under any
anticipated accident condition. Each component of the primary pump motors has been
analyzed for missile generation. Any fragments of the motor rotor would be contained
by the heavy stator. The same conclusion applies to the pump impeller because the small fragments that might be ejected would be contained by the heavy casing.
5.2.3.21 Regulatory Guide 1.14, Revision 1, August 1975 - Reactor Coolant Pump Flywheel Integrity The RCP Flywheel Inspection Program provides the ISI requirements for the RCP flywheel, which is in accordance with Regulat ory Guide 1.14, Revision 1, Position C.4.b.
Reference 39 demo nstrates compliance with the RCP motor flywheel design requirements given by Regulatory Gui de 1.14, Revision 1, in Position C.2.
An exception to the exam ination requirements gi ven by Regulatory Gui de 1.14, Revision 1, Positions C.4.b(1) and C.4.b(2) was granted ba sed on Reference 32 allowing either an ultrasonic volumetric or surface examination at ten year intervals. Subsequently, the examination frequency w as extended to an in terval not to exceed 20 years based on Reference 39.
In lieu of Position C.4.b(1) and C.4.b(2), a qualified in-place u ltrasonic testing examination over the volume from the inner-bore of the flywheel to the circle one-half of the outer radius or a surface examinat ion (MT and/or PT) of exposed surfaces of the removed flywheels may be conducted at an interval not to exceed 20 years.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-74 Revision 23 December 2016 5.2.3.22 Regulatory Guide 1.44, May 1973 - Control of the Use of Sensitized Stainless Steel Regulatory Guide 1.44, May 1973, describes methods for co ntrol of the application and processing of stainless steel to avoid severe sensi tization to diminish occurrences of stress corrosion cracking. The measures taken to avoid se nsitization are in general conformance with the rec ommendations of Re gulatory Guide 1.44, May 1973 (Reference
- 22) (refer to S ection 5.2.2.5).
5.2.3.23 Regulatory Guide 1.45, May 1973 - Reactor Coolant Pressure Boundary Leakage Detection Systems Means are provided to detect and, to the extent practical, identify the location of reactor
coolant leakage sources. Detection systems with diverse modes of operation are used
to ensure adequate surveillance with sufficient sensitivity so that increases in leakage
rate can be detected before the integrated leakage rate reaches a value that could
interfere with the safe operation of the plant. Section 5.2.3.23 discusses sources of reactor coolant leakage outside containment.
Regulatory Guide 1.45, May 1973 (Reference 25), described acceptable methods for selection of leakage detection systems for the RCPB. The construction permits for DCPP Unit 1 and Unit 2 were issued prior to the guidance of Regulatory Guide 1.45, May 1973. The RCPB leakage detection syst em meets the intent of Regulatory Guide 1.45, May 1973, to detect and monitor RCS leakage such that operators have sufficient time to take corrective actions (References 31 and 37).
5.2.3.23.1 Leakage Detection Methods Systems using diverse methods and modes of operation are provided to continuously
monitor environmental conditions within the containment, and to detect the presence of
radioactive and nonradioactive leakage to the containment. Once operation begins, background levels are established, thereby providin g a baseline for leakage detection.
Deviations from normal conditions indicate po ssible changes in leakage rates and are monitored in the control room and the auxiliary building. Indication s of leakage include changes in containment particulate and gaseous activity, containment sump level, containment condensation, and other volumetric measurement such as increased
coolant makeup demand. A list of systems available to detect these changes is
provided in Table 5.2-16.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-75 Revision 23 December 2016 5.2.3.23.1.1 Containment Radioactivity Monitors Containment radioactivity monitors continuously monitor the air particulate and gaseous
activity levels in the containment during normal plant operation. Leakage to the
containment from the RCPB will result in changes in airborne radioactivity levels that can be detected by this equipment. Detector sensitivity, in terms of leakage rates, depends on the radioactivity level in the reactor coolant itself.
The containment radioactivity monitors measure beta and/or gamma activity in the
containment by taking continuous air samples from the containment atmosphere. This
sample flow first passes through the air particulate monitor and then through the gas
monitor assembly. The sample is then retur ned to the containment. A complete description of the containment activity monitors, including sensitivity and control, indication, and alarm, is presented in Section 11.4.
5.2.3.23.1.2 Containment Sump Levels and Pump Operation Leakage from the primary system would result in reactor coolant flowing into one of the
containment sumps. Sump level and sump pump integrated flow is monitored to provide a measure of the overall leakage that remains in liquid state.
5.2.3.23.1.3 Containment Condensation Measurements The containment condensation measuring system provides a measure of the amount of
leakage vaporized (refer to Section 5.2.3.23.3). This system collects and measures moisture condensed from the containment atmosphere by the cooling coils of the fan cooler air circulation units. Moisture from leaks up to sizes permissible for continued plant operation will partially evaporate into the containment atmosphere and will be
condensed on the fan cooling coils. This system dependably and accurately measures
total vaporized leakage, including leakage from the cooling coils themselves. It measures the liquid runoff flowrate from the drain pans under each CFCU. The condensate measuring system consists of a vertical standpipe, valves, and
instrumentation installed in the drain piping of the reactor CFCU.
Depending on the number of reactor CFCUs in operation, the drainage flowrate from each unit due to normal condensation can be determined. Additional or abnormal leaks
will result in containment humidity and condensation runoff rate increases, and the
additional leakage can then be measured.
5.2.3.23.1.4 Other Methods of Detection (1) Charging Pump Operation During normal operation only one charging pump is operating. If a gross loss of reactor coolant should occur which was not detected by the DCPP UNITS 1 &
2 FSAR UPDATE 5.2-76 Revision 23 December 2016 methods previously described, the flowrate mismatch of the charging and letdown flows would i ndicate RCS leakage.
(2) Liquid Inventory Gross leakage can also be detected by an increase in the makeup rate to
the RCS. This is inherently a low-precision indication, because makeup to
the RCS is also required due to other process variables. A quantitative
measurement of leakage requires a test over a reasonable period of time
to establish changes in the physical inventory.
(3) Coolant Radiation Monitors The component cooling liquid monitor continuously monitors the CCW system for activity indicative of a leak of reactor coolant from either the RCS or the RHR system loop in the CCW system. In addition, condenser offgas monitors and SG blowdown radiation detectors are available to detect SG tube leakage.
(4) Containment Atmosphere Temperature and Pressure Measurement Various air temperature and pressure sensors would supplement
indications of RCS leakage. Containment temperature and pressure
fluctuate slightly during plant operation, but a rise above the normally
indicated range of values may indicate RCS leakage into the containment.
The accuracy and relevance of temperature and pressure measurements is a function of containment free volume and detector location. Alarm signals from these instruments would be valuable in recognizing rapid and
sizable energy releases to the containment.
Thermoswitches are installed in the leakoff piping from RCS valves with restricted
access during plant operation as a means of identifying the source of leakage (i.e., the
specific valve) from a packing or bellows fail ure. Identified indicating lights, located in a routinely inspected area, are actuated by the thermoswitches. A control room alarm is
provided for valve stem leakoff.
5.2.3.23.1.5 Visual and Ultrasonic Inspections Visual and ultrasonic inspections of the RCPB wil l be made periodically during plant shutdown periods. Limited access to the containment is possible for this purpose during
normal plant operation. The design of the reactor vessel and its arrangement in the
system provides accessibility during service life to the entire internal surface of the
vessel (except where access is limited by control rod drive or instrument penetrations).
Access is also provided to the entire primary piping system, except for the area of pipe
within the concrete biological shielding.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-77 Revision 23 December 2016 5.2.3.23.1.6 Reactor Coolant Sys tem Water Inventory Balance As prescribed by the Technical Specifications, a RCS water inventory balance shall be
performed at least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, with exceptions as noted in the Technical
Specifications. Tracking the RCS inventory in a consistent manner provides an
effective means of quantifying overall system leakages.
Data on other secondary methods of leak detection, such as pressurizer liquid level, VCT liquid level, charging pump flowrate, and PRT liquid level are provided in Table 5.2-16.
5.2.3.23.1.7 Indication in Control Room Positive indications in the control room of coolant leakage from the RCS to the
containment are provided by equipment that permits continuous monitoring of
containment air activity, containment sump level chang es, and of runoff from the condensate collecting pans under the cooling coils of the CFCUs. This equipment provides indication of normal background, which is indicative of a basic level of leakage
from the RCS and components. An increase in observed parameters is an indication of
leakage within the containment, and the equipment provided is capable of monitoring
this change.
As indicated in Table 5.2-16, numerous other forms of RCS leaka ge indication are provided in the control room or auxiliary building control area. Leakage detection
systems are provided and located in a manner such that for minor leakages the operator can identify the subsystem that is leaking and effectively isolate that leakage with no more than short-term interruption of t he operation of the complete system.
Figures 5.2-14 and 5.2-15 are examples of the correlative relationships between radioactivity leak detector indications and the corresponding volumetric leak flowrate.
This information is provided to the operator for a quick and easy interpretation of
leakage conditions, and forms the basis for determining operator action.
5.2.3.23.2 Limits for Reactor Coolant Leakage Operational leakage limiting conditions for RCS operation are presented in the Technical Specifications.
The Technical Specifications also present leakage limitations for the RCS PIVs listed in Table 5.2-23.
RCS PIVs protect low pressure ECCS systems such as the RHR system and the SI system from overpressurization and rupture of their low pressure piping which could result in a LOCA that bypasses the containment. Testing of these valves at least once
per refueling interval during startup ensures a low probability of gross failure. Each PIV
is required to be tested prior to returning the valve to service following maintenance, repair, or replacement work.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-78 Revision 23 December 2016 5.2.3.23.3 Unidentified Leakage The sensitivity and response time of RCPB leakage detection systems vary for different
methods of detection. However, the diverse systems available are required to have the
capability to detect continuous leakage rates as low as 1 gpm within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for unidentified leaks at the design conditions and assumptions, as recommended by
Regulatory Guide 1.45, May 1973 (Reference 25). The LBB analysis (Reference 42) demonstrates that this leak detection capability is sufficient to provide the margin of 10 on the leak rate in support of LBB (refer to Section 5.2.3.2, GDC 4, 1987).
The containment particulate monitor is the most sensitive instrument of those available for detection of reactor coolant leakage into the containment. This instrument is capable of detecting particulate radioactivity concentrations as low as 10
-11 µCi/cc. The sensitivity of the air particulate monitor to an increase in reactor coolant leakage rate is dependent on the magnitude of the normal leakage into the containment. The
sensitivity is greatest where normal leakage is low, as has been demonstrated by the
experience of Indian Point Unit No. 1, Yankee Rowe, and Dresden Unit 1. Based on data from these operating plants, it is expect ed that this unit will detect (at the 95 percent confidence level) an increase in containment air particulate activity resulting in a gross count rate equivalent to 1 x 10
-9 µCi/cc during normal full power operation.
As shown in Figure 5.2-9, this system has adequate response to detect a 1 gpm leak within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> assuming a reactor coolant particulate activity corresponding to as low as
0.1 percent fuel defects. The assumption of 0.1 percent fuel defects used in the design
calculation is less than the percentage of failed fuel assumed in the Environmental
Report (Reference 36) and follows the guidance of Regulatory Guide 1.45, May 1973 (References 25 and 37).
The containment radioactive gas monitor is inherently less sensitive (threshold at 10-7 µCi/cc) than the containment air particulate monitor, and would function in the event that significant reactor coolant gaseous activity results from fuel cladding defects. The sensitivity and range are such that gross count rates equivalent to from 10
-6 to 10-3 µCi/cc will be detected. This system is also adequate to detect a 1 gpm leak within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> assuming a reactor coolant gaseous activity corresponding to as low as 0.1
percent fuel defects as shown in Figure 5.2-9. The assumption of 0.1 percent fuel
defects used in the design calculation is les s than the percentage of failed fuel assumed in the Environmental Report and follows the guidance of Regulatory Guide 1.45, May 1973 (References 25 and 37).
The containment gaseous activity will result from any fission product gases (Kr-85, Xe-135) leaking from the RCS as well as from the argon-41 produced in the air around
the reactor vessel. Assuming a constant background radioactivity in the containment
atmosphere due predominantly to argon-41, and reactor coolant gaseous activity of 0.03 µCi/cc (corresponding to about 0.05 percent fuel defects), a 1-gpm coolant leak would double the fission product gas background in about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The occurrence of a DCPP UNITS 1 &
2 FSAR UPDATE 5.2-79 Revision 23 December 2016 leak of 2 to 4 gpm would double the background in less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. In these circumstances, this instrument is a useful backup to the air particulate monitor.
The adequacy of the containment particulate and radioactive gas monitors to detect a change in leakage during the initial period of plant operation will be limited by low
coolant activity levels. The gas detector will not be as sensitive as the other leakage
detection systems during this period because the argon-41 background will mask the
low level of gaseous activity from coolant leakage.
Within the containment, the average air temperature is held at 120°F or below in accordance with the Technical Specifications. The hot dry air promotes evaporation of
water leakage from hot systems, and the cooling coils of the fan cooler units provide a
significant surface area at or below the dewpoint temperature. Therefore, under
equilibrium conditions, the quantity of condensate collected by the cooling coils of the fan cooler units should be equal to the evaporated water leakage and steam leakage
from systems within the containment.
To determine abnormal leakage rate inside the containment based on condensation
measurements, it will first be necessary to determine the condensation rate from the fan
coolers during normal operation. With the initiation of an additional or abnormal leak, the containment atmosphere humidity will begin to increase but such an increase in
humidity is reduced by additional condensation on the fan cooler tubes. (assuming that there is no large heat addition to the containment that could cause the cooling water
temperature to increase.)
With the increasing specific and relative humidity, the heat removal capacity needed to cool the air-vapor mixture to its dewpoint decreases. Therefore, increases in available heat removal capacity (i.e., increases in the number of fans in operation) will result in
added condensate flow. Through accurate measurement of condensate flow from the
fan coolers, a reliable estimate of evaporated leakage inside the containment can be
made.
A preliminary estimate of the evaporated leakage can be obtained from the condensate
flow increase rate during the transient; a better estimate can be determined from the steady state condensate flow when equilibrium has been reached. After equilibrium is
attained, condensate flow from approximately 0.1 to 30 gpm per detector can be measured by this system.
Except for the condensate measuring system, the sensitivities of the RCPB leakage detection systems are not significantly affected during plant operation with concurrent
leaks from other sources. Condensation of moisture on the containment air cooler coils
will produce a scrubbing effect for particulate activity, but is not expected to appreciably
reduce particulate detector sensitivity.
When the plant is shut down, personnel can enter the containment to check visually for
leaks. The lack of escaping steam or water during hydrostatic tests has been widely DCPP UNITS 1 &
2 FSAR UPDATE 5.2-80 Revision 23 December 2016 used as a criterion for leaktightness of pressurized systems. Detection of the location of significant leaks would be aided by the presence of boric acid crystals near a leak. The
boric acid crystals are transported outside the RCS in the leaking fluid and then
deposited by the evaporation process. Sensitivities and response times of other
methods of leak detection are provided in Table 5.2-16 and in Figures 5.2-10 through
5.2-13.
5.2.3.23.4 Maximum Allowable Total Leakage As discussed above, the reactor coolant leakage detection systems provide the
capability for detecting extremely small lea kage rates from the RCPB during normal operation. Signals from the various leak detectors are displayed in the control room and
are used by the operators to determine if corrective action is required.
A limited amount of leakage is expected from the RCPB and from auxiliary systems
within the containment. Although it is desirable to maintain leakage at a minimum, a
maximum allowable total leakage rate is established and used as a basis for action by
the reactor operator to initiate corrective measures. Allowable total leakage rates for
the DCPP units are presented in the Technical Specifications. RCS identified leakage
is limited to 10 gpm by Technical Specification 3.4.13.
5.2.3.23.5 Differentiation Between Identified and Unidentified Leaks Generally, leakage into closed systems, or leakage into the containment atmosphere
from sources that are both specifically located and known either not to interfere with the
operation of the unidentified leakage monitoring systems or not to be from a flaw in the RCPB, are called identified leakages. Uncontained leaka ge to the containment atmosphere may be the result of a variety of possible leakages that are generally classified as unidentified leakages. Unidentified leakage is eventually collected in tanks
or sumps where the flowrate can be established and monitored during operation.
5.2.3.23.5.1 Leakage Location Capability Leakage detection systems have been designed to aid operating personnel, to the
extent possible, in differentiating between pos sible sources of detected leakage within the containment and in identifying the physical location of the leak. Containment entry
for visual inspection will, however, remain the only method of positively identifying the
source and magnitude of leakage detected by remote sensing systems.
The containment monitoring system provides the primary means of remotely identifying
the source and location of leakage within the containment. Increases in containment airborne activity levels detected by any of the monitor channels will indicate the RCPB
as the source of leakage. Additionally, the capability of drawing monitored samples
from several containment locations will allo w localization of the general area of leakage since activity levels will be somewhat higher in the vicinity of the leakage source.
Conversely, if the condensate measuring system detects increased containment DCPP UNITS 1 &
2 FSAR UPDATE 5.2-81 Revision 23 December 2016 moisture without a corresponding increase in airborne activity level, the indicated source of leakage would be judged to be a nonradioactive system, except when the reactor
coolant activity may be low.
Less sensitive methods of leakage detection, such as unexplained increases in reactor
plant makeup requirements to maintain pre ssurizer level, will also provide positive indication of the RCPB as the leakage source. Increases in the frequency of a particular containment sump pump operation will facilitate localization of the source to
components whose leakage would drain to that sump. Leakage rates of the magnitude necessary to be detectable by these latter methods are expected to be noted first by the
more sensitive radiation detection equipment.
5.2.3.23.5.2 Adequacy of Leakage Detection System The component cooling liquid monitor continuously monitors the component cooling
loop of auxiliary coolant for activity indicative of a leak of reactor coolant from either the
If an accident involving gross leakage from the RCS occurred, it would be detected by
the following methods:
(1) Pump Operation During normal operation, only one charging pump is operating. If a gross
loss of reactor coolant occurred which was not detected by previously
described methods, the difference between charging and letdown flowrate would indicate the leakage.
(2) Liquid Inventory Gross leaks might be detected by unscheduled increases in the amount of reactor coolant makeup water, which is required to maintain the normal
level in the pressurizer. Gross leakage would also be detected by a rise in the normal containment sump level.
(3) RHR Loop The RHR loop removes residual and sensible heat from the core and
reduces the temperature of the RCS during the second phase of plant
shutdown. Tube leaks from the RHR heat exchangers during normal
operation would be detected outside the containment by the component
cooling loop radiation monitors.
Leakage detection systems are provided and located in a manner such that the operator
can identify the subsystem, which is leaking and effectively isolate that leakage with no
more than short-term interruption of the operation of the complete system.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-82 Revision 23 December 2016 5.2.3.23.6 Sensitivity and Operability Tests Periodic testing of leakage detection systems will be conducted to verify the operability
and sensitivity of detector equipment. These tests include installation calibrations and alignments, periodic channel calibr ations, functional tests, and channel checks.
The containment monitoring system is calibrated on installation using typical isotopes of
interest. Subsequent periodic calibrations using detector check sources will consist of
single-point calibration to confirm detector sensitivity based on the known correlation
between the detector response and the check source standard. This procedure will
adequately measure instrument sensitivity since the geometry of the sampler cannot be
significantly altered after the initial cali bration. Channel checks to verify acceptable channel operability during normal operation and functional testing to verify proper
channel response to simulated signals will al so be conducted on a regular basis. A complete description of calibration and maintenance procedures and frequencies for the containment radiation monitor system is presented in Section 11.4. The condensate
measuring system will also be periodically te sted to ensure proper operation and verify sensitivity.
The equipment used, procedures involved, and frequency of testing, inspection
surveillance and examination of the structural and leaktight integrity of RCPB
components are described in detail in Section 5.2.3.14.
5.2.3.24 Regulatory Guide 1.97, Revision 3, May 1983 - Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants Instrumentation is provided to monitor RCS integrity following an accident.
Instrumentation related to RCPB is required to meet Regulatory Guide 1.97, Revision 3.
The requirements consist of continuous indication and recording of RCS level, RCS pressure, containment sump water level (wide and narrow ranges), containment pressure (normal range and wide range), high range containment area radiation monitor, condenser noble gas effluent radiation monitor. Refer to Table 7.5-6 for details.
Primary system relief valve continuous position indication and recording are provided for the PSVs (i.e., acoustic monitors). Refer to Section 7.5.2.8 for details. Position indication for the PORVs (i.e., valve position switches) is also provided.
5.2.3.25 Regulatory Guide 1.99, Revision 2, May 1988 - Radiation Embrittlement of Reactor Vessel Materials Regulatory Guide 1.99, Revision 2, provides general procedures for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used in the DCPP Unit 1 and Unit 2 reactor vessels. Refer to Section 5.2.2.4 for information NDT values.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-83 Revision 23 December 2016 Generic Letter 88-11, July 1988, recommends the use of Regulatory Guide 1.99, Revision 2. For the DCPP reactor vessels, PG&E has committed to use the methodology in Regulatory Guide 1.99, Revision 2.
5.2.3.26 NUREG-0737 (Items II.B.1, II.D.1, II.D.3, I I.K.2.13, and III.D.1.1), November 1980 - Clarification of TMI Action Plan Requirements Item II.B.1 - Reactor Vessel Head Vent System: A RVHVS is provided to exhaust non-condensable gases and/or steam from the RCS that could inhibit natural circulation core cooling. The configuration of the RCS vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve power supply or control system does not prevent isolation of the vent path.
Item II.D.1 - Performance Testing of Pressurized-Water Reactor Relief and Safety Valves: A program has been implemented for testing of the PSVs, PORVs and block valves to qualify these components under expected design transients.
Item II.D.3 - Valve Position Indication for PSVs and PORVs: Positive PSV and PORV position indication is provided in the in the control room. Refer to Section 7.5.2.8 for details. Item II.K.2.13 - Thermal Mechanical Report: An analysis has been performed to evaluate the effects of high pressure injection on vessel integrity for a SBLOCA. A generic report (Reference 40) provides a conservative bases for demonstrating that reactor vessel integrity is maintained for such an event. The conclusions of the generic report coupled with the requirements for protection of pressurized thermal shock demonstrate no loss of vessel integrity at EOL.
Refer to Section 5.2.3.17 for discussion of protection against pressurized thermal shock events.
Item III.D.1.1 - Integrity of Systems Outside Containment Likely to Contain Radioactive Material for Pressurized-Water Reactors: DCPP implements a program to reduce leakage from systems outside the containment that would or could contain highly radioactive fluids during a severe transient or accident. The systems, or portions of
systems, that are included in the leakage reduction program required by NUREG-0737, November 1980, and the reason for their inclusion, are as follows:
(1) The RHR and SI system that would circulate radioactive water from the RCS (2) The containment spray system (CSS) that would circulate radioactive water from the containment sump (3) The NSSS sampling system because of the highly radioactive fluids to be sampled DCPP UNITS 1 &
2 FSAR UPDATE 5.2-84 Revision 23 December 2016 (4) The gaseous radwaste system (GRS) because it could be used to collect highly radioactive gases from the RCS At intervals of approximately 24 months, operating pressure leak tests will be performed on appropriate portions of the SI system, the RHR system, the NSSS sampling system, and the CSS. Systems that normally contain liquids will be pressurized to normal operating pressure using systems pumps or hydro pumps. Each liquid system will be
visually inspected during its pressure test so that leakage from the system can be measured and corrected. Systems that normally contain gases will be pressurized with
a gas, and leakage will be determined usi ng a calibrated leakrate monitor. If gaseous systems have excessive leakage, then leaks will be located using appropriate leak detection methods such as the soap bubble. After initial criticality, leakage from the
GRS will be evaluated by monitoring the auxil iary building ventila tion exhaust with radiation detectors.
5.2.3.27 Generic Letter 1989-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance The RCPB PG&E Design Class I and positio n changeable MOVs are subject to the requirements of Generic Letter 89-10, June 1989, and associated Generic Letter 96-05, September 1996, and meet the requirements of the DCPP MOV Program Plan. The PORV block valves are included in the MOV testing program.
5.2.3.28 Generic Letter 1990-06, June 1990 - Enclosure B, Resolution of Generic Issue 94 - Additional Low-Temperature Overpressure Protection For Light-Water Reactors Pressure/temperature limit curves are generated in accordance with WCAP-14040 (Reference 41) and are documented in the Pressure and Temperature Limits Report (PTLR) per Technical Specification 5.6.6. An exemption from certain requirements of 10 CFR 50.60, and 10 CFR Part 50, Appendi x G allows the application of ASME Code Case N-514, Low Temperature Overpressure Protection, in determining the acceptable LTOP system setpoints.
RCS overpressure protection during startup and shutdown is provided by the LTOP system, which consists of two mutually redundant and independent systems. Each system receives reactor coolant pressure and temperature signals. When a
low-temperature, high-pressure transient occurs, it opens a pressurizer PORV until the
pressure returns to within acceptable limits. During normal operation, the system is off.
If the reactor coolant temperature is below the low temperature setpoint and the enable
switch on the main control board is not in the enable position, an alarm will sound on the
main annunciator. The operator can then enable the circuit before a water-solid
condition is reached, and the system is then ready to operate without further operator
action.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-85 Revision 23 December 2016 During startup, at the temperature at which the steam bubble is formed, the trip circuit is automatically defeated and the operator can disable the system later in the startup sequence.
The system is completely automatic after being manually enabled. Whenever the system is enabled and reactor coolant temperature is below the low temperature
setpoint, a high-pressure signal will trip it a utomatically and open the PORV until the pressure drops below the reset value.
Features of the LTOP control system include: indicating lights and annunciator alarm
when the system trips, indicating lights when the system is enabled, and annunciator
alarm when the isolation valve for the PORV is closed and the system is enabled.
The LTOP system relieves the RCS pressure transient given a single failure. Since the
two LTOP systems are mutually redundant and independent, failure of either one would
not affect the remaining system.
The system is testable at all times. The pressurizer PORVs are in series with
motor-operated block valves, which may be closed during testing. Test signals may be injected into the appropriate control circuits and the position of the valve monitored and
timed.
All LTOP components meet PG&E Design Class I. Refer to Chapter 7 for IEEE-279-1971 (Reference 21) criteria. The electrical portions of the system are powered from Class 1E 125-Vdc power sources. The air to the valves is backed by bottled nitrogen.
5.
2.4 REFERENCES
- 1. K. Cooper et al., Overpressure Protection for Westinghouse Pressurized Water Reactor, WCAP 7769, Revision 1, June 1972.
- 2. J. A. Nay, Topical Report, Process Instrumentation for Westinghouse Nuclear Steam Supply Systems, WCAP 7671, April 1971.
- 3. W. O. Shabbits, Dynamics Fracture Toughness Properties of Heavy Section A-533 Grade B Class 1 Steel Plate, WCAP-7623.
- 4. H. T. Corten, R. H. Sailors, Relationship Between Material Fracture Toughness Using Fracture Mechanics and Transition Temperature Tests, UILU-ENG 71-60010, August 1, 1972.
- 5. W. S. Hazelton et al., Basis for Heatup and Cooldown Limit Curves, WCAP-7924, July 1972.
- 6. MULTIFLEX 3.0, A Fortran IV Computer Program for Analyzing Thermal-Hydraulic-Structural System Dynamics Advanced Beam Model, WCAP-9735 DCPP UNITS 1 &
2 FSAR UPDATE 5.2-86 Revision 23 December 2016 Revision 2 (Proprietary) and WCAP-9736 Revision1 (Non-Proprietary), February 1998.
- 7. Deleted in Revision 23.
- 8. Diablo Canyon Power Plant - Inservice Inspection Program Plan - The Third 10-year Inspection Interval, Pacific Gas and Electric Company.
- 9. Structural Analysis of Reactor Coolant Loop/
Support System for the Diablo Canyon Nuclear Generating Station Unit No. 1, SD-117.
- 10. Deleted in Revision 23.
- 12. Schmittroth, E. A., FERRET Data Analysis Code, HEDL-TME-79-40, Hanford Engineering Development Laboratory, Richland, Washington, September 1979.
- 13. McElroy, W. N., et al, A Computer-Automated Iterative Method of Neutron Flux Spectra Determined by Foil Activation, AFWL-TR-67-41, Volumes I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.
- 14. RSIC Data Library Collection DLC-178, SNLRML Recommended Dosimetry Cross-Section Compendium, July 1994.
- 15. RSIC Computer Code Collection CCC-543, TORT-DORT Two- and Three-Dimensional Discrete Ordinates Transport, Version f2.8.14, January 1994.
- 16. RSIC Data Library Collection DLC-175, BUGLE-93, Production and Testing of the VITAMIN-B6 Fine Group and the BUGLE-93 Broad Group Neutron/Photon Cross-Section Libraries Derived from ENDF/B-VI Nuclear Data, April 1994.
- 17. H. P. Flatt and D. C. Baller, AIM
-5, A Multigroup One Dimensional Diffusion Equation Code, NAA-SR-4694, March 1960.
- 18. Documentation of Selected Westinghouse Structural Analysis Computer Codes, WCAP-8252, Revision 1, May 1977.
- 19. Deleted in Revision 23.
- 20. Deleted in Revision 22.
- 21. IEEE-Std-279, Criteria for Protection Systems for Nuclear Power Generating Stations, 1971.
DCPP UNITS 1 &
2 FSAR UPDATE 5.2-87 Revision 23 December 2016
- 22. Regulatory Guide 1.44, Control of the Use of Sensitized Stainless Steel, May 1973.
- 23. Safety Guide 14, Reactor Coolant Pump Flywheel Integrity (for Comment), October 1971.
- 24. NUREG-0737, Clarification of TMI Plan Requirements, U. S. Nuclear Regulatory Commission, November 1980.
- 25. Regulatory Guide 1.45, Reactor Coolant Pressure Boundary Leakage Detection Systems, May 1973.
- 26. PG&E Diablo Canyon Unit 2 Rea ctor Vessel Radiation Surveillance Program, WCAP 8783, December 1976.
- 27. Regulatory Guide 1.99, Radiation Damage to Reactor Vessel Materials, Revision 2, May 1988.
- 28. Supplemental Reactor Vessel Rad iation Surveillance Program For PG&E Diablo Canyon Unit 1 WCAP-13440, December 1992.
- 29. PG&E Letter to NRC No. DCL-92-072, Diablo Canyon Unit 1 - Supplemental Reactor Vessel Radiation Surveillance Program.
- 30. PG&E Letter to the NRC No. DCL-92-198 LER 1-92-015.
- 31. Letter from Sheri R. Peterson (NRC) to Gregory M. Rueger (PG&E), Leak-Before-Break Evaluation of Reactor Coolant System Piping for DCPP Units 1 and 2, March 2, 1993.
- 32. Strauch, P. L. et al, Topical Report on Reactor Coolant Pump Flywheel Inspection Examination, WCAP-14535A, November 1996
- 33. Diablo Canyon Units 1 and 2 Replacement Steam Generator Program - NSSS Licensing Report, WCAP-16638 (Proprietary), September 2007.
- 34. Deleted in Revision 23.
- 35. Deleted in Revision 23.
- 36. Letter from F. T. Searls (PG&E) to Atomic Energy Commission , dated August 9, 1971 , Enclosure "Environmental Report, Units 1 and 2 Diablo Canyon Site, Atomic Energy Commission Dockets 50-275, 50-323," dated July 1971.
- 37. Letter from Alan Wang (NRC) to John T. Conway (PG&E), "Diablo Canyon Power Plant, Unit Nos. 1 and 2 - Issuance of Amendments RE: Revision to DCPP UNITS 1 &
2 FSAR UPDATE 5.2-88 Revision 23 December 2016 Technical Specification 3.4.15, "RCS Leakage Detection Instrumentation," dated January 24, 2011.
- 38. Diablo Canyon Power Plant - Containment Inservice Inspection Program Plan-The Second 10-year Inspection Interval, Pacific Gas and Electric Company.
- 39. Straunch, P. L. et al, Extension of Reactor Coolant Flywheel Examination, WCAP 15666-A, Revision 1, October 2003.
- 40. T. A. Meyer, Summary Report on Reactor Vessel Integrity for Westinghouse Plants, WCAP-10019, December 1981.
- 41. Westinghouse Report WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 2, January 1996.
- 42. WCAP-13039, Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for DCPP Units 1 and 2, Westinghouse Electric Corporation, November 1991.
5.2.5 REFERENCE DRAWINGS Figures representing controlled engineering drawings a re incorporated by reference and are identified in Table 1.6-1. The contents of the drawings are controlled by DCPP procedures.
DCPP UNITS 1 &
2 FSAR UPDATE 5.3-1 Revision 23 December 2016 5.3 THERMAL HYDRAULIC SYSTEM DESIGN The overall objective of the reactor core thermal and hydraulic design is to provide
adequate heat transfer, compatible with the heat generation distribution in the core, such that the performance and safety criteria requirements of Chapter 4 are met under
all plant operating conditions.
5.3.1 ANALYTICAL METH ODS AND DATA The thermal and hydraulic design bases of the RCS are described in Sections 4.3 and 4.4 in terms of core heat generation rates, DNBR, analytical models, peaking factors, and other relevant aspects of the reactor.
5.3.2 OPERATING RESTRICTIONS ON REACTOR COOLANT PUMPS The No. 1 seal is a controlled-leakage, film-riding face seal. To establish face plate separation and equilibrium for startup of the RCPs, the operating procedures ensure that the pressure differential across the No. 1 seal will be at least 200 psid before starting the RCP. To ensure sufficient NPS H, the RCS pressure must be maintained at a minimum of 325 psig (with RCS temperature compatible with the pressure), with the VCT pressure high enough to provide an effective back pressure on the No. 1 seal of at least 15 psig.
5.3.3 TEMPERATURE-POWER OPERATING MAP The programmed relationship between RCS temperature and power for Unit 1 is shown in Figure 5.3-1. A similar relationship has been programmed for Unit 2 and the corresponding temperatures are also shown in Figure 5.3-1.
The effects of reduced core flow due to inoperative pumps are discussed in
Sections 5.5.1, 15.2, and 15.3.
Natural circulation capability of the system is shown in Section 4.4.3.
5.3.4 LOAD-FOLLOWING CHARACTERISTICS The RCS is designed on the basis of steady state operation at full power heat load. The
RCPs utilize constant-speed drives as described in Section 5.5 and the average coolant temperature is controlled to have a value that is a linear function of load, as described in
Section 7.7.
DCPP UNITS 1 &
2 FSAR UPDATE 5.3-2 Revision 23 December 2016 5.3.5 TRANSIENT EFFECTS Evaluation of transient effects is presented as follows:
Event FSAR Section
Complete loss of forced reactor coolant flow 15.3.4 Partial loss of forced reactor coolant flow 15.2.5 Loss of external electrical load and/or turbine trip 15.2.7 Loss of normal feedwater 15.2.8 Loss of offsite power 15.2.9 Accidental depressurization of the RCS 15.2.13 Component cyclic and transient design occurrences are contained in Table 5.2-4.
5.3.6 THERMAL AND HYDRAULIC CHARACTERISTICS
SUMMARY
TABLE The thermal and hydraulic characteristics are provided in Tables 4.1-1 and 5.1-1.
DCPP UNITS 1 &
2 FSAR UPDATE 5.4-1 Revision 23 December 2016 5.4 REACTOR PRESSUR E VESSEL AND A PPURTENANCES Section 5.4 discusses the design, material, fabrication, inspection, and quality
provisions that apply to the RPV and its appurtenances.
5.4.1 REACTOR PRESSURE VESSEL DESCRIPTION
5.4.1.1 Design Bases The RPV is an integral part of the RCP B and is designed to maintain its integrity under all anticipated modes of plant operation, including exposure to all foreseeable pressure
and temperature transients and neutron flux during the life of the plant, by ensuring that
all resulting stresses remain within allowab le values. The RPV supports the reactor core and CRDMs.
5.4.1.2 Design Transients Cyclic loads are introduced by normal power changes, reactor trip, startup, and shutdown operations. These design bases cycles are selected for fatigue evaluation
and constitute a conservative design envelope for the projected plant life. RPV analysis results in a usage factor that is less than 1.
Regarding the thermal and pressure transients involved in the LOCA, the RPV is analyzed to confirm that the delivery of cold emergency core cooling water to the vessel following a LOCA does not cause a loss of RPV integrity.
The design specifications require analysis to prove that the RPV is in compliance with the fatigue limits of ASME BPVC Section III-1965 through Winter 1966 Addenda (Unit 1) and Section III-1968 (Unit 2). The loadings and transients specified for the analysis are based on the most severe cond itions expected during service. The typical normal heatup and cooldown rates are less than the 100
°F per hour upset or faulted condition rate used for design evaluation purposes. These rates are reflected in the vessel design specifications (refer to Section 5.2).
5.4.1.3 Codes and Standards The manufacturer of the reactor vessels for Diablo Canyon Unit 1 and Unit 2 is Combustion Engineering, Inc., Chattanooga, Tennessee.
Refer to Table 5.2-3 for procurement information on RCS components. Pursuant to 10 CFR 50.55a(c), the applicable ASME requirements for RPV design, fabrication, and material specifications are ASME BPVC Section III-1965 through Winter 1966 Addenda for Unit 1 and Section III-1968 for Unit 2.
The RVCH was manufactured by AREVA.
Pursuant to 10 CFR 50.55a(c), the applicable ASME BPVC requirements for design, fabrication, and material specifications are the requirements of ASME BPVC Section III-2001 through 2003 Addenda.
DCPP UNITS 1 &
2 FSAR UPDATE 5.4-2 Revision 23 December 2016 5.4.1.4 Reactor Pressure Vessel Description The RPV is cylindrical with a welded hemispherical bottom head and removable, bolted, flanged, and gasketed hemispherical RVCH. The RPV flange and head are each sealed by two hollow metallic O-rings. Seal leakage is detected by means of two leakoff
channels: one between the inner and outer ring and one outside the outer O-ring. The RPV contains the core, core support st ructures, control rods, and other parts directly associated with the core.
The RVCH contains fifty-eight head adapters (nozzles). These head adapters are tubular members, attached by partial penetration welds to the underside of the closure
head. The upper end of the head adapters are welded to a CRDM latch housing or instrument adapter. The upper end of these items contains threads for the assembly of a CRDM rod drive travel housing or CET column. The RVCH also contains dedicated nozzles for the head vent and RVLIS.
Inlet and outlet nozzles are spaced evenly around the RPVs. Outlet nozzles are located on opposite sides of the RPV to facilitate optimum layout of the RCS equipment. The inlet nozzles are tapered from the coolant loop RPV interfaces to the RPV inside wall to reduce loop pressure drop.
The IHA is a multi-function structure located on top of the RVCH. The IHA includes the RVCH lift rig, the CRDM ventilation system (including fans, shrouds, and plenum), the CRDM missile shield, radiation shielding, the RPV stud tensioner hoist monorail, cable bridges, personnel access platforms, and ladders. The IHA also includes a seismic support structure, which is an integral part of the IHA that provides lateral structural support for the IHA and CRDMs. The seismic support structure assembly includes eight
seismic tie-rod restraints to transfer load fro m the IHA and the CRDMs to the reactor cavity walls. Figure 5.4-3 shows the major components included in the seismic support
structure (some items attached to the support st ructure are excluded for clarity).
The bottom head of the RPV contains penetration nozzles for connection and entry of the nuclear incore instrumentation. Each nozzle consists of a tubular member made of
an Inconel stainless steel composite tube. Each tube is attached to the inside of the
bottom head by a partial penetration weld.
Internal surfaces of the RPV that are in contact with primary coolant are weld overlaid with 5/32-inch minimum of stainless steel. The exterior of the RPV is insulated with canned stainless steel reflective sheets. The insulation is 3 inches thick and contoured
to enclose the top, sides, and bottom of the RPV.
A schematic of the RPV is shown in Figure 5.4-1 for Unit 1 and Figure 5.4-2 for Unit 2.
RPV principal design parameters for both Unit 1 and Unit 2 are provided in Table 5.4-1.
DCPP UNITS 1 &
2 FSAR UPDATE 5.4-3 Revision 23 December 2016 5.4.1.5 Inspection Provisions The internal surface of the RPV can be inspected using visual nondestructive techniques over the accessible areas. If nec essary, the core barrel can be removed, making the entire inside surface of the RPV accessible.
The RVCH is examined visually during each refueling. Periodic visual inspections of accessible outer CRDM penetration tubes and the gasket seating surface are performed. The transition area between the dome and head flange, which is the area of
highest stress of the RVCH, is accessible on the outer surface for visual inspection, surface examination, and ultrasonic testing. The closure studs, nuts, and washers can
be inspected periodically using visual, magnetic particle, and/or ultrasonic techniques.
Full-penetration welds in the following irradia ted areas of the installed RPV are available for visual and/or nondestructive inspection:
(1) RPV shell
(2) Primary coolant nozzles
(3) Bottom head
(4) Field welds between the RPV, no zzles, and the main coolant piping
The design considerations that have been incorporated into the system to permit the
above inspections are as follows:
(1) All reactor internals are completely removable. Appropriate tools, and the storage space required to permit these inspections, are provided.
(2) The RVCH is stored dry on the reactor operating deck during refueling to facilitate direct visual inspection.
(3) All RPV studs, nuts, and washers are removed to dry storage during refueling.
(4) Removable plugs are provided in the primary shield. The insulation covering the nozzle welds may be removed.
(5) A removable plug is provided in the lower core support plate to allow remote access for inspection of the bottom head without removal of the
lower internals.
The RPV presents access problems because of the radiation levels and remote underwater accessibility to this component.
Because of these limitations, several steps have been incorporated into the design and manufacturing procedures in preparation for DCPP UNITS 1 &
2 FSAR UPDATE 5.4-4 Revision 23 December 2016 the periodic nondestructive tests that are re quired by the ISI program, and in accordance with ASME BPVC Section XI-2001 through 2003 Addenda. These are:
(1) Shop ultrasonic examinations wer e performed on all internally clad surfaces to acceptance and repair standards that ensure an adequate cladding bond to allow later ultrasonic testing of the base metal from the
inside surface. The size of cladding bonding defect allowed is 3/4-inch by
3/4-inch.
(2) The design of the RPV shell in the core area is a clean, uncluttered, cylindrical surface to permit positioning of the ISI test equipment without
obstruction.
(3) After the shop hydrostatic testing, selected areas of the RPV were ultrasonically tested and mapped to facilitate the ISI program.
5.4.2 FEATURES FOR IMPROVED RELIABILITY RPV performance reliability is based on a conservative design, adequate protection measures, proper selection of materials, appropriate fabrication processes, quality
assurance program implementation, conservative operating procedures, and an
adequate ISI and material surveillance program. Section 5.2 addresses RPV design, overpressure protection, material selection, pressure and temperature operating
limitations, and surveillance programs. Fabrication and quality assurance measures are
discussed below.
5.4.3 PROTECTION OF CLOSURE STUDS Refueling procedures require the studs, nuts, and washers be removed from the RVCH and placed in storage racks during preparation for refueling. The storage racks are then
removed from the refueling cavity for maintenance and inspection prior to reactor
closure and refueling cavity flooding. Therefore, the RVCH studs are never exposed to the borated refueling cavity water.
The stud holes in the reactor flange are sealed with special plugs before removing the
RVCH, thus preventing leakage of the borated refueling water into the stud holes.
5.4.4 MATERIALS AND INSPECTIONS RPV materials are listed in Table 5.2-11. Construction, inspections and tests for the RPV and appurtenances are presented in Table 5.4-2. ISIs meet the requirements of ASME BPVC Section XI-2001 through 2003 Addenda, as referenced in 10 CFR 50.55a.
DCPP UNITS 1 &
2 FSAR UPDATE 5.4-5 Revision 23 December 2016 5.4.5 SPECIAL PROCESSES FOR FABRICATION AND INSPECTION 5.4.5.1 Fabrication Processes (1) Minimum preheat requirements were established for pressure boundary welds using low alloy weld materi al. Special preheat requirements were added for stainless steel cladding of low-stressed areas. Preheat was maintained until post-weld heat treatment, except for overlay cladding.
Limitations on preheat requirements (a) decrease the probabilities of weld
cracking by decreasing temperature gradients, (b) lower susceptibility to
brittle transformation, (c) prevent hydrogen embrittlement, and (d) reduce
peak hardness.
(2) On Unit 2, the use of severely sensitized stainless steel as a pressure boundary material was prohibited and eliminated either by choice of material or by programming the assembly method. This restriction on the
use of sensitized stainless steel provides the primary system with
preferential materials suitable for:
(a) Improved resistance to contaminants during shop fabrication, shipment, construction, and operation
(b) Application of critical areas.
Refer to Sections 5.2.2.5 and 5.2.3.22 for discussion of sensitization of RCS components.
(3) Galling prevention is accomplished by chrome plating of the surfaces of the guide studs in the RPV flange.
(4) Cracking prevention is accomplished by ensuring that the final joining beads are Inconel weld metal at all locations in the RPV where stainless steel and Inconel are joined.
(5) Core region shells fabricated of p late material have longitudinal welds and are angularly located away from t he peak neutron exposure experienced in the RPV.
5.4.5.2 Tests and Inspections Tests and inspections for the RPV and appurtenances are listed in Table 5.4-2. They
are discussed below.
DCPP UNITS 1 &
2 FSAR UPDATE 5.4-6 Revision 23 December 2016 5.4.5.2.1 Ultrasonic Examinations The following ultrasonic examinations were performed:
(1) During fabrication, angle beam inspection of 100 percent of plate material is performed to detect discontinuities that may be undetected by
longitudinal wave examination, in addition to the design code straight
beam ultrasonic test.
HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED (2) The RPV is examined after hydrot esting to provide a baseline map for use as a reference document in relation to later ISIs.
5.4.5.2.2 Penetrant Examinations The partial penetration welds for the CRDM head adapter are inspected by dye penetrant after the first layer of weld material, after each 1/4-inch of weld metal, and the
final surface. Bottom instrumentation tubes are inspected by dye penetrant after each
layer of weld metal. Core support block attachment welds are inspected by dye
penetrant after the first layer of weld metal and after each 1/2-inch of weld metal. This
is required to detect cracks or other defects, to lower the weld surface temperatures for cleanliness, and to prevent microfissures. All austenitic steel surfaces are 100 percent
dye penetrant tested after the hydrostatic test.
5.4.5.2.3 Magnetic Particle Examination (1) All surfaces of quenched and tempered materials are inspected on the inside diameter prior to cladding and the outside diameter is 100 percent inspected after hydrotesting. This serves to detect possible defects resulting from the forming and heat treatment operations.
(2) The attachment welds for the RPV supports, lifting lugs, and refueling seal ledge are inspected after the first layer of weld metal and after each 1/2-
inch of weld thickness. Where welds are back chipped, the areas are inspected prior to welding.
(3) All carbon steel surfaces are mag netic particle tested after the hydrostatic test. 5.4.6 QUALITY ASSURANCE SURVEILLANCE The surveillance program that calls for RPV quality assurance provisions to verify proper fabrication and to ensure that integrity is maintained throughout the plant's
lifetime, is listed in Table 5.4-2.
DCPP UNITS 1 &
2 FSAR UPDATE 5.4-7 Revision 23 December 2016 5.4.7 REACTOR PRESSURE VESSEL DESIGN DATA The RPV design parameters are presented in Table 5.4-1.
5.4.8 REACTOR PRESSURE VESSEL EVAULATION Section 5.2 presents an assessment of the stresses induced in the RPV during normal, upset, and faulted conditions, showing that in all cases they are below the respective
allowable stresses (refer to Tables 5.2-5, 5.2-6, and 5.2-7).
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-1 Revision 23 December 2016 5.5 COMPONENT AND SUBSYSTEM DESIGN This section discusses performance require ments and design features of the various components of the RCS and associated subsystems.
5.5.1 REACTOR COOLANT PUMPS Each unit has four identical RCPs, one in each loop.
5.5.1.1 Design Bases The RCP ensures adequate core cooling by forced circulation flow, and hence sufficient heat transfer, to maintain a DNBR greater than the applicable limit value (refer to Sections 4.4.
2.1 and 4.4.3.3) for all normal modes of operation. The required NPSH is, by conservative pump design, always less than that available by system design and
operation.
Sufficient pump rotation inertia is provided by a flywheel, in conjunction with the impeller
and motor assembly, to provide adequate flow during coastdown. This flow provides
the core with adequate cooling, follow ing an assumed loss of pump power.
The RCP motor has been tested without mechanical damage, at overspeeds up to and including 125 percent of normal speed (refer to Section 5.2.3.20).
The RCP is shown in Figure 5.5-1; its design parameters are provided in Table 5.5-1.
Code applicability and material requirements are provided in Tables 5.2-2 and 5.2-13, respectively.
5.5.1.2 Design Description The RCP is a vertical, single-stage, centrifugal, shaft seal pump designed to pump large
volumes of main coolant at high temperatures and pressures.
The pump consists of, from bottom to top, the hydraulic section, the shaft seal, and the motor. Each section is described as follows:
(1) The hydraulic section consists of an impeller, diffuser, casing, thermal barrier, heat exchanger, lower radial bearing, bolting ring, motor stand, and pump shaft.
(2) The shaft seal section consists of the No. 1 controlled leakage, film riding face seal, a shutdown seal (SDS) assembly, and the No. 2 and No. 3 rubbing face seals. The seals are contained within the main flange and seal housing.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-2 Revision 23 December 2016 (3) The motor section consists of a vertical solid-shaft, squirrel cage induction-type motor, and oil-lubricated double Kingsbury-type thrust bearing, two oil-lubricated radial bearings, and a flywheel.
Attached to the bottom of the pump shaft is the impeller. The reactor coolant is drawn up through the impeller, discharged through passages in the diffuser, and out through
the discharge nozzle in the side of the casing. A thermal barrier heat exchanger above
the impeller limits heat transfer between hot system water and pump internals. A weir
plate, installed in the pump discharge nozzle, prevents excessive flow of ECCS injection water into the casing in the event of an SBLOCA.
High-pressure seal injection water is introduced through the thermal barrier wall. A
portion of this water flows through the seals; the remainder flows downward into the
RCS, where it acts as a buffer to prevent system water from entering the radial bearing and seal section of the unit. The heat exchanger provides a means of cooling system water entering the pump radial bearing and seal section to an acceptable level in the
event that seal injection flow is l ost. The water-lubricated journal-type pump bearing, mounted above the thermal barrier heat exchanger, has a self-aligning spherical seat.
The RCP motor bearings are of conventional design. The radial bearings are the
segmented- pad-type and the thrust bearings are tilting pad Kingsbury bearings. All are oil-lubricated. The lower radial bearing and the thrust bearings are submerged in oil
and the upper radial bearing is fed oil from the oil flow off the outer surface of the thrust
runner.
The motor is an air-cooled squirrel cage induction motor. The insulation class of the motor is listed in Table 5.5-1. The rotor and stator are of standard construction and are cooled by air. A minimum of six RTDs are located throughout the stator to sense the winding temperature. The top of the motor consists of a flywheel and an anti-reverse rotation device.
Each RCP is equipped with a syst em to monitor shaft vibration. The system monitors pump shaft radial vibration, motor shaft radial vibration, and motor frame velocity. The
two pump shaft radial vibration probes are mounted in a horizontal plane above the seal
housing with one probe parallel to the pump discharge and the other perpendicular to
the pump discharge. The two motor shaft vibration probes are mounted in a horizontal
plane below the lower motor bearing with on e probe parallel to the pump discharge and the other perpendicular to the pump discharge. The two velocity probes are mounted in
a horizontal plane on the motor stand with on e probe parallel to the pump discharge and the other perpendicular to the pump discharge. A keyphasor probe is mounted below
the lower motor bearing and is used for spectral analysis and to measure pump speed.
In the event that the signal from a probe becomes invalid and becomes a nuisance
alarm the signal may be defeated, since the probes and cables are not accessible
during power operation.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-3 Revision 23 December 2016 The instrumentation monitors are mounted in a common rack located on the operating deck in containment. Alarms in the control room are provided by the rack in
containment. Vibration data from the instrument rack is collected and stored on a
server in the control room, and analyzed at a personal computer in the administration
building. The server and computer are shared by both units. The server or computer
may be turned off to support maintenance or power switching, as the vibration
equipment will still provide alarms and indicat ion. If the server is off, indication requires connection of test equipment to the local rack.
The RCP vibration monitoring system does not perform a PG&E Design Class I function.
As shown in Table 5.2-13, all parts of the pump in contact with the reactor coolant are
austenitic stainless steel except for seals, bearings, and special parts. CCW is supplied to the two oil coolers on the pump motor and to the pump thermal barrier heat
exchanger.
The pump shaft, seal housing, thermal barrier, bolting ring, and motor stand can be
removed from the casing as a unit without disturbing the reactor coolant piping. The
flywheel is available for inspection by removing the cover.
The performance characteristic, shown in Figure 5.5-2, is common to all of the fixed-
speed mixed-flow pumps, and the "knee" at about 45 percent design flow introduces no
operational restrictions since the pumps operate at full speed.
5.5.1.3 Design Evaluation This section discusses RCP design features incorporated to ensure safe and reliable operation while maintaining RCS integrity.
5.5.1.3.1 Pump Performance The RCPs are sized to equal or exceed the required flowrates (refer to Section 5.1.6).
Initial RCS tests confirm the total deliv ery capability. Thus, assurance of adequate forced circulation coolant flow is provided prior to initial plant operation.
The RTS ensures that pump operation is within the assumptions used for loss-of-coolant flow analyses, which also ensures t hat adequate core cooling is provided to permit an orderly reduction in power if flow from an RCP is lost during operation.
HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED An extensive test program was conducted for several years to develop the controlled
leakage shaft seal for pressurized water reactor applications. Long-term tests were
conducted on less than full-scale prototype seals as well as on full-size seals.
Operating plants continue to demonstrate the satisfactory performance of the controlled
leakage shaft seal pump design.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-4 Revision 23 December 2016 The support of the stationary member of the No. 1 seal (seal ring) is such as to allow large deflections, both axial and tilting, whi le still maintaining its controlled gap relative to the seal runner. Even if all the graphite were removed from the pump bearing, the shaft could not deflect far enough to cause opening of the controlled leakage gap. The "spring-rate" of the hydraulic forces associated with the maintenance of the gap is high
enough to ensure that the ring follows the runner under very rapid shaft deflections.
Testing of pumps with the No. 1 seal entirely bypassed (full reactor pressure on the
No. 2 seal) shows that relatively small leakage rates would be maintained for long
periods of time. The plant operator is warned of this condition by the increase in No. 1
seal leakoff, and has time to close this line and to conduct a safe plant shutdown
without significant leakage of reactor coolant to the containment. Thus, it may be
concluded that gross leakage from the pump does not occur, even if seals were to
suffer physical damage.
The effect of loss of offsite power on the pump itself is to cause an RCS pump trip, and
temporary stoppage in the supply of injection water to the pump seals and CCW to the thermal barrier for seal and bearing cooling if a generator trip results. The emergency
diesel generators are started automatically d ue to loss of offsite power, so that CCW flow is automatically restored to ensure cooling of the pump seals and bearings when
the reactor coolant temperature is above 150°F. Seal water injection flow is
subsequently restored by automatically restarting a charging pump on diesel generator
electrical power.
The SDS is housed within the No. 1 seal area and is a passive device actuated by high temperature resulting from a loss of seal injection and CCW cooling to the thermal barrier heat exchanger. The SDS is designed to function only when exposed to an elevated fluid temperature downstream of the RCP No. 1 seal, resulting from a loss of seal injection and CCW flow to the thermal barrier heat exchanger. SDS deployment limits leakage from the RCS through the RCP seal package. Leakage is limited when the SDS thermal actuator retracts due to intrusion of hot reactor coolant water into the seal area, which causes the SDS seal ring to constrict around the No. 1 seal sleeve.
5.5.1.3.2 Coastdown Capability It is important to reactor operation that the reactor coolant continues to flow for a short time after reactor trip. To provide this flow after a reactor trip, each RCP is provided with a flywheel. Thus, the rotating inertia of the pump, motor, and flywheel is employed
during the coastdown period to continue the reactor coolant flow.
An inadvertent actuation of the SDS on a rotating assembly will not have any measurable impact on RCP coastdown or on the pumps capability to provide sufficient cooling flow to the reactor core.
The pump is designed for the DE at the site. Bearing integrity is maintained as discussed below. It is, therefore, conclud ed that the coastdown capability of the pumps is maintained even under the most adverse case of a pump trip coincident with the DE.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-5 Revision 23 December 2016 5.5.1.3.3 Flywheel Integrity Integrity of the RCP flywheel is discussed in Sections 5.2.3.20 and 5.2.3.21.
5.5.1.3.4 Bearing Integrity The design requirements for the RCP bearings are primarily aimed at ensuring a long
life with negligible we ar, so as to give accurate alignment and smooth operation over long periods of time. To this end, the surface-bearing stresses are held at a very low
value, and, even under the most severe seismic transients, do not begin to approach loads which cannot be adequately carried for short periods of time.
Because there are no established criteria for short-term, stress-related failures in such bearings, it is not possible to make a meaningful quantification of such parameters as
margins to failure, safety factors, etc. A qualitative analysis of the bearing design, embodying such considerations, gives assurance of the adequacy of the bearing to
operate without failure.
High/low oil level in the motor bearings sign als an alarm in the control room. Each motor bearing contains embedded temperature detectors, and so initiation of failure, separate from loss of oil, is indicated and alarmed in the control room as a high bearing
temperature. Even if these indications are ignored and the bearing proceeds to fail, the
low melting point of Babbitt metal on the pad surfaces ensures that no sudden seizure
of the bearing occurs. In this event, the mot or continues to drive since it has sufficient reserve capacity to operate until it can be shut down.
The RCP shaft is designed so that its critical speed is well above the operating speed.
5.5.1.3.5 Locked Rotor The postulated case in which the pump impeller severely rubs on a stationary member
and then seizes, was evaluated (refer to Section 15.4.4 for the evaluation of this event on the RCS as a whole). The analysis showed that under such conditions, assuming instantaneous seizure of the impeller, the pump shaft fails in torsion just below the coupling to the motor, disengaging the flywheel and motor from the shaft. This constitutes a loss of coolant flow in the loop. Following such a postulated seizure, the
motor continues to run without any overspeed, and the flywheel maintains its integrity
since it is still supported on a shaft with two bearings.
There are no credible sources of shaft seizure other than impeller rubs. Any seizure of
the pump bearing is precluded by the graphite in the bearing. Any seizure in the seals
results in a shearing of the anti-rotation pin in the seal ring. The motor has adequate
power to continue pump operation even after the above occurrences. Indications of
pump malfunction in these conditions are first, by high-temperature signals from the
bearing water temperature detector, and second, by excessive No. 1 seal leakoff DCPP UNITS 1 &
2 FSAR UPDATE 5.5-6 Revision 23 December 2016 indications. Along with these signals, pump vibration levels are checked. When there are indications of a serious malfunction, the pump is shut down for investigation.
5.5.1.3.6 Critical Speed The RCPs are designed to operate below first critical speed. This results in a shaft
design that, even under the most severe postulated transient, gives very low stress
values.
Both the damped and lateral natural frequencies are determined by establishing a
number of shaft sections and applying weights and moments of inertia for each section
bearing spring and damping data.
The torsional natural frequencies are similarly determined. The lateral and torsional natural frequencies are greater than 120 and
110 percent of the running speed, respectively.
5.5.1.3.7 Missile Generation Each pump component is analyzed for missile generation. Any fragments of the motor rotor would be contained by the heavy stator. The same conclusion applies to the pump
impeller because the small fragments that might be ejected would be contained by the
heavy casing (refer to Section 5.2.3.20).
5.5.1.3.8 Pump Cavitation The minimum NPSH required by the RCP at running speed is approximately 170 feet (approximately 74 psi). For the controlled leakage seal to operate correctly, a differential pressure of approximately 200 psi across the seal is necessary. This results in a requirement for a minimum of 325 psi pressure in the primary loop before the RCP
may be operated. This 325 psi requirement is for initial fill and vent only. In normal
operation. This requirement is reflected in the operating instructions. At this pressure, the NPSH requirement is exceeded and no limitation on pump operation occurs from
this source.
5.5.1.3.9 Pump Overspeed Considerations The generator and the RCP remain electricall y connected for 30 seconds following turbine trip actuated by either the RTS or the turbine protection systems, except for certain trips caused by electrical or mechanical faults which require immediate tripping
of the generator. A complete load disconnect with turbine overspeed would result in an
overspeed potential for the RCP. The turbine control system and the turbine intercept
valves limit the overspeed to less than 120 percent, which is less than the design overspeed of the RCP. As additional backup, the main turbine has a mechanical overspeed protection trip usually set at about 110 percent.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-7 Revision 23 December 2016 The details of the turbine trip interface logic are shown in Figures 5.5-13 and 5.5-17.
The sequence of events following a generator trip, which transfers the ESFs onto the standby power supply, is discussed in Section 8.3.
5.5.1.3.10 Anti-reverse Rotation Device Each RCP is provided with an anti-reverse rotation device in the motor. This
anti-reverse mechanism consists of five pawls mounted on the outside diameter of the flywheel, a serrated ratchet plate mounted on the motor frame, a spring return for the ratchet plate, and three shock absorbers.
After the motor comes to a stop, a minimum of one pawl engages the ratchet plate and, as the motor tends to rotate in the opposite direction, the ratchet plate also rotates until
stopped by the shock absorbers. The rotor remains in this position until the motor is
energized again. After the motor comes up to speed, the ratchet plate is returned to its
original position by the spring return.
When the motor is started, the pawls initially drag over the ratchet plate. Once the motor reaches sufficient speed, centrifugal forces acting on the pawls produce enough
friction to prevent the pawls from rotating, and thus hold the pawls in the elevated
position until the motor is stopped.
5.5.1.3.11 Shaft Seal Leakage During normal operation, leakage along the RCP shaft is controlled by three shaft seals arranged in series so that reactor coolant leakage to the containment is essentially zero.
Charging flow is directed to each RCP via a 5-micron seal (maximum) water injection filter. It enters the pu mps through the thermal barrier and is directed down to a point between the pump shaft bearing and the thermal barrier cooling coils. Here the flow
splits and a portion flows down past the thermal barrier cooling cavity and labyrinth seals. The remainder flows up the pump shaft, cooling the lower bearing, and leaves
the pump via the No. 1 seal bypass line or the No. 1 seal leakoff line. There is also a
minor flow through the No. 2 seal.
Leakoff flow through the No. 1 seal from each pump is piped to a common manifold, and then, via a seal water return filter, through a seal water heat exchanger, to the VCT.
The VCT provides a back pressure of at least 15 psig on the No. 1 seal.
A small amount of No. 1 seal leakoff passes through the No. 2 seal. No. 2 seal leakoff
flows to the reactor coolant drain tank (RCDT).
The No. 3 seal is a double dam seal that divides seal flow into two paths. Part of the
flow is directed radially outward to join the No. 2 seal leakoff line and the second part
flows radially inward to the No. 3 seal leakoff line to the containment structure sump. A
standpipe is provided to ensure a back pressure of at least 7 feet of water on the No. 3
seal.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-8 Revision 23 December 2016 In the event of a loss of seal injection and CCW flow to the thermal barrier heat exchanger, reactor coolant begins to travel along the RCP shaft and displaces the cooler seal injection water. The SDS, designed to actuate only when exposed to an elevated fluid temperature downstream of the RCP No. 1 seal, deploys via retraction of a thermal actuator, which causes the SDS seal ring to constrict around the No. 1 seal sleeve. SDS deployment controls shaft seal leakage and limits the loss of reactor coolant via the RCP seal package to 1 gpm or less.
5.5.1.3.12 Spacer Couplers The installation of a removable spool piece, shown in Figure 5.5-3, in the RCP shaft facilitates the inspection and maintenance of the pump seal system without breaking
any of the fluid, electrical, or instrumentation connections to the motor, without removal of the motor.
5.5.1.4 Tests and Inspections Support feet are cast integral with the casing to eliminate a weld region. The design
enables disassembly and removal of the pump internals for normal access to the
internal surface of the pump casing.
The RCP quality assurance program is given in Table 5.5-2. Refer to Sections 5.1.8.19, 5.1.8.20, 5.2.3.14, and 5.2.3.15 for further discussion of testing and inspection of the RCS. HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED
5.5.1.4.1 Electroslag Welding
RCP casings fabricated by electroslag welding were qualified as follows:
(1) The electroslag welding procedure employing 2- and 3-wire technique was qualified in accordance with the requir ements of the ASME BPVC,Section IX, and Code Case 1355 (refer to Table 5.2-1) plus supplementary evaluations specified by Westinghouse.
(2) A separate weld test was made using the 2-wire electroslag technique to evaluate the effects of a stop and restar t of welding by this process. This evaluation was performed to establish proper procedures and techniques
as such an occurrence was anticipated durin g production applications due to equipment malfunction, power outages, etc.
(3) All of the weld test blocks in (1) and (2) above were radiographed using a 24 MeV betatron. The radiographic quality level obtained was between DCPP UNITS 1 &
2 FSAR UPDATE 5.5-9 Revision 23 December 2016 0.5 and 1 percent, as defined by ASTM E-94. There were no discontinuities evident in any of the electroslag welds.
The casting segments were surface conditi oned for 100 percent radiographic and penetrant inspections. The radiographic acceptance standards were ASTM E-186
Severity Level 2 except no Category D or E defectives were permitted for section
thicknesses up to 4-1/2 inches and ASTM E-280, Severity Level 2, for section
thicknesses greater than 4-1/2 inches. The edges of the electroslag weld preparations were machined. These surfaces were also penetrant inspected prior to welding. The
penetrant acceptance standards were those of the ASME BPVC,Section III, Paragraph N-627.
The completed electroslag weld surfaces were ground flush with the casting surface.
The electroslag weld and adjacent base material were then 100 percent radiographed in
accordance with ASME BPVC Case 1355. Also, the electroslag weld surfaces and adjacent base material were penetrant inspected in accordance with ASME BPVC,Section III, Paragraph N-627. Weld metal an d base metal chemical and physical properties were determined and certified. Heat treatment furnace charts were recorded
and certified, and are available at the NSSS vendor's facilities.
5.5.1.4.2 In-process Control of Variables
Many variables must be controlled to mai ntain desired quality welds. These variables and their relative importance are as follows:
(1) Heat Input vs. Output The heat input is determined by the product of volts and current and
measured by voltmeters and a mmeters, which are considered accurate and are calibrated every 30 days. During any specific weld these meters
are constantly monitored by the operators.
(2) Weld Gap Configuration The weld gap configuration is controlled by 1-1/4-inch spacer blocks. As
these blocks are removed, there is the possibility of gap variation. It has
been found that a variation from 1 to 1-3/4 inches is not detrimental to
weld quality as long as the current is adjusted accordingly.
(3) Flux Chemistry The flux used for welding is Arcos BV-I Vertomax. This is a neutral flux, the chemistry of which is specified by Arcos Corporation. The molten slag is kept at a nominal depth of 1-3/4 inches and may vary in depth by plus or minus 3/8 inch without affecting t he weld. This is measured with a stainless steel dipstick.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-10 Revision 23 December 2016 (4) Weld Cross-Section Configuration The higher the current or heat input and the lower the heat output, the greater the dilution of weld metal with base metal. This causes a rounder
barrel-shaped configuration compared to welding with lower heat input and higher heat output, which reduces the amount of dilution and provides
a more narrow barrel-shaped configuration. Configuration is also a
function of section thickness; the thinner the section, the rounder the
pattern produced.
5.5.1.4.3 Welder Qualification
Welder qualification is in accordance with ASME BPVC,Section IX rules.
5.5.2 STEAM GENERATORS Each RCS loop contains a vertical U-tube SG. The SGs provide high quality steam to the turbine. The tube and tubesheet boundary are designed to prevent the transfer of radioactivity generated within the core to the secondary system.
5.5.2.1 Design Bases SG design data are provided in Table 5.5-3. The design can sustain the transient conditions identified in Table 5.2-4. Estima tes of radioactivity levels anticipated in the secondary side of the SGs during normal operation and their bases for the estimates are discussed in Section 11.1. The transient analysis of a SGTR is discussed in Section 15.4.
When operating at 100 percent power, integral moisture separating equipment reduces
moisture content of the steam at the exit following transient conditions, the moisture content at the exit of the SGs is <0.25 percent:
- loading or unloading at a rate of 5 percent of full power steam flow per minute in the range from 15 to 100 percent of full load steam flow
- a step load change of 10 percent of full power in the range from 15 to 100 percent of full load steam flow
The SG tubesheet complex meets the stress limitations and fatigue criteria specified in ASME BPVC Section III-1998 through 2000 Addenda. Per Section 5.2.2.1.5.3, emergency conditions do not apply. Codes and materials requirements of the SG are listed in Tables 5.2-2 and 5.2-14, respectively. The SG design maximizes integrity against hydrodynamic excitation and vibration failure of the tubes for plant life.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-11 Revision 23 December 2016 The water chemistry in the reactor side is selected to provide the necessary boron content for reactivity control and to minimize corrosion of RCS surfaces. Water
chemistry for the primary coolant side is presented in Table 5.2-15.
5.5.2.1.1 Design Basis for the Steam Ou tlet Nozzle Flow Restrictor The design criterion for the steam nozzle flow restrictors is to limit steam flow in the event of an MSLB during normal operating conditions, in order to reduce pressure drop loadings on the SG internal components, as well as to limit the mass and energy release rate into the containment.
5.5.2.2 Design Description The SG, shown in Figure 5.5-4, is a vertical shell and U-tube design with evaporators having integral moisture separating equipment. The reactor coolant flows through the
inverted U-tubes, entering and leaving through the nozzles located in the hemispherical
bottom head of the SG. Steam is generated on the shell side and flows upward through the moisture separators to the outlet nozzle at the top of the vessel. The head is
divided into inlet and outlet chambers by a vertical partition plate extending from the
head to the tubesheet. Manways are provided for access to both sides of the divided
head.
The SG unit is primarily carbon steel. The heat transfer tubes and the divider plate are Inconel and the interior surfaces of the reactor coolant channel heads and nozzles are clad with austenitic stainless steel. The primary side of the tubesheet is weld clad with Inconel. Feedwater is introduced into the SGs through a feedwater nozzle located in the upper shell. The nozzle does not require a flow-limiting device because the feedring itself provides this function. The nozzle contains a welded thermal liner that minimizes the impact of rapid feedwater temperature transients on the nozzle. The feedwater distribution ring is welded to the feedwater nozzle to minimize the potential for draining the ring. The feedring is located above the elevation of the feed nozzle to minimize the time required to fill the feed nozzle during a cold water addition transient. The feedwater is discharged through spray nozzles installed on the top of the ring. These features reduce the thermal fatigue loading on the feedwater nozzle, eliminate steady-state thermal stratification in the feedwater nozzle and feedwater piping elbow at the feedwater nozzle entrance, and minimize the potential for bubble- collapse water hammer in the feedwater distribution ring. The feedwater piping elbow at the feedwater nozzle entrance also contains an elbow thermal liner that minimizes the effects of thermal stratification on the elbow-to-nozzle weld and the weld of the feedwater inlet thermal sleeve to feedwater nozzle.
The SG feedring is fabricated from alloy steel with a significant chromium content to provide enhanced erosion/corrosion resistance characteristics. The feedring has spray nozzles that are spaced around the feedring circumference to distribute the feedwater DCPP UNITS 1 &
2 FSAR UPDATE 5.5-12 Revision 23 December 2016 into the upper shell recirculating water pool. The spray nozzle perforations also act to prevent loose parts ingress from the feedwater system.
Subsequently, the water-steam mixture flows upward through the tube bundle and into the steam drum section. A set of centrifugal moisture separators, located above the tube bundle, removes most of the entrained water from the steam. The moisture separators recirculate flow that mixes with feedwater as it enters the downcomer formed by the shell and tube bundle wrapper. Steam dryers are employed to increase the steam quality to a minimum of 99.95 percent, which corresponds to a steam outlet
moisture content of 0.05 percent. The dryers can be inspected, or disassembled and removed, through one of two bolted and gasketed secondary manway access openings.
5.5.2.2.1 Design Description of the Steam Outlet Nozzle Flow Restrictor An integral flow restrictor is provided in each steam nozzle to limit flow in the event of an MSLB accident downstream of the steam nozzle. The flow restrictor consists of seven holes in the steam outlet nozzle forging, with Venturi type flow limiting inserts installed in each of these holes. The total minimum flow area is 1.4 ft 2 for the seven inserts. The Alloy 690 flow limiting inserts are welded to t he Alloy 690 cladding at the steam nozzle bottom. Materials, welding, and inspection requirement s applied in fabrication of the steam nozzle flow restrictor assemblies conform to ASME BPVC Section III-1998 through 2000 Addenda requirements.
The steam outlet nozzle flow restrictor assembly is shown in Figure 5.5-18.
5.5.2.3 Design Evaluation 5.5.2.3.1 Forced Convection The limiting case for heat transfer capability is the nominal 100 percent design thermal
duty. To ensure that this thermal duty will be met, the SGs are designed to operate with an effective fouling factor, or heat transfer resistance, that is greater than that experienced for comparable units in service. Adequate tubing area is selected to
ensure that the full design heat removal rate is achieved for these conditions.
The historical best estimate fouling factor applie d to Alloy 690-TT tubing is 0.00006 hr-ft 2-°F/Btu. The design fouling factor for the Diablo Canyon SGs is 0.00018 hr-ft 2-°F/Btu.
When added to the conduction resistance of the tubing, this additional resistance
accounts for approximately 17 percent margin for heat transfer, i.e., a 17 percent higher
heat transfer coefficient is expected compared to the design value. This margin
ensures that the SGs will provide sufficient heat transfer capability through the design life.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-13 Revision 23 December 2016 5.5.2.3.2 Natural Circulation Flow The driving head created by the change in coolant density as it is heated in the core and
rises to the outlet nozzle initiates convection circulation. This circulation is enhanced by
the fact that the SGs, which provide a heat sink, are at a higher elevation than the reactor core, which is the heat source. Thus, natural circulation is ensured for the
removal of decay heat during hot shutdown in the unlikely event of loss of forced
circulation. This was confirmed by DCPP Unit 1 testing.
5.5.2.3.3 Secondary System Fluid Flow Instability Prevention Undesirable perturbations in secondary side flow are postulated to result from events such as water hammer and circulation loop instability. Such events can compromise
the functional capability and mechanical inte grity of the secondary system. The SGs include design features intended to preclude these occurrences.
The potential for water hammer is mitigated by the inclusion of an upward-sloping
section of the feedwater ring header. This reduces the volume within the feedwater ring
assembly that could potentially be filled with steam, and also reduces the possibility of
thermal stratification in the feed flow. The SGs include top-discharge spray nozzles, which further reduce the possibility of steam pockets being trapped in the feedwater
ring, and also serve as a means to prevent loose parts from entering the SG through the feedwater system.
Instability in the circulation loop for the secondary fluid can result from a distribution of
pressure drops that favors two-phase flow, which is de-stabilizing and is found in the upper tube bundle and moisture separators, as opposed to single-phase flow, which is stabilizing and is found in the downcomer and lower tube bundle areas. A stability damping factor is determined in which a negative value indicates damped, stable circulation flow. The SGs are designed to provide damped, stable circulation over the full range of operating conditions, with sufficient margin to prevent increased two-phase
pressure drop, caused by conditions such as a partially blocked, broached tube support plate flow area, from causing instability.
5.5.2.3.4 Tube and Tubesheet Stress Analyses Tube and tubesheet stress analyses for the SGs confirm that the SG tubesheet will withstand the loading (quasi-static rather than shock loading) caused by LOCA. With the acceptance of the DCPP LBB analysis by the NRC (Reference 10), dynamic loading conditions resulting from pipe rupture events in the main RCL piping no longer have to be considered in the design basis analyses; only the much smaller dynamic loads
resulting from RCS branch line breaks have to be considered (refer to Section 3.6.2.1.1.1).
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-14 Revision 23 December 2016 5.5.2.3.5 Corrosion All volatile chemistry is used in the main steam, feedwater, and condensate systems to
provide improved corrosion protection and control.
The control measures exercised over the secondary water chemistry for the purpose of
inhibiting SG tube degradation consist of a program encompassing: (a) scheduled sampling and analyses of fluid systems for the critical control parameters, (b) recording, reviewing, and management of data, (c) identification of process sampling points, (d)
guidance for corrective actions for off-point chemistry, (e) identification of the authority responsible for the interpretation of data, and (f) the sequence and timing of
administrative events required to initiate corrective action.
Additional control measures for secondary water chemistry come from the turbine manufacturer. The program includes the mo nitoring of main steam purity. The SGs include a number of key design features that enhance operation, performance, and
maintenance. The design features and materials have been developed and selected to
minimize the potential for tube degradation. The design features enhance steam and
water flow by the tubes, which minimizes the potential for concentration of chemical
species that can be detrimental to tubing material.
The U-tubes are fabricated of nickel-chromium-iron (Ni-Cr-Fe) Alloy 690. The tubes undergo thermal treatment following tube-forming and annealing operations. The
thermal treatment subjects the tubes to elevated temperatures for a prescribed period of time to improve the microstructure of the material. Thermally treated Alloy 690 has been shown in laboratory tests and operating nuclear power plants to be very resistant to primary water stress corrosion cracking and outside diameter initiated stress corrosion cracking.
5.5.2.3.6 Design Evaluation for the Steam Outlet Nozzle Flow Restrictor In the event of an MSLB, steam flow rate from the SGs is restricted by the outlet nozzle Venturi inserts, which limit the steam blowdown rate from the SGs.
5.5.2.3.7 Flow-induced Vibration In the design of the SGs, the possibility of degradation of tubes due to either mechanical- or flow-induced excitation is thoroughly evaluated. This evaluation includes detailed analysis of the tube support systems as well as an extensive research program with tube vibration model tests.
In evaluating degradation due to vibration, consideration is given to sources of
excitation such as those generated by primary fluid flowing within the tubes, mechanically induced vibration, and secondary fluid flow on the outside of the tubes.
During normal operation, the effects of primary fluid flow within the tubes and
mechanically induced vibration are considered to be negligible and should cause little DCPP UNITS 1 &
2 FSAR UPDATE 5.5-15 Revision 23 December 2016 concern. Thus, the primary source of tube vibrations is the hydrodynamic excitation by the secondary fluid on the outside of the tubes. In general, three vibration mechanisms
have been identified:
(1) Vortex shedding (2) Fluidelastic excitation (3) Turbulence Vortex shedding does not provide detectable tube bundle vibration for the following reasons:
(1) Flow turbulence in the downcomer and tube bundle inlet region inhibits the formation of Von Karman's vortex train.
(2) The spatial variations of cross flow velocities along the tube preclude vortex shedding at a single frequency.
(3) Both axial and cross flow velocity components exist on the tubes. The axial flow component disrupts the Von Karman vortices.
The SG design is qualified by analyses (relying on theoretical calculations based on laboratory test data and operating SG experience), which demonstrate that no tubes will experience unacceptable degradation or wear due to vibration over the SG design life.
5.5.2.4 Tests and Inspections The SG quality assurance program is given in Table 5.5-5. Radiographic inspection and acceptance standards are in accordance with the requirements of ASME BPVC Section III-1998 through 2000 Addenda.
Liquid penetrant inspection was performed on weld deposited tubesheet cladding, channel head cladding, tube-to-tubesheet weldments, and weld deposit cladding.
Liquid penetrant inspection and acceptance standards are in accordance with the requirements of ASME BPVC Section III-1998 through 2000 Addenda.
Magnetic particle inspection was performed on all pressure boundary forgings (tubesheet, shell barrels, channel head, transition cone, elliptical head, and secondary-
side nozzles), and the following weldments:
- Nozzle to shell
- Upper lateral support lugs
- Instrument connections
- Temporary attachments after removal
- All accessible pressure-retaining welds after hydrostatic testing.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-16 Revision 23 December 2016 Magnetic particle inspection and acceptance standards were in accordance with the
requirements of ASME BPVC Section III-1998 through 2000 Addenda.
Ultrasonic examination was performed on all pressure boundary forgings (tubesheet, shell barrels, channel head, transition cone, elliptical head, primary nozzle safe ends, and secondary-side nozzles).
Manways provide access to both the primary and secondary sides of the SGs. Primary side inspection and maintenance is described in Section 5.5.2.5 and is typically
performed with nozzle dams in place to isolate the SG bowl from the RCS.
5.5.2.4.1 Tests and Inspections for the Steam Outlet Nozzle Flow Restrictor The flow restrictor Venturi inserts at the steam outlet are located inside the steam outlet nozzle and welded to the cladding.
Therefore, the flow restrictor inserts are not a
pressure boundary component. However, component integrity is ensured by
compliance with ASME BPVC Section III-1998 through 2000 Addenda requirements.
5.5.2.5 Steam Generator Tube Surveillance Program 5.5.2.5.1 Inservice Inspection SG tube inspection is performed in accordance with the Technical Specifications (Reference 6) and the DCPP surveillance test procedure. Eddy current non-destructive
testing is used to perform tube inspections. The SG tube surveillance program ensures that the structural and leakage integrity of this portion of the RCS will be maintained.
The program for ISI of SG tubes is based on NEI 97-06 (Reference 5). ISI of SG tubing is essential in order to maintain surveillance of the conditions of the tubes in the event
there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. ISI of SG tubing also provides a means of characterizing the nature and cause of any tube degradation
so that corrective measures can be taken.
Tube degradation will be detected during scheduled inservice SG tube examinations.
SG tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20 percent of the original tube wall thickness.
Plugging is required for all tubes with imperfections exceeding the plugging limit defined in the Technical Specifications. Degraded tubes may be left in service if non-destructive examination sizing techniques verify that the imperfection is less than the plugging limit (References 5 and 6).
5.5.2.5.2 Primary-to-Secondary Leakage The plant is expected to be operated in a manner such that the secondary coolant will
be maintained within those chemistry limits found to result in negligible corrosion of the DCPP UNITS 1 &
2 FSAR UPDATE 5.5-17 Revision 23 December 2016 SG tubes. DCPP Technical Sp ecifications limit primary to secondary leakage through an SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or an MSLB. DCPP has demonstrated that primary-to-secondary leakage of 150 gallons per day per SG can readily be detected during power operation. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during
which the leaking tubes will be located and plugged.
Refer to Section 15.5.18 for a radiological assessment of accident-induced leakage in any one SG following an MSLB.
5.5.3 REACTOR COOLANT PIPING Reactor coolant piping provides a flowpath connecting the major components of each
RCS loop. The RCS piping consti tutes a boundary to contain the coolant under operating temperature and pressure conditions and limit leakage (and radioactivity release) to the containment atmosphere. It contains pressurized water that is circulated at a flowrate and temperature consistent with reactor core thermal and hydraulic performance requirements.
5.5.3.1 Design Bases The RCS piping was d esigned and fabricated to accommodate the stresses due to the pressures and temperatures attained under all expected modes of plant operation or
system interactions. Code and material requirements are provided in Table 5.2-2 and
Section 5.2.2.3.
Materials of construction are specified to minimize corrosion/erosion and ensure
compatibility with the operating environment.
Refer to Section 5.2.2.1.3 for the codes and standards applicable to the RCL and pressurizer surge line piping for both Unit 1 and Unit 2.
5.5.3.2 Design Description Principal design data for the RCS piping for both Unit 1 and Unit 2 are provided in Table 5.5-6. The RCS piping was specified in the smallest sizes consistent with system
requirements. In general, high fluid velocities are used to reduce piping sizes. This design philosophy results in the reactor inlet and outlet piping diameters listed in Table
5.5-6. The line between the SG and the pump suction is larger to reduce pressure drop and improve flow conditions to the pump suction. To further improve pump suction conditions, a flow splitter is provided in the pipe bend upstream of the pump suction.
The reactor coolant piping is seamless forged, and fittings are cast. Cast sections of large 90° elbows are joined by electroslag welds. All materials are austenitic stainless steel. All smaller piping that is part of the RCPB, such as the pressurizer surge line, DCPP UNITS 1 &
2 FSAR UPDATE 5.5-18 Revision 23 December 2016 spray and relief line, loop drains, and connecting lines to other systems are also austenitic stainless steel. The nitrogen supply line for the PRT is carbon steel. All joints and connections are welded, except for the PORVs and PSVs, where flanged joints are used. Thermal sleeves are installed at points in the system where high thermal
stresses could develop due to rapid changes in fluid temperature during normal
operational transients. These points include:
(1) Charging connections at the primary loop from the CVCS (2) Both ends of the pressurizer surge line (3) Pressurizer spray line connection at the pressurizer
Thermal sleeves were not provided for the remaining injection connections of the ECCS
since these connections are not in normal use.
All piping connections from auxiliary systems were made above the horizontal centerline of the reactor coolant piping, with the exception of:
(1) RHR pump suction, which is 45
° down from the horizontal centerline. This enables the water level in the RCS to be lowered in the reactor coolant
pipe while continuing to operate the RHR system, should this be required
for maintenance.
(2) Loop drain lines and the connection for temporary level measurement of water in the RCS during refueling and maintenance operation.
(3) The differential pressure taps for flow measurement are downstream of the SGs on the first 90
° elbow. There are three flow transmitters at each elbow. The transmitters at each elbow are arranged so that they use a
common high-pressure tap (on the outside of the elbow) and separate low
pressure taps (on the inside of the elbow). Additional discussion is
included in Section 7.2.2.1.4.
Penetrations into the coolant flowpath were limited to the following:
(1) The spray line inlet connections extend into the cold leg piping in the form of a scoop so that the velocity head of the RCL flow adds to the spray driving force.
(2) The reactor coolant sample system taps protrude into the main stream to obtain a representative sample of the reactor coolant.
(3) The narrow range RCS temperature sensors (RTDs) are mounted in thermowells that extend into the hot and cold legs. The RTD bypass scoops and nozzles have been capped.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-19 Revision 23 December 2016 (4) The wide range RCS temperature sensors (RTDs) are mounted in thermowells that protrude into the hot legs and cold legs.
Signals from these instruments are used to compute the reactor coolant T (temperature of the hot leg, T hot , minus the temperature of the cold leg, T cold) and an average reactor coolant temperature (T avg). The T avg and T for each loop are indicated on the main control board. Chapter 7 further describes the temperature sensor arrangement.
The RCPB piping includes those sections of piping interconnecting the RPV, SG, and RCP. It also includes the following:
(1) Charging line and alternate charging line from the isolation valve up to the branch connections on the RCL (2) Letdown line and excess letdown line from the branch connections on the RCL to the isolation valve (3) Pressurizer spray lines from the reactor coolant cold legs to the spray nozzle on the pressurizer vessel (4) RHR lines to or from the RCLs up to the designated isolation or check valve (5) SI lines from the designated isolation or check valve to the RCLs (6) Accumulator lines from the designated isolation or check valve to the RCLs (7) Loop fill, loop drain, sample, and instrument lines to or from the designated isolation valve to or from the RCLs (8) Pressurizer surge line from one RCL hot leg to the pressurizer vessel inlet nozzle (9) Abandoned RTD scoop element, pressurizer spray scoop, sample connection with scoop, reactor coolant temperature element installation
boss, and the temperature element thermowell itself (10) All branch connection nozzles attached to RCLs (11) Pressure relief lines from nozzles on top of the pressurizer vessel up to and through the PORVs and PSVs (12) Seal injection water and labyrinth differential pressure lines to or from the RCP inside reactor containment DCPP UNITS 1 &
2 FSAR UPDATE 5.5-20 Revision 23 December 2016 (13) Auxiliary spray line from the isolation valve to the pressurizer spray line header (14) Sample lines from pressurizer to the isolation valve (15) Pressurizer loop seal drain lines to the pressurizer.
Details of the materials of construction and codes used in the fabrication of reactor
coolant piping and fittings are discussed in Section 5.2.
5.5.3.3 Design Evaluation 5.5.3.3.1 Piping Load and Stress Evaluation Piping loads and stress evaluation methodology for normal, upset, and faulted
conditions are described in Section 5.2.2.1.
5.5.3.3.2 Material Corrosion/Erosion Evaluation The water chemistry is selected to minimize corrosion. A periodic analysis of the
coolant chemical composition is performed to verify that the reactor coolant quality meets the specifications. The RCS water chemistry is presented in Section 5.2.2.3.4 and Table 5.2-15.
An upper limit of about 50 feet per second is specified for internal coolant velocity to
avoid the possibility of accelerated erosion.
All pressure-containing welds within the RCPB are available for examinati on and have removable insulation.
5.5.3.4 Tests and Inspections 5.5.3.4.1 Inservice Testing and Inspection Refer to Sections 5.1.8.19, 5.1.8.20, 5.2.3.14, and 5.2.3.15 for further discussion of testing and inspection of the RCS.
5.5.3.4.2 Piping Quality Assurance The RCS piping quality assurance program is given in Table 5.5-7.
5.5.3.4.3 Electroslag Weld Quality Assurance The 90° elbows used in the RCL piping were electroslag welded. A description of this procedure is contained in Section 5.5.1.
The following quality assurance actions for RCS piping were undertaken:
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-21 Revision 23 December 2016 (1) The electroslag welding procedure employing 1-wire technique was qualified in accordance with the r equirements of ASME BPVC Section IX, and Code Case 1355 plus supp lementary evaluation.
(2) The casting segments were surface conditioned for 100 percent radiographic and penetrant inspections. The acceptance standards were USAS Code Case N-10, and ASTM E-186, Severity Level 2, except no
Category D or E defectives were permitted.
5.5.4 MAIN STEAM LINE FLOW RESTRICTORS As described in Section 5.5.2.2.1, each SG has a flow restrictor located in the steam outlet nozzle to limit the steam blowdown from the SGs in the event of an MSLB. The flow restrictor consists of seven 6.03-inch ID venturi nozzles. In addition, a 16-inch flow
restrictor is installed in each main steam line outlet to measure steam flow.
The main steam line flow restrictors are welded into the inside of a length of main steam
pipe. Therefore, the 16-inch flow restrictors are not a pressure boundary component.
However, component integrity is ensured by compliance with ASME Code
requirements.
5.5.5 MAIN STEAM LINE ISOLATION SYSTEM Each main steam line has one isolation valve and one check valve, both of the swing
check type, located outside the containment. The isolation valves are held open by a
pneumatic actuator until a trip signal is received, as discussed in Section 6.2.4. For analysis of the ability of these valves to close under pipe break conditions (refer to Section 10.3.2.1).
5.5.6 RESIDUAL HEAT REMOVAL SYSTEM The RHR system is a dual function system that is aligned to immediately serve as the low-head portion of the ECCS in Modes 1 through 3 and is also aligned to provide normal plant cooldown function in Modes 4 through 6. This section discusses the design, functions, and requirements of the RHR system while performing the plant cooldown function.
The RHR system functions in conjunction with the high-head and intermediate-head portions of the ECCS to provide injection of borated water from the refueling water storage tank (RWST) into the RCS cold legs during the injection phase following a LOCA. During normal operation, the RHR system is lined up to perform this emergency
function.
In its capacity as the low-head portion of the ECCS, the RHR system provides long-term recirculation capability for core cooling following the injection phase of the LOCA. This DCPP UNITS 1 &
2 FSAR UPDATE 5.5-22 Revision 23 December 2016 function is accomplished by aligning the RHR system to take suction from the containment recirculation sump.
For a more complete discussion of the use of the RHR system as part of the ECCS, refer to Section 6.3.
The RHR system transfers heat from the RC S to the CCW system to reduce reactor coolant temperature to the cold shutdown temperature at a controlled rate during the
latter part of normal plant cooldown, and maintains this temperature until the plant is
started up again.
The RHR system can also be used to transfer refueling water between the RWST and the refueling cavity before and after the refueling operations.
During the recirculation phase of a LOCA, if both RHR pumps are in operation, one RHR pump may be used to provide flow from the containment recirculation sump to two containment spray rings for continued post-accident spray operation (refer to Section 6.2.2.2). 5.5.6.1 Design Bases
5.5.6.1.1 General Design Criterion 2, 1967 - Performance Standards The RHR system is designed to withstand the effects of, or be protected against, natural phenomena, such as earthquakes, tornadoes, flooding, winds, tsunamis, and other local site effects.
5.5.6.1.2 General Design Criterion 3, 1971 - Fire Protection The RHR system is designed and located to minimize, consistent with other safety requirements, the probability and effects of fires and explosions.
5.5.6.1.3 General Design Criterion 9, 1967 - Reactor Coolant Pressure Boundary The portion of the RHR system that is part of the RCPB is designed and constructed so as to have an exceedingly low probabi lity of gross rupture or significant leakage throughout its lifetime.
5.5.6.1.4 General Design Criterion 11, 1967 - Control Room The RHR system is designed to or contains instrumentation and controls that support actions to maintain the safe operational status of the plant from the control room or from an alternate location if control room access is lost due to fire or other causes.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-23 Revision 23 December 2016 5.5.6.1.5 General Design Criterion 12, 1967 - Instrumentation and Control Systems Instrumentation and controls are provided, as required, to monitor and maintain RHR system variables within prescribed operating ranges.
5.5.6.1.6 General Design Criterion 40, 1967 - Missile Protection The ESF (containment isolation) portion of the RHR system is designed to be protected against dynamic effects and missiles that might result from plant equipment failures.
5.5.6.1.7 General Design Criterion 49, 1967 - Containment Design Basis The RHR system supply line from RCS hot le g loop 4 is designed so that the containment structure can accommodate, without excee ding the design leakage rate, pressures and temperatures resulting from the largest credible energy release following a LOCA, including a considerable margin for effects from metal-w ater or other chemical reactions that could occur as a consequence of failure of ECCSs.
5.5.6.1.8 General Design Criterion 54, 1 971 - Piping Systems Penetrating Containment The RHR system supply line from RCS hot leg loop 4 piping that penetrates containment is provided with leak detection, isolation, redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating this system.
The piping is designed with a capability to test periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits. 5.5.6.1.9 General Design Criterion 55, 1971 - Reactor Coolant Pressure Boundary Penetrating Containment Each RHR system line that penetrates containment is provided with CIVs.
5.5.6.1.10 Residual Heat Removal System Safety Function Requirements (1) Overpressurization Protection Overpressure protection is provided for the RHR system when it is in operation (not isolated from the RCS), to prevent accidental overpressurization.
(2) Protection from Missiles The non-ESF PG&E Design Class I portion of the RHR system is designed to be protected against the effects of missiles which may result from plant equipment failure and from events and conditions outside the plant.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-24 Revision 23 December 2016 (3) Shared Function The normal plant cooldown function of the RHR system does not compromise its ESF safety function.
(4) Protection Against High Energy Pipe Rupture Effects The non-ESF PG&E Design Class I portion of the RHR system is designed and located to accommodate the dynamic effects of a postulated high-energy pipe failure to the extent necessary to assure that a safe shut down condition of the reactor can be accomplished and maintained.
(5) Protection from Moderate Energy Pipe Rupture Effects - Outside Containment The PG&E Design Class I portion of the RHR system located outside containment is designed to be protected against the effects of moderate energy pipe failure.
(6) Protection from Jet Impingement - Inside Containment The PG&E Design Class I portion of the RHR system located inside containment is designed to be protected against the effects of jet impingement which may result from high energy pipe rupture.
(7) Protection from Flooding Effects - Outside Containment The PG&E Design Class I portion of the RHR system located outside containment is designed to be protected from the effects of internal flooding.
5.5.6.1.11 10 CFR 50.49 - Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants RHR system components that require EQ are qualified to the requirements of 10 CFR 50.49. 5.5.6.1.12 10 CFR 50.55a(f) - Inservice Testing Requirements RHR system ASME Code components are tested to the requirements of 10 CFR 50.55a(f)(4) and 10 CFR 50.55a(f)(5) to the extent practical.
5.5.6.1.13 10 CFR 50.55a(g) - Inservice Inspection Requirements RHR system ASME Code components are inspected to the requirements of 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(g)(5) to the extent practical.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-25 Revision 23 December 2016 5.5.6.1.14 10 CFR Part 50 Appendix R (Sections III.G, III.J, and III.L) - Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979 Section III.G - Fire Protection of Safe Shutd own Capability: Fire protection of the RHR system is provided by a combination of physical separation, fire-rated barriers, and/or automatic suppression and detection.
Section III.J - Emergency Lighting: Emergency lighting or BOLs are provided in areas where operation of the RHR system may be required to safely shut down the unit following a fire.
Section III.L - Alternative and Dedicated Shut down Capability: Alternate shutdown capability is provided to achieve and maintain hot standby conditions, achieve cold shutdown conditions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and maintain cold shutdown conditions, thereafter.
5.5.6.1.15 Generic Letter 87-12, July 1987 - Loss of Residual Heat Removal While the Reactor Coolant System is Partially Filled RHR system operation when the RCS water level is below the top of the RPV was evaluated to identify and enhance configurational, operational, procedural, and training requirements to ensure that the RHR system continues to meet the licensing basis of the plant, and that no unanalyzed event, or threat to safety, exists in this condition.
5.5.6.1.16 Generic Letter 88-17, October 1988 - Loss of Decay Heat Removal 10 CFR 50.54(f)
DCPP implements the expeditiou s action and programmed enhancement recommendations of Generic Letter 88-17, October 1988, with respect to operation following placement of the NSSS on RHR system cooling, or following the attainment of NSSS conditions under which RHR system operation would be normally initiated, to ensure loss of decay heat removal does not occur.
5.5.6.1.17 Generic Letter 89-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance The RHR system MOVs meet the requirements of Generic Letter 89-10, June 1989, and associated Generic Letter 96-05, September 1996.
5.5.6.1.18 Generic Letter 95-07, August 1995 - Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves The RHR system PG&E Design Class I, power-operated gate valves meet the requirements of Generic Letter 95-07, August 1995.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-26 Revision 23 December 2016 5.5.6.1.19 Generic Letter 98-02, May 1998 - Loss of Reactor Coolant Inventory and Associated Potential for Loss of Emergency Mitigation Functions While in a Shutdown Condition The RHR system is administratively controlled, configurationally managed, and procedurally operated to preclude an inadvertent draindown event as described in Generic Letter 98-02, May 1998.
5.5.6.1.20 NRC Bulletin 88-04, May 198 8 - Potential Safety-Related Pump Loss The RHR system is designed such that the PG&E Design Class I pumps that share a common minimum flow recirculation line are not susceptible to the pump-to-pump interaction or dead-heading as described in NRC Bulletin 88-04, May 1988. In addition, the installed minimum flow capacity for RHR system PG&E Design Class I pumps is adequate for even a single pump in operation.
5.5.6.1.21 Branch Technical Position RSB 5-1, 1980 - Design Requirements of the Residual Heat Removal System The DCPP reactor design is such that it can be taken from normal operating conditions to cold shutdown using only PG&E Design Class I systems, with either only onsite or only offsite power, and with the most limiting single failure.
5.5.6.2 System Description The RHR system is designed to remove heat from the core and reduce the temperature of the RCS during the second phase of plant cooldown. During the first phase of
cooldown, the temperature of the RCS is reduced by transferring heat from the RCS to the SPCS via the SGs.
The RHR system is placed in operation when the nominal temperature and pressure of the RCS are 350°F and 390 psig, respectively. The cooldown calculation of Reference 12 assumes the RHR is placed in service no sooner than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reactor shutdown. Assuming that two RHR heat exchangers and two RHR pumps are
in service and that each heat exchanger is supplied with CCW at design flow and temperature, the analysis shows that the RHR system design is capable of reducing the
temperature of the reactor coolant to 140°F in less than 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after reactor
shutdown. The heat load handled by the RHR system during the cooldown transient
includes sensible and decay heat from the core and RCP heat.
RHR system design parameters are listed in Table 5.5-8. A schematic diagram of the
RHR system is shown in Figure 3.2-10.
The RHR system consists of two RHR he at exchangers, two RHR pumps, and the associated piping, valves, and instrumentation necessary for operational control. The
inlet line to the RHR system is connected to the hot leg of RCL 4, while the return lines DCPP UNITS 1 &
2 FSAR UPDATE 5.5-27 Revision 23 December 2016 are connected to the cold legs of each of the RCLs. These normal return lines are also the ECCS low-head injection lines (refer to Figure 6.3-4).
The RHR system suction line is isolated from the RCS by two MOVs in series while the discharge lines are isolated by two check valves in each line. These check valves are
not a part of the RHR system; they are shown as part of the ECCS. The isolation
valves inlet line pressure-relief valve and associated piping are located inside the
containment. The remainder of the system is located outside the containment.
During system operation, reactor coolant flows from the RCS to the RHR pumps, through the tube side of the RHR heat exchangers, and back to the RCS. The heat is transferred in the RHR heat exchangers to the CCW circulating through the shell side of the heat exchangers.
Coincident with RHR operations, a portion of the reactor coolant flow may be diverted
from downstream of the RHR heat exchang ers to the CVCS low-pressure letdown line for cleanup and/or pressure control. By regulating the diverted flowrate and the
charging flow, the RCS pressure can be controlled. Pressure regulation is necessary to
maintain the pressure range dictated by the fracture prevention criteria requirements of
the RPV and by the No. 1 seal differential pressure and NPSH requirements of the RCPs.
The RCS cooldown rate is manually controlled by regulating the reactor coolant flow
through the tube side of the RHR heat exchangers. A line containing a flow control
valve bypasses the RHR heat exchangers and is used to maintain a constant return
flow to the RCS. Instrumentation is provided to monitor system pressure, temperature, and total flow, and to activate an alarm on system low flow.
The RHR system is also used for filling the refueling cavity before refueling. After
refueling operations, water is pumped back to the RWST until the RPV water level is brought down to the desired level below the RPV flange. The remainder is removed via a drain connection at the bottom of the refueling canal.
When the RHR system is in operation, the water chemistry is the same as that of the
reactor coolant. Provision is made for the sampling system to extract samples from the flow of reactor coolant downstream of the RHR heat exchangers. A local sampling point is also provided on each RHR train between the pump and heat exchanger.
5.5.6.2.1 Component Description The materials used to fabricate RHR system components are in accordance with
applicable code requirements. All parts of components in contact with borated water are fabricated or clad with austenitic stainless steel or equivalent corrosion resistant
material.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-28 Revision 23 December 2016 RHR component applicable codes and classification are provided in Table 5.5-9.
Component parameters are listed in Table 5.5-10.
5.5.6.2.1.1 Residual Heat Removal Pumps Two pumps are installed in the RHR system. The two pumps are vertical, centrifugal
units with mechanical shaft seals. The pumps are sized to deliver sufficient reactor
coolant flow through the RHR heat exchang ers to meet the plant cooldown requirements. The use of two pumps ensures that cooling capacity is only partially lost
should one pump become inoperative.
The RHR pumps are protected from overheating and loss of suction flow by miniflow
bypass lines that provide flow to the pump suction at all times. A control valve located
in each miniflow line is regulated by a signal from the flow transmitters located in each
pump discharge header. The control valves open on low RHR pump discharge flow and
close when RHR flow has been established. To prevent pump to pump interaction as a
result of differences between pump flow characteristics, check valves were installed
downstream of the RHR heat exchangers.
During minimum flow operation the check valve will prevent the stronger pump from de ad heading or reversing flow into the weaker pump, thereby maintaining minimum required recirculation flow.
A pressure sensor in each pump discharge header provides a signal for an indicator in
the control room. A high pressure alarm is also actuated by the pressure sensor.
5.5.6.2.1.2 Residual Heat Removal Heat Exchangers Two RHR heat exchangers are installed in the system. The heat exchanger design is based on heat load and temperature differences between reactor coolant and CCW existing 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after reactor shutdown when the temperature difference between the two systems is small. The decay heat removal used in the cooldown analysis is given in
Table 5.5-8.
The RHR heat exchangers are part of the ECCS, supporting the recirculation mode in which long-term core cooling is provided during the accident recovery period. During
the emergency core cooling recirculation phase, water from the containment
recirculation sump flows through the tube side of the RHR heat exchangers, transferring heat from containment to the CCW system. Further discussion of the RHR heat exchangers in this mode is found in Section 6.3.2.4.4.
The most limiting RHR system heat exchanger design requirement is to remove decay
heat, sensible heat and RCP heat at the design flow rates starting four hours following reactor shutdown. Less limiting, the initial heat removal provided by the RHR heat
exchangers after a design basis LOCA occu rs after the RWST inventory has been injected into the reactor. Under these conditions, the RHR heat exchangers are in
service with a containment recirculation sump temperature well below the limiting DCPP UNITS 1 &
2 FSAR UPDATE 5.5-29 Revision 23 December 2016 condition. In addition to RHR heat exchangers, heat removal from containment following a LOCA is shared with the CFCUs.
The installation of two heat exchangers ensu res that the heat removal capacity of the system is only partially lost if one heat exchanger becomes inoperative.
The RHR heat exchangers are of the shell and U-tube type. Reactor coolant circulates
through the tubes, while CCW circulates through the shell. The tubes and other surfaces in contact with reactor coolant are austenitic stainless steel or equivalent corrosion resistant material. T he sh ell is carbon steel. The tubes are welded to the tubesheet to prevent leakage of reactor coolant.
5.5.6.2.1.3 Residual Heat Removal System Valves Valves that perform a modulating function are equipped with two sets of packing and an intermediate leakoff connection that discharges to the drain header.
Some manual valves and MOVs have backseats to facilitate repacking and to limit stem leakage when the valves are open. Leakoff connections are provided where required
by valve size and fluid conditions.
5.5.6.2.2 System Operation A discussion of RHR system operation during various reactor operating modes follows.
5.5.6.2.2.1 Reactor Startup Generally, during cold shutdown, residual heat from the reactor core is being removed
by the RHR system. The number of pumps and heat exchangers in service depends on
the RHR load at the time.
At initiation of plant startup, the RCS is completely filled, and the pressurizer heaters are
energized. The RHR pumps are operating, but a portion of the discharge is directed to
the CVCS via a line that is connected to the common header downstream of the RHR
heat exchanger. After the RCPs are running and the pressurizer steam bubble has
formed, the RHR pumps are stopped. Indication of steam bubble formation is provided
in the control room by the damping out of the RCS pressure fluctuations and by
pressurizer level indication. The RHR system is then isolated from the RCS and the
system pressure is controlled by normal letdown and the pressurizer spray and
pressurizer heaters.
An alternative to this startup process is a vacuum refill method of filling the RCS, described in Section 5.1.7.1. This may resul t in starting the RCPs after the pressurizer steam bubble is formed.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-30 Revision 23 December 2016 5.5.6.2.2.2 Power Generation and Hot Standby Operation During power generation and hot standby operation, the RHR system is not in service
but is aligned for operation as part of the ECCS.
5.5.6.2.2.3 Reactor Shutdown The initial phase of reactor cooldown is accomplished by transferring heat from the RCS
to the SPCS through the use of the SGs.
When the reactor coolant nominal temperature and pressure are reduced to 350°F and 390 psig, respectively, the second phase of cooldown starts with the RHR system being placed in operation. Data and procedure reviews indicate it will require more than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reactor shutdown to initiate RHR cooldown (Reference 12).
Startup of the RHR system includes a warm-up period during which time reactor coolant flow through the heat exchangers is limited to minimize thermal shock. The rate of heat
removal from the reactor coolant is manually controlled by regulating the coolant flow
through the RHR heat exchangers. By adjusting the control valves downstream of the
RHR heat exchangers, the mixed mean temperature of the return flows is controlled.
Coincident with the manual adjustment, the heat exchanger bypass valve contained in
the common bypass line is regulated to give the required total flow.
The reactor cooldown rate is limited by RCS equipment cooling rates based on
allowable stress limits, as well as the operating temperature limits of the CCW system.
As the reactor coolant temperature decreases, the reactor coolant flow through the RHR heat exchangers is increased.
As cooldown continues, the pressurizer is filled with water and the RCS is operated in the water-solid condition.
At this stage, pressure is controlled by regulating the charging flow rate and the
alternate letdown rate to the CVCS from the RHR system.
After the reactor coolant pressure is reduced and the temperature is 160°F or lower, the RCS may be opened for refueling or maintenance.
5.5.6.2.2.4 Refueling Several systems may be used during refueli ng to provide borated water from the RWST to the refueling cavity. These include the RHR system, CSS, SI system, refueling water purification system, and the CVCS (which includes the liquid holdup tanks [LHUTs]).
During this operation, the isolation valves to the RWST are opened.
The RVCH is removed. The refueling water is then pumped into the RPV and into the refueling cavity through the open RPV.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-31 Revision 23 December 2016 After the water level reaches the desired level, the RWST supply valves are closed, and RHR operation continues.
During refueling, the RHR system is maintained in service with the number of pumps
and heat exchangers in operation as required by the heat load.
Following refueling, the RHR pumps are us ed to drain the refueling cavity to the top of the RPV flange, and reduce the level in the RPV to the desired level below the top of the RPV flange by pumping water from the RCS to the RWST.
5.5.6.3 Design Evaluation Design features of the RHR system ensure safe and reliable system performance as
discussed below.
5.5.6.3.1 System Availability and Reliability The system is provided with two RHR pumps and two RHR heat exchangers arranged
in separate flowpaths. If one of the two pumps or one of the two heat exchangers is not
operable, safe cooldown of the plant is not compromised, although the time required for
cooldown is extended.
To ensure reliability, the two RHR pumps are connected to two separate electrical buses so that each pump receives power from a differe nt source. If a total loss of offsite power occurs while the system is in service, each bus is automatically transferred to a separate emergency diesel power supply.
5.5.6.3.2 Radiological Considerations The highest radiation levels experienced by the RHR system are those that would result from a LOCA. Following a LOCA, the RHR system is used as part of the ECCS. During the recirculation phase of emergency core cooling, the RHR system is designed to
operate for up to a year pumping water from the containment recirculation sump, cooling it, and returning it to the containment to cool the core.
Since the RHR system is located outside the containment, except for some valves and
piping, most of the system is not subjected to the high levels of radioactivity in the
containment post-accident environment. To ensure continued operation of the RHR system components, the valve motor operators, the RHR pump motors, and the RHR pump seals have been evaluated for operation in post-accident environments (refer to Section 5.5.6.4.11). Refer to Section 3.11 for details of the evaluation.
The operation of the RHR system does not involve a radiation hazard for the operators
since the system is controlled remotely from the con trol room. If maintenance of the system is necessary, the portion of the system requiring maintenance is isolated by DCPP UNITS 1 &
2 FSAR UPDATE 5.5-32 Revision 23 December 2016 remotely operated valves and/or manual valves with stem extensions, which allow operation of the valves from a shielded locati on. The isolated piping is drained and flushed before maintenance is performed.
5.5.6.4 Safety Evaluation 5.5.6.4.1 General Design Criterion 2, 1967 - Performance Standards All RHR components are located within the PG&E Design Class I auxiliary and containment buildings. These buildings, or applicable portions thereof, are designed to withstand the effects of winds and tornadoes (refe r to Section 3.3), floods and tsunamis (refer to Section 3.4), external missiles (refer to Section 3.5), earthquakes (refer to Section 3.7), and other natural phenomena, to protect the RHR SSCs, ensuring their design functions will be performed.
PG&E Design Class I SSCs of the RHR system are seismically qualified to ensure their design functions can be performed following an earthquake, as described in Section 3.7. 5.5.6.4.2 General Design Criterion 3, 1971 - Fire Protection The RHR system is designed and located to minimize the probability and effects of fires and explosions per the fire protect ion guidelines of Branch Technical Position APCSB 9.5-1 (refer to Appendix 9.5B, Table B-1).
Appendix 9.5A provides the fire hazards analysis and is organized by fire zone.
Power is removed from the RHR suction line isolation valve motors when the RHR system is isolated (refer to Section 5.5.6.4.10).
5.5.6.4.3 General Design Criterion 9, 1967 - Reactor Coolant Pressure Boundary The portion of the RHR system that is part of the RCPB is designed and constructed so as to have an exceedingly low probabi lity of gross rupture or significant leakage throughout its lifetime. The RCPB is designed to accommodate the system pressures and temperatures attained under all expected modes of plant operation, including all anticipated transients, and to maintain the stresses within applicable stress limits (refer to Section 5.2.3.3).
5.5.6.4.4 General Design Criterion 11, 1967 - Control Room Instrumentation, alarms, and controls are provided in the control room for operators to monitor and maintain RHR system parameters. Instrumentation and controls for the RHR system are further discussed in Sections 5.5.6.4.5 and 6.3.3.4. Cold shutdown from outside the control room is accomplished with the use of RHR system indicators and controls which are located outside the control room. Interlocks with MOVs are discussed in Section 7.6.2.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-33 Revision 23 December 2016 5.5.6.4.5 General Design Criterion 12, 1967 - Instrumentation and Control Systems Instrumentation and controls are provided, as required, to monitor and maintain RHR variables within prescribed operating ranges.
Specifically, the RHR system variables that are monitored are: RHR heat exchanger outlet temperature; RHR pump-motor temperature and discharge pressure; and RHR system flow to the RCS hot and cold legs. An RHR pump discharge header flow meter controls the RHR pump minimum flow. Further discussion of these instruments and controls is provided by Section 6.3.3.4. 5.5.6.4.6 General Design Criterion 40, 1967 - Missile Protection The provisions taken to protect the ESF (containment isolation) portion of the RHR system from damage that might result from missiles and dynamic effects associated with equipment and high-energy pipe failures are discussed in Sections 3.5, 3.6, and 6.2.4. 5.5.6.4.7 General Design Criterion 49, 1967 - Containment Design Basis The RHR system supply line from the RCS hot leg loop 4 containment penetration, including the system piping and valves required for containment isolation, is designed to withstand the pressures an d temperatures that could res ult from a L OCA without exceeding containment design leakage rates. Refer to Sections 3.8.2.1.3 and 6.3.3.17 for additional justification. Refer to Section 6.3 and Table 6.2-39 for penetrations that are part of ECCS.
5.5.6.4.8 General Design Criterion 54, 1 971 - Piping Systems Penetrating Containment The RHR system supply line from the RCS h ot leg loop 4 isolation valves required for containment closure are periodically tested as part of the IST Program Plan for operability in accordance with GDC 54, 1971. Test connections are provided in the piping of applicable penetrations to verify valve leakages are within prescribed limits.
Testing of the components required for the CIS is discussed in Section 6.2.4. Refer to Section 6.3 and Table 6.2-39 for penetrations that are part of ECCS.
5.5.6.4.9 General Design Criterion 55, 1971 - Reactor Coolant Pressure Boundary Penetrating Containment The RHR system is designed such that each RHR line that is part of the RCPB that penetrates containment is provided with CIVs in compliance with GDC 55, 1971. Refer to Section 6.2.4.2.1 and Table 6.2-39 for penetration and configuration details with regards to GDC 55, 1971.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-34 Revision 23 December 2016 5.5.6.4.10 Residual Heat Removal System Safety Function Requirements (1) Overpressurization Protection The inlet line to the RHR system is equipped with a pressure relief valve sized to relieve the combined flow of both charging pumps into the RCS and thus prevents exceeding
the RHR system design pressure.
Each discharge line to the RCS is equipped with a pressure relief valve located in the
ECCS (refer to Figure 3.2-9, Sheets 5 and 6 and Figure 3.2-10, Sheets 1 and 2). They relieve the maximum possible back-leakage through the valves separating the RHR
system from the RCS.
The design of the RHR system includes the following features for valves on the inlet line
between the high-pressure RCS and the lower pressure RHR system:
(1) To prevent both RHR suction line isolation valves from opening as a result of fire damage to electrical cables, ac power is removed from the operators of the indicated MOVs for plant conditions during which the RHR system is isolated.
(2) The isolation valve adjoining the RCS is interlocked with a pressure signal to prevent it from being opened whenever the RCS pressure is greater than a set value.
(3) The second isolation valve, the one adjoining the RHR system, is similarly interlocked with a pressure signal to prevent opening if RCS pressure is
above a set value, and a pressurizer temperature signal to prevent
opening if it exceeds a set value.
(4) The RHR suction valves interlock relays are powered from the solid state protection system (SSPS) output cabinets. To maintain the ability to open the RHR suction valve(s) when the SSPS output cabinet(s) are de-
energized in Mode 6 or defueled, a jumper(s) is used to lock-in the RHR
suction valve(s) open permissive. This defeats the applicable RHR
system overpressurization/temperature protection. Jumper installation is
limited to Mode 6 and defueled only.
Refer to Section 7.6 for a more complete discussion of the permissive interlocks on these isolation valves.
(2) Protection from Missiles The provisions taken to protect the non-ESF PG&E Design Class I portion of the RHR system from missiles are discussed in Section 3.5. The RHR system design is such that DCPP UNITS 1 &
2 FSAR UPDATE 5.5-35 Revision 23 December 2016 physical protection is adequately provided a gainst physical hazards in areas through which the system is routed.
(3) Shared Function The safety function performed by the RHR system is not compromised by its normal function during plant cooldown. The valves associated with the RHR system are normally aligned to allow immediate use of this system in its ESF mode of operation.
The system has been designed in such a manner that two redundant flow circuits are
available, ensuring the availab ility of at least one train for safety purposes.
The normal plant cooldown function of the RHR system is accomplished through a
suction line arrangement that is independent of any safety function. The normal
cooldown return lines are arranged in parallel redundant circuits and are utilized also as
the low-head SI lines to the RCS. Utilization of the same return c ircuits for the safety function as well as for normal cooldown, lends assurance to the proper functioning of
these lines for safety purposes.
(4) Protection Against High Energy Pipe Rupture Effects The provisions taken to protect the non-ESF PG&E Design Class I portion of the RHR system from damage that might result from dynamic effects associated with a postulated rupture of high-energy piping are discussed in Section 3.6. The RHR system design is such that physical protection is ade quately provided against physical hazards in areas through which the system is routed.
(5) Protection from Moderate Energy Pipe Rupture Effects - Outside Containment The provisions taken to provide protection of the PG&E Design Class I portion of the RHR system located outside containment from the effects of moderate energy pipe failure are discussed in Section 3.6.
(6) Protection from Jet Impingement - Inside Containment The provisions taken to provide protection of the PG&E Design Class I portion of the RHR system located inside containment from the ef fects of jet impingement which may result from high energy pipe rupture are discussed in Section 3.6.
(7) Protection from Flooding Effects The provisions taken to provide protection of the PG&E Design Class I portion of the RHR system from flooding that might result from the effects associated with a postulated rupture of piping are discussed in Section 3.6.
If a pipe rupture or equipment failure occurs in a RHR pump compartment, overflow from one pump compartment would drain through a 14-inch line to the pipe trench DCPP UNITS 1 &
2 FSAR UPDATE 5.5-36 Revision 23 December 2016 rather than flood the adjacent compartment. Addition ally, the maximum calculated flood level within each compartment is below the elevation of the RHR pump motors.
5.5.6.4.11 10 CFR 50.49 - Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants The RHR SSCs required to function in harsh environments under accident conditions are qualified to the applicable environmental conditions to ensure that they will continue to perform their safety functions. Refer to Section 6.3 for RHR SSCs that are also a part of ECCS. Section 3.11 describes the DCPP EQ Program and the requirements for the environmental design of electrical and related mechanical equipment. The affected equipment includes flow and pressure transmitters, valve motors and operators, and switches, and is listed on the EQ Master List.
Refer to Section 5.5.6.3.2 for additional information.
5.5.6.4.12 10 CFR 50.55a(f) - Inservice Testing Requirements The IST requirements for the RHR system are contained in the DCPP IST Program Plan. 5.5.6.4.13 10 CFR 50.55a(g) - Inservice Inspection Requirements The RHR system is inspected in accordance with ASME BPVC Section XI-2001 through 2003 Addenda, as stated in the DCPP ISI Program Plan.
5.5.6.4.14 10 CFR Part 50 Appendix R (Sections III.G, III.J, and III.L) - Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979 Section III.G - Fire Protection of Safe Shutd own Capability: Tables 9.5G-1 and 9.5G-2 for DCPP Unit 1 and Unit 2, respectively, list the minimum RHR equipment required to bring the plant to a cold shutdown condition as defined by 10 CFR Part 50, Appendix R, Section III.G. These SSCs are provided fire protection features appropriate to the requirements of Section III.G. The actions necessary for cold shutdown for fires in certain fire areas involves manually aligning valves. Refer to RHR Safety Function Requirements, Section 5.5.6.4.10(1)(1).
Section III.J - Emergency Lighting: Emergency lighting or BOLs are provided in areas where operation of the RHR system may be required to safely shutdown the unit following a fire (refer to Appendices 9.5D and 9.5B, Item D.5).
Section III.L - Alternative and Dedicated Shutdown Capability: Safe shutdown capabilities are provided in the control room and at alternate locations via local operation (refer to Section 7.4 and Appendix 9.5H). The ability to safely shut down the plant following a fire in any fire area is summarized in Section 4.0 of Appendices 9.5A and 9.5E.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-37 Revision 23 December 2016 5.5.6.4.15 Generic Letter 87-12, July 1987 - Loss of Residual Heat Removal While the Reactor Coolant System is Partially Filled The RHR system design and operating procedures ensure that decay heat removal is provided and that the integrity of the RCPB is ensured during operation with a partially filled RCS. Training is provided to operators to ensure RCS inventory and RHR flow are maintained during operation in this condition.
5.5.6.4.16 Generic Letter 88-17, October 1988 - Loss of Decay Heat Removal 10 CFR 50.54(f)
There are limited periods during plant operation when the RCS may be operated with reduced inventory while irradiated fuel is in the RPV. Examples include refueling outages or maintenance evolutions. A reduced inventory condition, including mid-loop conditions, exists whenever the RCS level is lower than 111 feet elevation, which is three feet below the RPV flange. This section describes the administrative controls and instrumentation relied upon during reduced inventory operations.
Procedures and administrative controls have been implemented to assure that containment closure will be achieved prior to the time at which core uncovery could result from a loss of decay heat removal coup led with the inability to initiate alternate core cooling or addition of water to the RCS inventory. Procedures have been implemented to avoid operations that deliberately or knowingly lead to perturbations to the RCS and/or to systems that are necessary to maintain the RCS in a stable and controlled condition while the RCS is in a reduced inventory condition.
At least two independent, continuous temperature indications that are representative of the core exit conditions are available whenever the RCS is in a mid-loop condition and the RVCH is located on top of the RPV.
At least two independent, continuous RCS water level indications are provided whenever the RCS is in a reduced inventory condition. Water level indications are periodically checked and recorded by an operator or automatically and continuously monitored and alarmed.
At least two available or operable means of addin g additional inventory to the RCS are available that are in addition to pumps that are a part of the normal decay heat removal systems. Specifically, prior to draining to mid-loop, one charging pump, gravity fill makeup from the RWST, and an SI pump with associated hot leg flow path to the RCS are available.
Analyses have been performed to suppleme nt existing information and to develop a basis for other actions.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-38 Revision 23 December 2016 A pressurization analysis for shutdown conditions was performed to evaluate, for low-to-high decay heat shutdown conditions, the thermal hydraulic response, particularly the maximum RCS pressure limits, if no operator recovery actions were taken to limit or prevent boiling in the RCS (References 13 and 14). The results of these analyses are used to determine acceptable RCS vent path configurations used during outage conditions as a contingency to mitigate RCS pressurization upon a postulated loss of RHR. Typical RCS vent path openings capable of use include the RPV flange, one or more PSVs or PORVs, SG primary hot leg manways, or combinations of these openings depending on the decay heat load.
Other analyses performed include the pressurization and integrity of containment after a loss of RHR while at mid-loop, and a level instrumentation analysis in order to understand its behavior during reduced inventory.
5.5.6.4.17 Generic Letter 89-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance RHR system MOVs 8701 and 8702 are subject to the requirements of Generic Letter 89-10, June 1989, and associated Generic Letter 96-05, September 1996, and meet the requirements of the DCPP MOV Program Plan.
5.5.6.4.18 Generic Letter 95-07, August 1995 - Pressure Locking and Thermal Binding of Safety- Related Power-Operated Gate Valves RHR system power operated gate valves 8701 an d 8702, which were determined to be susceptible to pressure lockin g, were modified by installing bonnet cavi ty leakoffs with block valves to the high pre ssure inlet lines to prevent pressure locking. No power-operated gate valves in the RHR system were found susc eptible to thermal binding.
5.5.6.4.19 Generic Letter 98-02, May 1998 - Loss of Reactor Coolant Inventory and Associated Potential for Loss of Emergency Mitigation Functions While in a Shutdown Condition DCPP Unit 1 and Unit 2 are vulnerable to the potential draindown of the RCS to the RWST, and render the RHR system inoperable, if valve 8741 were inadvertently opened with RHR in service in Mode 4. Procedural controls are in place to prevent this from occurring.
5.5.6.4.20 NRC Bulletin 88-04, May 198 8 - Potential Safety-Related Pump Loss The RHR system is designed to ensure the installed miniflow capacity is adequate for even a single pump in operation and to prevent dead-heading of either RHR pump due to pump-to-pump interaction during miniflow operation.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-39 Revision 23 December 2016 5.5.6.4.21 Branch Technical Position RSB 5-1, 1980 - Design Requirements of the Residual Heat Removal System The DCPP reactor design is such that it can be taken from normal operating conditions to cold shutdown using only PG&E Design Class I systems, with either only onsite or only offsite power, and with the most limiting single failure.
DCPP conducted tests with supporting analysis to: (a) confirm that adequate mixing of borated water was achieved under natural circulation conditions with an estimation of the times required to achieve such mixing; (b) confirm that the cooldown under natural circulation was achieved within the limits specified in the emergency operating procedures; and (c) confirm no credible sin gle failure would preclude achieving cold shutdown conditions.
5.5.6.5 Tests and Inspections Periodic visual inspections and preventive maintenance are conducted during plant operation according to normal industrial practice.
The instrumentation channels for the RHR pump flow instrumentation devices are
calibrated on a nominal 24-month frequency.
The RHR pumps are tested by starting them periodically.
5.5.6.6 Instrumentation Applications Refer to Section 5.5.6.4.5 for the instrumentation applications related to the RHR system. 5.5.7 REACTOR COOLANT CLEANUP SYSTEM
The CVCS provides reactor coolant cleanup and is discussed in Section 9.3. The
radiological considerations are discussed in Chapter 11.
5.5.8 MAIN STEAM LINE AND FEEDWATER PIPING Main steam line piping is covered in Section 10.3. Feedwater piping is covered in
Section 10.4.
5.5.9 PRESSURIZER The pressurizer provides a point in the RCS where liquid and vapor are maintained at
equilibrium temperature and pressure under saturated conditions for pressure control
purposes.
During an insurge, the spray system, fed from two cold legs, condenses steam in the
vessel to prevent the pressurizer pressure from reaching the setpoint of the PORVs.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-40 Revision 23 December 2016 During an outsurge, flashing of water to s team and generation of steam by automatic actuation of the heaters helps keep the pressure above the low-pressure reactor trip setpoint. Heaters are also energized, on high water level during insurges, to heat the
subcooled surge water entering the pressurizer from the RCL.
5.5.9.1 Design Bases The general configuration of the pressurizer is shown in Figure 5.5-8. The design data
of the pressurizer are provided in Table 5.5-12. Codes and material requirements are
provided in Section 5.2.
5.5.9.1.1 Pressurizer Surge Line The surge line is sized to limit the pressure drop between the RCS and the PSVs with maximum allowable discharge flo w from the PSVs. Overpressure of the RCS does not exceed 110 percent of the design pressure. The surge line and the thermal sleeves at
each end are designed to withstand the thermal stresses resulting from volume surges, which occur during operation.
5.5.9.1.2 Pressurizer The pressurizer volume (refer to Table 5.5-12) satisfies the following requirements:
(1) The combined saturated water volume and steam expansion volume is sufficient to provide the desired pressure response to system volume
changes. (2) The water volume is sufficient to prevent the heaters from being uncovered during a step load increase of 10 percent of full power.
(3) The steam volume is large enough to accommodate the surge resulting from the design step load reduction from full load with reactor control and
steam dump without the water level reaching the high level reactor trip
point. (4) The steam volume is large enough to prevent water relief through the PSVs following a loss of load with the high water level initiating a reactor trip. (5) The pressurizer does not empty following reactor and turbine trip.
(6) The emergency core cooling signal is not activated during reactor trip and turbine trip.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-41 Revision 23 December 2016 5.5.9.2 Design Description The pressurizer is designed to accommodate positive and negative reactor coolant
surges caused by RCS transients.
5.5.9.2.1 Pressurizer Surge Line The pressurizer surge line connects the pressurizer to one reactor hot leg. The line
enables continuous coolant volume/pressure adjustment s between the RCS and the pressurizer.
5.5.9.2.2 Pressurizer The pressurizer is a vertical, cylindrical vessel with essentially hemispherical top and
bottom heads constructed of carbon steel, with austenitic stainless steel cladding on all
surfaces exposed to the reactor coolant.
The surge line nozzle and removable electric heaters are installed in the bottom head.
The heaters are removable for maintenance or replacement. A thermal sleeve is
provided to minimize stresses in the surge line nozzle. A screen at the surge line
nozzle and baffles in the lower section of the pressurizer prevent a cold insurge of water
from flowing directly to the steam/water interface and assist mixing.
Spray line nozzles and PORV/PSV connections are located in the top head of the vessel. Spray flow is modulated by automatically controlled air-operated valves. The
spray valves can also be operated manually by a switch in the control room.
A small continuous spray flow is provided through a manual bypass valve around the
power-operated spray valves to ensure that the pressurizer liquid is homogeneous with
the coolant and to prevent excessive cooling of the spray piping. During an outsurge from the pressurizer, flashing of water to steam and generating of steam by automatic actuation of the heaters keep the pressure above the minimum allowable limit. During
an insurge from the RCS, the spray system, which is fed from two cold legs, condenses steam in the vessel to prevent the pressurizer pressure from reaching the setpoint of the
PORVs for normal design transients. Heaters are energized on high water level during insurge to heat the subcooled surge water that enters the pressurizer from the RCL.
The heaters are further discussed in Section 8.3.
5.5.9.2.2.1 Pressurizer Support The skirt-type support, shown in Figure 5.5-12, is attached to the lower head and extends for a full 360
° around the vessel. The lower part of the skirt terminates in a bolting flange with bolt holes to secure the vessel to its structural steel framework. The
skirt-type support is provided with ventilation holes around its upper perimeter to ensure free convection of ambient air past the heater plus connector ends for cooling.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-42 Revision 23 December 2016 5.5.9.2.2.2 Pressurizer Instrumentation Refer to Chapter 7 for details of the instrumentation associated with pressurizer
pressure, level, and temperature.
5.5.9.2.2.3 Spray Line Temperatures Temperatures in the spray lines from two loops are measured and indicated. Alarms
from these signals are actuated by low spray water temperature. Low temperature
conditions indicate insufficient flow in the spray lines.
5.5.9.2.2.4 Safety and Relief Valve Discharge Temperatures Temperatures in the PSV and PORV discharge lines are measured and indicated. An increase in a discharge line temperature is an indication of leakage through the
associated valve.
5.5.9.3 Design Evaluation The pressurizer is designed to provide safe and reliable RCS pressure control.
5.5.9.3.1 System Pressure RCS pressure is maintained by the steam bubble in the pressurizer. During normal
operation, the pressurizer maintains RCS pressure by automatic operation of
pressurizer heaters and spray. When the pressurizer is f illed with water (i.e., near the end of the second phase of plant cooldown and during initial system heatup, if the
vacuum refill method of filling the RCS is not used as described in Section 5.1.7.1), RCS pressure is maintained by the RHR, CVCS, and LTOP systems. Safety limits are
established to control the rate of t emperature change in the pressurizer. These safety
limits are administratively controlled to ensure that RCS pressure and temperature do
not exceed the maximum transient value allowed under ASME BPVC Section III, and thereby ensure continued integrity of the RCPB.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-43 Revision 23 December 2016 5.5.9.3.2 Pressurizer Performance The pressurizer has a minimum free internal volume. The normal operating water
volume at full load conditions is 60 percent of the minimum free internal vessel volume.
Under part load conditions, the water volume in the vessel is reduced for proportional reductions in plant load to 22 percent of free vessel volume at zero power level. During controlled between 22 percent and 35 percent of the indicated level. During shutdown in Section 5.2.3.28, the administrative controls and requirements of the PTLR take precedence. Pressurizer performance has been analyzed for the various plant operating transients discussed in Section 5.2.2.1. The design pressure was not exceeded with the pressurizer design parameters listed in Table 5.5-12.
5.5.9.3.3 Pressure Setpoints The RCS design and operating pressure together with the safety, PORV, and pressurizer spray valves setpoints, and the protection system setpoint pressures are
listed in Section 5.2.2.2. The design pressure allows for operating transient pressure changes. The selected design margin considers core thermal lag, coolant transport
times and pressure drops, instrumentation and control response characteristics, and
system relief valve characteristics.
5.5.9.3.4 Pressurizer Spray Two separate, automatically controlled spray valves with remote-manual overrides are
used to initiate pressurizer spray. In parallel with each spray valve is a manual throttle
valve that permits a small, continuous flow through both spray lines to reduce thermal stresses and thermal shock when the spray valves open, and to help maintain uniform water chemistry and temperature in the pressurizer. Spray flow is not normally initiated
if the temperature difference between the pressurizer and spray fluid exceeds 320°F.
Temperature sensors with low alarms are provided in each spray line to alert the
operator to insufficient bypass flow. The layout of the common spray line piping to the
pressurizer forms a water seal that prevents the steam buildup back to the control valves. The spray rate is selected to prevent the pressurizer pressure from reaching the
operating setpoint of the PORVs during a step reduction in power level of 10 percent of full load.
The pressurizer spray lines and valves are la rge enough to provide adequate spray using as the driving force the differential pressure between the surge line connection in
the hot leg and the spray line connection in the cold leg. The spray line inlet
connections extend into the cold leg piping in the form of a scoop so that the velocity head of the RCL flow adds to the spray driving force. The spray valves and spray line connections are arranged so that the spray will operate when one RCP is not operating.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-44 Revision 23 December 2016 The line may also be used to assist in equalizing the boron concentration between the RCLs and the pressurizer.
A flowpath from the CVCS to the pressurizer spray line is also provided. This additional
facility provides auxiliary spray to the vapor space of the pressurizer during cooldown if the RCPs are not operating. The thermal sleeves on the pressurizer spray connection
and the spray piping are designed to withstand the thermal stresses resulting from the
introduction of cold spray water.
5.5.9.3.5 Pressurizer Design Analysis The occurrences for pressurizer design cycle analysis are defined as follows:
(1) For design purposes, the temperature in the pressurizer vessel is always assumed to equal saturation temperature for the existing RCS pressure, except in the pressurizer steam space subsequent to a pressure increase.
In this case, the temperature of the steam space will exceed the saturation
temperature since an isentropic compression of the steam is assumed.
(2) The temperature shock on the spray nozzle is assumed to equal the temperature of the nozzle minus the cold leg temperature, and the
temperature shock on the surge nozzle is assumed to equal the
pressurizer water space temperature minus the hot leg temperature.
(3) Pressurizer spray is assumed to be initiated instantaneously reaching its design value as soon as the RCS pressure increases 40 psi above the nominal operating pressure. Spray is assumed to be terminated as soon as the RCS pressure falls below the normal operating pressure-plus
40 psi-level.
(4) Unless otherwise noted, pressurizer spray is assumed to be initiated once during each transient condition.
The pressurizer surge nozzle is also assumed to be subject to one temperature transient per transient
condition, unless otherwise noted.
(5) At the end of each transient, except the faulted conditions, the RCS is assumed to return to a load condition con sistent with the plant heatup transient.
(6) Temperature changes occurring as a result of pressurizer spray are assumed to be instantaneous. Temperature changes occurring on the
surge nozzle are also assumed to be instantaneous.
(7) Whenever spray is initiated in the pressurizer, the pressurizer water level is assumed to be at the no-load level.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-45 Revision 23 December 2016 5.5.9.4 Tests and Inspections The pressurizer is designed and constructed in accordance with ASME BPVC Section III-1965 through Summer 1966 Addenda. Peripheral support rings are furnished for the insulation modules. The pressurizer quality assurance program is
given in Table 5.5-13. Refer to Sections 5.1.8.19, 5.1.8.20, 5.2.3.14, and 5.2.3.15 for further discussion of testing and inspection of the RCS.
5.5.10 PRESSURIZER RELIEF TANK The PRT accommodates the pressurizer and other relief valve discharges.
5.5.10.1 Design Bases Design data for the PRT are provided in Ta ble 5.5-14. Codes and materials applicable to the tank are discussed in Section 5.2.
The tank design is based on the requirement to condense and cool a discharge of
pressurizer steam equal to 110 percent of the volume above the full power pressurizer
water level setpoint. The tank is not designed to accept a continuous discharge from the
pressurizer. The volume of water in the tank (refer to Table 5.1-1) is capable of absorbing the heat from the assumed discharge, with an initial temperature of 120°F
and increasing to a final temperature of 200°F. The tank is cooled, when necessary, by
manual spraying of cool water into the tank and draining the warm mixture to the RCDT.
5.5.10.2 Design Description The PRT is a horizontal, cylindrical vessel with elliptical ends, which condenses and cools the discharge from the PSVs and PORVs. Discharge from smaller relief valves located inside and outside the containment is also piped to the PRT. The PRT normally contains water and a predominantly nitrogen atmosphere. Provision is made to permit the gas in the tank to be periodically monitored for hydrogen and/or oxygen
concentrations. Through its connection to the GRW system, the PRT provides a means for removing any noncondensable gases from the RCS that might collect in the
pressurizer vessel.
Steam is discharged through a sparger pipe under the water level. This condenses and cools the steam by mixing it with water that is near ambient temperature. The PRT is equipped with an internal spray and a drain that are used to cool the PRT following a discharge. A flanged nozzle is provided on the PRT for the pressurizer discharge line connection. The PRT is protected against a discharge exceeding its design pressure by two rupture disks that discharge into the reactor containment.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-46 Revision 23 December 2016 5.5.10.2.1 Pressurizer Relief Tank Pressure The PRT pressure transmitter supplies a signal for an indicator with a high-pressure
alarm. Also, the PRT pressure transmitter provides a signal to close the air-operated
valve to the GRW system vent header on high pressure.
5.5.10.2.2 Pressurizer Relief Tank Level The PRT level transmitter supplies a signal for an indicator with high and low level
alarms. 5.5.10.2.3 Pressurizer Relief Tank Water Temperature The temperature of the water in the PRT is indicated in the control room. An alarm
actuated by high temperature informs the operator that cooling of the tank contents is
required.
5.5.10.3 Design Evaluation The volume of water in the PRT is capable of absorbing heat from the pressurizer discharge during a step load decrease of 10 percent. Water temperature in the PRT is maintained at the nominal containment temperature.
The rupture disks on the PRT have a relief capacity equal to the combined capacity of the PSVs. The PRT design pressure is twice the calculated pressure resulting from the maximum design PSV discharge described above. The PRT and rupture disks holders are also designed for full vacuum to prevent tank collapse if the contents cool following a discharge without nitrogen being added. The discharge piping from the PSVs and PORVs to the PRT is sufficiently large to prevent back pressure at the PSVs from exceeding 20 percent of the setpoint pressure at full flow.
5.5.11 VALVES The PG&E Design Class I function of the val ves within the RCPB listed in Table 5.2-9 is to act as pressure-retaining components and leaktight barriers during normal plant
operation and accidents.
5.5.11.1 Design Bases As noted in Section 5.2, all RCS valves incl uding those in connected systems, out to and including the second isolation valve, are normally closed or capable of automatic or remote manual closure. Valve closure time must be such that for any postulated component failure outside the system boundary, the loss of reactor coolant event would
not prevent orderly reactor shutdown and cooldown assuming makeup is provided by
normal makeup systems. Normal makeup systems are those systems normally used to
maintain reactor coolant inventory under startup, hot standby, operation, or cooldown DCPP UNITS 1 &
2 FSAR UPDATE 5.5-47 Revision 23 December 2016 conditions. If the second of two normally open check valves is considered as the pressure boundary, means are provided to periodically assess back-flow leakage of the
first valve when closed. For a check valve to qualify as the system pressure boundary, it must be located inside the containment.
RCPB valves are listed in Table 5.2-9. Materials of construction are specified to minimize corrosion/erosion and to ensure compatibility with the environment. Design
parameters are provided in Table 5.5-15.
Valves are designed and fabricated in accordance with either USAS B16.5, MSS-SP-66, or ASME BPVC Section III-1968 or 1974. To the extent practicable, valve leakage is minimized by design.
5.5.11.2 Design Description Gate valves are either wedge design or parallel disk and are essentially straight
through. The wedge may be either split or solid. All gate valves have a backseat, outside screw and yoke. Globe valves, "T" and "Y" style, are full-ported with outside
screw and yoke construction. Ball valves are V-notch design for equal percentage flow
characteristics. Check valves are spring-loaded lift piston types for sizes 2 inches and
smaller, and swing type for sizes 2-1/2 inches and larger. All check valves containing radioactive fluid are stainless steel and do not have body penetrations other than the
inlet, outlet, and bonnet. The check hinge is serviced through the bonnet. The RHR
heat exchanger outlet check valves have hinge pin covers.
All valves in the RCS that are in contact with the coolant are constructed primarily of stainless steel. Other materials in contact with the coolant, such as for hard surfacing and packing, are special materials.
All RCPB manual and MOVs that are 3 inches and larger are provided with double-packed stuffing boxes and stem intermediate lantern gland leakoff connections. Some
of the throttling control valves, regardless of size, are provided with double-packed
stuffing boxes and with stem leakoff connections. All leakoff connections are piped to a
closed collection system. Leakage to the atmosphere is essentially zero for these
valves.
Each accumulator check valve is designed with a low-pressure drop configuration with all operating parts contained within the body. The disk has unlimited rotation to provide
a change of seating surface and alignment after each valve opening.
Valves at the RHR system interface are provided with interlocks that meet the intent of
Reference 1. These interlocks are discussed in detail in Sections 5.5.6 and 7.6.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-48 Revision 23 December 2016 5.5.11.3 Design Evaluation Stress analysis of the RCL/support system, discussed in Sections 3.9 and 5.2, ensure acceptable stresses for all valves in the RCPB. Reactor coolant chemistry parameters are specified to minimize corrosion. Periodic analyses of coolant chemical composition, discussed in the DCPP Equipment Control Guidelines, ensure that the reactor coolant meets these specifications. The upper-limit coolant velocity of about 50 feet per second
minimizes erosion. Valve leakage is minimized by design features as discussed above.
5.5.11.4 Tests and Inspections Hydrostatic, seat leakage, and operation tests are performed on RCPB valves in accordance with ASME BPVC Section XI, Subsection IWV, (hydrostatic) and the IST Program Plan (all other testing), as required by the Technical Specifications and 10 CFR 50.55a. Refer to Sections 5.1.8.19, 5.1.8.20, 5.2.3.14, and 5.2.3.15 for further discussion of testing and inspection of the RCS.
There are no full-penetration welds within valve body walls. Valves are accessible for disassembly and internal visual inspection.
5.5.12 SAFETY AND RELIEF VALVES The pressurizer is equipped with PSVs and PORVs for overpressure protection and control. Their use is described in Section 5.2.2.2.
5.5.12.1 Design Bases The combined capacity of the PSVs is designed to accommodate the maximum surge resulting from complete loss of load. This objective is met without reactor trip or any
operator action, provided the main steam safety valves open as designed when steam pressure reaches the main steam safety valve setting. The PORVs are designed to limit pressurizer pressure to a value below the fixed high-pressure reactor trip setpoint.
The PORVs are also credited to prevent pressurizer overfill in a spurious SI event (refer to Section 15.2.15.3).
5.5.12.2 Design Description The PSVs are totally enclosed pop type. The valves are spring loaded, self-actuated, and have back pressure compensation features.
The pressurizer is equipped with three PORVs, each with a corresponding PORV block valve. The PORVs are air-operated and actuated by Class 1E 125-Vdc solenoid valves that are energized-to-open, spring-to-close. The circuits to the solenoid valves are supplied with redundant interlocks that prevent energization below normal operating
pressure. Control power is Class 1E 125-Vdc from the station batteries (refer to Section 8.3.2). Indication is powered from the Class 1E 120-Vac instrument power supply DCPP UNITS 1 &
2 FSAR UPDATE 5.5-49 Revision 23 December 2016 system. The PORV block valves are shown schematically in Figure 3.2-7. Each of the three valves is powered from a separate Class 1E 480-V bus.
Positive indication of PORV position is obtained by a direct, stem-mounted indicator, which mechanically actuates limit switches at the full-open and full-closed valve stem positions. Acoustic monitors located in the downstream piping provide indication of
PSV positions. The acoustic position indicat ion is seismically qualified to the DDE and HE and environmentally qualified. Sections 3.10 and 3.11 discuss equipment qualification. An alarm is provided in the control room to signal if a PORV is not fully
closed.
The 6-inch pipes connecting the pressurizer nozzles to their respective PSVs are shaped in the form of a loop seal. This arrangement is necessary to accommodate
thermal movement and the collection of condensate for the water loop seal. However, the PSVs have been converted from water-seated to steam-seated, and the water loop seal was eliminated by continuously draining the condensate back to the pressurizer liquid space. With the elimination of the water loop seal, hydraulic loading due to the
presence of water in the loop seal is no longer a concern.
The PORVs are quick-opening, operated automatically or by remote control. Remotely operated stop valves are provided to isolate the PORVs if excessive leakage develops.
Temperatures in the PSV and PORV discharge lines are measured and indicated. An increase in a discharge line temperature is an indication of leakage through the
associated valve. Design parameters for the pressurizer spray control, PSV, and PORVs are provided in Table 5.5-16.
5.5.12.3 Design Evaluation The PSVs prevent RCS pressure from exceeding 110 percent of design pressure. The pressurizer PORVs prevent actuation of the fixed high-pressure trip for all design transients up to and including the design step load decrease, with steam dump but
without reactor trip. The PORVs also limit undesirable opening of the spring-loaded PSVs.
The mounting of these valves is designed to accommodate the magnitude and direction
of thrust of the PSV discharges. In addition, the physical layout is such as to limit the piping reaction loads on these valves.
5.5.12.4 Tests and Inspections PSVs and PORVs, as well as the corresponding PORV block valves, were tested on a prototypical basis to demonstrate their ability to open and close under expected operating conditions for design basis transients and accidents. Qualification criteria include provisions for the associated circuitry, piping, and supports as well as the valves
themselves.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-50 Revision 23 December 2016 Each pressurizer PORV will be demonstrated operable at least once per 24 months by performing a channel calibration of the actuation instrumentation. This frequency
interval is subject to Surveillance Requirement 3.0.2 of the Technical Specifications.
The only other testing performed on PSVs and PORVs, other than operational tests and inspections, is the required hydro static, seat leakage, and operation tests. These tests ensure that the valves will operate as designed. Refer to Sections 5.1.8.19, 5.1.8.20, 5.2.3.14, and 5.2.3.15 for further discussion of testing and inspection of the RCS.
There are no full-penetration welds within the valve body walls. Valves are accessible
for disassembly and internal visual inspection.
5.5.13 COMPONENT SUPPORTS RCS component supports are designed to maintain safe and reliable component and system operation.
5.5.13.1 Design Bases Component supports allow virtually unrestrained lateral thermal movement of the loop during plant operation and provide restraint to the loops and components during
accident conditions. The loading combinations and stress limits are discussed in Section 5.2. The design maintains the integrity of the RCPB for normal and accident conditions and satisfies the requirements of the piping code. Results of piping and supports stress evaluation are presented in Section 5.2.2.1.10.4 and Table 5.2-5 and Section 5.2.2.1.10.5 and Table 5.2-8, respectively.
5.5.13.2 Design Description The support structures for the SG lower supp orts and the RCP supports are primarily welded structural steel sections. The SG upper supports consist of a steel ring with lateral bumpers and four snubbers per SG. The primary equipment supports consist of both linear-type components (tension and compression struts, columns, and beams) and plate and shell components.
The RPV supports incorporate a closed, grout-filled steel box ring-type structure.
Attachments to the supported equipment are the nonintegral type that are bolted to or
bear against the components. The supports-to-concrete attachments are either
embedded anchor bolts or fabricated assemblies. The supports permit virtually
unrestrained thermal growth of the supported systems but restrain vertical, lateral, and rotational movement resulting from seismic and pipe break loadings. This is
accomplished using spherical bushings in the SG columns for vertical support and structural frames, hydraulic snubbers, and struts for lateral support.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-51 Revision 23 December 2016 The principal support material is welded and bolted structural steel that is subjected to Charpy V-notch impact tests in accordance with ASTM Standard Method A370.
Material properties are discussed in Section 5.2.2.3.
The supports for the various components are described in the following paragraphs.
5.5.13.2.1 Reactor Support The reactor is supported on a massive concrete structure that also serves as a
biological shield. Forces are transmitted from the reactor to the concrete support structure by an octagonal closed steel box that provides support at four of the eight
reactor nozzles as shown in Figure 5.5-9. The bearing plates below the reactor nozzle
support shoes contain cooling water passages to control the temperature of the
supporting concrete. The reactor support resists seismic loads and coolant loop (hot
and cold leg) piping reactions. The reactor support system allows the reactor to expand
radially over the supports but resists translational and torsional movement by the
combined tangential restraining action of each nozzle support.
5.5.13.2.2 Steam Generator Supports The SGs are supported by two independent upper and lower structural systems as shown in Figures 5.5-10 and 5.5-11 and described below:
(1) Vertical Supports Four vertical pipe columns for each SG provide full vertical restraint while allowing free movement radially with respect to the reactor. These are bolted at the top to the SG and at the bottom to the concrete structure.
Spherical ball bushing s at the top and bottom of each of column allow unrestrained lateral movement of the SG during heatup and cooldown.
(2) Horizontal Supports Horizontal supports restrain the SGs at two levels:
(a) At elevation 140 feet, where the reinforced concrete slab acts as a rigid diaphragm supporting horizontal forces (predominantly seismic) generated at this level.
(b) At elevation 111 feet (the channel head), where support pads are provided on the SG.
The horizontal supports permit slow radial movement due to thermal expansion while maintaining a positive restraint against sudden loads such as an earthquake or pipe
rupture. This is accomplished through the use of four hydraulic snubbers that have a normal/upset allowable load rating of 1450 kips each at elevation 140 feet attached to a DCPP UNITS 1 &
2 FSAR UPDATE 5.5-52 Revision 23 December 2016 ring shimmed to the SG at 20 locations around the circumference. The faulted allowable snubber load is 2050 kips.
The support pads at elevation 111 feet are keyed and shimmed to a sliding frame that is
sandwiched between two rigid stationary frames anchored to massive concrete walls.
The sliding frame is provided with a bumper system to transfer load to the stationary
frames. The frame system fo r each of two sets of SGs is interconnected so that pipe rupture loads in one loop are distributed between two frame systems.
5.5.13.2.3 Reactor Coolant Pump Supports The RCPs are supported on structural steel frames restrained horizontally at
elevation 106 feet 5-1/2 inches by a system of steel struts anchored to rigid concrete walls as shown in Figures 5.5-10 and 5.5-11. Thermal expansion is permitted by low
friction support pads and oversized mounting holes. The support pads are keyed and
shimmed to the frame. This support system resists vertical and lateral loads due to all
plant operating conditions.
5.5.13.2.4 Pressurizer Support The pressurizer is bolted to a structural steel frame, providing vertical and lateral
support at its base at elevation 113 feet 2 inches as shown in Figure 5.5-12. Additional
lateral support is provided by rigid guides embedded in the concrete slab near the CG of the vessel at elevation 139 feet, in conjunction with lugs projecting from the vessel shell.
The upper support allows the pressurizer to expand radially and vertically, but resists
torsional and translational horizontal movements.
5.5.13.2.5 Crossover Pipe Restraint The crossover leg is restrained at elevation 96 feet by a system of two sets of steel
bumpers located at the elbows of the pipe as shown in Figure 5.5-10. Each set consists
of a bumper strapped to the pipe, which bears on a rigid bumper anchored to a concrete
pad at elevation 94 feet. The restraint resists blowdown loads from a rupture of the crossover pipe. The crossover pipe restraints were deactivated by removing shims.
The bumpers strapped to the pipe and the rigid bumpers were left intact and are
abandoned in place.
5.5.13.3 Design Evaluation Detailed evaluation ensures the design ad equacy and structural integrity of the RCL and the primary equipment supports system. Th e detailed evaluation is made by comparing the analytical results with established criteria for acceptability. Structural analyses are
performed to demonstrate design adequacy for safety and reliability of the plant in case
of a large or small seismic disturbance and/or LOCA conditions. Loads (thermal, weight and pressure) that the system is expected to encounter often during its lifetime are DCPP UNITS 1 &
2 FSAR UPDATE 5.5-53 Revision 23 December 2016 applied and stresses are compared to allowable values, as described in Section 5.2.2.1.
The stress limits for component supports are provided in Tables 5.2-8 and 5.2-8a.
5.5.14 REACTOR VESSEL HEA D VENT SYSTEM 5.5.14.1 Design Bases The basic function of the RVHVS is to remove noncondensable gases from the RVCH.
This system is designed to mitigate a possible condition of inadequate core cooling or
impaired natural circulation resulting from the accumulation of noncondensable gases in the RCS. The design of the RVHVS is in accordance with the requirements of NUREG-
0578, July 1979 (Reference 7) and the subsequent definitions and clarifications in NUREG-0737, November 1980 (Reference 8) (refer to Section 5.1.8.26, Item II.B.1).
5.5.14.2 Design Description The RVHVS removes noncondensable gase s or steam from the RCS via remote-manual operations from the control room. The system discharges at the RVCH, into a well-ventilated area of the containment, to ensure optimum dilution of combustible
gases. The RVHVS is designed to vent a volume of hydrogen at system design
pressure and temperature approximately equivalent to one-half of the RCS volume in 1
hour.
The flow diagram of the RVHVS is shown in Figure 5.5-14. The RVHVS consists of two parallel flowpaths with redundant isolation valves in each flowpath. The venting
operation uses only one of these flowp aths at any time. Equipment design parameters are listed in Table 5.5-17. Isolation valve limit switch position indication is provided in
the control room.
The active portion of the system consists of four 1 inch open/close solenoid operated isolation valves connected to a dedicated RVCH penetration, located near the center of the RVCH. The use of two valves in series in each flowpath minimizes the possibility of RCPB leakage. The isolation valves in one flowpath are powered by one Class 1E power supply, and the valves in the second flowpath are powered by a second Class 1E power supply. The isolation valves are fail closed, normally closed, active valves.
Device qualification is described in Sections 3.10 and 3.11.
If one single active failure prevents a venting operation through one flowpath, the
redundant path is available for venting. Similarly, the two isolation valves in each
flowpath provide a single failure method of isolating the venting system. With
two valves in series, the failure of any one valve or power supply will not inadvertently
open a vent path. These valves are energized-to-open, spring-to-close. Thus, the
combination of PG&E Design Cla ss I train assignments and valve failure modes will not prevent vessel head venting or venting isolation with any single active failure.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-54 Revision 23 December 2016 The RVHVS has two normally deenergized valves in series in each flowpath. This arrangement eliminates the possibility of a spuriously opened flowpath due to the
spurious movement of one valve. As such, power lockout to any valve is not considered
necessary.
The RVHVS is connected to a RVCH vent nozzle penetration. The reactor vent piping utilizes a 3/8-inch orifice prior to branching into two redundant flowpaths. The system is
designed to limit the blowdown from a break downstream of the orifices such that loss
through a severance of one of these lines is sufficiently small to allow operators to
execute an orderly plant shutdown.
A break of the RVHVS line upstream of the orifices would result in an SBLOCA of not greater than 1 inch diameter. Such a break is similar to those analyzed in Reference 2.
Since a break in the head vent line would behave similarly to the hot leg break case
presented in Reference 2, the results presented therein are applicable to a RVHVS line break. This postulated vent line break results, therefore, in no calculated core
uncovery.
All piping and equipment from the housing to second isolation valve are designed and fabricated in accordance with ASME BPVC Section III-2001 through 2003 Addenda, Class 1 requirements. The remainder of the pipi ng is PG&E Design Class II.
5.5.14.3 Supports The RVHVS piping is supported to ensure that the resulting loads and stresses on the piping and on the vent connection to the housing are acceptable. All supports and support structures comply with the requirements of th e ASME BPVC Section III-2001 through 2003 Addenda, Subsection NF.
5.5.15 REFERENCES
- 1. IEEE-Std-279, Criteria for Protection Systems for Nuclear Power Generating Station, 1971.
- 2. Report on Small Break Accidents for Westinghouse NSSS System, WCAP-9600, June 1979.
- 3. Deleted in Revision 18.
- 4. Deleted in Revision 19.
- 5. NEI 97-06, Steam Generator Program Guidelines, latest revision.
- 6. Technical Specifications, Diablo Canyon Power Plant Units 1 and 2, Appendix A to License Nos. DPR-80 and DPR-82, as amended.
DCPP UNITS 1 &
2 FSAR UPDATE 5.5-55 Revision 23 December 2016
- 7. Nuclear Regulatory Commission, TMI Short-Term Lessons Learned Requirements, NUREG-0578, July 1979.
- 8. Nuclear Regulatory Commission, Clarification of TMI Plan Requirements, NUREG-0737, November 1980.
- 9. Deleted in Revision 19.
- 10. Letter from Sheri R. Peterson (NRC) to Gregory M. Rueger (PG&E), Leak-Before-Break Evaluation of Reactor Coolant System Piping for DCPP
Units 1 and 2, March 2, 1993,
- 11. Deleted in Revision 19.
- 12. Westinghouse Calculation SE/FSE-C-PGE-0013, RHRS Cooldown Performance at Uprated Conditions, June 5, 1996.
- 13. Toby Burnett, et al., Systems Evaluation for Reactor Flange Venting for the Diablo Canyon Power Plant, Westinghouse Technical Report, August 1992.
- 14. E. R. Frantz, et al., RCS Pressurization Analysis for Diablo Canyon Shutdown Scenarios, Westinghouse Technical Report, April 3, 1997.
5.5.16 REFERENCE DRAWINGS Figures representing controlled engineering drawings a re incorporated by reference and are identified in Table 1.6-1. The contents of the drawings are controlled by DCPP procedures.
DCPP UNITS 1 &
2 FSAR UPDATE 5.6-1 Revision 23 December 2016 5.6 INSTRUMENTATION REQUIREMENTS 5.6.1 REACTOR COOLANT SYSTEM The RPV, pressurizer, and each of the RCLs are monitored by process control instrumentation. This instrumentation provides the input signals to the following control, display, and protection functions that are described in Chapter 7:
(1) Reactor trip (Section 7.2)
(a) RCS temperatures (overtemperature T, overpower T) (b) Pressurizer pressure (low and high pressure trips) (c) Pressurizer level (d) RCS flow (e) RCP breaker position (2) Engineered safety feat ures actuation (Section 7.3)
(a) Pressurizer pressure (3) PG&E Design Class I functions for safe shutdown (Section 7.4)
(a) Decay heat removal (RCS loop temperatures) (b) RCS pressure control (pressurizer level and pressure)
(4) PG&E Design Class I display information (Section 7.5)
(a) RCS temperatures (b) Pressurizer level (c) Pressurizer pressure (d) RCS pressure (e) RCS flow (f) RCP motor amps (g) PSV position (h) PORV position (i) RPV level (j) Subcooling margin (k) Incore temperatures (5) Other safety features (Section 7.6)
(a) RCS pressure (RHR valve interlock) (b) Pressurizer temperature (RHR valve interlock)
(6) Control systems not required for safety (Section 7.7)
DCPP UNITS 1 &
2 FSAR UPDATE 5.6-2 Revision 23 December 2016 (a) Reactor control system (T avg control) (b) Plant control system interlocks (overtemperature turbine runback) (c) Pressurizer pressure control (d) Pressurizer level control (e) Steam dump control (T avg based) (f) Incore temperatures Refer to Section 5.5.1.2 for a discussion of RCP vibration monitoring.
The RCS design and operating pressure together with the PSV, PORV, and pressurizer spray valves nominal setpoints, and the protection system nominal setpoint pressures are listed in Table 5.2-10. The design pressure allows for operating transient pressure
changes. The selected design margin considers core thermal lag, coolant transport
times and pressure drops, instrumentation and control response characteristics, and
system relief valve characteristics.
5.6.1.1 Inadequate Core Cooling Instrumentation To meet the requirements for supplementing existing instrumentation to unambiguously indicate inadequate core cooling (refer to Sections 5.1.1.26 and 5.1.8.26, Item II.F.2), a subcooling meter and a reactor vessel water level measurement are provided.
Inadequate core cooling detection instrumentation is discussed in more detail in Section
7.5.2.2. The subcooling meters are a subset of RVLIS and provide the operator with
on-line indication of the core coolant temperature and pressure margins to saturation
conditions. The reactor vessel water level is determined by the reactor vessel head level system by measuring the pressure drop between the upper and lower plena in the vessel.
Each subcooling meter (train A or B) has wide range temperature inputs from two each
of the RCS hot legs and the hottest incore thermocouple associated with that train. Two
pressure measurements (one per train) are input from the hot legs. The subcooling
meter displays consist of a digital meter on the main control board (train B), a recorder
to provide a redundant display (train A/PAM1), and the indication on each RVLIS display (PAM3 and PAM4). All the indications provide the temperature margin to
saturation of the RCS. In addition to temperature margin, the RVLIS displays also
provide the pressure margin.
The reactor vessel level measurement is used in combination with the existing core exit
thermocouples and the subcooling meter. Differential pressure between the top of the
reactor vessel and the bottom of the reactor vessel on two narrow-range and two
wide-range instruments is measured. The system functions as follows: with the RCPs off, the pressure drop between the top and t he bottom of the vessel indicates the collapsed liquid level (the equivalent liquid level without voids in the two-phase region) in the vessel. This is read on the narrow-range instrument in terms of feet of liquid.
With the RCPs running, the pressure drop (in feet of liquid) from the top to the bottom of DCPP UNITS 1 &
2 FSAR UPDATE 5.6-3 Revision 23 December 2016 the vessel when compared to the measurement with the same combination of running pumps during normal, single phase RCS con dition, provides an approximate indication of the void fraction in the vessel. This is read on the wide-range instrument as percent
of full flow differential pressure with the vessel filled with water.
5.6.1.2 Loose Parts Monitoring A loose parts and vibration monitoring system is provided for early detection of possible loose parts in the RCS and to reduce their probability of causing damage to RCS components.
Accelerometers (piezoelectric crystals) are located in areas where loose parts are most likely to become entrapped. Redundant accelerometers are installed on the top and the bottom of the RPV and on the lower head of each of the four SGs. Signals from the accelerometers are transmitted by high-temperature leads to preamplifiers located in the containment. From the preamplifiers, the signals are sent to the data acquisition and control panel located in the control room. All components are designed to remain operational over the life of the plant in the temperature, humidity, and radiation environment in which they are installed.
When the output of an individual transducer channel exceeds an adjustable setpoint:
(1) The condition activates a local alarm at the control cabinet.
(2) The output of the alarmed channel is evaluated for validity and logged before being transmitted to the main control board annunciator.
The output of the transducers can be audiomonitored by the operator at the control panel. The alarm monitoring of the selected channel continues during audiomonitoring.
In the event that the output of a loose part channel exceeds the alarm value, the record of the event will be available to the operator and plant staff for analysis. The event will be compared with other previously recorded signatures of the RCS. If necessary, consultants will be contacted to further evaluate the event. This analysis, together with other plant instrumentation, will form the basis for judgment of the effects and significance of the loose parts event.
The sensitivity of the loose parts channe ls is such that a loose part striking the RPV or SGs with as little as one-half-foot-pound of energy produces signals of sufficient strength to be detected over the normal background signals.
5.6.2 RESIDUAL HEAT REMOVAL SYSTEM Process control instrumentation for the RHR system is provided for the following
purposes:
DCPP UNITS 1 &
2 FSAR UPDATE 5.6-4 Revision 23 December 2016 (1) Furnish input signals for monitoring and/or alarming purposes for:
(a) Temperature indications (b) Pressure indications (c) Flow indications (2) Furnish input signals for control purposes of such processes as follows:
(a) Control valve in the RHR pump bypass line so that it opens at flows below a preset limit and closes at flows above a preset limit (b) RHR isolation valves control circui try (refer to Section 7.6 for the description of the interlocks) (c) Control valve in the RHR heat exchanger bypass line to control temperature of reactor coolant returning to reactor loops during
plant cooldown (d) RHR pump circuitry for starting RHR pumps on "S" signal (e) RHR pump trip on low RWST level 5.
6.3 REFERENCES
- 1. Deleted in Revision 22.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.0-1 Sheet 1 of 4 APPLICABLE DESIGN BASIS CRITERIA Revision 23 December 2016 CRITERIA TITLE APPLICABILITY Reactor Coolant System Reactor Coolant System Reactor Coolant Pressure Boundary Section 5.1 5.2 1. General Design Criteria Criterion 2, 1967 Performance Standards X X Criterion 3, 1971 Fire Protection X Criterion 4, 1967 Sharing of Systems X Criterion 4, 1987 Environmental and Dynamic Effects Design Bases X X Criterion 6, 1967 Reactor Core Design X Criterion 9, 1967 Reactor Coolant Pressure Boundary X X Criterion 11, 1967 Control Room X X Criterion 12, 1967 Instrumentation and Controls X X Criterion 13, 1967 Fission Process Monitors and Controls X Criterion 15, 1967 Engineered Safety Features Protection Systems X Criterion 16, 1967 Monitoring Reactor Coolant Pressure Boundary X Criterion 21, 1967 Single Failure Definition X Criterion 33, 1967 Reactor Coolant Pressure Boundary Capability X Criterion 34, 1967 Reactor Coolant Pressure Boundary Rapid Propagation Failure Prevention X Criterion 35, 1967 Reactor Coolant Pressure Boundary Brittle Fracture Prevention X Criterion 36, 1967 Reactor Coolant Pressure Boundary Surveillance X Criterion 40, 1967 Missile Protection X Criterion 49, 1967 Containment Design Basis X Criterion 51, 1967 Reactor Coolant Pressure Boundary Outside Containment X Criterion 54, 1971 Piping Systems Penetrating Containment X Criterion 55, 1971 Reactor Coolant Pressure Boundary Penetrating Containment X Criterion 56, 1971 Primary Containment Isolation X
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.0-1 Sheet 2 of 4 APPLICABLE DESIGN BASIS CRITERIA Revision 23 December 2016 CRITERIA TITLE APPLICABILITY Reactor Coolant System Reactor Coolant System Reactor Coolant Pressure Boundary Section 5.1 5.2 2. System Safety Functional Requirements Protection from Missiles and Dynamic Effects X X Reactor Heat Removal X RCS Thermal-Hydraulic Requirements X RCS Coolant Functional Properties X RCS Pressure and Volume Control X Steam Flow Restriction X RCP Coastdown X Pressurizer Relief Tank X 3. 10 CFR Part 50 50.49 Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants X 50.55a Codes and Standards X 50.55a(f)
Inservice Testing Requirements X X 50.55a(g)
Inservice Inspection Requirements X X 50.60 Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation X 50.61 Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events X 50.63 Loss of All Alternating Current Power X Appendix G Fracture Toughness Requirements X Appendix H Reactor Vessel Material Surveillance Program Requirements X Appendix R Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979 X 4. Regulatory Guides DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.0-1 Sheet 3 of 4 APPLICABLE DESIGN BASIS CRITERIA Revision 23 December 2016 CRITERIA TITLE APPLICABILITY Reactor Coolant System Reactor Coolant System Reactor Coolant Pressure Boundary Section 5.1 5.2 Safety Guide 14, October 1971 Reactor Coolant Pump Flywheel Integrity X Regulatory Guide 1.14, Revision 1, August 1975 Reactor Coolant Pump Flywheel Integrity X Regulatory Guide 1.44, May 1973 Control of the Use of Sensitized Stainless Steel X Regulatory Guide 1.45, May 1973 Reactor Coolant Pressure Boundary Leakage Detection Systems X Regulatory Guide 1.89, November 1974 Environmental Qualification of Class 1E Equipment for Nuclear Power Plants X Regulatory Guide 1.97, Revision 3, May 1983 Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident X X Regulatory Guide 1.99, Revision 2, May 1988 Radiation Embrittlement of Reactor Vessel Materials X Regulatory Guide 1.121, August 1976 Bases for Plugging Degraded PWR Steam Generator Tubes X 5. NRC NUREG NUREG-0737, November 1980 Clarification of TMI Action Plan Requirements X X 6. NRC Generic Letters Generic Letter 83-37, November 1983 NUREG-0737 Technical Specifications X Generic Letter 88-05, March 1988 Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants X Generic Letter 89-10, June 1989 Safety-Related Motor-Operated Valve Testing and Surveillance X
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.0-1 Sheet 4 of 4 APPLICABLE DESIGN BASIS CRITERIA Revision 23 December 2016 CRITERIA TITLE APPLICABILITY Reactor Coolant System Reactor Coolant System Reactor Coolant Pressure Boundary Section 5.1 5.2 Generic Letter 90-06, June 1990 Resolution of Generic Issue 70, "Power-Operated Relief Valve and Block Valve Reliability" and Generic Issue 94, "Additional Low-Temperature Over Pressure Protection for Light-Water Reactors" Pursuant to 10 CFR 50.54(f)
X X Generic Letter 95-07, August 1995 Pressure Locking and Thermal Binding of Safety-Related Power-Operated Valves X 7. NRC Bulletins NRC Bulletin 88-09, July 1988 Thimble Tube Thinning in Westinghouse Reactors X NRC Bulletin 88-11, December 1988 Pressurizer Surge Line Thermal Stratification X 8. Branch Technical Position Branch Technical Position ASB 10-2, March 1978 Design Guidelines for Avoiding Water Hammers in Steam Generators X
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.1-1 Sheet 1 of 2 Revision 22 May 2015 SYSTEM DESIGN AND OPERATING PARAMETERS (c) Unit 1 Unit 2 Plant design life, years (a) 50 50 Nominal operating pressure, psig 2,235 2,235 Total system volume, including
pressurizer and surge line, ft 3 12,064 +/- 100
12,169 +/- 100 System liquid volume, including 11,082 - 11,337 (f) 11,187 - 11,448 (d) pressurizer water, ft 3 (nominal)
Total heat output , Btu/hr 11,687 x10 6 11,687x 10 6
System thermal and hydraulic data (f)
Minimum Measured Flow (RCS total flow), gpm 359,200 362,500 Core Bypass Flow, %
7.5 9.0 Mechanical Design Flow (MDF), gpm/loop 99,600 102,000 Thermal Design Flow, lb/hr 132.9 x 10 6 - 135.1 x 10 6 (f) 134.0 x 10 6 - 136.3 x 10 6 (d)
Reactor vessel Inlet temp, °F 531.7 - 544.5 (f) 531.9 - 545.1 (d) Outlet temp, °F 598.3 - 610.1 (f) 598.1 - 610.1 (d) Steam generator Inlet temp, °F 598.3 - 610.1 (f) 598.1 - 610.1 (d) Outlet temp, °F 531.4 - 544.2 (f) 531.6 - 544.8 (d) Design Fouling Factor , hr-ft 2-°F/BTU 0.00018 0.00018 Reactor coolant pump Inlet temp, °F 531.4 - 544.2 (f) 531.6 - 544.8 (d) Outlet temp, °F 531.7 - 544.5 (f) 531.9 - 545.1 (d) Steam pressure, psia 730 - 821 (f) (g) 731 - 825 (d) (h)
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.1-1 Sheet 2 of 2 Revision 22 May 2015 Steam flow, lb/hr (total) 14.64 x 10 6 - 14.89 x 10 6 (f)(g) 14.64 x 10 6 - 14.90 x 10 6 (d) (h) Feedwater inlet temp, °F 425.0 - 435.0 425.0 - 435.0 Pressurizer spray rate, maximum, gpm 800 800
Pressurizer heater capacity, kW (b) 1800 1800 Pressurizer relief tank volume, ft 3 1800 1800 Best Estimate Operating Data (c) NSSS Power, MWt 3425 3425 (e): Reactor Vessel Avg. Temp., o F 565.0 565.0 RCS Flow, gpm/loop 94,900 95,500 Reactor Coolant Pump developed head, ft 282.3 266.6 48.2 42.46
38.5 38.9 7.2 7.3 (b) Secondary Side Performance Parameters:
Reactor Vessel Avg. Temp., o F 577.3 577.6 RCS Flow, gpm/loop 94,900 95,500 Steam Generators Steam pressure, psia 874 878 Steam flow, lb/hr x 10 6 14.920 14.924 Best Estimate Fouling Factor, hr-ft 2-ºF/BTU 0.00006 0.00006 (a) Although DCPP useful life is expected to be 40 years, the RCS design conservatively assumes that integrity must be maintained during 50 years.
(b) See Table 5.5-12.
(c) 0% SGTP, NSSS rated power (d) Design value corresponding to full power, 565.0 - 577.6ºF vessel average temperature.
(e) Best Estimate calculations were performed to maximize Best Estimate Flow and system/component pressure drops. (f) Design value corresponding to full power, 565.0 - 577.3ºF vessel average coolant temperature. (g) If a high steam pressure is more limiting for analysis purposes, a greater steam pressure of 881 psia, steam temperature of 529.4°F, and steam flow of 14.93x10 6 lb/hr total should be assumed for Unit 1. This is to envelop the possibility that the plant could operate with better than expected steam generator performance. (h) If a high steam pressure is more limiting for analysis purposes, a greater steam pressure of 885 psia, steam temperature of 530.0°F, and steam flow of 14.93x10 6 lb/hr total should be assumed for Unit 2. This is to envelop the possibility that the plant could operate with better than expected steam generator performance.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.2-1 Sheet 1 of 2 Revision 23 December 2016 ASME CODE CASES FOR WESTINGHOUSE PWR CLASS A COMPONENTS HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Code Case (b) Title
1141 Foreign Produced Steel 1332 Requirements for Steel Forgings 1334 Requirements for Corrosion Resistant Steel Bars 1335 Requirements for Bolting Material 1337 Requirements for Special Type 403 Modified Forgings or Bars (Section III) 1344 Requirements for Nickel-Chromium Age-Hardenable Alloys 1345 Requirements for Nickel-Molybdenum-Chromium-Iron Alloys 1355 Electroslag Welding
1358 (a) High Yield Strength Steel for Section III Construction 1360 (a) Explosive Welding 1361 Socket Welds 1364 Ultrasonic Transducers SA-435 (Section II) 1384 Requirements for Precipitation Hardening Alloy Bars & Forgings 1388 Requirements for Stainless Steel - Precipitation Hardening 1390 Requirements for Nickel-Chromium Age-Hardenable Alloys for Bolting 1395 SA-508, Class 2 Forgings - Modified Manganese Content 1401 Welding Repair to Cladding 1407 Time of Examination
1412 (a) Modified High Yield Strength Steel 1414 (a) High Yield Strength Cr-Mo 1423 Plate: Wrought Type 304 with Nitrogen Added 1433 Forgings: SA-387 1434 Class BN Steel Casting (Postweld Heat Treatment for SA-487) 1448 Use of Case Interpretations of ANSI B31 Code for Pressure Piping 1456 Substitution of Ultrasonic Examination 1459 Welding Repairs to Base Metal 1461 Electron Beam Welding 1470 External Pressure Charts for Low Alloy Steel 1471 Vacuum Electron Beam Welding of Tube Sheet Joints 1474 Integrally Finned Tubes (Section III) 1477 B-31.7, ANSI 1970 Addenda N-20-4 SB-163 Nickel-Chromium-Iron Tubing at a Specified Minimum Yield Strength of 40,000 psi 1487 Evaluation of Nuclear Piping for Faulted Conditions 1492 Postweld Heat Treatment 1493 Postweld Heat Treatment 1494 Weld Procedure Qualification Test 1498 SA-508, Class 2, Minimum Tempering Temperature 1501 Use of SA-453 Bolts in Service Below 800 degrees F without Stress Rupture Tests 1504 Electrical and Mechanical Penetration Assemblies
1505 (a) Use of 26 Cr, 1 Mo Steel 1508 Allowable Stresses, Design Stress Intensity and/or Yield Strength Values 1514 Fracture Toughness Requirements 1515 Ultrasonic Examination of Ring Forgings for Shell Section of Section III - Class I Vessels 1516 Welding of Non-Integral Seats in Valves for Section III Application 1517 Material Used in Pipe Fittings 1519 Use of A-105-71 in lieu of SA-105 1521 Use of H. Grades SA-240, SA-479, SA-336, and SA-358 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.2-1 Sheet 2 of 2 Revision 23 December 2016 Code Case (b) Title
1522 ASTM Material Specifications 1523 Plate Steel Refined by Electroslag Remelting 1524 Piping 2" NPS and Smaller 1525 Pipe Descaled by Other Than Pickling 1526 Elimination of Surface Defects 1527 Integrally Finned Tubes 1528 High Strength SA-508 Class 2 and SA-541 Class 2 Forgings for Section III Construction of Class I Components 1529 Material for Instrument Line Fittings 1531 Electrical Penetrations, Special Alloys for Electrical Penetrations Seals 1534 Overpressurization of Valves 1535 Hydrostatic Test of Class I Nuclear Valves 1539 Metal Bellows and Metal Diaphragm Steam Sealed Valves, Class 1, 2, and 3 1542 Requirements for Type 403 Modified Forgings of Bars for Bolting Material 1544 Radiographic Acceptance Standards for Repair Welds 1545 Test Specimens from Separate Forgings for Class 1, 2, 3, and MC.
1546 Fracture Toughness Test for Weld Metal Section 1547 Weld Procedure Qualification Tests; Impact Testing Requirements, Class I 1522 Design by Analysis of Section III Class I Valves
1556 (a) Penetrameters for Film Side Radiographs in Table T-320 of Section V 1567 Test Lots for Low Alloy Steel Electrodes 1568 Test Lots for Low Alloy Steel Electrodes 1571 Materials for Instrument Line Fittings; For SA-234 Carbon Steel Fittings 1573 Vacuum Relief Valves 1574 Hydrostatic Test Pressure for Safety Relief Valves
(a) Westinghouse has performed a review of these specific code cases and knows of no specific application made to components for Diablo Canyon Units 1 and 2.
(b) Code cases adopted for use at DCPP are specified in the introduction to the Inservice Inspection Program Plan.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.2-2 Sheet 1 of 2 Revision 23 December 2016 EQUIPMENT CODE AND CLASSIFICATION LIST
Code Unit 1 Unit 2 Component Class (d) Code Addenda Code Addenda Reactor Coolant System Reactor vessel A ASME III 1965 thru Winter 1966 ASME III 1968 none Reactor vessel closure head A ASME III 2001 thru 2003 ASME III 2001 thru 2003 Control rod drive mechanism housing A ASME III 2001 thru 2003 ASME III 2001 thru 2003 Steam generator (tube side) A ASME III 1998 thru 2000 ASME III 1998 thru 2000 (shell side)
C (a) ASME III 1998 thru 2000 ASME III 1998 thru 2000 Pressurizer A ASME III 1965 thru Summer 1966 ASME III 1965 thru Summer 1966 Reactor coolant piping (b)(c), fittings N/A ASA B31.1 1955 none USAS B31.1.0 1967 none Surge pipe, fittings N/A ASA B31.1 1955 none USAS B31.1.0 1967 none Reactor coolant thermowells N/A ASA B31.1 none ASA B31.1 none Safety valves N/A ASME III 1968 Article 9 ASME III 1968 Article 9 Relief valves N/A USAS B16.5 none USAS B16.5 none Valves to reactor coolant system boundary USAS B16.5 or None USAS B16.5 or None MSS-SP-66 or MSS-SP-66 or N/A ASME III 1968 ASME III 1968 or 1974 (e) or 1974 (e) Piping to reactor coolant system boundary A ANSI B31.7 1969 1970 ANSI B31.7 1969 1970 Pressurizer relief tank C ASME III 1968 thru Summer 1968 ASME III 1968 thru Summer 1968 Reactor coolant pump standpipe orifice N/A No Code No Code Reactor coolant pump standpipe N/A ASME VIII 1968 None ASME VIII 1968 None DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.2-2 Sheet 2 of 2 Revision 23 December 2016 Code Unit 1 Unit 2 Component Class (d) Code Addenda Code Addenda Reactor coolant pump (f) Casing A ASME III 1965 thru Summer 1966 ASME III 1965 thru Summer 1966 Main flange A ASME III 1965 thru Summer 1966 ASME III 1965 thru Summer 1966 Thermal barrier A ASME III 1965 thru Summer 1966 ASME III 1965 thru Summer 1966 #1 seal housing A ASME III 1965 thru Summer 1966 ASME III 1965 thru Summer 1966 #2 seal housing A ASME III 1965 thru Summer 1966 ASME III 1965 thru Summer 1966 Pressure retaining bolting A ASME III 1965 thru Summer 1966 ASME III 1965 thru Summer 1966 Remaining parts N/A ASME III 1965 thru Summer 1966 ASME III 1965 thru Summer 1966 Reactor coolant pump motor oil coolers B ASME III 1965 thru summer 1966 ASME III 1965 thru Summer 1966 (a) Code design requirements are in excess of the requirement dictated by the applicable safety class.
(b) Reactor coolant system piping subassemblies inspected to ASME I as required by California law.
(c) Classification for other piping and associated valves in the reactor coolant system boundary shall be as defined by the systems engineering flow diagrams for the appropriate safety class.
(d) ASME Code Classes A, B, and C were re-designated 1, 2 and 3 in Editions starting in 1971. Refer to Chapter 3.
(e) A small number of valves was purchased to ASME Section III, 1974 requirements.
(f) RCP code dates based on P.O. date of 3/29/67, however, RCP stress report is based on ASME III 1968 Edition with addenda thr ough Summer 1970.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 20 November 2011 TABLE 5.2-3 PROCUREMENT INFORMATION COMPONENTS WITHIN REACTOR COOLANT SYSTEM BOUNDARY Component Purchase Order Dates
Unit 1 Unit 2 Reactor vessel 3/27/67 12/27/68 Replacement RVCH 7/28/06 7/28/06 CRDM housing 7/28/06 7/28/06 Original steam generator 11/22/66 4/6/67 Replacement steam generator 8/12/04 8/12/04 Pressurizer 4/24/67 4/24/67 Reactor coolant pump 3/29/67 3/29/67
Reactor coolant pipe, fittings, and fabrication 5/2/67 (a) 10/7/68 (a) 1/16/68 (b) 11/20/69 (b)
Surge pipe, fittings, and fabrication 5/2/67 (a) 10/7/68 (a) Piping to reactor coolant system boundary fabrication and installation 5/25/70 5/25/70
(a) Purchase of pipe.
(b) Fabrication of pipe.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.2-4 Sheet 1 of 2 Revision 23 December 2016
SUMMARY
OF REACTOR COOLANT SYSTEM DESIGN TRANSIENTS Normal Conditions Occurrences
- 1. RCS heatup and cooldown at 100°F/hr 250 (each)(e) 2. Unit loading and unloading at 5% of full power/min 18,300 (each)
- 3. Step load increase and decrease of 10% of full power 2,500 (each)
- 4. Large step load decrease 250 5. Steady state fluctuations infinite Upset Conditions
- 1. Loss of load (above 15% full power), without immediate turbine or reactor trip 100 (e) 2. Loss of all offsite power 50 (e) 3. Partial loss of flow 100 (e) 4. Reactor trip from full power 500 (e) 5. Inadvertent auxiliary spray (differential temperature > 320
°F 12 (e) 6. Design earthquake 20
- 7. Cold Overpressurization (LTOP) 10 Faulted Conditions (a) 1. RCPB pipe break (d) 1 2. Steam line break 1 3. Double design earthquake (b) 1 4. 7.5M Hosgri earthquake (b) 1 Test Conditions
- 1. Turbine roll test 10 (e) 2. Hydrostatic test conditions
- a. Primary side 10 (e) b. Secondary side 10 (e) 3. Leak tests (for closures)
- a. Primary side 60 (e) b. Secondary side 10
- 4. Tube leak tests (secondary side pressurized as follows) 200 psig 400 400 psig 200 600 psig 120 840 psig 80 Component Specific Analysis (g) Normal Conditions Occurrences
- 1. Pressurizer heatup at 100°F/hr and cooldown at 200°F/hr 250 2. Steam Generator hot standby operation/feedwater cycling (f) 18,300
- 3. Pressurizer boron concentration equalization 32,000 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.2-4 Sheet 2 of 2 Revision 23 December 2016 (a) In accordance with the ASME Boiler and Pressure Vessel Code, faulted conditions are not included in fatigue evaluations. (b) See Section 3.7.
(c) Deleted in Revision 22. (d) With the acceptance of the DCPP leak-before-break analysis by the NRC, dynamic loading conditions resulting from pipe rupture events in the main reactor coolant loop piping no longer have to be considered in the design
basis analyses; only the loads resulting from RCS branch line breaks have to be considered. (e) These limits were contained in Technical Specifications (Table 5.7-1) prior to License Amendment 135 (Improved Technical Specifications) (f) Applies to steam generator only.
(g) These transients apply to the specific component listed, and are provided to clarify the applicable transient.
The number of occurrences represents the applicable number of cycles for the component, consistent with the occurrences identified at the system level. These are not in addition to the number of occurrences identified at the system level.
DCPP UNITS 1 & 2 FSAR UPDATE Sheet 1 of 2 Revision 23 December 2016 TABLE 5.2-5 STRESS LIMITS FOR PG&E QUALITY/CODE CLASS I LOOP PIPING AND VALVES Condition / Loading Combinations Loop Piping (a) (f) Valves 1. Normal (Deadweight + Pressure) S h See Section 3.9.2 (Thermal) S a See Section 3.9.2
- 2. Upset 1.2 S h See Section 3.9.2 (Normal +/- DE loads)
- 3. Faulted - 1 1.8 S h (b) See Section 3.9.2 (Deadweight +Pressure +/- DDE)
- 4. Faulted - 2 (Deadweight + Pressure +/- DDE/Hosgri (c) + LOCA (d)) 3.6 S h (b) See Section 3.9.2
- 5. Faulted - 3 2.4 S h (b) See Section 3.9.2 (Deadweight + Pressure +/- Hosgri)
- 6. Faulted - 4 3.6 S h (b) See Section 3.9.2 (Deadweight + Pressure +/- DDE/Hosgri (c) + Other Pipe Rupture (e))
(a) S h = allowable stress from USAS B31.1 Code for power piping S a = 1.25 S c + 0.25 S h DCPP UNITS 1 & 2 FSAR UPDATE Sheet 2 of 2 Revision 23 December 2016 TABLE 5.2-5 STRESS LIMITS FOR PG&E QUALITY/CODE CLASS I LOOP PIPING AND VALVES S c = allowable stress at cold (ambient) temperature (b) See Table 5.2-7 for additional faulted condition stress limits for loop piping.
(c) The more limiting between the DDE loads and the Hosgri loads. Formal analysis of the (LOCA + Hosgri) faulted load combination is in progress. This analysis is being tracked in the DCPP corrective action program.
(d) Loss of Coolant Accident (LOCA) Loads - The original stress analysis considered main coolant pipe ruptures. With the acceptance of DCPP leak-before-break analysis by the NRC, only the loads resulting from branch line breaks are considered.
(e) Main steam line or feedwater line rupture.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.2-6 Sheet 1 of 2 Revision 23 December 2016 LOAD COMBINATIONS AND STRESS CRITERIA FOR PRIMARY EQUIPMENT (a) CONDITION LOAD COMBINATION STRESS CRITERIA (e)(j) Design Deadweight + Pressure
+/- DE P m S m P L + P b 1.5 S m Normal Deadweight + Pressure + Thermal P L + P b + P e + Q 3 S m (b)
Upset - 1 Deadweight + Pressure + Thermal
+/- DE U T 1.0 (b) P L + P b + P e + Q 3 S m (b) Upset - 2 Deadweight + Pressure + Thermal U T 1.0 (b) P L + P b + P e + Q 3 S m (b) Faulted - 1 Deadweight + Pressure
+/- DDE Table 5.2-7 (k) Faulted - 2 Deadweight + Pressure
+/- (DDE or Hosgri (d,h))+LOCA (d, g) Table 5.2-7 Faulted - 3 Deadweight + Pressure
+/- Hosgri Table 5.2-7 Faulted - 4 Deadweight + Pressure +/- DDE or Hosgri (h) + Other Pipe Rupture (f, i) Table 5.2-7 (a) Reactor coolant pressure boundary components of the steam generators, reactor vessel, reactor coolant pumps, pressurizer. (b) Based on elastic analysis. For simplified elastic-plastic analysis, the stress limits of the 1971 ASME Code Section III, NB-3228.3 apply. (c) Deleted (d) Seismic faulted conditions (DDE or Hosgri) and LOCA combined by ABSUM or SRSS method (SRSS subject to the conditions and limitations of NUREG-0484). (e) For definition of stress criteria terms, see Additional Notes.
(f) Main steam line or feedwater line rupture as applicable. (g) Loss of Coolant Accident (LOCA) Loads - The original stress analysis considered main coolant pipe ruptures. With the acceptance of the DCPP leak-before-break analysis by the NRC, only the loads resulting from RCS branch line breaks are considered. (h) The more limiting between the DDE loads and the Hosgri loads. Formal analysis of the (LOCA +
Hosgri) faulted load combination is in progress. This analysis is being tracked in the DCPP corrective action program. (i) DDE or Hosgri and Other Pipe Rupture combined by ABSUM, unless otherwise noted. (j) For steam generators, stress limits are taken from Appendix F of ASME III. (k) For the reactor vessel, the Faulted-1stress limit is:
P m (or P L) 1.2S m or S y*P m (or P L) + P b 1.8S m or 1.5 S y* (*For elastic analysis, use the greater of the values specified).
P m = General membrane; average primary stress across solid section. Excludes discontinuities and concentrations. Produced only by mechanical loads.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.2-6 Sheet 2 of 2 Revision 23 December 2016 LOAD COMBINATIONS AND STRESS CRITERIA FOR PRIMARY EQUIPMENT P L = Local membrane; average stress across any solid section. Considers discontinuities, but not concentrations. Produced only by mechanical loads.
P b = Bending; component of primary stress proportional to distance from centroid of solid section.
Excludes discontinuities and concentrations. Produced only by mechanical loads.
P e = Expansions; stresses which result from the constraint of "free end displacement" and the effect of anchor point motions resulting from earthquakes. Considers effects of discontinuities, but not local stress concentration. (Not applicable to vessels). Q = Membrane Plus Bending; self-equilibrating stress necessary to satisfy continuity of structure.
Occurs at structural discontinuities. Can be caused by mechanical loads or by differential
thermal expansion. Excludes local stress concentrations.
U T = Cumulative usage factor.
S m = Stress intensity from ASME Section III at temperature S y = Yield stress at temperature DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 5.2-6a LOAD COMBINATIONS AND ACCEPTANCE CRITERIA FOR REPLACEMENT PRIMARY EQUIPMENT (Note 1) NOTES: 1. RVCH, CRDM pressure housings (pressure retaining components), CETNA, and Vent/RVLIS nozzle 2. Load Case Description DL Dead Load (or Dead Weight) I Impulse P Pressure T Thermal Expansion (considered if applicable)
DE Design Earthquake DDE Double Design Earthquake HE Hosgri Earthquake LOCA Loss of Coolant Accident Load (Note 4) 3. Impulse loads apply only to CRDMs 4. For CRDMs, LOCA loads are applied where the CRDM attaches to the RVCH. 5. Seismic faulted conditions (DDE or HE) and LOCA combined by ABSUM or SRSS method (SRSS subject to the conditions and limitations of NUREG-0484). 6. Other pipe ruptures; i.e., main steam line break and feedwater line break, do not impact these components and, therefore, are not included in the load combinations. 7. Formal analysis of the (LOCA + Hosgri) faulted load combination for the affected equipment is in progress. This analysis is being tracked in the DCPP corrective action program. LOAD CONDITION LOAD COMBINATION (Note 2 & 6)
ACCEPTANCE CRITERIA Design DL + P +/- DE ASME Boiler and Pressure Vessel (B&PV) Code,
Section III, Division 1, 2001
Edition through 2003
Addenda - Subsections
NCA and NB Normal (ASME
Service Level A)
DL + P +/- I (Note 3) + T Upset (ASME Service
Level B) DL + P + T +/- DE DL + P + T Faulted (ASME Service Level D)
DL + P +/- (HE or DDE)
DL + P +/- DDE + LOCA (Note 4 & 5)
DL + P +/- HE + LOCA (Note 4, 5, 7)
ASME B&PV Code,Section III, Division 1, 2001
Edition through 2003
Addenda - Appendix F DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.2-7 Revision 23 December 2016 FAULTED CONDITION STRESS LIMITS FOR PG&E QUALITY/CODE CLASS I COMPONENTS System (or Subsystem) Component Stress Limits for Stress Limits for Analysis Analysis Vessels and Pumps (f) Loop Piping (h) Test P m P m + P b P Elastic Smaller of Smaller of 3.6S h (g) 0.8 L T (c)(d) 2.4 S m and 0.70 S u 3.6 Sm and 1.05 S u (b) Elastic
Plastic Larger of 0.70 S u or Larger of 0.70 S ut or Larger of 0.70 S ut or 0.8 L T (c)(d) S y + 1/3 (S u - S y)(c) S y + 1/3 (S ut - S y)(c) S y + 1/3 (S ut - S y)(c)
Limit Analysis 0.9L 1 (a)(c) 0.9L 1 (a)(c) 0.9L 1 (a)(c) 0.8 L T (c)(d)
Plastic Larger of 0.70 S u Larger of 0.70 S ut Larger of 0.70 S ut Plastic or or or 0.8 L T (c)(d)
Elastic S y + 1/3 (S u - S y) S y + 1/3 (S ut - S y) S y + 1/3 (S ut - S y) S y = Yield stress at temperature S ut = ultimate stress from true stress-strain curve at temperature S u = Ultimate stress from engineering stress strain curve at temperature S m = stress intensity from ASME Section III at temperature S h = Allowable stress from USAS B31.1 Code (a) L 1 = Lower bound limit load with an assumed yield point equal to 2.3 S m or 1.5 S y , as applicable. (b) These limits are based on a bending shape factor of 1.5 for simple bending cases with different shape factors; the limits will be changed proportionally. (c) When elastic system analysis is performed, the effect of component deformation on the dynamic system response should be checked.
(d) L T = The limits established for the analysis need not be satisfied if it can be shown from the test of a prototype or model that the specified loads (dynamic or static equivalent) do not exceed 80% of L T , where L T is the ultimate load or load combination used in the test. In using this method, account shall be taken of the size effect and dimensional tolerances (similitude relationships) that may exist between the actual component a nd the tested models
to ensure that the loads obtained from the test are a conservative representation of the load carrying capability of the actual component under
postulated loading for faulted conditions. (e) Deleted (f) For steam generators, stress limits are taken from Appendix F of ASME Section III.
(g) 3.6 S h limit applies to DDE or Hosgri + pipe break (LOCA, main steam line, or feedwater line rupture) faulted conditions only. (h)
DCPP UNITS 1 & 2 FSAR UPDATE Sheet 1 of 2 Revision 23 December 2016 TABLE 5.2-8 LOADING COMBINATIONS AND ACCEPTANCE CRITERIA FOR PRIMARY EQUIPMENT (a) SUPPORTS CONDITION LOADING COMBINATIONS LINEAR-TYPE COMPONENT SUPPORT STRESS LIMITS (i)(j) Normal Deadweight + Temperature + Pressure 1969 AISC Specification, Part 1 Upset Deadweight + Temperature + Pressure +/- DE 1969 AISC Specification, Part 1 Faulted - 1 Deadweight + Pressure
+/- DDE 1969 AISC Specification, Part 2 (c) or S y after load redistribution, whichever is higher Faulted - 2 Deadweight + Pressure
+/- DDE or Hosgri (b,g) + LOCA (b,f) 1969 AISC Specification, Part 2 (c) or S y after load redistribution, whichever is higher Faulted - 3 Deadweight + Pressure +/- Hosgri 1969 AISC Specification, Part 2 (c) or S y after load redistribution, whichever is higher Faulted - 4 Deadweight + Pressure +/- DDE or Hosgri (g) + Other Pipe Rupture (d,h) 1969 AISC Specification, Part 2 (c) or S y after load redistribution, whichever is higher (a) Steam generators, reactor vessel, reactor coolant pumps, pressurizer. (b) Seismic faulted conditions (DDE or Hosgri) and LOCA combined by ABSUM or SRSS method (SRSS subject to the conditions and limitations of NUREG-0484).
DCPP UNITS 1 & 2 FSAR UPDATE Sheet 2 of 2 Revision 23 December 2016 TABLE 5.2-8 LOADING COMBINATIONS AND ACCEPTANCE CRITERIA FOR PRIMARY EQUIPMENT SUPPORTS (c) For supports qualified by load test, allowable loads = 0.8 times L t per UFSAR Table 5.2-7. (d) Main steam line or feedwater line rupture whichever is more limiting. (e) Deleted (f) LOCA - Loss of Coolant Accident Loads - The original stress analysis considered main coolant pipe ruptures. With the accep tance of the DCPP leak-before-break analysis by the NRC, only the loads resulting from RCS branch line breaks are considered. (g) The more limiting between the DDE loads and the Hosgri loads. (h) DDE or Hosgri and Other Pipe Rupture combined by ABSUM, unless otherwise noted.
(i) Stress Limits are also applicable to bolts and anchor bolts. When using S y , connection strength shall exceed the strength of the weakest connected member, shear yield shall be taken as Sy/3 0.5 , and combined tension/shear shall meet the elliptical interaction relationship. (j) For plate and shell-type component supports, allowable stress limits shall be per ASME BPVC Section III, Sub-Section NF, 19 80 Edition including Winter 82 Addenda.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 5.2-8a LOAD COMBINATIONS AND ACCEPTANCE CRITERIA FOR INTEGRATED HEAD ASSEMBLY (IHA)
[PG&E Design Class I Support Structure Components]
LOAD CONDITION LOAD COMBINATION (Notes 1 &,2)
ACCEPTANCE CRITERIA Design DL + P ASME Boiler and Pressure Vessel (B&PV) Code,
Section III, Division 1, 2001
Edition through 2003
Addenda - Subsection NF Normal (ASME
Service Level A)
DL + ML DL + P + T Upset (ASME Service
Level B) DL + P +/- DE
DL + P + T +/- DE Faulted (ASME
Service Level D)
DL + P + T +/- (HE or DDE)
DL + P + T +/- (DDE 2+LOCA 2)1/2 DL + P + T +/- (HE 2+LOCA 2)1/2 ASME B&PV Code,Section III, Division 1, 2001
Edition through 2003
Addenda - Appendix F Faulted (Missile
Shield and Support)
DL + P + T +/- (DDE 2+MI 2)1/2 DL + P + T +/- (HE 2+ MI 2)1/2 NOTES: 1. Load Case Description DL Dead Load P Pressure T Thermal (Note 2) ML Maintenance Load (live loads on walkways during maintenance activities) MI Missile impact load (missile shield and support only)
DE Design Earthquake DDE Double Design Earthquake HE Hosgri Earthquake LOCA Loss of Coolant Accident Load (Note 3) 2. The IHA offers no resistance to reactor vessel thermal growth and, therefore, sustains no stress due to such growth. The temperature load symbol is included in
the above table since this load was considered as part of the IHA design criteria.
Applicable service and accident temperatures are considered when determining
material properties and material allowable stress values.
- 3. The response spectra input used for the IHA LOCA analysis is the envelope of the Unit 1 and 2 LOCA response spectra associated with a pressurizer surge line break, residual heat removal (RHR) line break, and accumulator line break. LOCA motions
are at the reactor head and were therefore applied where the IHA is attached to the
reactor head.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.2-9 Sheet 1 of 3 Revision 23 December 2016 ACTIVE AND INACTIVE VALVES IN THE REACTOR COOLANT PRESSURE BOUNDARY (a) Type Valve Valve A-Active System Valves I.D. Number Location and Figure Number Type Size, in.
Actuation I-Inactive RCS 8000 A, B, C Pressurizer Gate 3 Motor A Figure 3.2-7, Sheets 3 & 4 RCS 8010 A, B, C Pressurizer Relief 6 P A Figure 3.2-7, Sheets 3 & 4 RCS 8078 A,B,C,D Reactor vessel head vent Globe 1 Solenoid A Figure 5.5-14
RCS PCV-455 A,B Pressurizer spray Ball 4 Air I Figure 3.2-7, Sheets 3 & 4
RCS PCV-455 C Pressurizer Globe 2 Air A PCV-456 Figure 3.2-7, Sheets 3 & 4
RCS PCV-474 Pressurizer Globe 2 Air I Figure 3.2-7, Sheets 3 & 4
CVCS LCV-459 RCS cold leg loop 2 Globe 2 Air A Figure 3.2-8, Sheets 5 & 6
CVCS LCV-460 RCS cold leg loop 2 Globe 2 Air A Figure 3.2-8, Sheets 5 & 6
CVCS 8145 CVCS pressurizer auxiliary spray Globe 2 Air A 8148 Figure 3.2-8, Sheets 5 & 6
CVCS 8166 RCS excess letdown Globe 1 Air I 8167 Figure 3.2-8, Sheets 1B & 2
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.2-9 Sheet 2 of 3 Revision 23 December 2016 Type Valve Valve A-Active System Valves I.D. Number Location and Figure Number Type Size, in.
Actuation I-Inactive CVCS 8367 A, B, C, D CVCS seal water injection Figure 3.2-8, Sheets 1, 1A, 1B, 1C &
2 Check 2 P I CVCS 8372 A, B, C, D CVCS seal water injection Check 2 P I Figure 3.2-8, Sheets 1, 1A, 1B, 1C &
2 CVCS 8377 CVCS pressurizer auxiliary spray Check 2 P I Figure 3.2-8, Sheets 5 & 6
CVCS 8378 A CVCS charging line to loop 3 Check 3 P I 8379 A Figure 3.2-8, Sheets 5 & 6
CVCS 8378 B CVCS charging line to loop 4 Check 3 P I 8379 B Figure 3.2-8, Sheet 5 & 6
RHR 8701 RHR isol. hot leg loop 4 Gate 14 Motor I (b) Figure 3.2-10, Sheets 1 & 2
RHR 8702 RHR isol. hot leg loop 4 Gate 14 Motor I (b) Figure 3.2-10, Sheets 1 & 2
RHR 8740 A, B RCS hot leg Check 8 P I Figure 3.2-10, Sheets 1 & 2
SIS 8818 A, B, C, D SIS cold leg Check 6 P I Figure 3.2-9, Sheets 5 & 6
SIS 8819 A, B, C, D SIS cold legs Check 2 P I Figure 3.2-9, Sheets 5 & 6
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.2-9 Sheet 3 of 3 Revision 23 December 2016 Type Valve Valve A-Active System Valves I.D. Number Location and Figure Number Type Size, in.
Actuation I-Inactive SIS 8820 SIS boron injection containment
isolation, Figure 3.2-9, Sheets 3 & 4 Check 3 P I SIS 8900 A, B, C, D SIS cold leg Check 1 1/2 P I Figure 3.2-9, Sheets 3 & 4
SIS 8905 A, B, C, D RCS hot legs Check 2 P I Figure 3.2-9, Sheets 5 & 6
SIS 8948 A, B, C, D RCS cold leg Check 10 P I Figure 3.2-9, Sheets 1 & 2
SIS 8949 A, B, C, D RCS hot legs Check 6 P I Figure 3.2-9, Sheets 5 & 6
SIS 8956 A, B, C, D RCS cold leg Check 10 P I Figure 3.2-9, Sheets 1 & 2
(a) As defined in 10 CFR 50.2, valves are listed first by system (RCS, CVCS, RHR, and SIS) and then by valve I.D. number.
(b) For the postulated Hosgri earthquake this valve is considered active.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 11A April 1997 TABLE 5.2-10 REACTOR COOLANT SYSTEM NOMINAL PRESSURE SETPOINTS (PSIG)
Design pressure 2485 Operating pressure 2235 Safety valves 2485 Power relief valves 2335 Pressurizer spray valves (begin to open) 2260 Pressurizer spray valves (full open) 2310 High-pressure reactor trip 2385 High-pressure alarm 2310 Low-pressure reactor trip (typical, but variable) 1950 Low-pressure alarm 2210 Hydrostatic test pressure 3107 Backup heaters on (pressurizer) 2210 Proportional heaters (begin to operate) 2250
Proportional heaters (full operation) pressurizer 2220 Pressurizer power relief valve interlock 2185
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 5.2-11 REACTOR VESSEL MATERIALS Section Materials
Pressure plate Unit 1: A-533 Grade B Class 1 Unit 2: SA-533 Grade B Class 1
Pressure forgings Unit 1: A-508 Class 2 Unit 2: SA-508 Class 2
RVCH Forging SA-508 Grade 3 Class 1
Primary nozzle safe ends Stainless steel Type 316 Forging
Cladding, stainless Type 304 or equivalent (Combination of Types 308, 308L, 309, 309L, and 312)
Stainless weld rod Types 308L, 308, and 309
O-ring head seals Inconel 718
CRDM housings Inconel 690 and stainless Type 304
Lower tube SB-167
Studs SA-540 Grade B-23 and B-24
Instrumentation nozzles Inconel 600
Thermal insulation Stainless steel
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.2-12 Sheet 1 of 2 Revision 23 December 2016 PRESSURIZER, PRESSURIZER RELIEF TANKS, AND SURGE LINE MATERIALS Pressurizer Unit 1 Unit 2 Shell SA-533, Grade A SA-533, Grade A (Class 1) (Class 2)
Heads SA-216, Grade WCC SA-533, Grade A
(Class 2)
Support skirt SA-516, Grade 70 SA-516, Grade 70 Nozzle weld ends SA-182, F316 SA-182, F316L Inst. tube coupling SA-182, F316 SA-182, F316 Cladding, stainless Type 304 or Type 304 or equivalent equivalent Nozzle forgings SA-508, Class 2 Mn-Mo
Nozzle Weld Overlay N/A First pass 309L, ERNiCr-3 over
dissimilar metal weld Remainder of overlay ERNiCrFe-7 (Automatic GTAW)
ERNiCrFe-7A (Manual GTAW)
Internal plate SA-240, Type 304 SA-240, Type 304 Inst. tubing SA-213, Type 304 316 SA-213, Type 304 316 Heater well tubing SA-213, Type 316 seamless SA-213, Type 316 seamless Heater well adaptor SA-182, F316 SA-182, F316
Pressurizer Relief Tank
Shell ASTM A-285, Grade C ASTM A-285, Grade C Heads ASTM A-285, Grade C ASTM A-285, Grade C Internal coating Amercoat 55 Amercoat 55
Surge Line
Pipes ASTM A-376, Type 316 ASTM A-376, Type 316
Fittings (14 inch elbows) ASTM A-403, WP316 ASTM A-403, WP316 Nozzles ASTM A-182, Grade F316 ASTM A-182, Grade F316
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.2-12 Sheet 2 of 2 Revision 23 December 2016 Valves Unit 1 Unit 2 Pressure-containing parts ASTM A-351, Grade CF8M ASTM A-351, Grade CF8M ASTM A-182, Grade F ASTM A-182, Grade F and ASME SA-351, Grade CF3M (for RCS-8029) and ASME SA-351, Grade
CF3M (for RCS-8029)
DCPP UNITS 1 & 2 FSAR UPDATE Revision 19 May 2010 TABLE 5.2-13 REACTOR COOLANT PUMP MATERIALS Shaft ASTM A-182, Grade F347 Impeller ASTM A-351, Grade CF8 Casing ASTM A-351, Grade CF8 Flywheel ASTM A-533, Grade B, Class I
DCPP UNITS 1 & 2 FSAR UPDATE Revision 19 May 2010 TABLE 5.2-14 STEAM GENERATOR MATERIALS Pressure forgings ASME SA 508, Grade 3, Class 2
Cladding Stainless steel Types 309L, 308L
Tubesheet cladding Alloy 690 weld material
Tubes Alloy 690 TT
DCPP UNITS 1 & 2 FSAR UPDATE Revision 12 September 1998 TABLE 5.2-15 REACTOR COOLANT WATER CHEMISTRY SPECIFICATION Parameter Steady State Transient Limit Conductivity, Mho/cm @ 25 C (a)(c) ---- pH @ 25 C (a)(c) ---- Oxygen, ppm (b) 0.10 1.0 Chloride, ppm 0.15 1.5 Fluoride, ppm 0.15 1.5 Hydrogen, cc(STP)/kg power > 1 MWt (c) ---- normal target band (c) ---- Total suspended solids, ppm (c) ---- Li-7, ppm as Li (c) ---- Boric acid, ppm as B (c) ---- Silica, ppm (c) ---- Aluminum, ppm (c) ---- Calcium, ppm (c) ---- Magnesium, ppm (c) ---- Sulfur compounds, ppm (c) ---- (a) Varies with boric acid and lithium hydroxide concentration.
(b) Limit is not applicable with Tavg 250 F. During startup, hydrazine may be used to achieve RCS concentrations of up to 10 ppm when the coolant temperature is between 150 and 180 F and the oxygen exceeds 0.1 ppm.
(c) Chemical Control Limits and Actions Gui delines for the Primary Systems are listed in plant procedures.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.2-16 Sheet 1 of 4 Revision 23 December 2016 REACTOR COOLANT BOUNDARY LEAKAGE DETECTION SYSTEMS Radioactivity Detection Systems Detector Location or Process
Medium
Type
Range Approximate Time to Detect 1-gpm Leak
Identified
Leak Detection (c)
Instrument Class (a) Indicator
in Control
Room Containment Air G-M 10
-1 to 10 4 mR/hr Less responsive than other detection systems No II Yes Incore inst area Air G-M 10
-1 to 10 4 mR/hr Less responsive than other detection systems No II Yes Containment air
particulate Air NaI Scintillator 10 to 10 6 cpm See Fig. 5.2-9 No II (b) Yes Containment
radiogas Air G-M 10 to 10 6 cpm See Fig. 5.2-9 No II (b) Yes Plant vent radiogas Air Beta Scintillator 10 to 5E6 cpm Less responsive than other detection systems No IB Yes Condenser air
ejector Air Beta Scintillator 10 to 5E6 cpm See Fig. 5.2-10 Yes II Yes Component cooling
liquid Liquid NaI Scintillator 10 to 10 6 cpm See Fig. 5.2-12 No IC Yes Steam generator
blowdown Liquid NaI Scintillator 10 to 10 6 cpm See Fig. 5.2-11 Yes II Yes DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.2-16 Sheet 2 of 4 Revision 23 December 2016 Other Detection Systems Detector Location or Process
Medium
Type
Range and Repeatability (e)
Approximate Time to
Detect 1-gpm Leak (q)
Identified (c) Leak Detection
Instrument Class (a) Indicator
in Control
Room Containment (d) condensation Liquid Change in time required to
accumulate
fixed volume see note (m) 1 hr (g)(h)(l) No II Yes Containment
sumps Liquid Liquid level and quantity of
liquid 1 to 48 in. W.C.
(n) 1 to 35 in. W.C. (p) +/-1 in. <1 hr (h) No II Yes Reactor vessel
flange leakoff Liquid Temperature 50 to 300
°F +/-5 °F <30 sec (f) Yes II Yes Reactor coolant
drain tank Liquid Liquid level and quantity of
liquid 0-100%
+/-2%
<20 min (h) Yes II No Pressurizer relief
valve discharge Liquid Temperature 50 to 400
°F +/-7 °F <30 sec (f) Yes II Yes Pressurizer relief
tank Liquid Yes II Yes Liquid level 0 to 100 %
+/-2% <12 hrs (h)
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.2-16 Sheet 3 of 4 Revision 23 December 2016 Systems Used to Quantify Leakage (i) Detector System
Medium Type Range/Sensitivity
Instrument Class Indicated in Control Room Pressurizer level Liquid Liquid level 0 to 100%
(g)(j) 125 gal/% level IA Yes Volume control tank level Liquid Liquid level 0 to 100%
(g)(j) 19 gal/% level II Yes Charging pump flow Liquid Flow 0 to 200 gpm (k) +/- 10% span when flow
>60 gpm (channel uncertainty value)
IB Yes Pressurizer relief tank level Liquid Liquid level 0 to 100% (h) 128 gal/% level II Yes (a) The PG&E Design Class I instrumentation (i.e., Instrument Class IA and Instrument Class IB Category 1) is capable of performing its nuclear safety function during and after a Double Design Earthquake (DDE) and/or Hosgri Earthquake (HE). Class IC instrument systems r efer to maintenance of pressure boundary integrity of Category I fluid systems. Also refer to UFSAR Sections 3.2 and 3.10.2.
(b) These units were not constructed to withstand DDE accelerations; however, they will be housed in a PG&E Design Class I stru cture and protected from external damage associated with a seismic event. Therefore, it is considered that these units can be returned to operational status within 36
hours of a DDE.
(c) Leakage is defined as identified or unidentified in accordance with Regulatory Guide 1.45.
(d) Containment condensation measures moisture condensed by the fan cooler drip collection system.
(e) Repeatability, including the operators ability to read the same value at another time, is included in this column; this is a true measure of ability to detect a change in system conditions over a period of time.
(f) Automatically alarmed.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.2-16 Sheet 4 of 4 Revision 23 December 2016 (g) Requires operator action - (i.e., close valve, start-stop pump, etc., and operator monitoring and logging).
(h) Requires operator monitoring and logging to note changes in rate, level, flow, etc.
(i) Systems listed here would be used to quantify true leakage rate in the event systems listed on Sheets 1 & 2 above detected an unidentified leak.
These systems also provide additional capability for detecting leak rates of 1-gpm within short periods of time.
(j) Normal variations in process variable or automatic control systems will mask this change. Operator must take action as in (g) above to detect leakage.
(k) Insufficient accuracy/repeatability to ever detect a 1-gpm change in flowrate.
(l) Dependent on initial conditions. May take longer for fan cooler drip level if humidity is initially low.
(m) Level switches (HI and HI-HI) are provided in each CFCU drain line. The level switches have a fixed location in each drain line providing a repeatable alarm. The time intervals between the receipt of the HI level and HI-HI level alarms are monitored and logged by th e operator. Alarm intervals less than a conservative pre-defined value directs the operator to perform an RCS water inventory balance to quantify the RCS leakage rate.
(n) This range refers to the containment structure sumps.
(o) Not used.
(p) This range refers to the reactor cavity sump.
(q) This column refers to the capability of the detection system to sense a leak.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 5.2-17A DCPP UNIT 1 REACTOR VESSEL TOUGHNESS DATA Minimum Average 50 Ft-lb/35 Mil Upper Shelf Plate Material Cu Ni P NDTT Tempt
°F RT NDT Ft-lb Component No. Type (Wt%) (Wt%) (Wt%) °F Long Trans °F Long Trans Repl. Cl. Hd. 07W89-1-1 SA508,CL1 0.05 0.82 0.005 -50 (b) 50 (b) - 211 Ves. Sh. Flg. B4101 A508,CL2 -- 0.75 0.010 35 (a) -5 15 (a) 35 99 (a) Inlet Noz. B4103-1 A508,CL2 -- 0.66 0.013 60 (a) 17 37 (a) 60 77 (a) Inlet Noz. B4103-2 A508,CL2 -- 0.67 0.013 60 (a) 27 47 (a) 60 75 (a) Inlet Noz. B4103-3 A508,CL2 -- 0.68 0.010 43 (a) 10 30 (a) 43 108 (a) Inlet Noz. B4103-4 A508,CL2 -- 0.66 0.010 48 (a) 2 22 (a) 48 106 (a) Outlet Noz. B4104-1 A508,CL2 -- 0.74 0.011 60 (a) -13 7 (a) 60 77 (a) Outlet Noz. B4104-2 A508,CL2 -- 0.76 0.006 43 (a) -3 17 (a) 43 74 (a) Outlet Noz. B4104-3 A508,CL2 -- 0.71 0.012 54 (a) -12 8 (a) 54 86 (a) Outlet Noz. B4104-4 A508,CL2 -- 0.68 0.008 60 (a) 30 50 (a) 60 84 (a) Upper Shl. B4105-1 SA533B,CL1 0.12 0.56 0.010 10 68 88 (a) 28 80 (a) Upper Shl. B4105-2 SA533B,CL1 0.12 0.57 0.008 0 49 69 (a) 9 74 (a) Upper Shl. B4105-3 SA533B,CL1 0.14 0.56 0.010 0 54 74 (a) 14 81 (a) Inter. Shl. B4106-1 SA533B,CL1 0.125 0.53 0.013 -10 57 40 -10 134 116 Inter. Shl. B4106-2 SA533B,CL1 0.120 0.50 0.013 -10 36 57 -3 132 114 Inter.Shl. B4106-3 SA533B,CL1 0.086 0.476 0.011 10 70 90 (a) 30 119 77 (a) Lower Shl. B4107-1 SA533B,CL1 0.13 0.56 0.011 -10 59 75 15 127 110 Lower Shl. B4107-2 SA533B,CL1 0.12 0.56 0.010 -10 64 80 20 127 103 Lower Shl. B4107-3 SA533B,CL1 0.12 0.52 0.010 -50 52 38 -22 135 116 Bot. Hd. Seg. B4111-1 SA533B,CL1 0.15 0.51 0.014 -20 33 53 (a) -7 82 (a) Bot. Hd. Seg. B4111-2 SA533B,CL1 0.12 0.53 0.009 -40 16 36 (a) -24 90 (a) Bot. Hd. Seg. B4111-3 SA533B,CL1 0.13 0.50 0.009 -40 21 41 (a) -19 85 (a) Bot. Hd. Seg. B4110 SA553B,CL1 0.06 0.44 0.010 -10 60 80 (a) 20 75 (a)
(a) Estimated per NRC Standard Review Plan Section 5.3.2.
(b) An NDTT value of -40F was used in the vendor analysis.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 5.2-17B DCPP UNIT 2 REACTOR VESSEL TOUGHNESS DATA Minimum Average 50 Ft-lb/35 Mil Upper Shelf Plate Material Cu Ni P NDTT Tempt
°F RT NDT Ft-lb Component No. Type (Wt%) (Wt%) (Wt%) °F Long Trans °F Long Trans Repl. Cl. Hd. 06W255-1-1 SA508,CL1 0.05 0.82 0.005 - 20 - 211 Inlet Noz. B5461-1 A508,CL2 0.09 0.70 0.012 -20 23 43 (a) -17 116 75 Inlet Noz. B5461-2 A508,CL2 0.09 0.70 0.012 2 18 (a) -20 119 77 (a) Inlet Noz. B5461-3 A508,CL2 0.10 0.82 0.013 45 -25 (a) -40 127 83 (a) Inlet Noz. B5461-4 A508,CL2 0.10 0.81 0.013 48 -28 (a) -40 129 84 (a) Outlet Noz. B5462-1 SA508,CL2 0.11 0.67 0.010 4 16 (a) -44 145 94 Outlet Noz. B5462-4 SA508,CL2 0.11 0.67 0.009 10 10 (a) -40 137.5 89 (a) Outlet Noz. B5462-2 SA508,CL2 0.11 0.67 0.009 -40 14 34 (a) -26 135.5 88 (a) Outlet Noz. B5462-3 SA508,CL2 0.11 0.67 0.009 -50 17 37 (a) -23 131.5 85 (a) Upper Shl. B5453-1 SA533B,CL1 0.11 0.60 0.014 0 85 88 28 92 82 Upper Shl. B5453-3 SA533B,CL1 0.11 0.60 0.012 -10 45 65 (a) 5 136.5 86.5 (a b) Upper Shl. B5011-1R SA533B,CL1 0.11 0.65 0.015 10 40 60 (a) 0 110 72 (a) Inter. Shl. B5454-1 SA533B,CL1 0.14 0.65 0.010 -40 14 112 52 128 91 Inter. Shl. B5454-2 SA533B,CL1 0.14 0.59 0.012 0 60 127 67 113 99 Inter. Shl. B5454-3 SA533B,CL1 0.15 0.62 0.013 -40 30 93 33 129 90 Lower Shl. B5455-1 SA533B,CL1 0.14 0.56 0.010 -20 42 45 -15 134 112 Lower Shl. B5455-2 SA533B,CL1 0.14 0.56 0.011 0 25 45 0 137 122 Lower Shl. B5455-3 SA533B,CL1 0.10 0.62 0.010 0 55 75 15 128 100 Bot. Hd. Seg. B5009-2 SA533B,CL1 0.13 0.57 0.011 -10 110 130 (a) 70 85 55 (a) Bot. Hd. Seg. B5009-3 SA533B,CL1 0.13 0.60 0.009 12 8 (a) -20 131 84 Bot. Hd. Seg. B5009-1 SA533B,CL1 0.13 0.58 0.010 0 88 108 (a) 48 95 62 (a) Bot. Hd. Seg. B5010 SA533B,CL1 0.14 0.63 0.011 -30 20 40 (a) -20 114 74
(a) Estimated per NRC Standard Review Plan Section 5.3.2.
(b) Westinghouse Letter LTR-PCAM-09-26, Revision 1, "Diablo Canyon Units 1 and 2 Reactor Vessel Extended Beltline Material Properties Search," June 3, 2009
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 5.2-18A IDENTIFICATION OF UNIT 1 REACTOR VESSEL BELTLINE REGION BASE MATERIAL Material Composition, Wt.%
Component Plate No.
Heat No. Spec. No.
C Mn P S Si Ni Mo Cu Inter shell B4106-1 C2884-1 SA533B,CL1 0.25 1.34 0.013 0.015 0.21 0.53 0.45 0.125 Inter shell B4106-2 C2854-2 SA533B,CL1 0.18 1.32 0.013 0.015 0.23 0.50 0.46 0.120 Inter shell B4106-3 C2793-1 SA533B,CL1 0.20 1.33 0.011 0.012 0.25 0.476 0.46 0.086 Lower shell B4107-1 C3121-1 SA533B,CL1 0.25 1.36 0.011 0.014 0.24 0.56 0.48 0.13 Lower shell B4107-2 C3131-2 SA533B,CL1 0.24 1.32 0.010 0.013 0.23 0.56 0.46 0.12 Lower shell B4107-3 C3131-1 SA533B,CL1 0.19 1.38 0.010 0.013 0.26 0.52 0.46 0.12
DCPP UNITS 1 & 2 FSAR UPDATE Revision 21 September 2013 TABLE 5.2-18B IDENTIFICATION OF UNIT 2 REACTOR VESSEL BELTLINE REGION BASE MATERIAL Material Composition, Wt.%
Component Plate No.
Heat No. Spec. No.
C Mn P S Si Ni Mo Cu Inter shell B5454-1 C5161-1 SA533B,CL1 0.21 1.30 0.010 0.015 0.19 0.65 0.46 0.14 Inter shell B5454-2 C5168-2 SA533B,CL1 0.25 1.38 0.012 0.016 0.21 0.59 0.55 0.14 Inter shell B5454-3 C5161-2 SA533B,CL1 0.23 1.32 0.013 0.015 0.20 0.62 0.45 0.15 Lower shell B5455-1 C5175-1 SA533B,CL1 0.21 1.38 0.010 0.018 0.19 0.56 0.56 0.14 Lower shell B5455-2 C5175-2 SA533B,CL1 0.22 1.40 0.011 0.018 0.19 0.56 0.56 0.14 Lower shell B5455-3 C5176-1 SA533B,CL1 0.23 1.34 0.010 0.014 0.20 0.62 0.56 0.10
DCPP UNITS 1 & 2 FSAR UPDATE Revision 15 September 2003 TABLE 5.2-19A FRACTURE TOUGHNESS PROPERTIES OF UNIT 1 REACTOR VESSEL BELTLINE REGION BASE MATERIAL Initial EOL (a) Material T NDT ( F) RT NDT ( F) USE (b) (ft-lb) Fluence (c) (N/cm 2) RT NDT (d) ( F) USE (d) (ft-lb) Upper Shell Plate B4105-1 10 28 (e) 80 (e) 1.64E+17 89 74 B4105-2 0 9 (e) 74 (e) 1.64E+17 70 68 B4105-3 0 14 (e) 81 (e) 1.64E+17 77 74 Inter Shell Plate B4106-1 10 116 7.93E+18 115 90 B4106-2 3 114 7.93E+18 113 90 B4106-3 10 30 (e) 77 (e) 7.93E+18 139 63 Lower Shell Plate B4107-1 -10 15 110 7.93E+18 133 87 B4107-2 -10 20 103 7.93E+18 131 82 B4107-3 22 116 7.93E+18 88 93
_______________________
(a) End of license for 40 operating years, September 2021.
(b) Upper shelf energy.
(c) Fluence at vessel wall 1/4 thickness location.
(d) Per Regulatory Guide 1.99, Revision 2.
(e) Estimated from data in the longitudinal direction per NRC Standard Review Plan Section 5.3.2.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 5.2-19B FRACTURE TOUGHNESS PROPERTIES OF UNIT 2 REACTOR VESSEL BELTLINE REGION BASE MATERIAL Initial EOL (a) Material T NDT (°F) RT NDT (°F) USE (b) (ft-lb) Fluence (c) (N/cm 2) RT NDT (d) (°F) USE (d) (ft-lb) Upper Shell Plate B5453-1 0 28 82 1.81E+17 74 75 B5453-3 -10 5 (e) 86.5 (f) 1.81E+17 65 82 B5011-1R 10 0 (e) 72 (e) 1.81E+17 60 66 Inter Shell Plate B5454-1 -40 52 91 8.75E+18 166 69 B5454-2 0 67 99 8.75E+18 180 76 B5454-3 -40 33 90 8.75E+18 173 68 Lower Shell Plate B5455-1 15 112 8.75E+18 114 86 B5455-2 0 0 122 8.75E+18 129 94 B5455-3 0 15 100 8.75E+18 112 81
(a) End of license for 40 operating years, April 2025.
(b) Upper shelf energy.
(c) Fluence at vessel wall 1/4 thickness location.
(d) Per Regulatory Guide 1.99, Revision 2.
(e) Estimated from data in the longitudinal direction per NRC Standard Review Plan Section 5.3.2.
(f) Westinghouse Letter LTR-PCAM-09-26, Revision 1, "Diablo Canyon Units 1 and 2 Reactor Vessel Extended Beltline Material Properties Search," June 3, 2009
DCPP UNITS 1 & 2 FSAR UPDATE Revision 15 September 2003 TABLE 5.2-20A IDENTIFICATION OF UNIT 1 REACTOR VESSEL BELTLINE REGION WELD METAL Weld Weld Wire Flux Average Deposit Composition, Wt.%
Weld Location Process Type Heat No. Type Lot No. C Mn P S Si Mo Ni CR Cu Upper shell Sub-Arc B-4 Mod. 13253 Linde 1092 37740.18 1.30 0.020 0.013 0.24 0.45 0.73 0.19 0.25 to inter shell circle seam 8-442
Inter shell Sub-Arc B-4 Mod. 27204 Linde 1092 37240.14 1.36 0.016 0.025 0.45 0.48 1.018 0.06 0.203 long seams 2-442 A, B, & C
Inter shell Sub-Arc B-4 Mod. 21935 Linde1092 38690.14 1.38 0.015 0.010 0.15 0.54 0.704 -- 0.183 to lower shell circle seam 9-442
Lower shell Sub-Arc B-4 Mod. 27204 Linde 1092 37740.14 1.36 0.016 0.025 0.45 0.48 1.018 0.06 0.203 long seams 3-442 A, B, & C
DCPP UNITS 1 & 2 FSAR UPDATE Revision 15 September 2003 TABLE 5.2-20B IDENTIFICATION OF UNIT 2 REACTOR VESSEL BELTLINE REGION WELD METAL Weld Weld Wire Flux Average Deposit Composition, Wt.%
Weld Location Process Type Heat No. Type Lot No. C Mn P S Si Mo Ni CR Cu Nozzle shell Sub-Arc B-4 Mod. 21935 Linde 1092 3889 0.14 1.38 0.015 0.010 0.15 0.540.704 - 0.183 to inter shell circle seam 8-201
Inter shell Sub-Arc B-4 Mod. 21935 Linde 1092 3869 0.13 1.41 0.018 0.010 0.16 0.550.87 0.03 0.22 long seams (Tandem) B-4 Mod. 12008 2-201 A, B, & C
Inter shell Sub-Arc B-4 10120 Linde 0091 3458 0.14 1.12 0.011 0.008 0.18 0.480.082 - 0.046 to lower shell circle seam 9-201
Lower shell Sub-Arc B-4 33A277 Linde 124 3878 0.11 1.17 0.015 0.011 0.26 0.500.165 0.06 0.258 long seams 3-201 A, B, & C
DCPP UNITS 1 & 2 FSAR UPDATE Revision 20 November 2011 TABLE 5.2-21A FRACTURE TOUGHNESS PROPERTIES OF UNIT 1 REACTOR VESSEL BELTLINE REGION WELD METAL Initial EOL (a) Material RT NDT (°F) USE (b) (ft-lb) Fluence (c) (N/cm 2) RT NDT (d) (°F) USE (d) (ft-lb) Upper Shell Long.
Welds 1-442 A,B,C
-20 86 (f) <1.64+17 69 74 Upper Shell to Inter.
Shell Weld 8-442 -56 (e) 111 (g) <1.64E+17 40 93 Inter. Shell Long.
Welds 2-442 A,B -56 (e) 91 (h) 5.35E+18 194 66 2-442 C -56 (e) 91 (h) 2.87E+18 157 69 Inter. Shell to Lower Shell Weld 9-442 -56 (e) 109 (I) 7.93E+18 166 75 Lower Shell Long.
Welds 3-442 A,B -56 (e) 91 (h) 4.46E+18 182 67 3-442 C -56 (e) 91 (h) 7.93E+18 218 63 (a) End of license for 40 operating years, September 2021.
(b) Upper shelf energy.
(c) Fluence at vessel wall 1/4 thickness location.
(d) Per Regulatory Guide 1.99, Revision 2.
(e) Generic value per 10 CFR 50.61.
(f) CE Vessel Weld Test Report, April 9, 1968.
(g) WCAP 10492, Analysis of Capsule T, Salem 2 Surveillance Program, March 1984.
(h) WCAP 15958, Rev. 0, "Analysis of Capsule V from PG&E Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program,"
January 2003. (i) PG&E Letter DCL-95-176, August 16, 1995, and PG&E Letter DCL-98-094, July 6, 1998.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 20 November 2011 TABLE 5.2-21B FRACTURE TOUGHNESS PROPERTIES OF UNIT 2 REACTOR VESSEL BELTLINE REGION WELD METAL Initial EOL (a) Material RT NDT (°F) USE (b) (ft-lb) Fluence (c) (N/cm 2) RT NDT (d) (°F) USE (d) (ft-lb) Upper Shell Long.
Welds 1-201 A,B,C -50 118 (f) <1.81E+17 14 97 Upper Shell to Inter.
Shell Weld 8-201 -56 (e) 109 (g) <1.81E+17 37 95 Inter. Shell Long.
Welds 2-201 A,B -50 118 (f) 5.61E+18 165 78 2-201 C -50 118 (f) 6.08E+18 170 76 Inter. Shell to Lower Shell Weld 9-201 -56 (e) 125 (h) 8.75E+18 35 102 Lower Shell Long.
Welds 3-201 A,B -56 (e) 88 (h) 6.08E+18 121 56 3-201 B -56 (e) 88 (h) 5.61E+18 118 57 (a) End of license for 40 operating years, April 2025.
(b) Upper shelf energy.
(c) Fluence at vessel wall 1/4 thickness location.
(d) Per Regulatory Guide 1.99, Revision 2.
(e) Generic value per 10 CFR 50.61. (f) WCAP 15423, "Analysis of Capsule V from PG&E Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program,"
September 2000. (g) PG&E Letter DCL-95-176, August 16, 1995. (h) Average of three Charpy tests at +10
°F, CD weld wire/flux qualification test.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 21 September 2013 TABLE 5.2-22 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE UNIT 1 Lead Fluence at Capsule Removal Capsule (f)(g) Location Factor (d) Center (n/cm 2)(d) Time (Plant EFPY)
(a) S 320° 3.48 2.83E+18 1.25 (Tested,1R1)
Y 40° 3.45 1.05E+19 5.86 (Tested, 1R5)
T 140° 3.45 1.05E+19 5.86 (Removed, 1R5)
Z 220° 3.45 1.05E+19 5.86 (Removed, 1R5)
V 320° 2.26 1.36E+19 14.3 (Tested 1R11)
C (b) 140° 3.47 1.22E+19 15.9 (Removed 1R12)
D (b) 220° 3.47 1.22E+19 15.9 (Removed 1R12)
B (b) 40° 3.47 3.44E+19 (projected) 33.0 (Planned 1R23)
A (b) 184° 1.32 Standby Standby U 356° 1.24 Standby Standby X 176° 1.24 Standby Standby W 4° 1.24 Standby Standby UNIT 2 Lead Fluence at Capsule Removal Capsule Location Factor (d) Center (n/cm 2)(d) Time (EFPY)
(a) U 56° 5.20 3.30E+18 1.02 (Tested, 2R1)
X 236° 5.39 9.06E+18 3.16 (Tested, 2R3)
Y 238.5° 4.56 1.53E+19 7.08 (Tested, 2R6)
W (e) 124° 5.35 2.78E+19 11.49 (Removed, 2R9)
V (e) 58.5° 4.57 2.38E+19 11.49 (Tested, 2R9)
Z (e) 304° 5.35 2.78E+19 11.49 (Removed, 2R9)
(a) Approximate full power years from plant startup. (b) Four supplemental capsules installed at 5.86 EFPY (EOC5).
(c) Deleted in Revision 16.
(d) Approximate values taken from WCAP-17299-NP (Rev. 0) for Units 1 and 2. (e) Capsule EFPY for Unit 2 capsules removed in 2R9; W = 61.5, V = 52.5, and Z = 61.5 (f) Unit 1 capsules T, U, W, X, and Z are Type 1 (base metal only)
(g) Unit 1 capsules S, V, and Y are Type 2 (base metal and weld)
DCPP UNITS 1 & 2 FSAR UPDATE Revision 16 June 2005 TABLE 5.2-23 REACTOR COOLANT SYSTEM PRESSURE BOUNDARY ISOLATION VALVES VALVE NUMBER FUNCTION
- 1. 8948 A, B, C, and D Accumulator, RHR, and SIS first off check valves from RCS cold legs 2. 8819 A, B, C, and D SIS second off check valves from RCS cold legs 3. 8818 A, B, C, and D RHR second off check valves from RCS cold legs 4. 8956 A, B, C, and D Accumulator second off check valves from RCS cold legs 5. 8701 and 8702 RHR suction isolation valves 6. 8949 A, B, C, and D RHR and SIS first off check valves from RCS hot legs 7. 8905 A, B, C, and D SIS second off check valves from RCS hot legs
- 8. 8740 (a) A and B RHR second off check valves from RCS hot legs
____________________
(a) 8703 may be used to satisfy Technical Specification 3.4.14 Required Actions A.1 or A.2.1
when in Condition A for valves 8740A and 8740B.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 5.4-1 REACTOR VESSEL DESIGN PARAMETERS (BOTH UNITS)
Design/operating pressure, psig 2485/2235 Design temperature, °F 650 Overall height of vessel and closure head, ft-in.
(bottom head OD of control rod mechanism latch housing mating surface) 47-9 Thickness of insulation, min, in.
3 Number of reactor closure head studs 54 Diameter of reactor closure head/studs, in.
7 ID of flange, in.
205 ID at shell, in.
173 Inlet nozzle ID, in.
27-1/2 Outlet nozzle ID, in.
29 Cladding thickness, min, in.
5/32 Lower head thickness, min, in.
5-1/4 Vessel beltline thickness, min, in.
8-1/2 Closure head thickness, in.
7 DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 5.4-2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED REACTOR VESSEL CONSTRUCTION QUALITY ASSURANCE PROGRAM Forgings RT (a) UT (a) PT (a) MT (a) 1. Flanges - Yes - Yes 2. Studs - Yes - Yes 3. Instrumentation tubes - Yes Yes - 4. Main nozzles - Yes - Yes 5. Nozzles safe ends - Yes Yes - 6. CRDM and Thermocouple Nozzles - Yes Yes - 7. RVHVS and RVLIS Nozzles Yes - Yes -
Plates - Yes - Yes Weldments 1. Main seam Yes Yes (c) - Yes 2. Instrumentation tube connection - - Yes - 3. Main nozzles Yes Yes (c) - Yes 4. Cladding - Yes (b) Yes - 5. Nozzle to safe ends weld Yes - - Yes
- 6. Nozzle to safe ends weld overlay (Unit 2) Yes Yes (c) Yes - 7. All ferritic welds accessible after hydrotest - - - Yes
- 8. All nonferritic welds accessible after hydrotest - - Yes -
- 9. Seal ledge - - - Yes
- 10. Head lift lugs - - - Yes
- 11. Core pads welds - Yes Yes Yes
- 12. CRDM and Thermocouple Nozzle Connections - - Yes -
- 13. RVHVS and RVLIS Nozzle Connections - - Yes -
(a) RT - Radiographic; UT - Ultrasonic; PT - Dye penetrant; MT - Magnetic particle (b) UT of cladding bond-to-base metal (c) UT after hydrotest
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 5.5-1 REACTOR COOLANT PUMP DESIGN PARAMETERS (BOTH UNITS)
Design pressure, psig 2,485 Design temperature, °F 650 Capacity per pump, gpm 88,500 Developed head, ft 277 NPSH required, ft 170 Suction temperature, °F 545 RPM nameplate rating 1,180 Discharge nozzle, ID, in. 27-1/2 Suction nozzle, ID, in. 31 Overall unit height, ft-in. 28-6.7
Water volume, ft 3 56 Moment of inertia, ft-lb 82,000 Weight, dry, lb 198,000 Motor Type AC induction single-speed, air-cooled Power, HP 6,000 Voltage, volts 11,500 Insulation class B, F or H Thermalastic Epoxy F Megaseal Epoxy Phase 3 Starting Current, amps 1,700 Input (hot reactor coolant), kW 4,371 Input (cold reactor coolant), kW 5,790 Seal water injection, gpm 8 Seal water return, gpm 3
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 5.5-2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED REACTOR COOLANT PUMP QUALITY ASSURANCE PROGRAM RT (a) UT (a) PT (a) MT (a) Castings Yes - Yes - Forgings
- 1. Main shaft - Yes Yes -
- 2. Main studs - Yes Yes -
- 3. Flywheel (rolled plate) - Yes Yes (for the bore)
- 1. Circumferential Yes - Yes -
- 2. Instrument connections - - Yes -
(a) RT - Radiographic UT - Ultrasonic PT - Dye penetrant MT - Magnetic particle
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.5-3 Sheet 1 of 2 Revision 23 December 2016 STEAM GENERATOR DESIGN DATA (a) Unit 1 Unit 2 Number of steam generators 4 4 Design pressure, reactor coolant/steam, psig 2,485/1085 2,485/1085 Reactor coolant hydrostatic test pressure (tube side-cold), psig 3,107 3,107 Design temperature, reactor coolant/steam, °F 650/600 650/600 Reactor coolant flow, (per SG) lb/hr 33.2 x 10 6 33.5 x 10 6 Total heat transfer surface area, ft 2 54,240 54,240 Heat transferred, Btu/hr 2,920 x 10 6 2,920 x 10 6
Steam conditions at full load Outlet nozzle:
Steam flow, lb/hr 3.7 x 10 6 3.7 x 10 6 Steam temperature, °F 519 519 Steam pressure, psia 805 (c) 805 (c) Maximum moisture carryover, wt % 0.05 0.05 Feedwater, temperature, °F 435 435 Overall height, ft-in. 68-2 68-2
Shell OD, upper/lower, in. 175-3/8 /135-3/8 175-3/8/135-3/8
Number of U-tubes (b) 4,444 4,444 U-tube outer diameter, in.
0.75 0.75 Tube wall thickness, (minimum), in.
0.043 0.043 Number of manways/ID, in.
4/18 4/18 Number of handholes/ID, in.
4/6 4/6 Number of inspection ports/ID, in.
8/2.5 8/2.5 Number of tube upper bundle inspection
ports/ID, in.
2/4 2/4 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.5-3 Sheet 2 of 2 Revision 23 December 2016 Rated Load Unit 1 Unit 2 Reactor coolant water volume, ft 3 1016 1016 Primary side fluid heat content, Btu 26.0 x 10 6 26.0 x 10 6 Secondary side water volume, ft 3 2100 2100 Secondary side steam volume, ft 3 3700 3700 Secondary side fluid heat content, Btu 6.0 x 10 7 6.0 x 10 7 (a) Quantities are for each steam generator.
(b) The actual number of active tubes (i.e., those contributing to the heat transfer surface area) may be less than the number given due to the plugging and/or removal of some
tubes.
(c) Warranted exit pressure at SG end-of-life (e.g., 10% SGTP and design fouling conditions).
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.5-5 Sheet 1 of 2 Revision 23 December 2016 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED STEAM GENERATOR QUALITY ASSURANCE PROGRAM (BOTH UNITS)
RT (a) UT (a) PT (a) MT (a) ET (a) Tubesheet
- 1. Forging - Yes - Yes -
- 2. Cladding - Yes (b) Yes - -
Channel Head
- 1. Forging Yes - Yes -
- 2. Cladding - Yes Yes - -
Secondary Shell and Head
- 1. Forgings - Yes - Yes -
Tubes - Yes - - Yes Nozzles (Forging) - Yes - Yes -
- 1. Shell, circumferential Yes Yes (d) - Yes -
- 2. Cladding, (channel head-tubesheet joint cladding
restoration) - Yes Yes - -
- 3. Feedwater nozzle to shell Yes - - Yes -
- 4. Support brackets - - - Yes -
- 5. Tube to tubesheet - - Yes - -
- 6. Instrument connections (primary and secondary) - - - Yes -
- 7. Temporary attachments - - - Yes -
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.5-5 Sheet 2 of 2 Revision 23 December 2016 RT (a) UT (a) PT (a) MT (a) ET (a) after removal Weldments (Cont'd)
- 8. After hydrostatic test (all welds where accessible) - - - Yes -
- 9. Primary nozzle safe ends Yes Yes Yes - -
- 10. Steam nozzle safe ends Yes - - -
- 11. Feedwater nozzle safe
ends Yes Yes Yes - -
(a) RT - Radiographic UT - Ultrasonic PT - Dyepenetrant MT - Magnetic particle ET - Eddy current
(b) Flat surfaces only
(c) Weld deposit areas only (d) Welds subject to ASME Section XI ISI
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.5-6 Revision 23 December 2016 REACTOR COOLANT PIPING DESIGN PARAMETERS (BOTH UNITS)
Reactor inlet piping, ID, in. 27.5
Reactor inlet piping, nominal/min wall thickness, in. 2.38/2.22
Reactor outlet piping, ID, in. 29
Reactor outlet piping, nominal/min wall thickness, in. 2.50/2.33
Coolant pump suction piping, ID, in. 31
Coolant pump suction piping, nominal/min wall thickness, in. 2.66/2.50
Pressurizer surge line piping, Unit 1/Unit 2 ID, in. 11.50/11.19
Pressurizer surge line piping, Unit 1/Unit 2 nominal wall thickness, in. 1.25/1.41
Water volume, all loops and surge line, ft 3 1500 Design/operating pressure, psig 2485/2235
Design temperature, °F 650 Design temperature (pressurizer surge line)
°F 680 Design pressure, pressurizer relief line From pressurizer to
safety valve, 2485 psig, 680° F Design temperature, pressurizer relief lines From safety valve to pressurizer
relief tank, 600 psig, 450
°F
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.5-7 Revision 23 December 2016 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED REACTOR COOLANT PIPING QUALITY ASSURANCE PROGRAM (BOTH UNITS)
RT (a) UT (a) PT (a) Fittings and Pipe (Castings) Yes - Yes Fittings and Pipe (Forgings) - Yes Yes Weldments
- 1. Circumferential Yes - Yes
- 2. Nozzle to piperun (except no RT for nozzles less than 4 inches) Yes - Yes
- 3. Instrument connections - - Yes
(a) RT - Radiographic UT - Ultrasonic PT - Dye penetrant
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.5-8 Revision 14 November 2001 DESIGN BASES FOR RESIDUAL HEAT REMOVAL SYSTEM OPERATION (BOTH UNITS)
Residual heat removal system startup No sooner than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reactor
shutdown Number of Trains in Operation 2
Reactor coolant system initial pressure, psig 390
Reactor coolant system initial temperature, F 350 Component cooling water design temperature, F 95 Cooldown time, hours after reactor shutdown <20
Reactor coolant system temper ature at end of cooldown, F 140 Decay heat generation used in cooldown analysis, Btu/hr 75.5 x 10 6
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.5-9 Revision 23 December 2016 RESIDUAL HEAT REMOVAL SYSTEM CODES AND CLASSIFICATIONS (BOTH UNIT 1 and UNIT 2)
Components Code
Residual heat removal pump Draft ASME Code for Pumps and Valves for Nuclear Power-1968, Class II Residual heat exchanger (tube side) ASME BPVC Section III-1968, Class C
(shell side) ASME BPVC Section VIII-1968
Piping ANSI B31.7-1969 with 1970 Addendum, Class II for PG&E Design Class I portions ANSI B31.1-1967 with 1970 Addendum for non-PG&E Design Class I portions
Valves ANSI B16.5-1968
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.5-10 Sheet 1 of 2 Revision 12 September 1998 RESIDUAL HEAT REMOVAL SYSTEM COMPONENT DATA (BOTH UNITS)
Residual Heat Removal Pump Number 2 (per unit)
Design pressure, psig 700 Design temperature, F 400 Design flow, gpm 3000 Design head, ft 350 Net positive suction head, ft Available 36.3 Required 11.0 Residual Heat Exchanger Number 2 (per unit)
Design heat removal capacity, Btu/hr 34.15 x 10 6 Tube-side Shell-side
Design pressure, psig 630 150 Design temperature, F 400 250 Design flow, lb/hr 1.48 x 10 6 2.48 x 10 6 Inlet temperature, F 137 95 Outlet temperature, F 114 108.8 Material Austenitic stainless steel Carbon steel
Fluid Reactor coolant Component cooling water DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.5-10 Sheet 2 of 2 Revision 12 September 1998
Piping and Valves Design pressure, psig 2485 (a) Design temperature, F 650 (a) Design pressure, psig 700 Design temperature, F 400 Suction side relief valve
Relief pressure, psig 450 Relief capacity, gpm 900
Discharge side relief valve
Relief pressure, psig 600 Relief capacity, gpm 20
Material Austenitic stainless steel
(a) Valves and piping that are part of the reactor coolant pressure boundary.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.5-11 Revision 18 October 2008 RECIRCULATION LOOP LEAKAGE Type of Leakage Control and Unit Leakage to Leakage to No. of Leakage Rate Atmosphere, Drain Tank, Items Units Used in the Analysis cc/hr cc/hr Residual heat removal 2 Mechanical seal 20 0 pumps (low-head safety with leakoff of injection) one drop/min
Centrifugal charging pump 2 Same as residual 40 0 (CCP1 and CCP2) heat removal pump
Safety injection 2 Same as residual 40 0 heat removal pump Flanges:
- a. Pump 12 Gasket-adjusted to 0 0 zero leakage following any test
- b. Valves bonnet body 40 10 drops/min/flange 1200 0 (larger than 2 in.) used in analysis (30 cc/hr)
- c. Control valves 6 180 0
- d. Heat exchangers 2 240 0
Valves - stem leakoffs 40 Backseated, double 0 40 packing with leak-off of 1 cc/hr/in.
stem diameter
Miscellaneous small 50 Flanged body packed 50 0 valves stems - 1 drop/min used
Miscellaneous large Double-packing 40 0 valves (larger than 2 in.) 1 cc/hr/in. stem diameter
TOTALS 1910 40
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.5-12 Revision 12 September 1998 PRESSURIZER DESIGN DATA Design/operating pressure, psig 2485/2235 Hydrostatic test pressure (cold), psig 3107 Design/operating temperature, F 680/653 Water volume, full power, ft 3 1080 Steam volume, full power, ft 3 720 Surge line nozzle diameter, in. 14 Shell ID, in. 84
Electric heaters capacity, kW (a) 1800 Heatup rate of pressurizer using heaters only, F/hr 55 Maximum spray rate, gpm 800
(a) Initial heater capacity limit; 150 kW is the minimum required capacity for each backup group that can be supplied by emergency vital power (2 groups).
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 5.5-13 Revision 23 December 2016 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED PRESSURIZER QUALITY ASSURANCE PROGRAM (BOTH UNITS)
Heads RT (a) UT (a) PT (a) MT (a) ET (a) 1. Plates Yes - - Yes -
- 2. Cladding - - Yes - -
Shell 1. Plates - Yes - Yes -
- 2. Cladding - - Yes - -
Heaters
- 1. Tubing (b) - Yes Yes - - 2. Center of element - - - - Yes Nozzle - Yes Yes - -
- 1. Shell, longitudinal Yes - - Yes -
- 2. Shell, circumferential Yes - - Yes -
- 3. Cladding - - Yes - -
- 4. Nozzle safe end (forging) Yes - Yes - -
- 5. Instrument connections - - Yes - -
- 6. Support skirt - - - Yes -
- 7. Temporary attachments after removal - - - Yes -
- 8. All welds and plate heads after - - - Yes -
hydrostatic test
Final Assembly
- 1. All accessible exterior surfaces - - - Yes -
after hydrostatic test
(a) RT - Radiographic; UT - Ultrasonic; PT - Dye penetrant; MT - Magnetic particle; ET - Eddy current (b) Or a UT and ET
DCPP UNITS 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 5.5-14 PRESSURIZER RELIEF TANK DESIGN DATA Design pressure, psig 100
Rupture disk release pressure, psig 100 5% Design temperature, F 340 Total rupture disk relief capacity 1.6 x 10 6 lb/hr at 100 psig
DCPP UNITS 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 5.5-15 REACTOR COOLANT SYSTEM BOUNDARY VALVE DESIGN PARAMETERS Design pressure, psig 2485
Nominal operating pressure, psig 2235
Preoperational plant hydrotest, psig 3107
Design temperature, F 650 DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 5.5-16 PRESSURIZER VALVES DESIGN PARAMETERS Pressurizer Spray Control Valves Number 2
Design pressure 2485
Design temperature, °F 650 Design flow for valves full open, each, gpm 400
Pressurizer Safety Valves
Number 3
Maximum relieving capacity, ASME rated flow, lb/hr 420,000 (per valve)
Set pressure, psig 2485
Fluid Saturated steam
Backpressure:
Normal, psig 3 to 5 Expected during discharge, psig 350
Pressurizer Operated Power Relief Valves (a)
Number 3
Design pressure, psig 2485
Design temperature, °F 650 Relieving capacity at 2,350 psig, lb/hr (per valve) 210,000
Fluid Saturated steam (a) PORVs are credited with liquid discharge for spurious operation of the safety injection system at power events (refer to Section 15.2.15).
DCPP UNITS 1 & 2 FSAR UPDATE Revision 20 November 2011 TABLE 5.5-17 REACTOR VESSEL HEAD VENT SYSTEM EQUIPMENT DESIGN PARAMETERS Valves Number (includes six manual valves) 10 Design pressure, psig 2485
Design temperature, °F 650 Piping Vent line, nominal diameter, in. 1
Design pressure, psig 2485
Design temperature, °F 620 Revision 11 November 1996 FIGURE 5.1-2 PUMP HEAD - FLOW CHARACTERISTICS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE This figure depicts information utilized in the original plant design and is not intended to
be updated. For current plant information, refer to Figure 5.1-2A.
Revision 22 May 2015 FSAR UpdateUnits 1 and 2Diablo Canyon Site(New) Figure 5.1-2ASafety Analysis -RCS Flow ParametersComment [MLF28]: Added for Clarity -Referto Applicability Determination Matrix Item 20.5.1-10Revision 19May2010FSARUPDATE
UNITS 1 and 2 DIABLO CANYON SITEFIGURE5.1-2ASAFETYANALYSIS-RCSFLOWPARAMETERS Revision 22 May 2015 Revision 11 November 1996 FIGURE 5.2-1 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIALS FOR THE REACTOR VESSEL UNIT 1 DIABLO CANYON SITE FSAR UPDATE SUPPORT LUMPED MASS FIGURE 5.2-2 REACTOR COOLANT LOOP MODEL FOR STATIC AND LOCA Revision 23 December 2016 FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 5.2-2A REACTOR COOLANT 4-LOOP MODEL SUPPORT LUMPED MASS STEAM GENERATOR (TYP OF EACH LOOP)
REACTOR COOLANT PUMP (TYP OF EACH LOOP)
REACTORVESSELNote: In the DE and DDE analyses, one of the RCPs is represented by a five-mass, lumped model.
Revision 23 December 2016 FIGURE 5.2-3 THRUST RCL MODEL SHOWING HYDRAULIC FORCE LOCATIONS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 19 Ma y 2010 Revision 11 November 1996 FIGURE 5.2-4 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIAL FOR THE REACTOR VESSEL UNIT 2 DIABLO CANYON SITE FSAR UPDATE FIGURE 5.2-7 LOWER BOUND FRACTURE TOUGHNESS A533 GRADE B CLASS 1 (REF WCAP-7623)
UNIT 1 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 Revision 11 November 1996 FIGURE 5.2-8 TRANSITION TEMPERATURE CORRELATION BETWEEN K ld (DYNAMIC) AND C v FOR A SERIES OF UNIRRADIAYED STEELS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 5.2-9 CONTAINMENT MONITOR RESPONSE TIME VERSUS PRIMARY LEAKRATE UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 5.2-10 AIR EJECTOR RADIOGAS MONITOR RESPONSE TIME VERSUS PRIMARY LEAKRATE UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 5.2-11 BLOWDOWN LIQUID MONITOR RESPONSE TIME VERSUS PRIMARY LEAKRATE UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 5.2-12 CONTAINMENT COOLING WATER LIQUID MONITOR RESPONSE TIME VERSUS PRIMARY LEAKRATE UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 5.2-13 CONTAINMENT AREA MONITOR RESPONSE TIME VERSUS PRIMARY LEAKRATE UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE 0.0010.010.1 1 10 1000001101001,00010,000100,000COUNTS/(MINUTE) 2PRIMARY LEAKRATE (GPM) 25 10-3 10-1 10-2 0.1% Fuel Defect 0.2% Fuel Defects 1% Fuel Defects Revision 19 Ma y 2010 FIGURE 5.2-14 CONTAINMENT RADIOGAS MONI TOR COUNT RATE VERSES PRIMARY LEAKRATE AFTER EQUILIBRIUM UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 5.2-15 CONTAINMENT PARTICULATE MONITOR COUNT RATE VERSUS PRIMARY LEAKRATE AFTER EQUILIBRIUM UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 5.2-16 SURVEILLANCE CAPSULE ELEVATION VIEW UNIT 1 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 5.2-17 SURVEILLANCE CAPSULE PLAN VIEW UNIT 1 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 5.2-18 SURVEILLANCE CAPSULE ELEVATION VIEW UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 5.2-19 SURVEILLANCE CAPSULE PLAN VIEW UNIT 2 DIABLO CANYON SITE FSAR UPDATE 530 540 550 560 570 580 590 600 610 6200102030405060708090100% POWERTEMPERATURE O F THOT LEG TAVERAGE TCOLD LEG NOTE 1: UNIT 1 AND UNIT 2 DESIGN VALUE RANGES FOR FULL POWER.
NOTE 2: THE PLOTS SHOWN ARE FOR THE MAXIMUM T HOT LEG , TAVERAGE , AND TCOLD LEG TEMPERATURES AT FULL POWER.
Revision 21 September 2013 FIGURE 5.3-1 HOT LEG, COLD LEG, AND AVERAGE REACTOR COOLANT LOOP TEMPERATURE AS A FUNCTION OF PERCENT FULL POWERUNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE UNIT 1 NOTE 1 598.3ºF to 610.1ºF UNIT 2 598.1ºF to 610.1ºF UNIT 1 565.0ºF to 577.3ºF UNIT 2 565.0ºF to 577.6ºF UNIT 1 531.7ºF to 544.5ºF UNIT 2 531.9ºF to 545.1ºF NOTE 2 Revision 20 November 2011 FIGURE 5.4-1 REACTOR VESSEL UNIT 1 DIABLO CANYON SITE FSAR UPDATE Revision 20 November 2011 FIGURE 5.4-2 REACTOR VESSEL UNIT 2 DIABLO CANYON SITE FSAR UPDATE DCPP UNITS 1 & 2 FSAR UPDATE
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 5.4-3 INTEGRATED HEAD ASSEMBLY SEISMIC SUPPORT STRUCTURE ASSEMBLY Revision 22 CR 3.7(6) SAPN 50636029-15 Pin connection Between seismic tie-rods and IHA seismic support brackets Pin connection Between seismic tie-rods and seismic wall support brackets Revision 22 May 2015 Revision 11 November 1996 FIGURE 5.5-1 REACTOR COOLANT CONTROLLED LEAKAGE PUMP UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 5.5-2 REACTOR COOLANT PUMP ESTIMATED PERFORMANCE CHARACTERISTICS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 5.5-3 REACTOR COOLANT PUMP SPOOL PIECE AND MOTOR SUPPORT STAND UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE
Revision 11 November 1996 FIGURE 5.5-8 PRESSURIZER UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 5.5-9 REACTOR SUPPORT UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 19 Ma y 2010 UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE FIGURE 5.5-10 STEAM GENERATOR AND REACTOR COOLANT PUMP SUPPORTS
- Crossover pipe restraints Inactive.
FIGURE 5.5-11 COMPONENT SUPPORTS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008 Revision 11 November 1996 FIGURE 5.5-12 PRESSURIZER SUPPORT UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE
FIGURE 5.5-14 SCHEMATIC FLOW DIAGRAM OF THE REACTOR VESSEL HEAD VENT SYSTEM UNITS 1 AND 2 DIABLO CANYON SITE F S AR UPDATE Revision 20 November 2011 UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE FIGURE 5.5-18 SEVEN NOZZLE RSG OUTLET FLOW RESTRICTOR Revision 19 Ma y 2010