DCL-17-038, Diablo Canyon Power Plant, Units 1 & 2, Revised Updated Final Safety Analysis Report, Rev. 23, Chapter 6, Engineered Safety Features (Redacted)
Text
DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 ENGINEERED SAFETY FEATURES CONTENTS Section Title Page i Revision 23 December 2016 6.1 GENERAL 6.1-1 6.
1.1 INTRODUCTION
6.1-1 6.1.2
SUMMARY
DESCRIPTION 6.1-2 6.
1.3 REFERENCES
6.1-5 6.2 CONTAINMENT SYSTEMS 6.2-1 6.2.1 CONTAINMENT FUNCTIONAL DESIGN 6.2-1 6.2.1.1 Design Bases 6.2-1 6.2.1.2 Containment Subcompartment Analysis 6.2-2 6.2.1.3 Safety Evaluation 6.2-17 6.2.1.4 Tests and Inspections 6.2-19 6.2.1.5 Instrumentation Applications 6.2-19 6.2.1.6 Materials 6.2-20 6.2.2 CONTAINMENT HEAT REMOVAL SYSTEMS 6.2-20 6.2.2.1 Design Bases 6.2-21 6.2.2.2 System Description 6.2-26 6.2.2.3 Safety Evaluation 6.2-30 6.2.2.4 Tests and Inspections 6.2-50 6.2.2.5 Instrumentation Applications 6.2-51 6.2.2.6 Materials 6.2-51 6.2.3 CONTAINMENT AIR PURIFICATION AND CLEANUP SYSTEMS 6.2-51 6.2.3.1 Design Bases 6.2-52 6.2.3.2 System Description 6.2-54 6.2.3.3 Safety Evaluation 6.2-56 6.2.3.4 Tests and Inspections (Historical) 6.2-65 6.2.3.5 Instrumentation Applications 6.2-66 6.2.3.6 Materials 6.2-66 6.2.4 CONTAINMENT ISOLATION SYSTEM 6.2-66 6.2.4.1 Design Bases 6.2-66 6.2.4.2 System Description 6.2-70 6.2.4.3 System Design 6.2-74 DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 ENGINEERED SAFETY FEATURES CONTENTS (Continued)
Section Title Page ii Revision 23 December 2016 6.2.4.4 Safety Evaluation 6.2-76 6.2.4.5 Tests and Inspections 6.2-84 6.2.4.6 Materials 6.2-84 6.2.5 COMBUSTIBLE GAS CONTROL IN CONTAINMENT 6.2-84 6.2.5.1 Design Bases 6.2-86 6.2.5.2 System Description 6.2-89 6.2.5.3 Safety Evaluation 6.2-93 6.2.5.4 Tests and Inspections 6.2-103 6.2.5.5 Instrumentation Applications 6.2-103 6.2.5.6 Materials 6.2-104 6.
2.6 REFERENCES
6.2-105 6.2.7 REFERENCE DRAWINGS 6.2-109 6.3 EMERGENCY CORE COOLING SYSTEM 6.3-1 6.3.1 DESIGN BASES 6.3-1 6.3.1.1 General Design Criterion 2, 1967 - Performance Standards 6.3-1 6.3.1.2 General Design Criterion 3, 1971 - Fire Protection 6.3-1 6.3.1.3 General Design Criterion 11, 1967 - Control Room 6.3-2 6.3.1.4 General Design Criterion 12, 1967 - Instrumentation and
Control Systems 6.3-2 6.3.1.5 General Design Criterion 21, 1967 - Single Failure Definition 6.3-2 6.3.1.6 General Design Criterion 37, 1967 - Engineered Safety
Features Basis for Design 6.3-2 6.3.1.7 General Design Criterion 38, 1967 - Reliability and Testability
of Engineered Safety Features 6.3-2 6.3.1.8 General Design Criterion 40, 1967 - Missile Protection 6.3-2 6.3.1.9 General Design Criterion 41, 1967 - Engineered Safety
Features Performance Capability 6.3-2 6.3.1.10 General Design Criterion 42, 1967 - Engineered Safety
Features Components Capability 6.3-3 6.3.1.11 General Design Criterion 43, 1967 - Accident Aggravation
Prevention 6.3-3 6.3.1.12 General Design Criterion 44, 1967 - Emergency Core
Cooling Systems Capability 6.3-3 DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 ENGINEERED SAFETY FEATURES CONTENTS (Continued)
Section Title Page iii Revision 23 December 2016 6.3.1.13 General Design Criterion 45, 1967 - Inspection of Emergency Core Cooling Systems 6.3-3 6.3.1.14 General Design Criterion 46, 1967 - Testing of Emergency Core Cooling Systems Components 6.3-3 6.3.1.15 General Design Criterion 47, 1967 - Testing of Emergency Core Cooling Systems 6.3-3 6.3.1.16 General Design Criterion 48, 1967 - Testing of Operational Sequence of Emergency Core Cooling Systems 6.3-4 6.3.1.17 General Design Criterion 49, 1967 - Containment Design Basis 6.3-4 6.3.1.18 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment 6.3-4 6.3.1.19 General Design Criterion 55, 1971 - Reactor Coolant Pressure Boundary Penetrating Containment 6.3-4 6.3.1.20 General Design Criterion 56, 1971 - Primary Containment Isolation 6.3-4 6.3.1.21 Emergency Core Cooling System Safety Function Requirements 6.3-4 6.3.1.22 10 CFR 50.46 - Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Plants 6.3-5 6.3.1.23 10 CFR 50.49 - Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants 6.3-5 6.3.1.24 10 CFR 50.55a(f) - Inservice Testing Requirements 6.3-5 6.3.1.25 10 CFR 50.55a(g) - Inservice Inspection Requirements 6.3-5 6.3.1.26 Safety Guide 1, November 1970 - Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps 6.3-5 6.3.1.27 Regulatory Guide 1.79, June 1974 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors 6.3-5 6.3.1.28 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 6.3-5 6.3.1.29 NUREG-0737 (Items I.C.1, I.D.2, I I.B.2, II.F.1, I I.F.2, II.K.3.30, II.K.3.31, III.D.1.1), November 1980 - Clarification of TMI Action Plan Requirements 6.3-6 6.3.1.30 Generic Letter 89-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance 6.3-7 DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 ENGINEERED SAFETY FEATURES CONTENTS (Continued)
Section Title Page iv Revision 23 December 2016 6.3.1.31 Generic Letter 95-07, August 1995 - Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves 6.3-7 6.3.1.32 Generic Letter 96-06, September 1996 - Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions 6.3-7 6.3.1.33 Generic Letter 97-04, October 1997 - Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps 6.3-7 6.3.1.34 Generic Letter 98-04, July 1998 - Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System After a loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment 6.3-7 6.3.1.35 Generic Letter 2004-02, September 2004 - Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors 6.3-8 6.3.1.36 Generic Letter 2008-01, January 2008 - Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems 6.3-8 6.3.1.37 IE Bulletin 79-06A (Position 8), April 1979 - Review of Operational Errors and System Misalignments Identified During the Three Mile Island Incident 6.3-8 6.3.1.38 IE Bulletin 80-18, July 1980 - Maintenance of Adequate Minimum Flow Thru Centrifugal Charging Pumps Following Secondary Side High Energy Line Rupture 6.3-8 6.3.1.39 NRC Bulletin 88-04, May 1988 - Potential Safety-Related Pump Loss 6.3-8 6.3.1.40 NRC Bulletin 88-08, June 1988 - Thermal Stresses in Piping Connected to Reactor Coolant Systems 6.3-9 6.3.1.41 NRC Bulletin 2003-01, June 2003
- Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors 6.3-9 6.3.1.42 Branch Technical Position EICSB 18, November 1975 -
Application of the Single Failure Criterion to Manually-Controlled Electrically-Operated Valves 6.3-9
6.3.2 SYSTEM DESCRIPTION 6.3-9 6.3.2.1 Range of Coolant Ruptures and Leaks 6.3-10 DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 ENGINEERED SAFETY FEATURES CONTENTS (Continued)
Section Title Page v Revision 23 December 2016 6.3.2.2 Fission Product Decay Heat 6.3-10 6.3.2.3 Reactivity Required for Cold Shutdown 6.3-10 6.3.2.4 Equipment and Component Descriptions 6.3-10 6.3.2.5 Motor-Operated Valves and Controls 6.3-17 6.3.2.6 Schematic Piping and Instrumentation Diagrams 6.3-18 6.3.2.7 ECCS Flow Diagrams 6.3-18 6.3.2.8 Applicable Codes and Classifications 6.3-18 6.3.2.9 Materials 6.3-18 6.3.2.10 Design Pressures and Temperatures 6.3-19 6.3.2.11 Coolant Quantity 6.3-19 6.3.2.12 Accumulator Availability 6.3-20 6.3.2.13 Dependence on Other Systems 6.3-20 6.3.2.14 Lag Times 6.3-20 6.3.3 SAFETY EVALUATION 6.3-21 6.3.3.1 General Design Criterion 2, 1967 - Performance Standards 6.3-21 6.3.3.2 General Design Criterion 3, 1971 - Fire Protection 6.3-22 6.3.3.3 General Design Criterion 11, 1967 - Control Room 6.3-22 6.3.3.4 General Design Criterion 12, 1967 - Instrumentation and Control Systems 6.3-22 6.3.3.5 General Design Criterion 21, 1967 - Single Failure Definition 6.3-25 6.3.3.6 General Design Criterion 37, 1967 - Engineered Safety
Features Basis for Design 6.3-29 6.3.3.7 General Design Criterion 38, 1967 - Reliability and
Testability of Engineered Safety Features 6.3-38 6.3.3.8 General Design Criterion 40, 1967 - Missile Protection 6.3-39 6.3.3.9 General Design Criterion 41, 1967 - Engineered Safety
Features Performance Capability 6.3-39 6.3.3.10 General Design Criterion 42, 1967 - Engineered Safety
Features Components Capability 6.3-41 6.3.3.11 General Design Criterion 43, 1967 - Accident Aggravation
Prevention 6.3-41 6.3.3.12 General Design Criterion 44, 1967 - Emergency Core
Cooling Systems Capability 6.3-42 6.3.3.13 General Design Criterion 45, 1967 - Inspection of
Emergency Core Cooling Systems 6.3-44 6.3.3.14 General Design Criterion 46, 1967 - Testing of Emergency
Core Cooling Systems Components 6.3-44 DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 ENGINEERED SAFETY FEATURES CONTENTS (Continued)
Section Title Page vi Revision 23 December 2016 6.3.3.15 General Design Criterion 47, 1967 - Testing of Emergency Core Cooling Systems 6.3-45 6.3.3.16 General Design Criterion 48, 1967 - Testing of Operational Sequence of Emergency Core Cooling Systems 6.3-45 6.3.3.17 General Design Criterion 49, 1967 - Containment Design Basis 6.3-46 6.3.3.18 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment 6.3-46 6.3.3.19 General Design Criterion 55, 1971 - Reactor Coolant Pressure Boundary Penetrating Containment 6.3-46 6.3.3.20 General Design Criterion 56, 1971 - Primary Containment Isolation 6.3-46 6.3.3.21 Emergency Core Cooling System Safety Function Requirements 6.3-46 6.3.3.22 10 CFR 50.46 - Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Plants 6.3-47 6.3.3.23 10 CFR 50.49 - Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants 6.3-47 6.3.3.24 10 CFR 50.55a(f) - Inservice Testing Requirements 6.3-48 6.3.3.25 10 CFR 50.55a(g) - Inservice Inspection Requirements 6.3-48 6.3.3.26 Safety Guide 1, November 1970 - Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps 6.3-48 6.3.3.27 Regulatory Guide 1.79, June 1974 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors 6.3-48 6.3.3.28 Regulatory Guide 1.97, Revision 3, May 1983 -
Instrumentation for Li ght-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident 6.3-51 6.3.3.29 NUREG-0737 (Items I.C.1, I.D.2, II.B.2, II.F.1 , II.K.3.30, II.K.3.31, III.D.1.1), November 1980 - Clarification of TMI Action Plan Requirements 6.3-51 6.3.3.30 Generic Letter 89-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance 6.3-52 6.3.3.31 Generic Letter 95-07, August 1995 - Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves 6.3-53 DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 ENGINEERED SAFETY FEATURES CONTENTS (Continued)
Section Title Page vii Revision 23 December 2016 6.3.3.32 Generic Letter 96-06, September 1996 - Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions 6.3-53 6.3.3.33 Generic Letter 97-04, October 1997 - Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps 6.3-53 6.3.3.34 Generic Letter 98-04, July 1998 - Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System After a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment 6.3.54 6.3.3.35 Generic Letter 2004-02, September 2004 - Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors 6.3-55 6.3.3.36 Generic Letter 2008-01, January 2008 - Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems 6.3-63 6.3.3.37 IE Bulleting 79-06A (Position 8), April 1979 - Review of Operational Errors and System Misalignments Identified During the Three Mile Island Incident 6.3-63 6.3.3.38 IE Bulletin 80-18, July 1980 - Maintenance of Adequate Minimum Flow Thru Centrifugal Charging Pumps Following Secondary Side High Energy Line Rupture 6.3-64 6.3.3.39 NRC Bulletin 88-04, May 1988 - Potential Safety-Related Pump Loss 6.3-64 6.3.3.40 NRC Bulletin 88-08, June 1988 - Thermal Stresses in Piping Connected to Reactor Coolant Systems 6.3-64 6.3.3.41 NRC Bulletin 2003-01, June 2003 - Potential Impact of Debris Blockage on Emergency Sump Recirculation and Pressurized-Water Reactors 6.3-65 6.3.3.42 Branch Technical Position EICSB 18, November 1975 -
Application of the Single Failure Criterion to Manually-Controlled Electrically-Operated Valves 6.3.65
6.3.4 TESTS AND INSPECTIONS 6.3-66 6.3.4.1 Quality Control 6.3-66 6.3.4.2 Preoperational System Tests (Historical) 6.3-66 6.3.4.3 Containment Recirculation Sump and Screen Inspection 6.3-67
DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 ENGINEERED SAFETY FEATURES CONTENTS (Continued)
Section Title Page viii Revision 23 December 2016 6.3.5 INSTRUMENTATION APPLICATIONS 6.3-67 6.
3.6 REFERENCES
6.3-67 6.4 HABITABILITY SYSTEMS 6.4-1 6.4.1 CONTROL ROOM 6.4-1 6.4.1.1 Design Bases 6.4-1 6.4.1.2 System Description 6.4-4 6.4.1.3 Safety Evaluation 6.4-5 6.4.1.4 Tests and Inspections 6.4-9 6.4.1.5 Instrumentation Applications 6.4-10 6.4.2 TECHNICAL SUPPORT CENTER 6.4-10 6.4.2.1 Design Bases 6.4-10 6.4.2.2 System Description 6.4-11 6.4.2.3 Safety Evaluation 6.4-12 6.4.2.4 Tests and Inspections 6-4-13 6.4.2.5 Instrumentation Applications 6-4-14 6.
4.3 REFERENCES
6.4-14 6.5 AUXILIARY FEEDWATER SYSTEM 6.5-1 6.5.1 DESIGN BASES 6.5-1 6.5.1.1 General Design Criterion 2, 1967 - Performance Standards 6.5-1 6.5.1.2 General Design Criterion 3, 1971 - Fire Protection 6.5-1 6.5.1.3 General Design Criterion 4, 1967 - Sharing of Systems 6.5-1 6.5.1.4 General Design Criterion 11, 1967 - Control Room 6.5-1 6.5.1.5 General Design Criterion 12, 1967 - Instrumentation and
Control Systems 6.5-1 6.5.1.6 General Design Criterion 21, 1967 - Single Failure Definition 6.5-1 6.5.1.7 General Design Criterion 37, 1967 - Engineered Safety
Features Basis for Design 6.5-2 6.5.1.8 General Design Criterion 38, 1967 - Reliability and Testability
of Engineered Safety Features 6.5-2 6.5.1.9 General Design Criterion 40, 1967 - Missile Protection
(Dynamic Effects) 6.5-2 DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 ENGINEERED SAFETY FEATURES CONTENTS (Continued)
Section Title Page ix Revision 23 December 2016 6.5.1.10 General Design Criterion 41, 1967 - Engineered Safety Features Performance Capability 6.5-2 6..5.1.11 General Design Criterion 54,1971 - Piping Systems Penetrating Containment 6.5-2 6.5.1.12 General Design Criterion 57, 1971 - Closed System Isolation Valves 6.5-2 6.5.1.13 Auxiliary Feedwater System Safety Function Requirements 6.5.1.14 10 CFR 50.49 - Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants 6.5-3 6.5.1.15 10 CFR 50.55a(f) - Inservice Testing Requirements 6.5-3 6.5.1.16 10 CFR 50.55a(g) - Inservice Inspection Requirements 6.5-3 6.5.1.17 10 CFR 50.62 - Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants 6.5-3 6.5.1.18 10 CFR 50.63 - Loss of all Alternating Current Power 6.5-3 6.5.1.19 10 CFR Part 50 Appendix R (Sections III.G, J, and L) - Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979 6.5-3 6.5.1.20 Regulatory Guide 1.97, Revision 3 - Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident 6.5-4 6.5.1.21 NUREG-0737 (Items II.E.1.1 and II.E.1.2), November 1980 - Clarification of TMI Action Plan Requirements 6.5-4 6.5.1.22 Generic Letter 89-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance 6.5-4
6.5.2 SYSTEM DESCRIPTION 6.5-4 6.5.2.1 Equipment and Component Descriptions 6.5-5 6.5.2.2 Design Conditions 6.5-8 6.5.2.3 Applicable Codes and Classifications 6.5-11
6.5.3 SAFETY EVALUATION 6.5-11 6.5.3.1 General Design Criterion 2, 1967 - Performance Standards 6.5-11 6.5.3.2 General Design Criterion 3, 1971 - Fire Protection 6.5-12 6.5.3.3 General Design Criterion 4, 1967 - Sharing of Systems 6.5-12 6.5.3.4 General Design Criterion 11, 1967 - Control Room 6.5-13 6.5.3.5 General Design Criterion 12, 1967 - Instrumentation and Control Systems 6.5-14 6.5.3.6 General Design Criterion 21, 1967 - Single Failure Definition 6.5-14 DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 ENGINEERED SAFETY FEATURES CONTENTS (Continued)
Section Title Page x Revision 23 December 2016 6.5.3.7 General Design Criterion 37, 1967 - Engineered Safety Features Basis for Design 6.5-15 6.5.3.8 General Design Criterion 38, 1967 - Reliability and Testability of Engineered Safety Features 6.5-19 6.5.3.9 General Design Criterion 40, 1967 - Missile Protection (Dynamic Effects) 6.5-19 6.5.3.10 General Design Criterion 41, 1967 - Engineered Safety Features Performance Capability 6.5-20 6.5.3.11 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment 6.5-20 6.5.3.12 General Design Criterion 57, 1971 - Closed System Isolation Valves 6.5-21 6.5.3.13 Auxiliary Feedwater System Function Requirements 6.5-21 6.5.3.14 10 CFR 50.49 - Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants 6.5-21 6.5.3.15 10 CFR 50.55a(f) - Inservice Testing Requirements 6.5-21 6.5.3.16 10 CFR 55.55a(g) - Inservice Inspection Requirements 6.5-22 6.5.3.17 10 CFR 50.62 - Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants 6.5-22 6.5.3.18 10 CFR 50.63 - Loss of All Alternating Current Power 6.5-22 6.5.3.19 10 CFR Part 50 Appendix R (Sections III.G, J, and L) - Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979 6.5-22 6.5.3.20 Regulatory Guide 1.97, Revision 3 - Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident 6.5-23 6.5.3.21 NUREG-0737 (Items II.E.1.1, and II.E.1.2), November 1980 -
Clarification of TMI Action Plan Requirements 6.5-23 6.5.3.22 Generic Letter 89-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance 6.5-24
6.5.4 TESTS AND INSPECTIONS 6.5-24
6.5.5 INSTRUMENTATION REQUIREMENTS 6.5-24
6.
5.6 REFERENCES
6.5-24
6.5.7 REFERENCE DRAWINGS 6.5-25 DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 ENGINEERED SAFETY FEATURES CONTENTS (Continued)
Section Title Page xi Revision 23 December 2016
6.2D APPENDIX 6.2D - ANALYSIS OF LONG-TERM LOSS-OF-COOLANT ACCIDENTS AND MAIN STEAMLINE BREAK EVENTS 6.2D-1
DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 TABLES Table Title xii Revision 23 December 2016 6.1-1 Applicable Design Basis Criteria
6.2-1 Deleted in Revision 18.
6.2-2 Deleted in Revision 18.
6.2-3 Deleted in Revision 18.
6.2-4 Deleted in Revision 18.
6.2-5 Deleted in Revision 11.
6.2-6 Deleted in Revision 18.
6.2-7 Deleted in Revision 18.
6.2-8 Deleted in Revision 11.
6.2-9 Deleted in Revision 11.
6.2-10 Deleted in Revision 11.
6.2-11 Deleted in Revision 18.
6.2-12 Deleted in Revision 18.
6.2-13 Deleted in Revision 18.
6.2-14 Deleted in Revision 23.
6.2-15 Deleted in Revision 23.
DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 TABLES (Continued)
Table Title xiii Revision 23 December 2016 6.2-16 Deleted in Revision 23.
6.2-17 Deleted in Revision 23.
6.2-18 Deleted in Revision 23.
6.2-19 Deleted in Revision 23.
6.2-20 Deleted in Revision 23.
6.2-21 Deleted in Revision 23.
6.2-22 Deleted in Revision 23.
6.2-23 Deleted in Revision 23.
6.2-24 Deleted in Revision 23.
6.2-25 Containment Heat Removal Systems Design Code Requirements
6.2-26 Containment Heat Removal Systems Design Parameters
6.2-27 Single Failure Analysis - Containment Heat Removal Systems
6.2-28 Deleted in Revision 18.
6.2-29 Spray Additive System Design Parameters
6.2-30 Spray Additive System - Codes Used in System Design
6.2-31 Deleted in Revision 11.
6.2-32 Deleted in Revision 11.
6.2-33 Deleted in Revision 11.
6.2-34 Deleted in Revision 11.
6.2-35 Deleted in Revision 11.
DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 TABLES (Continued)
Table Title xiv Revision 23 December 2016 6.2-36 Parameters and Results for Spray Iodine Removal Analysis During Injection Phase Operation
6.2-37 Spray Fall Heights in the Containment
6.2-38 Spray Additive System Single Failure Analysis
6.2-39 Containment Piping Penetrations and Valving
6.2-40 Operating Conditions for Containment Isolation
6.2-41 Post-LOCA Temperature Transient Used for Aluminum and Zinc Corrosion
6.2-42 Parameters Used to Determine Hydrogen Generation
6.2-43 Core Fission Product Energy After Operation with Extended Fuel Cycles
6.2-44 Fission Product Decay Deposition in Sump Solution 6.2-45 Summary of Hydrogen Accumulation Data
6.2-46 Deleted in Revision 2.
6.2-47 Containment Reflective Insulation
6.2-48 Containment Conventional Insulation
6.2-49 Deleted in Revision 11.
6.2-50 Deleted in Revision 11.
6.2-51 Deleted in Revision 8.
6.2-52 Deleted in Revision 8.
6.2-53 Deleted in Revision 11
6.2-54 Deleted in Revision 11.
6.2-55 Deleted in Revision 23.
DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 TABLES (Continued)
Table Title xv Revision 23 December 2016 6.2-56 Deleted in Revision 23.
6.2-57 Containment Subcompartment Analyses Double-Ended Pressurizer Surge Line Break At The Hot Leg Conne ction Mass And Energy Releases
6.2-58 Containment Subcompartment Analyses Double Ended Pressurizer Spray Line Break At The Top Of The Pressurizer Mass And Energy Releases
6.2-59 Containment Subcompartment Analyses Do uble-Ended Residual Heat Removal Suction Line Break At The RCS Hot Leg Connection Mass And
Energy Releases
6.2-60 Containment Subcompartment Analyses Double-Ended Accumulator Line Break At The RCS Cold Leg Connection Mass And Energy Releases
6.2-61 Containment Subcompartment Analyses Node Descriptions For Subcompartment Model 6.2-62 Containment Subcompartment Analyses Volumes For Subcompartment Mode 6.2-63 Containment Subcompartment Analyses Subcompartment Model Flow Path Characteristics
6.2-64 Containment Subcompartment Analyses LOCA Peak Differential Pressures
6.2-65 Containment Subcompartment Analyses Main Steam Line Break Outside The Crane Wall At The Containment Penetration Mass And Energy Releases
6.2-66 Containment Subcompartment Analyses Main Feedwater Line Break Inside The Steam Generator Enclosure At The Inlet To The Steam Generator Mass
And Energy Released
6.2-67 Containment Subcompartment Analyses Non-LOCA Peak Differential Pressures
6.3-1 Emergency Core Cooling System - Component Parameters
6.3-2 Emergency Core Cooling System - Design Code Requirements
6.3-3 Materials of Construction - Emergency Core Cooling System Components DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 TABLES (Continued)
Table Title xvi Revision 23 December 2016
6.3-4 Deleted in Revision 11A
6.3-5 Safety Injection to Recirculation Mode; Sequence and Timing of Manual Changeover
6.3-6 Normal Operating Status of Emergency Core Cooling System Components for Core Cooling
6.3-7 Sequence and Delay Times for Startup of ECCS
6.3-8 Emergency Core Cooling System Shared Functions Evaluation
6.3-9 Maximum Potential Recirculation Loop Leakage External to Containment
6.3-10 ECCS Relief Valve Data
6.3-11 Net Positive Suction Heads for Post-DBA Operational Pump 6.3-12 ECCS Motor-Operated Valves with Electric Power Removed During Normal Power Plant Operation
6.3-13 Single Active Failure Analys is for Emergency Core Cooling System Components
6.3-14 Emergency Core Cooling System Recirculati on Piping Passive Failure Analysis 6.5-1 Criteria for Auxiliary Feedwater S ystem Design Basis Conditions
6.5-2 Summary of Assumptions - AFW System Design Verification
6.5-3 Summary of Sensible Heat Sources (For Plant Cooldown by AFW System)
DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 FIGURES Figure Title xvii Revision 23 December 2016 6.2-1 Deleted in Revision 18.
6.2-2 Deleted in Revision 18.
6.2-3 Deleted in Revision 18.
6.2-4 Deleted in Revision 18.
6.2-5 Deleted in Revision 11.
6.2-6 Deleted in Revision 11.
6.2-7 Deleted in Revision 11.
6.2-8 Deleted in Revision 11.
6.2-9 Deleted in Revision 11.
6.2-10 Containment Spray Pump Performance Curve
6.2-11(a) Deleted in Revision 22.
6.2-11A (a) Deleted in Revision 22.
6.2-12 Containment Spray Nozzle Cutaway
6.2-13 Containment Spray Headers Plan View
6.2-14 Comparison of Spray Removal Model and CSE Results (Run A6)
6.2-15 Containment Recirculation Sump pH vs.
Time After LOCA Begins
6.2-16 Containment Equilibrium Elemental Iodine Partition Coefficient vs. Time for Minimum Sump pH Case
6.2-17 Containment Isolation System DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 FIGURES (Continued)
Figure Title xviii Revision 23 December 2016 6.2-18 Penetration Diagram Legend
6.2-19 Penetration Diagram
6.2-20 Containment Hydrogen Purge System Purge Stream
6.2-21 Containment Hydrogen Purge System Supply Stream
6.2-22 Containment Hydrogen Purge System, Hydrogen Analyzer Stream
6.2-23 Model B Electric Hydrogen Recombiner - Cutaway
6.2-24 Aluminum and Zinc Corrosion Rate Design Curve
6.2-25 Results of Westinghouse Capsule Irradiation Tests
6.2-26 Post-LOCA Containment Hydrogen Concentration
6.2-27 Post-LOCA Hydrogen Accumulation
6.2-28 Post-LOCA Hydrogen Production
6.2-29 Post-LOCA Hydrogen Accumulation from Corrosion of Material Inside Containment with No Recombiner
6.2-30 Deleted in Revision 9.
6.2-31 Deleted in Revision 2.
6.2-32 Deleted in Revision 18.
6.2-33 Deleted in Revision 23.
6.2-34 Deleted in Revision 23.
6.2-35 Deleted in Revision 23.
6.2-36 Deleted in Revision 23.
6.2-37 Deleted in Revision 23.
DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 FIGURES (Continued)
Figure Title xix Revision 23 December 2016 6.2-38 Deleted in Revision 23.
6.2-39 Deleted in Revision 23.
6.2-40 Deleted in Revision 23.
6.2-41 Deleted in Revision 23.
6.2-42 Deleted in Revision 23.
6.2-43 Deleted in Revision 23.
6.2-44 Deleted in Revision 23.
6.2-45 Deleted in Revision 22.
6.2-46 Deleted in Revision 23.
6.2-47 Deleted in Revision 11.
6.2-47A Deleted in Revision 11.
6.2-48 Deleted in Revision 11.
6.2-49 Deleted in Revision 8.
6.2-50 Deleted in Revision 8.
6.2-51 Deleted in Revision 23.
6.2-52 Deleted in Revision 23.
6.2-53 Containment Subcompartment Analyses Location Of Nodes Corresponding To The Subcompartment Model
6.2-54 Containment Subcompartment Analyses Portion Of Subcomparment Model From Elevation 91 To 117 Feet
6.2-55 Containment Subcompartment Analyses Portion Of Subcompartment Model From Elevation 117 To 140 Feet
DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 FIGURES (Continued)
Figure Title xx Revision 23 December 2016 6.2-56 Containment Subcompartment Analyses Portion Of Subcompartment Model Above Elevation 140 Feet
6.2-57 Containment Subcompartment Analyses Pressurizer Enclosure Model From Elevation 140 Feet To 176 Feet
6.2-58 Containment Subcompartment Analyses Pressurizer Enclosure Model From Elevation 147 Feet 9 Inches To 166 Feet
6.2-59 Containment Subcompartment Analyses Pressurizer Enclosure Model From Elevation 140 Feet To 147 Feet 9 Inches
6.2-60 Containment Subcompartment Analyses Steam Generator Enclosure Model From Elevation 140 Feet To 184 Feet
6.2-61 Containment Subcompartment Analyses Steam Generator Enclosure Model From Elevation 140 Feet To 151 Feet 11-1/2 Inches 6.2-62 Containment Subcompartment Analyses Steam Generator Enclosure Model From Elevation 151 Feet 11-1/2 Inches To 171 Feet 6.2-63 Containment Subcompartment Analyses Nodes And Flow Paths For Subcompartment Model
6.2-64 Containment Subcompartment Analyses Nodes And Flow Paths For Subcompartment Model
6.2-65 Containment Subcompartment Analyses Nodes And Flow Paths For Subcompartment Model
6.2-66 Containment Subcompartment Analyses Maximum Differential Pressure For Pressurizer Surge Line Break At Hot Leg Connection In Node 3
6.2-67 Containment Subcompartment Analyses Maximum Differential Pressure For Rhr Suction Line Break At Hot Leg Connection In Node 6
6.2-68 Containment Subcompartment Analyses Maximum Differential Pressure For Pressurizer Spray Line Break At The Top Of The Pressurizer In Node 38
6.2-69 Containment Subcompartment Analyses Maximum Differential Pressure For An Accumulator Line Break At The Tank Nozzle In Node 17 DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 FIGURES (Continued)
Figure Title xxi Revision 23 December 2016
6.2-70 Containment Subcompartment Analyses Maximum Differential Pressure For Main Feedwater Line Break At The Steam Generator Inlet Nozzle Node 36
6.2-71 Containment Subcompartment Analyses Maximum Differential Pressure Main Steam Line Break At The Containment Penetration Node 10
6.2-72 Containment Subcompartment Analyses Maximum Differential Pressure For Pressurizer Surge Line Break At Hot Leg Connection In Node 3
6.2-73 Containment Subcompartment Analyses Maximum Differential Pressure For Pressurizer Surge Line Break At Hot Leg Connection In Node 3
6.2-74 Containment Subcompartment Analyses Maximum Differential Pressure For Pressurizer Surge Line Break At Hot Leg Connection In Node 3
6.3-1 Residual Heat Removal Pump Performance Curves (Typical)
6.3-2 Centrifugal Charging Pumps 1 & 2 Performance Curves (Typical)
6.3-3 Safety Injection Pump Performance Curves (Typical)
6.3-4 Alignment of ECCS-related Components During Injection Phase of Emergency Core Cooling
6.3-5 Alignment of ECCS-related Components During Recirculation Phase of Emergency Core Cooling
6.3-6 (a) Arrangement of Containment Recirculation Sump Strainer (Unit 1)
6.3-7 (a) Arrangement of Containment Recirculation Sump Strainer (Unit 2)
6.5-1 (a) Auxiliary Feedwater System
6.5-2 (a) Long Term Cooling Water System
6.5-3 Auxiliary Feedwater Flow for Plant Shutdown
NOTE:
DCPP UNITS 1 &
2 FSAR UPDATE Chapter 6 FIGURES (Continued)
Figure Title xxii Revision 23 December 2016 (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 xxiii Revision 23 December 2016 Chapter 6 APPENDICES Appendix Title 6.2A Deleted in Revision 18.
6.2B Deleted in Revision 11.
6.2C Deleted in Revision 19.
6.2D REANALYSIS OF LONG-TERM LOSS-OF-COOLANT ACCIDENTS AND MAIN STEAMLINE BREAK EVENTS
6.3A Deleted in Revision 22.
DCPP UNITS 1 &
2 FSAR UPDATE 6.1-1 Revision 22 May 2015 Chapter 6 ENGINEERED SAFETY FEATURES 6.1 GENERAL
6.
1.1 INTRODUCTION
Engineered safety features (ESF) systems ar e provided to reduce the safety and radiological consequences of possible Diablo Canyon Power Plant (DCPP) accidents.
These systems cool the reactor core during a loss-of-coolant accident (LOCA) or main steam line break (MSLB), absorb energy released during accidents, contain solids, liquids, or gases released during accidents, and/or absorb radioactive materials that
could otherwise be released from the plant buildings. These systems are standby
systems, in that they are called upon to perform their ESF functions only in the event of unexpected severe plant accidents. They pr ovide protection beyond the systems and plant design features that are primarily intended for the prevention of accidents.
The principles and guidelines used in the design, construction, and operation of the ESF
systems described in Chapter 6 are specified in the individual sections of Chapter 6 and Table 6.1-1. Refer to Section 3.2 and Table 3.2-4 for a comparison of the PG&E Quality/Code classes to the recommendations of ANSI Standard N18.2, August 1970 Draft.
The methods used to evaluate ESF performance are primarily contained in this chapter and Chapter 15.
The DCPP Technical Specifications (Reference 1) establish limiting conditions for maintenance of ESF components. Maintenance of a particular component is permitted
if the remaining components meet the minimum requirements for operation and the
following conditions are also met:
(1) The remaining equipment has been demonstrated to be in operable condition.
(2) A suitable limit is placed on the total time required to complete maintenance to return the component to an operable condition.
ESF systems meet redundancy requirements, thus maintenance of active components
is possible during operation without impairment of the safety function. Routine servicing
and maintenance of equipment of this type that is not required more frequently than on
an outage basis would generally be scheduled for periods of refueling and maintenance outages. Any continued reactor operation du ring outages of individual ESF components will conform to reasonable, experienced judgment and industry practices, thus ensuring safe operation.
DCPP UNITS 1 &
2 FSAR UPDATE 6.1-2 Revision 22 May 2015 This chapter provides detailed descriptions of the DCPP ESFs and evaluates their
performance under postulated accident conditions. Specifically, information is provided
which shows that:
(1) The concept upon which the operation of each system is predicted has been proven sufficiently by experience, and/or by tests under simulated
accident conditions, and/or by conservative extrapolations from present
knowledge.
(2) The system will function during the period required and will accomplish its intended purpose.
(3) The system has been designed with adequate consideration of component and system reliability, component and system redundancy, and separation
of components and portions of systems.
(4) Provisions have been made for periodic tests, inspections, and surveillance to ensure that the sy stems will be dependable and effective when called upon to function.
6.1.2
SUMMARY
DESCRIPTION The ESFs provided at DCPP are the following:
(1) Containment Systems The steel-lined, reinforced concrete containment structure, including the
concrete cylindrical wall, base, and dome, is designed to prevent
significant release to the environs of radioactive materials that could be
released into the containment as a result of accidents inside the
containment (refer to Sections 6.2.1 and 6.2.4).
(2) The Emergency Core Cool ing System (ECCS)
The ECCS provides water to cool the core in the event of an accidental loss of primary reactor coolant water.
The ECCS also supplies dissolved boron into the cooling water to provide shutdown margin (refer to Section 6.3).
(3) The Containment Spray System (CSS)
The primary function of the CSS is to help limit the peak temperature and pressure in the containment in the event of a LOCA or MSLB (refer to Section 6.2.2). The CSS, in conjunction with the spray additive system (SAS), also helps to limit the offsite radiation levels following the DCPP UNITS 1 &
2 FSAR UPDATE 6.1-3 Revision 22 May 2015 postulated LOCA by removing airborne iodine from the containment atmosphere during the injection phase.
(4) The Containment Fan Cool er System (CFCS)
The CFCS functions in conjunction with the CSS to limit the temperature and pressure in the containment structure in the event of a LOCA or MSLB (refer to Section 6.2.2). The CFCS also provides mixing of the sprayed and unsprayed regions of the containment atmosphere to improve airborne fission product removal (refer to Section 6.2.3). The CFCS function of mixing the containment atmosphere for hydrogen control is discussed below.
(5) The Spray Additive System The SAS functions by adding sodium hydroxide, an effective iodine scrubbing solution, to the CSS water to reduce the content of iodine and other fission products in the containment atmosphere and prevent the re-evolution of the iodine in the recirculated core cooling solution following a LOCA (refer to Section 6.2.3).
(6) Containment Combustible Gas Control The long-term buildup of gaseous hydrogen in the containment following a LOCA is primarily controlled by ensuring a mixed containment atmosphere and providing equipment for monitoring hydrogen concentrations. The CFCS is the primary means credited for containment atmosphere mixing (refer to Section 6.2.5).
(7) The Fuel Handling Buildin g Ventilation System (FHBVS)
The FHBVS provides a significant reduction in the amounts of volatile radioactive materials that could be released to the atmosphere in the
event of a major fuel handling accident (refer to Section 9.4.4).
(8) The Auxiliary Building Ventilation System(ABVS)
The ABVS provides the capability for significant reductions in the amounts of volatile radioactive materials that could be released to the atmosphere
in the event of leakage from the residual heat removal (RHR) system recirculation loop following a LOCA (refer to Section 9.4.2).
(9) The Control Room Ve ntilation System(CRVS)
The CRVS permits continuous occupancy of the control room and technical support center (TSC) under design basis accidents by providing DCPP UNITS 1 &
2 FSAR UPDATE 6.1-4 Revision 22 May 2015 the capability to control infiltration of volatile radioactive material (refer to Sections 9.4.1 and 6.4.1).
(10) The Auxiliary Feedwater (AFW) System The AFW system supplies water to the secondary side of the steam generators for reactor decay heat removal, when the main feedwater system is unavailable (refer to Section 6.5).
Instrument air is used in most of the ESF systems. In some cases nitrogen is used.
Bottled air or nitrogen is provided, as required, via the backup air/nitrogen supply system. ESF devices are designed to maintain a safe position or move to a safe position on loss of air or nitrogen pressure. Thus, the air and nitrogen systems are not
needed to ensure device initial positioning or safe operation and are PG&E Design Class II. Detailed analyses of the compressed air and backup air/nitrogen supply systems, and their relation to PG&E Design Class I devices, are found in Section 9.3.1 (refer to Tables 3.9-9 and 6.2-39 for a listing of such devices).
All ESF remotely operated valves have position indication on the control board in two
places. Red and green indicator lights are located next to the manual control station, showing open and closed valve positions.
The ESF positions of these valves are displayed on the monitor light panels (four panels), which consist of an array of white
lights.
Three of the light panels are de-energized during normal operation; applicable portions of these are energized concurrent with a Phase A containment isolation, a Phase B containment isolation, a safety injection signal, a containment ventilation isolation, a steam generator high-high level, a main steam isolation, or a feedwater isolation. The remaining panel is always energized. The design of these arrays is such that the white
lights will be dark when the valves are in their normal or required positions for power
operation, or their correct position after automatic actuation. These light panels can be
tested during normal operation with switches on the control panel. These monitor lights
thus enable the operator to quickly assess th e status of the ESF systems. These indications are derived from contacts integral to the valve operators. In the case of the
accumulator isolation valves, redundancy of position indication is provided by valve
stem-mounted limit switches (the stem-mounted switches are independent of the limit
switches in the motor operators), which actu ate an annunciator on the control board when the valves are not correctly positioned. Refer to Section 7.6 for additional
information.
Pump motor power feed breakers indicate that they have closed by energizing
indicating lights on the control board in order to enable additional monitoring of in-
containment conditions in the post-LOCA recovery period.
DCPP UNITS 1 &
2 FSAR UPDATE 6.1-5 Revision 22 May 2015 6.
1.3 REFERENCES
- 1. 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 6.2-1 Revision 23 December 2016 6.2 CONTAINMENT SYSTEMS Containment systems enclose the reactor and most plant systems and equipment that
operate at high temperatures and pressures and may contain radioactive materials.
This section describes and evaluates the design of the containment systems and confirms their capability to fulfill their intended objectives.
6.2.1 CONTAINMENT FUNCTIONAL DESIGN The containment structure and subcompartments are designed to sustain the resulting
pressures and temperatures from gross failure up to and including a loss-of-coolant
accident (LOCA). Long-term mass and energy releases and containment integrity from
a LOCA or main steam line break (MSLB) are analyzed in Appendix 6.2D for the
evaluation of the resulting peak containment pressure. Short-term mass and energy releases and subcompartment integrity analyses are addressed in the following
sections.
6.2.1.1 Design Bases
6.2.1.1.1 General Design Criterion 4, 1987
- Environmental and Dynamic Effects Design Bases The containment is designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including LOCAs. The dynamic effects associated with
postulated reactor coolant system (RCS) primary loop pipe ruptures are excluded from the Diablo Canyon Power Plant (DCPP) design basis for subcompartment analysis.
6.2.1.1.2 General Design Criterion 10, 1967 - Containment
The containment is designed to sustain the initial effects of gross equipment failures, such as a LOCA, without loss of required integrity and, together with other engineered safety features (ESFs), to retain the functional capability of the containment to protect the public.
6.2.1.1.3 General Design Criterion 49, 1967 - Containment Design Basis
The containment is designed to accommodate the pressures and temperatures resulting from the largest credible energy release follo wing a LOCA without exceeding the design leakage rate.
6.2.1.1.4 General Design Criterion 54, 1967
- Containment Leakage Rate Testing The containment is designed with the capability for integrated leakage rate testing to be conducted at design pressure after installation of all penetrations to verify its conformance with required performance.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-2 Revision 23 December 2016 6.2.1.1.5 General Design Criterion 55, 1967 - Containment Periodic Leakage Rate Testing The containment is designed so that integrated leakage rate testing can be done at the design pressure periodically during the plants lifetime.
6.2.1.1.6 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment The containment is designed as a barrier to maintain control over plant radioactive effluents, whether gaseous, liquid, or solid meeting the radiological limits of 10 CFR Part 100. Appropriate holdup capacity is provided for retention of gaseous effluents, particularly where unfavorable environmental conditions can be expected to require operational limitations upon the release of radioactive effluents to the environment.
6.2.1.1.7 10 CFR Part 50, Appendix J, Option B - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors The containment is designed to allow for conductance of a performance-based containment leakage rate testing program for Type A containment integrated leak rate
tests (ILRT) and Type B testing for the air lock door seals.
6.2.1.1.8 10 CFR Part 50, Appendix K, Part I.A - ECCS Evaluation Models, Sources of Heat during the LOCA The containment is designed to accommodate the largest credible energy release following a postulated pipe break taking into account the heat sources listed in
Paragraph I.A of 10 CFR Part 50, Appendix K.
6.2.1.1.9 Regulatory Guide 1.163, September 1995 - Performance-Based Containment Leak-Test Program The containment is designed to allow the use of a performance-based leak-test program, including the leakage-rate test methods, procedures, and analyses as
required by Regulatory Guide 1.163, September 1995.
6.2.1.2 Containment Subcompartment Analysis Short-term mass and energy (M&E) release calculat ions from primary (i.e., loss-of-coolant accidents or LOCAs) and secondary side (non-LOCAs) piping are performed to support pressurization analyses for subcompartments inside the containment including the reactor cavity region, the loop compartments, the pressurizer enclosure and the steam generator enclosure. These analyses are performed to ensure that the structural elements in the immediate proximity of the break location can maintain their structural DCPP UNITS 1 &
2 FSAR UPDATE 6.2-3 Revision 23 December 2016 integrity during the short pressure pulse (generally less than 3 seconds) that accompanies a high-energy pipe break within the region.
There are two reasons for the specified duration. The first is that the initial pulse that occurs upon the opening of the break is typically the most limiting for a subcompartment. The second is that the global containment pressure will approach an individual subcompartment pressure rather quickly and the goal of the subcompartment analysis is to determine the peak differential pressure across walls, supports, components, and structures. When compon ents are being considered, the peak differential pressure is commonly used to determine an asymmetric load. The peak differential pressure can often occur within milliseconds of the break opening time.
The double-ended guillotine and s plit breaks that have been postulated in primary coolant loop piping in the loop compartments, the reactor cavity region, the reactor vessel annulus, the pipe annulus, and for the steam generator analysis and that are currently presented in Reference 62 are no longer necessary for Diablo Canyon Unit 1 and Unit 2 due to the approval (Reference 61) and subsequent application of leak-before-break (LBB) analysis and monitoring. The only break that remains from the original design basis subcompartment analyses is the double-ended spray line break inside the pressurizer enclosure. Therefore, the short-term M&E scope only includes generating LOCA M&E releases for large branch line breaks (refer to Section 6.2.1.2.1) and non-LOCA M&E releases for double-ended breaks of the secondary side (refer to Section 6.2.1.2.3).
6.2.1.2.1 Short-Term LOCA Mass and Energy Releases A reactor coolant system (RCS) model that bounds bot h Diablo Canyon Unit 1 and Unit 2 was developed. All M&E releases were generated with the SATAN-V code (Reference 59) that has been approved for use in generating short-term LOCA M&E releases.
The piping analysis for the large branch lines was used to determine the break locations that required analysis as part of the subcompartment response. All breaks were conservatively analyzed as full double-ended guill otine breaks. The short-term LOCA M&E releases are calculated with the SATAN-V computer program documented in Reference 59.
6.2.1.2.1.1 Input Parameters and Assumptions The release of M&E created by a postulated break in the RCS piping are principally a function of the calculated break area, system pressure, and RCS fluid density, which is a function of the operating conditions of temperature and pressure. The LOCA short-term M&E are relatively insensitive to initial RCS conditions of flow and power except as they relate to coolant density, where higher coolant density and system pressure lead to increased LOCA M&E releases. Lower flow rates and higher power lead to lower cold leg temperatures (and higher coolant density) for the same average coolant DCPP UNITS 1 &
2 FSAR UPDATE 6.2-4 Revision 23 December 2016 temperature. Therefore, the limiting LOCA M&E are calculated for conditions consistent with a minimum thermal design flow at norm al operating conditions at full power. RCS temperatures are nominal values reduced by the known instrument uncertainty in order to assure the lowest operating temperatures.
To conservatively calculate LOCA short-term M&E for Diablo Canyon Units 1 and 2, the following operating conditions were considered in establishing the limiting temperature and pressures:
Unit 1 and Unit 2 Bounding Conditions
- 1. Initial RCS conditions associated with a minimum thermal design flow of 87,700 gpm per loop. 2. Reactor core power of 3411 MWt. 3. A nominal RCS hot full power (HFP) vessel T AVG of 565.0°F. This provides lowest possible RCS TCOLD of 531.7°F and THOT of 598.1°F minus a temperature uncertainty of 4.3°F 4. A nominal RCS pressure of 2250 psia plus a pressure uncertainty of 42 psia.
Thus, the LOCA M&E releases were generated with a minimum TCOLD of 527.4°F, a minimum T HOT of 593.8°F, and a pressurizer pressure of 2292 psia, including uncertainties, in order to bound the operation of both Unit 1 and Unit 2.
6.2.1.2.1.2 Description of Analyses and Evaluations The M&E releases from the postulated break must be determined prior to evaluating the containment subcompartment responses. The SATAN-V computer code is used to calculate short-term blowdown transients. This code uses a control volume approach to model the behavior of the piping system, includes critical flow calculation and applies an implicit representation of the depressurization.
Due to the short-term length of the transient, the break flow remains subcooled throughout the 3 second transient and thus only the Zaloudek correlation is relied upon in the calculation. The LOCA short-term M&E for the Diablo Canyon Units 1 and 2 used the code and methodology described in Reference 59 and the operating parameters described above. The large branch lines that are considered are:
- a. Pressurizer Surge Line b. Pressurizer Spray Line c. RHR Suction Line d. Accumulator Lines Furthermore, some of these pipes can be postulated to break in more than one location.
Therefore, the locations that are considered are:
- 1. Pressurizer Surge Line Break at the Hot Leg Connection
- 2. Pressurizer Surge Line Break at the Pressurizer Nozzle DCPP UNITS 1 &
2 FSAR UPDATE 6.2-5 Revision 23 December 2016
- 3. Pressurizer Spray Line Break at the Top of the Pressurizer
- 4. Pressurizer Spray Line Break at the RCS Cold Leg Connection
- 5. RHR Suction Line Break at the Hot Leg Nozzle
- 6. Accumulator Line Break at the RCS Cold Leg Connection
- 7. Accumulator Line Break at the Tank Tables 6.2-57 through 6.2-60 document the time histories of the limiting LOCA short-term M&E releases.
Discussion of Breaks Analyzed Pressurizer Surge Line Break at the Hot Leg Connection The Unit 1 pressurizer surge line is 14 inch outside diameter (OD), Schedule 140, and the Unit 2 surge line is 14 inch OD, Schedule 160. The pipe schedule for the Unit 1 surge line produces a larger break area for the same nominal pipe size and thus a larger M&E release for Unit 1. Therefore, the Unit 1 configuration is bounding and is used in the analysis. The pressurizer surge line break is the largest break to be analyzed among the breaks that could occur when LBB analysis and monitoring is applied to the primary loop piping and thus is expected to generate the largest differential pressures in the loop compartment, across the primary shield wall, and across the crane wall. The pressurization resulting from this break could be used to determine asymmetric loads on components within this region. This break is modeled as a hole in the RCS hot leg on one side and the surge line piping on the opposite side that is connected to the pressurizer tank. The volume and hydraulic resistance of the surge line are modeled. Table 6.2-57 provides the LOCA M&E used in the subcompartment pressurization analysis.
Pressurizer Surge Line Break at the Pressurizer Nozzle This break is at the bottom of the pressurizer tank at an approximate elevation of 113 feet. This elevation puts this break in the lower region of the containment very near the elevation for the surge line connection at the RCS hot leg and thus would affect the same structures and components. This break differs from surge line break at the RCS hot leg in the placement of the surge line v olume and in piping hydraulic resistance. In the pressurizer surge line break at the pressurizer tank, the surge line remains attached to the RCS hot leg and provides an increased resistance to break flow from the RCS.
Thus, the mass and energy release is reduced compared to a surge line break at the RCS hot leg connection and, therefore, is not limiting.
Pressurizer Spray Line Break at the Top of the Pressurizer The pressurizer spray line is inside the pressurizer enclosure and a break in the pressurizer spray line will result in pressurizing the enclosure. Thus, this case is analyzed to assure that the enclosure design limit is not exceeded. These results may also be incorporated in the determination of the asymmetric loads on the pressurizer.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-6 Revision 23 December 2016 The spray lines in both Diablo Canyon Unit 1 and Unit 2 are made of 4 inch Schedule 120 piping. The spray line break at the top of the pressurizer has less M&E release than a spray line break at the RCS cold leg. This is due to having steam at the top of the pressurizer and the flow resistance of the portion of piping still attached to the cold leg that limits the M&E release rate. However, due to the small volume at the top of the pressurizer enclosure, to accommodate the spray line break release, this analysis is deemed necessary to assure the design criteria are met. Table 6.2-58 provides the LOCA M&E used in the pressurizer enclosure subcompartment analysis.
Pressurizer Spray Line Break at the RCS Cold Leg Connection This break is located below the pressurizer enclosure at the RCS cold leg elevation and thus has a negligible effect in pressurization of the pressurizer enclosure. The spray line piping at this elevation is 4 inch Schedule 120 and the break is modeled as having occurred at the cold leg connection. Thus, one side of the break is modeled as a hole with a diameter equal to 4 inch Schedule 120 pipe at the RCS cold leg and the other side is the pressurizer spray line attached to the pressurizer. This break occurs in the same compartment as the much larger surge line break, and is therefore bounded by the surge line M&E release.
RHR Suction Line Break at the RCS Hot Leg Connection This is the line used to cool the RCS continually starting from mode 4 (T < 350°F) until the RCS temperature is again returned to above 350°F, usually during a shut down or refueling. The RHR suction piping is 14 inch Schedule 160 pipe which is valved-out during power operation. Thus, the break is modeled as a single-ended break at the RCS hot leg. The M&E release from the isolated RHR high pressure piping up to the isolation valve is calculated as a steady release rate based on the break flow correlation using the initial pressure and temperature as the input condition. The volume of piping between the RCS hot leg connection and the RHR suction isolation valve is less than 10 ft 3. The blowdown of this portion of the piping section would occur in less than 0.05 seconds. Table 6.2-59 provides the LOCA M&E used in the subcompartment pressurization analysis.
Accumulator Line Break at the RCS Cold Leg Connection This break is located between the accumulator injection line check valve closest to the RCS and the RCS cold leg connection.
Thus, both the accumulator tank and the RCS blow down through the double-ended break. The accumulator lines at the cold leg connection are 10 inch Schedule 140 piping and thus the break size is based on this piping diameter and schedule. The choice of accumulator injection lines among the four accumulators does not affect the results. The transient is limited to 3 seconds in duration and the injection line and the RCS cold leg do not experience flashing and thus the break flow remains subcooled liquid. Flas hing does not occur internal to the piping since the pressure remains above the saturation pressure for the 3 second period.
Since the flow is always critically li mited, the small differences in accumulator injection DCPP UNITS 1 &
2 FSAR UPDATE 6.2-7 Revision 23 December 2016 line piping lengths and hydraulic resistances do not affect the results. Table 6.2-60 presents the LOCA M&E used in the subcompartment pressurization analysis.
Accumulator Nozzle Break at the Tank Separate LOCA M&E releases were not generated for a break at the accumulator tank nozzle since this would result in a blowdown of only the accumulator tank. The RCS side of the accumulator line at this connection is protected from blowing down by two check valves in series. However, for the purposes of subcompartment pressurization response, the LOCA M&E at these locations were conservatively analyzed with the same LOCA M&E at the RCS cold leg connection. Subcompartment pressures were computed for these break cases.
6.2.1.2.1.3 LOCA Mass and Energy Releases Conclusion LBB analysis and monitoring as applied to the primary loop piping at Diablo Canyon Unit 1 and Unit 2 has been used to eliminate primary loop piping, leaving select primary side branch line breaks as limiting pipe break events. Branch lines analyzed were the pressurizer surge line at two locations, pressurizer spray line at two locations, RHR suction line and the most limiting of the four accumulator injection lines. The location of the break within any analyzed piping run w as chosen to maximize the release of M&E.
Consideration was given to the location of check valves, isolation valves, hydraulic resistance, or other restrictions to the release of M&E in selecting the most conservative break locations. The limiting M&E releases resulting from these analyses are provided in Tables 6.2-57 through 6.2-60 and are used in the analysis of the post-LOCA pressurization of the containment subcompartments.
6.2.1.2.2 LOCA Subcompartment Pressurization A conservative representation of the Diablo Canyon Unit 1 and Unit 2 containment was nodalized in order to determine the pressurization in certain regions that would result from a localized pipe rupture. A plant specific model with 38 nodes was created that bounds both units. Table 6.2-61 summarizes the nodalization based on subcompartment boundaries within the containment. Structural elements (walls, ceilings, and floors) and components were used to define the boundaries for the subcompartments. Table 6.2-62 provides conservative minimum volumes for each of the 38 nodes in the network. Table 6.2-63 provides individual flow path characteristics for the nodal network that conservatively minimizes the flow from the upstream node to the downstream node. The minimum volume of each node, combined with the flow path characteristics that minimize communication between the nodes, create a subcompartment network that will conservatively maximize the differential pressure between the nodes. The spatial presentation of the 38 nodes by elevation is provided in Figures 6.2-53 through Figure 6.2-62. The TMD (Reference 60) nodal flow path connections are shown schematically in Figure 6.2-63, Figure 6.2-64, and Figure 6.2-
- 65.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-8 Revision 23 December 2016 The LOCA subcompartment analysis described below and in Sections 6.2.1.2.2.1, 6.2.1.2.2.2, and 6.2.1.2.2.3 was performed to determine subcompartment pressures and resulting differential pressures on various structural elements within containment.
During the early stages of a LOCA, pressure differentials may be briefly established in
the containment. While the geometry of the containment, except for the net free volume, has no direct effect upon the containment peak pressure, indirect
considerations such as the design of structural supports of ESF equipment and the
prevention of missile generation make it desirable to calculate the differential pressure
transients caused by different breaks.
Application of LBB has determined that breaks in the RCS branch lines must still be considered for the design of containment structural elements. RCS branch lines that are postulated to break resulting in a LOCA are 1) the pressurizer surge line, 2) the pressurizer spray line, 3) the RHR suction line and 4) the accumulator injection lines.
There are other smaller branch lines connected to the reactor coolant system that could result in a LOCA but these smaller branch lines are bounded by consideration of the four branch lines identified above. There are various locations within containment that need to be considered for these breaks. For instance, the pressurizer surge line can break at the connection to the RCS hot leg or at the bottom of the pressurizer tank.
Therefore, both locations are considered. In addition, due to the application of LBB, the double-ended ruptures of secondary side piping could become limiting and also need to be considered.
The compartment responses with the 38 node model to the M&E releases from the four branch lines above were analyzed using the Transient Mass Distribution (TMD) (Reference 60) computer code with an unaugmented homogeneous critical mass flowrate correlation. The analyses using the TMD code (Reference 60), with the compressibility factor and non-augmented critical flow correlation, were performed to
determine the response of the postulated ruptures. The nodalization sensitivity study indicates that the model selected is adequate to determine the asymmetrical
pressurization forces that may act upon a vessel. The NRC evaluated the TMD code (Reference 60) and found the methods acceptable provided that non-augmented critical
flow relationships are used rather than augmented.
The TMD mathematical model used to calculate the flows and pressures throughout the
containment is based upon time-dependent equations of conservation of mass, conservation of energy, conservation of momentum, and state. Flow inertia effects
between the volumes are also calculated. The model calculates critical flow conditions
for application under high-pressure differentials. A 100 percent entrainment of the water
emerging from the break is assumed. Subcompartment vent discharge flows are
considered as unrecoverable pressure losses and, consequently, vent discharge
coefficients are not used. A vent discharge coefficient of unity is conservatively assumed when vent flow is critical.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-9 Revision 23 December 2016 Calculated values of peak differential pressure and the absolute pressure at the time of peak differential pressure are tabulated in Table 6.2-64 for compartments within the containment.
6.2.1.2.2.1 Loop Compartment Analysis The loop compartment is the region outside the reactor cavity wall but inside the crane wall from elevation 91 feet up to elevation 140 feet. This region is defined by 6 nodes with boundaries that bisect each of the four steam generators as seen in Figures 6.2-53 through Figure 6.2-55.
If a postulated LOCA occurs in the loop compartment region, the steam mass that enters this space must be vented to the rest of the containment. The flowpaths potentially available for such venting are through the operating deck at elevation 140, through the crane wall, and into the reactor vessel annulus, as well as to the adjacent loop compartments. The first of these routes permits steam, air, and water to directly enter the upper dome of containment, the second grants access to the annular spaces between the crane wall and the containment shell, and the third allows entry to the space surrounding the reactor vessel. The M&E release rates used in the analysis are given, as a function of time after the postulated break (refer to Section 6.2.1.2.1.2).
The limiting break in the loop compartments is a break of the pressurizer surge line at the hot leg connection (refer to Figures 6.2-72, 6.2-66 and 6.2-74). Additional breaks analyzed, but not limiting for the loop compartments, are a break in the accumulator injection line (refer to Figure 6.2-69) at the reactor coolant system cold leg connection and a break in the RHR suction line at the reactor coolant system hot leg (refer to Figure 6.2-67).
The limiting pressure differential acting across the crane wall occurred for a break in the pressurizer surge line located in TMD node 3. The limiting pressure differential acting across the 140 elevation operating deck and the loop compartment floor above the lower reactor cavity also occurred due to a break in the pressurizer surge line in TMD node 3. The peak differential pressures, with the absolute pressures at the time of peak differential pressure, are included in Table 6.2-64.
As shown in Table 6.2-64, the peak differential pressure calculated for each structural element (defining the bounds of the subcompartments) in the loop compartment region is less than the design value.
6.2.1.2.2.2 Pressurizer Enclosure Analysis The pressurizer enclosure is shown in Figures 6.2-56 through Figure 6.2-59. The enclosure begins at the 140 foot elevation and extends upward to elevation 176 feet.
The analysis assumed a double-ended rupture of the pressurizer spray line, since this
would result in the maximum differential pressure across the pressurizer enclosure DCPP UNITS 1 &
2 FSAR UPDATE 6.2-10 Revision 23 December 2016 walls. In the event of such a rupture, fluid from the break would be vented to both the containment dome and the loop compartment regions. The M&E release rates used in the analysis are given, as a function of time after the postulated break (refer to Section 6.2.1.2.1.2).
The limiting break for the pressurizer enclosure is a break in the pressurizer spray line at the top of the pressurizer in TMD node 38 (refer to Figure 6.2-68).The peak differential pressure, with the absolute pressure at the time of peak differential pressure, is included in Table 6.2-64.
As shown in Table 6.2-64, the peak differential pressure calculated with the TMD code for the pressurizer enclosure is less than the design value.
6.2.1.2.2.3 Reactor Cavity Analysis The reactor cavity region is represented by three nodes in the subcompartment model (refer to Figures 6.2-53 and 6.2-54), with two nodes surrounding the cylindrical portion of the reactor vessel and one node in the lower reactor cavity below the reactor vessel.
The reactor cavity region is defined from elevation 61.5 feet up to approximately elevation 114 feet.
The bounding break for the reactor cavity analysis assumes a break of the pressurizer surge line at the hot leg connection (refer to Figure 6.2-73).This break will pressurize the loop compartment region of containment and steam will flow into the reactor cavity through the area around the hot leg and cold leg nozzles where the loop piping penetrates the primary shield wall. The M&E release rates used in the analysis are given as a function of time after the postulated break (refer to Section 6.2.1.2.1.2).
The peak differential pressure across the primary shield wall occurred for the above break in TMD node 3. The peak differential pressures, with the absolute pressures at the time of peak differential pressure, are included in Table 6.2-64.
As shown in Table 6.2-64, the peak differential pressure calculated with the TMD code for the primary shield wall is less than the design value.
6.2.1.2.2.4 LOCA Subcompartment Pressurization Conclusions The LOCA breaks were reviewed and selections were made to analyze the most limiting break locations. The most limiting break location for the LOCA analysis is a break in the pressurizer surge line at the RCS hot leg connection, which yielded a result of 5.90 psi pressure differential between a loop compartment and the annulus region between nodes 3 and 12. Table 6.2-64 provides a summary of the peak differential pressure results for the LOCA transients.
6.2.1.2.3 Short-Term Non-LOCA Mass and Energy Releases High-energy lines inside containment includ e not only primary-side piping but also DCPP UNITS 1 &
2 FSAR UPDATE 6.2-11 Revision 23 December 2016 secondary-side main steam lines and main feedwater lines. Postulated breaks in these secondary-side lines are commonly referred to as non-LOCA breaks. The short-term M&E releases from these non-LOCA breaks are included in the containment subcompartment pressurization analyses, which are performed to ensure that the structural elements can maintain their structural integrity (refer to Section 6.2.1.2).
6.2.1.2.3.1 Input Parameters To conservatively calculate the non-LOCA short-term M&E release rates for Diablo Canyon Unit 1 and Unit 2, the following data was considered in establishing the limiting M&E releases.
Unit 1 and Unit 2 Bounding Conditions
- Initial RCS temperature of 547°F (no-load)
- Steam pressure of 1020 psia at no-load saturated temperature conditions
- Maximum main feedwater temperature of 435.0°F 6.2.1.2.3.2 Description of Analyses and Evaluations The non-LOCA short-term M&E releases from the postulated secondary-side breaks must be determined prior to evaluating the containment subcompartment response.
The Moody critical flow correlation is used to determine short-term blowdown M&E releases using the calculation technique do cumented in Appendix E of ANSI/ANS-58.2-1980 (Reference 63). With no consideration for steam line and feedwater line isolation due to the short duration of the M&E release s, the break flowrates from the forward and reverse directions are determined. The break flows are held constant with a step change in flow when a segment of piping is determined to have been emptied.
Calculations of the non-LOCA short-term M&E releases do not credit protection system actuations or details such as system depressurization. All non-LOCA short-term M&E releases are generated using the Moody critical mass flux correlation (Reference 64).
Therefore, the M&E releases represent a conservative upper-bound value for use in short-term pressurization calculations.
Breaks analyzed were full double-ended guillotine breaks of the main steam line at the outlet nozzle of the steam generator, the main steam line at the containment penetration, the main feedwater line connection at the inlet to the steam generator, and the main feedwater line at the containment penetration. Breaks postulated in these pipes bound other secondary-side breaks because of the break area and the system pressure.
Discussion of Breaks Analyzed Main Steam Line Break Inside the Steam Generator Enclosure
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-12 Revision 23 December 2016 This break is modeled at the outlet nozzle of the steam generator assuming no-load power conditions to maximize the pressure in the main steam system. The total break area for a double-ended steam line break is not equal to a double-ended severance of the main steam line because one end of the break is limited via the integral flow restrictor in the discharge nozzle of the faulted steam generator.
This break releases directly into the upper dome of containment, which causes negligible differential pressures, and is bounded by other break cases.
Main Feedwater Line Break Inside the Steam Generator Enclosure This break is modeled at the inlet to the steam generator assuming no-load power conditions in the steam generator to maximize the pressure in the steam generator; and assuming full-power conditions in the main feedwater system to maximize the enthalpy of the feedwater fluid. The main feedwater temperature is assumed at the high end of the full-power feedwater temperature window to maximize the saturation pressure in the main feedwater system. The break area for a double-ended main feedwater line break is equal to a double-ended severance of the main feedwater line.
The main feedwater line break short-term M&E release data is presented in Table 6.2-66 and was used in the subcompartment pressurization analysis.
Main Steam Line Break Outside the Crane Wall This break is modeled at the containment penetration assuming no-load power conditions to maximize the pressure in the main steam system. The initial break area for a double-ended main steam line break is equal to a double-ended severance of the main steam line since the break location is downstream of the discharge nozzle of the steam generator. Following the initial blowd own of the piping between the break location and the steam generator, the break is limited via the integral flow restrictor at the discharge nozzle of the faulted steam generator. Since this break location is a significant distance from the faulted steam generator, the timing of the release rates considers the length of main steam line piping between the break and the steam generator.
The main steam line break short-term M&E release data is presented in Table 6.2-65 and was used in the subcompartment pressurization analysis.
Main Feedwater Line Break Outside the Crane Wall This break is modeled at the containment penetration assuming no-load power conditions in the steam generator to maximize the pressure in the steam generator; and assuming full-power conditions in the main feedwater system to maximize the enthalpy of the feedwater fluid. The main feedwater temperature is assumed at the high end of the full-power feedwater temperature window to maximize the saturation pressure in the main feedwater system. The break area for a double-ended main feedwater line break DCPP UNITS 1 &
2 FSAR UPDATE 6.2-13 Revision 23 December 2016 is equal to a double-ended severance of the main feedwater line. Since the break location is a significant distance from the faulted steam generator, the timing of the release rates considers the length of feedwater line piping between the break and the steam generator.
This postulated break is situated next to the main steam line break outside the crane wall. The main steam line break produces a larger M&E release. Therefore, the main feedwater line break is bounded by the main steam line break at this location.
6.2.1.2.3.3 Non-LOCA Mass and Energy Releases Conclusion The secondary-side breaks analyzed are the double-en ded guillotine break of the main steam line and the double-ended guillotine b reak of the main feedwater line at the steam generator inside the steam generator enclosure and at the containment penetration outside the crane wall. Consideration was given to the location of check valves in the main feedwater system to determine the piping volume that would be involved with a particular break location. The M&E releases that are generated for these break locations are provided in Tables 6.2-65 and Table 6.2-66 and are used in the analysis of the subcompartment pressurization.
6.2.1.2.4 Non-LOCA Subcompartment Pressurization The following containment subcompartments were analyzed for the non-LOCA breaks.
The non-LOCA breaks were analyzed using the Transient Mass Distribution (TMD) (Reference 60) computer code. The noding displayed in Figures 6.2-53 through 6.2-62 provide the noding of the containment subcompartments and are referenced for identification of the break location.
Main Steam Line Break Inside the Steam Generator Enclosure This break is at the top of the steam generator (steam outlet nozzle, refer to Figure 6.2-53). The M&E releases enter the upper part of the containment dome (node 7) and would not be expected to result in a very significant pressurization or differential pressure that would provide a load on a wall, component, or support. The maximum differential pressure for a double-ended main steam line break in node 7 is 0.52 psi occurring at 0.302 second (3/10th of a second). This break is bounded by the Main Feedwater line break inside the steam generator enclosure.
Main Feedwater Line Break Inside the Steam Generator Enclosure This break at the side of the steam generator (node
- 36) is at the main feedwater nozzle (refer to Figure 6.2-53, Figure 6.2-60, Figure 6.2-62, and Figure 6.2-65). The M&E releases enter this small volume and result in a pressurization or differential pressure that would provide a load on the steam generator shield wall, and the steam generator enclosure walls. The M&E release rates used in the analysis are given as a function of time after the postulated break (refer to Section 6.2.1.2.3).
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-14 Revision 23 December 2016 The peak differential pressure for a double-ended main feedwater line break occurs in node 36 (refer to Figure 6.2-70). The peak differential pressure with the absolute pressure at the time of peak differential pressure is included in Table 6.2-67.
As shown in Table 6.2-67, the peak differential pressure calculated with the TMD code for the steam generator shield wall is less than the design value.
Main Steam Line Break Outside the Crane Wall This break in the main steam line is at the containment penetration (refer to Figure 6.2-53 and Figure 6.2-55). The M&E releases enter the annulus region between the containment shell and the crane wall.
The resulting pressurization would produce an inward localized load on the crane wall or an outward localized load on the containment shell. The M&E release rates used in the analysis are given as a function of time after the postulated break (refer to Section 6.2.1.2.3).
The peak differential pressure for a double-ended main steam line break occurs in node10 (refer to Figure 6.2-71). The peak differential pressure with the absolute pressure at the time of peak differential pressure is included in Table 6.2-67.
As shown in Table 6.2-67, the peak differential pressure calculated with the TMD code for the crane wall is less than the design value.
Main Feedwater Line Break Outside the Crane Wall This break in the main feedwater line is at the containment penetration (refer to Figure 6.2-53 and Figure 6.2-55). The M&E releases enter the annulus region between the containment shell and the crane wall.
The resulting pressurization would produce an inward localized load on the crane wall or an outward localized load on the containment shell. The maximum differential pressure for a double-ended main feedwater line break in node 10 is 2.33 psi occurring at 0.033 second (1/30th of a second). This break is bounded by the Main Steam line break outside the crane wall.
6.2.1.2.4.1 Non-LOCA Subcompartment Pressurization Conclusions The non-LOCA breaks were reviewed and selections were made to analyze the limiting main steam line break and main feedwater line break locations. The most limiting non-LOCA break is a main feedwater line break at the inlet connection to the side of the steam generator in a localized space defined as node 36. The maximum differential pressure between the break node and the surrounding nodes in the containment model is 19.54 psi occurring instantaneously with the initiation of the break. Table 6.2-67 provides a summary of the peak differential pressure results for the non-LOCA transients.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-15 Revision 23 December 2016 6.2.1.2.5 Asymmetric Subcompartment Pressurization Loads Asymmetric subcompartment pressurization loads, were developed for the reactor coolant loop (RCL) piping system analysis. LOCA and non-LOCA postulated breaks were reviewed and selections were made to analyze the most limiting break locations as described in Section 6.2.1.2, 6.2.1.2.2, and 6.2.1.2.4.
LOCA and non-LOCA postulated pipe breaks include breaks in the pressurizer surge, pressurizer spray, RHR suction, accumulator injection, main steam and main feedwater lines. For each line, breaks were postulated at various locations resulting in multiple break cases.
From these localized pipe breaks, time-history pressures were developed and maximum differential pressures were established for the different subcompartments within containment. These pressure differentials will potentially cause asymmetric pressure loads across the components or structures contained in between the compartments.
The force acting on a component from the asymmetric pressure can be calculated as the pressure differential multiplied by the equivalent area of the component.
Numerous break cases were evaluated to determine the various pressure differentials across compartments. In order to develop a limiting time-history pressure differential function; typically time-history pressure differentials were enveloped creating a composite time-history pressure differential function for the different breaks. These composite functions were used in the dynamic asymmetric pressurization analyses.
Reactor Coolant Loop (RCL) Piping System The reactor coolant loop piping, reactor coolant pump, and lower portions of the steam generator and pressurizer components are all contained within the lower containment compartments, nodes 1 through 6 (Figures 6.2-53 and 6.2-54).
For the most limiting breaks in the lower containment compartments, the maximum pressure differentials were enveloped for the pressurizer surge, RHR suction, and accumulator injection break cases.
The reactor pressure vessel is contained within nodes 31, 32, and 33 (Figure 6.2-53).
For the most limiting breaks outside the reactor cavity region, the maximum pressure differentials in the horizontal (nodes 31-32) direction and in the vertical direction (nodes 33-7) were enveloped for the pressurizer surge, RHR suction, and accumulator injection break cases.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-16 Revision 23 December 2016 Steam Generator The lower portions of the steam generator are contained within the lower containment compartment nodes 1 through 6 (Figures 6.2-53 and 6.2-54) The upper portions of the steam generator are contained within nodes 34, 35, 36, 37, and 7 (Figures 6.2-53, 6.2-56, 6.2-60, 6.2-61 and 6.2-62).
The feedwater nozzle break at the steam generator results in larger horizontal asymmetric pressurization loads than the other postulated breaks. The development of time-history forcing functions, including the locations and directions of those forcing functions were developed to represent the asymmetric pressurization load and applied in the feedwater line pipe break analysis.
To evaluate the vertical asymmetric pressurization effect on the steam generator, the maximum differential pressures between the lower containment compartment nodes (nodes 1 through 6) were compared with the upper containment compartment node (node 7) for the pressurizer surge, RHR suction, and the accumulator injection break cases. The most limiting break cases for the pressurizer surge, RHR suction, and accumulator injection breaks occurring in the lower containment region were used to determine the asymmetric pressurization effects on the upper portion of the steam generator, with the exception for the feedwater break case, which was evaluated in a separate analysis for evaluating the feedwater break as described above.
Pressurizer The pressurizer is contained within the lower containment compartment, node 3 (Figure 6.2-53 and 6.2-54), and the pressurizer compartment, nodes 19 through 30 and node 38 (Figure 6.2-56, 6.2-57, 6.2-58 and 6.2-59).
To evaluate the vertical asymmetric pressurization effect on the pressurizer, the maximum differential pressure between the lower containment compartment, node 3, was compared with the pressurizer upper compartment, node 38. The vertical upward asymmetric force on the pressurizer was determined from the maximum pressure differential from a surge line break. The vertical downward asymmetric force on the pressurizer was determined from the maximum pressure differential from a spray line break. To evaluate the horizontal asymmetric pressurization effect on the pressurizer, horizontal forces were developed for breaks in the pressurizer surge and spray lines.
The horizontal and vertical forces are applied to the pressurizer analysis to represent the asymmetric pressurization effects.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-17 Revision 23 December 2016 Dynamic Asymmetric Pressurization Analyses A RCL piping system analysis was performed to evaluate the dynamic time-history asymmetric subcompartment pressurization effects from breaks in the pressurizer surge, RHR suction, and accumulator injection lines in the lower containment region. A dynamic time-history analysis was selected to evaluate the asymmetric pressure effects as opposed to a static analysis so that the results from the asymmetric pressure analysis could be combined on a time history basis with the results from the LOCA time history analysis. The dynamic time-history asymmetric pressurization analysis was performed using the WESTDYN piping analysis software.
The time-history asymmetric pressure forcing functions were developed by taking the time-history pressure differential functions adjusted for direction of the force and multiplying them by the equivalent areas of the various components. The forcing functions are applied to the RPV, steam generator and the RCP in the WESTDYN model. The WESTDYN input model represents a four loop model of the RCL piping system which includes the RPV. The WEST DYN four loop seismic model was used, however, the asymmetric pressure forces are only applie d to a single loop in the model since postulated breaks only occur in one loop at a time. The results represent an enveloped maximum of the time-history data points that can be conservatively added to the results of the LOCA analysis to obtain the combined LOCA plus asymmetric pressurization effects. The maximum enveloped primary equipment support loads and primary equipment nozzle loads are calculated.
The asymmetric pressure forces are then used in the reactor coolant loop and the NSSS component (RPV, SG, Pressurizer, and RCP) and supports evaluations (Refer to Section 5.2) 6.2.1.3 Safety Evaluation
6.2.1.3.1 General Design Criterion 4, 1987 - Environmental and Dynamic Effects Design Bases
The application of LBB is not applicable to the long-term mass and energy release
containment integrity analysis in Appendix 6.2D.
LBB is approved as a part of the DCPP license basis for the RCS primary loop piping (refer to Section 3.6.2.1.1.1).
The containment subcompartments are designed to sustain the short-term effects of postulated branch line failures as demonstrated in Section 6.2.1.2.2 and Table 6.2-64.
6.2.1.3.2 General Design Criterion 10, 1967 - Containment The containment is designed to sustain the effects of LOCA and MSLB events as demonstrated by the containment integrity analysis in Appendix 6.2D. This
demonstrates that the containment, supported by the containment heat removal system DCPP UNITS 1 &
2 FSAR UPDATE 6.2-18 Revision 23 December 2016 (CHRS), will retain its functional capability.
Refer to Section 6.2.2 for the discussion of the CHRS.
6.2.1.3.3 General Design Criterion 49, 1967 - Containment Design Basis
The results of the containment integrity analysis demonstrate that the containment pressure and temperature are maintained below the design conditions following limiting LOCA events (refer to Appendix 6.2D). This demonstrates that the containment, supported by the containment heat removal system (CHRS), will retain its functional
capability to not exceed the design leakage rate in order to protect the public. Refer to
Section 6.2.2 for a discussion of the CHRS.
6.2.1.3.4 General Design Criterion 54, 1967 - Containment Leakage Rate Testing
Containment leakage rate testing was conducted after completion of construction and installation of all penetrations to verify its conformance with required performance (refer to Section 6.2.1.4.1).
6.2.1.3.5 General Design Criterion 55, 1967 - Containment Periodic Leakage Rate Testing Periodic ILRT of the containment is performed as part of the DCPP Containment Periodic Leakage Rate Testing Program (refer to Sections 6.2.1.3.7 and 6.2.1.3.9).
6.2.1.3.6 General Design Criterion, 70, 1967 - Control of Releases of Radioactivity to the Environment The containment, in conjunction with the containment isolation system (CIS) (refer to Section 6.2.4.4.12), is designed to be a barrier to maintain control over plant radioactive effluents, whether gaseous, liquid, or solid.
The containment is designed to withstand the effects of a LOCA (refer to Se ction 6.2.1.3.3), ensuring that the offsite radiological exposures resulting from a LOCA are within the limits of 10 CFR Part 100 (refer to
Section 15.5).
6.2.1.3.7 10 CFR Part 50, Appendix J, Option B - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors 10 CFR Part 50, Appendix J, Option B, Type A and B testing, as modified by approved exemptions, are performed in accordance with Technical Specification 5.5.16, Containment Leakage Rate Testing Program. The ILRT is performed at a P a , calculated peak containment internal pressure, of 43.5 psig.
The maximum allowable leakage rate
is not greater than 0.10% of the containment air weight per day. For the discussion of
Type C testing (testing of containment penetrations) refer to Section 6.2.4.4.12.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-19 Revision 23 December 2016 6.2.1.3.8 10 CFR Part 50, Appendix K, Part I.A - ECCS Evaluation Models, Sources of Heat during the LOCA The initial conditions for the LOCA and MSLB events of the containment integrity analyses are established at the max imum calculated power for the reactor with additional margins for instrument error for the sources of heat listed in accordance with 10 CFR Part 50, Appendix K, Part I.A (refer to Section 6.2D.3.1.4).
6.2.1.3.9 Regulatory Guide 1.163, September 1995 - Performance-Based Containment Leak-Test Program The DCPP Containment Leakage Rate Testing Program utilizes a performance-based approach, consistent with Regulatory Guide 1.163, September 1995, to comply with the
requirements of 10 CFR Part 50, Appendix J, Option B (refer to Section 6.2.1.3.7).
6.2.1.4 Tests and Inspections 6.2.1.4.1 Preoperational Testing HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED
Following completion of each containment structure, a structural integrity test was performed by pressurizing the containment with air to 115 percent of design pressure, or 54 psig. These tests were performed as described in Section 3.8.2.1.7.
During the depressurization phase of the structural integrity tests, overall integrated leakage rate tests were performed in accordance with the requirements of Appendix J to 10 CFR Part 50 (refer to Section 3.8.2.1.7).
6.2.1.4.2 Inservice Surveillance Periodic leakage rate testing will be performed over the life of the units in accordance
with the requirements of Appendix J to 10 CFR Part 50, Option B, as modified by
approved exemptions (refer to Section 6.2.1.
3.7). Applicable surve illance requirements for such testing are included in the Technical Specifications. A leakage detection
system has been installed to measure leakrate for the personnel air lock door seals to
ensure compliance with the Technical Specifications (Reference 46). Periodic testing of
the containment isolation valves is discussed in Sections 6.2.4.4.12 and 6.2.4.5.
6.2.1.5 Instrumentation Applications Pressure inside the containment is continuously monitored by independent pressure
transmitters located at widely separated points outside the containment. Instruments
with range of -5 to 55 psig and 0 to 200 psig are available. Section 7.3.2 describes
containment pressure as an input to the ESFs actuation system and Section 7.5.2
describes containment pressure display instrumentation.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-20 Revision 23 December 2016 Other instrumentation available for monitoring conditions within the containment include:
(1) Containment water level monitors (described in Sections 6.3.3.4.4 and 7.5.2.1.3)
(2) Containment hydrogen monitors (described in Section 6.2.5.5)
(3) Temperature detectors positioned at various locations within the containment air volume (described in Section 9.4.5.3.3)
(4) Containment radiation and plant vent monitors (described in Sections 9.4.2 and 11.4.2.1.2)
6.2.1.6 Materials Containment structural heat sink materials used for containment integrity analyses
following a LOCA or main steam line break are listed in Table 6.2D-19; corresponding material properties are listed in Table 6.2D-20 (refer to Section 6.2D.3.2.4). A current
record of paint used on containment heat sink structures and equipment is maintained
in engineering files.
6.2.2 CONTAINMENT HEAT REMOVAL SYSTEMS The containment heat removal systems (CHRS) are the containment fan cooler system (CFCS) and the containment spray system (CSS). The functional performance objectives of the CHRS are:
(1) The CFCS limits the containment ambient temperature during normal plant operating conditions (refer to Section 9.4.5);
(2) The CFCS and CSS reduce the containment ambient temperature and pressure following a loss-of-coolant accident (LOCA) or main steam line
break (MSLB) inside containment. While performing this cooling function, the CHRS also helps limit offsite r adiation levels by reducing the pressure differential between containment and outside atmosphere, thus reducing
the driving force for leakage of fission products from the containment
atmosphere; (3) The CFCS provides mixing of the sprayed and unsprayed regions of the containment to improve airborne fission product removal (refer to Section
6.2.3);
(4) The CSS removes airborne fission products from the containment atmosphere following a LOCA (refer to Section 6.2.3);
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-21 Revision 23 December 2016 (5) The CSS, in conjunction with the spray additive system (SAS), prevents the re-evolution of the iodine in the recirculated core cooling solution (i.e.
sump water) following a LOCA (refer to Section 6.2.3);
(6) The CFCS provides a mixed atmosphere for hydrogen control (refer to Section 6.2.5).
Used in conjunction with one another during the injection phase, one containment spray
pump and two containment fan cooler units (CFCUs) will provide the heat removal
capability to maintain the post-accident containment atmospheric pressure and
temperature below the design values of 47 psig and 271°F, respectively. The CFCS is
credited for long-term containment pressure and temperature control throughout the
injection and recirculation phases following a LOCA or MSLB. The CSS is credited only for operation during the spray injection phase following a LOCA or MSLB; it is not
required for operation during the recirculation phase for mitigating the effects of a
LOCA. The physical SSC design bases, testing, and inspection requirements of the
CFCS and CSS are discussed in this section.
6.2.2.1 Design Bases 6.2.2.1.1 General Design Criterion 2, 1967 - Performance Standards The CFCS and CSS are designed to withstand the effects of, or are protected against, natural phenomena, such as earthquakes, flooding, tornadoes, winds, and other local
site effects.
6.2.2.1.2 General Design Criterion 10, 1967 - Containment
The CFCS and CSS are designed to aid other ESFs in retaining the functional capability of the containment to protect the public in the event of gross equipment failures, such as a large coolant boundary break.
6.2.2.1.3 General Design Criterion 11, 1967 - Control Room
The CFCS and CSS are designed to support actions to maintain and control the safe operational status of the plant from the control room.
6.2.2.1.4 General Design Criterion 12, 1967 - Instrumentation and Control Systems The CFCS and CSS are provided with instrumentation and controls as required to monitor and maintain the CHRS variables within prescribed operating ranges.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-22 Revision 23 December 2016 6.2.2.1.5 General Design Criterion 15, 1967 - Engineered Safety Features Protection Systems The CFCS and CSS are provided with instrumentation for sensing accident conditions.
6.2.2.1.6 General Design Criterion 19, 1971 - Control Room
The CFCS and CSS, in conjunction with the SAS, are designed to limit radiation exposure to personnel to permit access and occupancy of the control room under accident conditions without personnel receiving radiatio n exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.
6.2.2.1.7 General Design Criterion 21, 1 967 - Single Failure Definition The CFCS and CSS are designed to tolerate a single failure during the period of recovery following an accident without loss of their protective functions, including
multiple failures resulting from a single event, which is treated as a single failure.
6.2.2.1.8 General Design Criterion 37, 1967 - Engineered Safety Features Basis for Design The CFCS and CSS are designed to provide back-up to the safety functions provided by the core design, the reactor coolant pressure boundary, and their protection systems.
6.2.2.1.9 General Design Criterion 38, 1967 - Reliability and Testability of Engineered Safety Features The CFCS and CSS are designed to provide high functional reliability and ready testability.
6.2.2.1.10 General Design Criterion 40, 1967 - Missile Protection The CFCS and CSS are designed to be protected against dynamic effects and missiles that might result from plant equipment failures.
6.2.2.1.11 General Design Criterion 41, 1967 - Engineered Safety Features Performance Capability The CFCS and CSS are designed to provide sufficient performance capabilities to accommodate a partial loss of installed capacity, including a single failure of an active component, and still perform their required safety functions.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-23 Revision 23 December 2016 6.2.2.1.12 General Design Criterion 42, 1967 - Engineered Safety Features Components Capability The CFCS and CSS are designed so that the capability of each component and system to perform its required function is not impaired by the effects of a LOCA.
6.2.2.1.13 General Design Criterion 49, 1967 - Containment Design Basis The CFCS and CSS are designed so that the containment structure can accommodate, without exceeding the design leakage rate, the pressures and temperatures resulting
from the largest credible mass and energy releases following a 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.
6.2.2.1.14 General Design Criterion 52, 1967 - Containment Heat Removal Systems The CFCS and CSS are designed as two systems of different principles, each with full capacity, for active heat removal from the containment under accident conditions.
6.2.2.1.15 General Design Criterion 54, 1 971 - Piping Systems Penetrating Containment The CSS 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.
6.2.2.1.16 General Design Criterion 56, 1971 - Primary Containment Isolation The CSS contains valving in piping that penetrates containment and that is connected directly to the containment atmosphere. Remote manual isolation valves are provided
outside containment and automatic (check) valves are provided inside containment to
ensure containment integrity is maintained.
6.2.2.1.17 General Design Criterion 58, 1967 - Inspection of Containment Pressure-Reducing Systems The CFCS and CSS are designed to facilitate the periodic physical inspection of all important components, such as fans, pumps, valves, dampers, and spray nozzles.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-24 Revision 23 December 2016 6.2.2.1.18 General Design Criterion 59, 1967 - Testing of Containment Pressure-Reducing Systems The CFCS and CSS are designed so that active components, such as fans, pumps, valves, and dampers can be tested periodically for operability and required functional
performance.
6.2.2.1.19 General Design Criterion 60, 1967 - Testing of Containment Spray Systems The CSS is designed to allow periodic testing of the delivery capability of the system at a position as close to the spray nozzles as is practical.
6.2.2.1.20 General Design Criterion 61, 1967 - Testing of Operational Sequence of Containment Pressure-Reducing Systems Components The CFCS and CSS are designed to provide the capabilities to test under certain conditions as close to the design as practical, the full operational sequence that would
bring the CHRS into action, including the transfer to alternate power sources.
6.2.2.1.21 General Design Criterion 62, 1967 - Inspection of Air Cleanup Systems The CFCS and CSS are designed to facilitate physical inspection of all critical parts, such as ducts, fans, pumps, valves, dampers, and spray nozzles.
6.2.2.1.22 General Design Criterion 63, 1967 - Testing of Air Cleanup Systems Components The CFCS and CSS are designed so that active components, such as fans, pump, valves, and dampers, can be tested periodically for operability and required functional
performance.
6.2.2.1.23 General Design Criterion 64, 1967 - Testing of Air Cleanup Systems The CFCS is designed to provide for in situ periodic testing and surveillance to ensure trapping materials have not deteriorated beyond acceptable limits.
6.2.2.1.24 General Design Criterion 65, 1967 - Testing of Operational Sequence of Air Cleanup Systems The CFCS and CSS are designed with the capability to test, under conditions as close to design as practical, the full operational sequence that would bring the CHRS into action, including the transfer to alternate power sources and design air flow delivery capability.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-25 Revision 23 December 2016 6.2.2.1.25 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment The CFCS and CSS, in conjunction with the SAS, are designed with provisions for maintaining control of the plants radioactiv e gaseous effluents to meet the radiological limits of 10 CFR Part 100.
6.2.2.1.26 10 CFR 50.49 - Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants CFCS and CSS components that require environmental qualification (EQ) are qualified to the requirements of 10 CFR 50.49.
6.2.2.1.27 10 CFR 50.55a(f) - Inservice Testing Requirements
CFCS and CSS 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.
6.2.2.1.28 10 CFR 50.55a(g) - Inservice Inspection Requirements
CFCS and CSS 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.
6.2.2.1.29 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 CFCS and CSS provide instrumentation to monitor containment isolation valve position, containment spray flow, containment pressure, containment temperature during and following an accident, and a two-step process for monitoring heat removal by
the CFCS.
6.2.2.1.30 NUREG-0737 (Items II.F.1, III.D.
1.1), November 1980 - Clarification of TMI Action Plan Requirements Item II.F.1 - Additional Accident Monitoring Instrumentation:
Position (4) - The CSS provides instrumentation to continuously monitor and record containment pressure in the control room.
Item III.D.1.1 - Integrity of Systems Outside Containment Likely to Contain Radioactive Material for Pressurized-Water Reactors and Boiling-Water Reactors:
The CSS is designed to allow for the perform ance of leak rate tests as part of the containment isolation valve surveillance test program.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-26 Revision 23 December 2016 6.2.2.1.31 Generic Letter 89-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance The CSS PG&E Design Class I and position changeable motor-operated valves (MOVs) meet the requirements of Generic Letter 89-10, June 1989, and associated Generic Letter 96-05, September 1996.
6.2.2.1.32 Generic Letter 97-04, October 1997
- Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps The CSS is evaluated to assure adequate NPSH is available to the containment spray pumps under all design basis accident (DBA) scenarios.
6.2.2.2 System Description The CHRS is designed to provide sufficient heat removal capability to maintain the post-accident containment atmospheric pressure and temperature below the design values
of 47 psig and 271°F, respectively (refer to Section 3.8.2.1.1). The containment atmospheric temperature after a MSLB does briefly increase greater than the
containment design temperature, but, as discussed in Section 6.2D.4.2.4, there is no
explicit design temperature limit for a MSLB. Heat energy sources considered are
described in Section 6.2D.2.1.12.
The CFCS also functions during normal operating conditions to limit the containment
ambient temperature to 120°F and is described in Section 9.4.5.
The CFCS, shown schematically in Figure 9.4-4, consists of five identical fan coolers, each including cooling coils, fan and drive motor, locked-open air flow dampers and pressure relief dampers, duct distribution system, instrumentation, and control. During
operation of the units, air is drawn into the cooling coils, cooled, and discharged back through the ductwork to the containment atmosphere.
The design parameters for the CHRS comp onents and materials are listed in
Table 6.2-26. Codes and standards used as a basis for the design of the components
are given in Table 6.2-25.
Ductwork distributes the cooled air to the various containment compartments and areas.
During normal and post-accident operations, the flow sequence through each air fan
cooler is as follows: locked-open normal and accident air flow dampers, cooling coils, fan, and distribution ductwork.
Airflow through the exhaust ducting, towards the fan, can occur when the fan is idle.
Incorporated into the fan/motor coupling is an anti-reverse rotation device (AARD) that
precludes the fan motor from rotating backwards. This device replaces backdraft
dampers previously installed in the fan discharge duct.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-27 Revision 23 December 2016 The CSS, shown schematically in Figure 3.2-12, consists of two pumps, spray ring
headers and nozzles, valves, and connecting piping.
Following a LOCA, water from the refueling water storage tank (RWST) is initially used for containment spray. Later, water
recirculated from the containment sump can be supplied by the residual heat removal (RHR) pumps for recirculation spray. If no component failures affect the RHR train
capability, the emergency procedures direct the initiation of recirculation sprays.
However, single failures that result in the loss of one RHR train cause the decision of
how to divide the recirculation flow between spray and core injection to be made by the
technical support center in charge of accident mitigation.
6.2.2.2.1 Component Descriptions 6.2.2.2.1.1 Containment Spray System 6.2.2.2.1.1.1 Refueling Water Storage Tank The RWST serves as a source of emergency borated cooling water for the injection
phase. The RWST will normally be aligned to the suction of the emergency core
cooling pumps and the containment spray pumps. During a MSLB inside containment
or LOCA the containment spray pumps will continue to take suction from the RWST
until the RWST low-low level signal is provided. Refer to Section 6.3 for additional
details on the RWST.
6.2.2.2.1.1.2 Containment Spray Pumps
The containment spray pumps are of the horizontal centrifugal type and are driven by electric motors. The motors are powered from separate Class 1E, 4.16-kV buses.
6.2.2.2.1.1.3 Spray Nozzles The spray nozzles are of the hollow cone de sign having an open throat and 3/8 inch spray orifice and are not subject to clogging by particles less than 1/4 inch in size.
Refer to discussion in Section 6.2.2.3.8.1.2.
6.2.2.2.1.1.4 Containment Spray System Piping and Valves The piping and valves for the CSS have design pressures and temperatures of 240 psig
and 200°F, respectively.
6.2.2.2.1.2 Containment Fan Cooler System The design data of the CFCS are presented in Table 6.2-26. During normal operation, the number of units running will depend on the amount of cooling required in the
containment. The operator uses the containment temperature and pressure readings to
determine how many fan cooler units should be operating. At full power operation, four DCPP UNITS 1 &
2 FSAR UPDATE 6.2-28 Revision 23 December 2016 or fewer units are usually required, while at cold shutdown only one unit may be needed (refer to Section 9.4.5). Limiting conditions for operation are included in the Technical
Specifications.
6.2.2.2.1.2.1 Cooling Coils The coils are fabricated with copper plate fins on copper tubes. Each fan cooler
consists of 12 individual coils mounted in two banks, 6 coils high. These banks are located one behind the other for horizontal s eries air flow, and the tubes of the coil are horizontal with vertical fins.
The cooling coil assembly consists of one bank of WC-36114-4H (1/2 water velocity
circuiting) coils four rows deep and one bank of WC-36114-6T (1/3 water velocity
circuiting) coils six rows deep. Each coil is provided with a drain pan and drain piping to prevent flooding of the coil face area during accident conditions. This condensate is
drained to the containment sump. The monitoring of this condensate for the leakage
detection system of the RCS pressure boundary is discussed in Section 5.2.3.23.1.2.
The component cooling water system (CCW) supplies cooling water to the containment fan coolers and is discussed in Section 9.2.2.
6.2.2.2.1.2.2 Fans The five containment cooling fans are of the centrifugal, nonoverloading direct-drive
type. The fan bearings have a specially designed seal, are heavy duty, and are
selected for the proper thrust and axial loads. Special lubricant is used to ensure protection during accident operation (refer to Section 6.2.2.3.26).
6.2.2.2.1.2.3 Enclosure Each of the five fan cooler enclosure assemblies consists of four prefabricated modular
units. Modules 1, 2, and 3 are located on the inlet side of the cooling coils. Module 4 is
located on the outlet side of the cooling coils and serves as a plenum for the fan inlet.
The CFCS was modified to remove the moisture separators and HEPA filters.
6.2.2.2.1.2.4 Anti-Reverse Rotation Device An AARD is incorporated into the fan/motor coupling that prevents the fan motor from rotating backwards. This device protects the fan motor from airflow that could cause an
over-current trip condition when the idle fan starts.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-29 Revision 23 December 2016 6.2.2.2.1.2.5 Pressure Relief Damper Each fan cooler unit is equipped with a pressure relief damper. These dampers are
normally closed counterweighted devices that open progressively as the pressure
differential across them exceeds 0.25 psi.
6.2.2.2.1.2.6 Ductwork Vacuum relief dampers and pressure relief dampers are provided along the ductwork to
limit the differential pressure acting on the duct to 0.2 psi during accident conditions.
The debris screen, flexible connection, duct branch, and vacuum relief dampers are
required to function from the beginning of an accident to full closure of the containment
purge isolation valves (assuming the valve is open for containment purge when the
accident occurs). This is to ensure that debris generated during an accident will not
lodge in the seat of the containment isolation valve to prevent its full closure.
Ducts are constructed of galvanized sheet steel. Bolted flanges are provided with
gaskets suitable for 300°F service for the PG&E Design Class I portion of the ducts and suitable for 250°F for the PG&E Design Class II portion of the ducts. All longitudinal seams are tack welded or riveted and sealed.
6.2.2.2.1.2.7 Air Flow Dampers Dampers are locked in their normal operating position. The air flow dampers are an
integral part of the fan coolers. Each damper is constructed of steel painted with corrosion-inhibiting paint, with multiple blade s and edge seals to minimize leakage.
6.2.2.2.1.2.8 Motors for Fan Coolers A two-speed, single-winding motor is used to drive each fan cooler. The motor operates
at the high speed during normal operation and at the low speed during post-accident
operation.
The motor unit is provided with an integral air-to-water heat exchanger.
The motor heat exchanger consists of the co oling coil, the housing, and two pressure equalization dampers (relief valves to vent containment pressure into the heat
exchanger).
The motor heat exchanger circulates component cooli ng water. The joints of the motor heat exchanger cooling coil are brazed with a high-temperature alloy.
The motor heat exchanger cooling coil is mounted within the motor heat exchanger housing and is generally of the same type construction as the main coils. The plenum
after the coil has a condensate drain connection.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-30 Revision 23 December 2016 The fan cooler motor heat exchanger housing is equipped with two pressure
equalization dampers to relieve pressure differentials resulting from a post-accident
pressure transient. The valves begin to open at 5 inches water gauge. The valves are
designed for a maximum pressure differential of 30 inches of water. Each valve has a
maximum flow area of 7.07 square inches. The valves and body flapper plates are
electrolytically nickel-coated carbon steel.
The valve brackets, links, shafts, springs, and fasteners are Type 316 stainless steel. The seat seals are a silicone rubber O-ring.
Each valve is subjected to a certification test before shipment to ensure proper opening
pressure and leaktightness.
The motor, motor heat exchanger, and fan are mounted on a common base for extra
rigidity.
6.2.2.2.1.3 Electrical Supply The CFCS fan motors and CSS valves are powered from separate Class 1E 480-V
buses and the containment spray pumps are powered from separate Class 1E 4.16-kV buses (refer to Section 8.3.1.1.3).
6.2.2.3 Safety Evaluation
6.2.2.3.1 General Design Criterion 2, 1967 - Performance Standards
With the exception of the RWST and connected piping, the CFCS and CSS components are located within the PG&E Design Class I aux iliary and containment buildings. The applicable portions of these buildings are designed to withstand the effects of winds and tornadoes (refer to Section 3.3), floods and tsunamis (refer to Section 3.4), external
missiles (refer to Section 3.5), and earthquakes (refer to Section 3.7). The design of the
auxiliary and containment build ings will protect the CFCS and CSS components against natural phenomena and local site effects, ensuring their design functions will be
performed.
The evaluation of the performance standards for the RWST is provided in Section
6.3.2.2.2.
The CFCS fan coolers are designed to PG&E Design Class I criteria.
The fan cooler discharge ductwork and supports, up to and including the backdraft
dampers' frame, are PG&E Design Class I to maintain the integrity and the design heat
removal capability of the operating fan coolers. Except for the short section of branch
duct between the containment purge exhaust isolation valve and its debris screen, the
ducting is PG&E Design Class II and is seis mically designed. This duct section must maintain its integrity during an accident up to the time full closure of the containment
purge isolation valve has been attained to en sure that debris generated during an accident will not lodge in the seat of the containment isolation valve to prevent its full DCPP UNITS 1 &
2 FSAR UPDATE 6.2-31 Revision 23 December 2016 closure. The duct branch bounded by the containment purge isolation valve and its debris screen, including the flexible connectio n, is classified as PG&E Design Class I.
The vacuum relief dampers are classified as PG&E Design Class I and the pressure relief dampers are classified as PG&E Design Class II. This greatly minimizes the collapsing/damage to the distribution ductwork from pressure transients during an
accident condition.
The HE and DDE piping analyses for the piping located inside containment downstream of sealed-open isolation valves 9006 A and B take credit for the empty piping configuration that exists during normal plant operation.
6.2.2.3.2 General Design Criterion 10, 1967 - Containment The CFCS and CSS are designed to support containment integrity to sustain the effects of a LOCA or MSLB. The CHRS maintains the containment within its maximum design
conditions therefore retaining its functional capability. The containment atmospheric
temperature after a MSLB does briefly increase greater than the containment design
temperature, but, as discussed in Section 6.2D.4.2.4, there is no explicit design
temperature limit for a MSLB. The containment integrity analyses are demonstrated in
Appendix 6.2D.
6.2.2.3.3 General Design Criterion 11, 1967 - Control Room
The CFCS and CSS are designed with remote-manual operation for each train of CSS and CFCUs in the control room.
The CFCS and CSS are designed with control room indication including containment pressure, containment isolation valve position, containment spray pump discharge flow, heat removal by containment fan heat remov al system through: CFCU motor speed indicating lights, CCW flow indication, containment pressure indication, and containment atmosphere temperature indication (refer to Section 7.5.2).
Therefore, indication and controls are provided in the main control room to maintain
safe operational status of the CFCS and CSS.
6.2.2.3.4 General Design Criterion 12, 1967 - Instrumentation and Control Systems The CFCS and CSS are provided with instrumentation and controls to monitor and
maintain CHRS variables within prescribed operating ranges. In addition to actuating
based on signals generated by the engineered safety features actuation system (ESFAS), which receives input from containment pressure-related instrumentation (refer
to Section 7.3.2), CHRS instrumentation requirements include the capabilities for
measurement and local indication of suction pressure which can be used to evaluate
NPSH and discharge pressure in the containment spray pumps and containment spray
pump flow to containment. To ensure that water flows to the safety injection system DCPP UNITS 1 &
2 FSAR UPDATE 6.2-32 Revision 23 December 2016 (SIS) after a LOCA and determine when to shift from the injection to the recirculation mode, RWST level indication and alarm are provided. High-high containment pressure (P signal) coincident with a safety injection signa l (S signal) will automatically initiate the CSS as discussed in Section 6.2.2.3.5.
The containment fan cooler bearings are monitored for vibration. Similarly, the containment fan cooler motor assembly bea rings and windings are monitored to ensure that vibration and temperature limits are not exceeded.
The following instrumentation associated with the containment fan coolers enables
additional monitoring of in-containment conditions in the post-LOCA recovery period:
(1) The CCW discharge flow for the containment fan coolers is indicated in the control room and alarmed if the flow is low.
(2) The CCW exit temperatures are indicated locally outside the containment building.
(3) Bearing temperatures are indicated and alarmed on the plant process computer.
6.2.2.3.5 General Design Criterion 15, 1967 - Engineered Safety Features Protection Systems The CHRS is designed with instrumentation that measures containment pressure. The CSS will be actuated through ESFAS by a "P" signal (high-high containment pressure actuation setpoint) on coincidence of two-out-of-four high-high containment pressure signals. Coincidence of "S" (safety injection) and "P" signals starts the containment
spray pumps and opens the discharge valves to the spray headers. Refer to Section 7.3 for a description of the ESFAS.
6.2.2.3.6 General Design Criterion 19, 1971 - Control Room
The CSS is designed to deliver borated water from the RWST to the containment atmosphere during the injection phase following a LOCA. The CSS, in conjunction with
the CFCS, reduces the containment ambient temperature and pressure following a
LOCA or MSLB, thus reducing the driving force for leakage of fission products from the
containment atmosphere. The CFCS provides mixing of the sprayed and unsprayed
regions of the containment to improve airborne fission product removal. The SAS, as
discussed in Section 6.2.3, injects sodium hydroxide to the suction of the CSS for
delivery to the containment atmosphere to prevent the re-evolution of the iodine in the
recirculated core cooling solution. The CSS removes iodine from the containment
atmosphere to ensure that radiological expo sures for control room personnel are within 5 rem whole body. Refer to Section 15.5 for radiological consequences of plant
accidents. Therefore, the CSS, in conjunction with the CFCS and SAS, limits the
control room doses to 5 rem whole body.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-33 Revision 23 December 2016 6.2.2.3.7 General Design Criterion 21, 1 967 - Single Failure Definition The CFCS and CSS are designed such that no single failure in either train will prevent the CHRS from performing its design function. The CSS is comprised of two completely independent trains of containme nt spray. Each CSS train of pumps and valves are powered by separate Class 1E 4.16-kV and 480-V power supplies, respectively. The
CFCS is comprised of five independent fan coolers; two are powered from Class 1E
480-V Bus F, two are powered from Class 1E 480-V Bus G, and one is powered from
Class 1E 480-V Bus H.
Any single failure will still leave sufficient CS S and CFCS capability to together mitigate DBAs. Used in conjunction with one another during the injection phase, one
containment spray pump and two containment fan cooler units will provide the heat
removal capability to maintain the post-accident containment pressure below the design
value of 47 psig. CHRS design parameters are listed in Table 6.2-26.
A single failure analysis on all active components of the CHRS was performed to show that the failure of any single component will not prevent performance of the design
function. This analysis is summarized in Table 6.2-27.
6.2.2.3.8 General Design Criterion 37, 1967 - Engineered Safety Features Basis for Design The CSS is credited only for operation during the injection phase following a LOCA and MSLB. The CFCS is credited for containment pressure and temperature control throughout the injection and recirculation phases following a LOCA or MSLB.
The CSS and CFCS limit the effects of post blowdown energy additions to the containment during the injection phase following a LOCA. For a detailed description of the analytical methods and models used to assess the performance capability of the
CHRS, refer to the containment integrity analysis presented in Appendix 6.2D.
The CHRS provides a backup to the safety provided by the core design, the reactor coolant pressure boundary, and their protection systems. As discussed in this section, the CHRS is designed to withstand any size reactor or secondary coolant pressure
boundary break, including a LOCA or MSLB.
6.2.2.3.8.1 Containment Spray System
The CSS, in conjunction with the CFCS, reduces the containment ambient temperature and pressure following a LOCA or MSLB, thus reducing the driving force for leakage of
fission products from the containment atmosphere. The CSS removes airborne iodine
from the containment atmosphere following a LOCA (refer to Section 6.2.3). The CSS, in conjunction with the spray additive system (SAS), prevents the re-evolution of the DCPP UNITS 1 &
2 FSAR UPDATE 6.2-34 Revision 23 December 2016 iodine in the recirculated core cooling solution (also known as sump water) following a LOCA (refer to Section 6.2.3).
6.2.2.3.8.1.1 Containment Spray Pumps The containment spray pumps are designed to perform at rated capacity against a total head composed of containment design pressure, nozzle elevation head, and the line
and nozzle pressure losses. Adequate NPSH is available for operation of the
containment spray pumps throughout the entire spray injection phase considering the
low-low level in the RWST. A pe rformance curve for the containment spray pumps is shown in Figure 6.2-10.
During the spray injection phase, a single containment spray pump delivers approximately 2466 gpm to the spray header, at containment design pressure (47 psig).
If both of the two spray pumps are available in this phase, approximately 4932 gpm is
delivered to the containment spray headers from the RWST. This flowrate is not ECCS
dependent.
6.2.2.3.8.1.2 Containment Spray Nozzles The CSS nozzles produce a mean drop diameter of 700 microns at rated system conditions (40 psi p and 15.2 gpm per nozzle). The minimum fall path for CSS water droplets is conservatively assumed to be the distance from the lowest spray ring to the operating deck. The heat transfer calculations presented in Section 6.2.1 and Appendix
6.2D show that essentially all spray droplets reach thermal equilibrium at containment
design temperature and pressure in a distance considerably less than the minimum fall
path. The spray solution is stable and soluble at all temperatures of interest in the containment and will not precipitate or otherwise interfere with nozzle performance. If containment spray is used in the recirculation phase, the containment recirculation
sump screens (refer to Figures 6.3-6 and 6.3-7) limit particle size to preclude the
possibility of clogging (refer to Section 6.3). Nozzle design and performance characteristics are listed in Table 6.2-26 and in Figure 6.2-12. A plan view of the
containment spray headers showing the location of the spray nozzles is given in
Figure 6.2-13. An evaluation of the containment spray pattern, including droplet size
and distribution, is discussed in Section 6.2.3.
6.2.2.3.8.1.3 Containment Spray System Phases of Operation The CSS may operate over an extended period and under the environmental conditions existing following a LOCA or an MSLB.
The system operation can be divided into the following two distinct phases.
6.2.2.3.8.1.3.1 Injection Phase
The CSS will be actuated by a "P" signal, eith er manually from the control room, or on coincidence of two-out-of-four high-high cont ainment pressure signals. Coincidence of DCPP UNITS 1 &
2 FSAR UPDATE 6.2-35 Revision 23 December 2016 "S" and "P" signals starts the containment spray pumps and opens the discharge valves to the spray headers. The "P" signal alone will open the valves associated with the spray additive tank. During the spray injection phase, the pumps are drawing borated
water from the RWST and mixing it with NaOH solution from the spray additive tank (refer to Section 6.2.3). Spray injection wi ll continue until the RWST low-low level is
reached, at which time the containment spray pumps are manually tripped and isolated.
6.2.2.3.8.1.3.2 Recirculation Phase If the CSS is used in the spray recirculation phase, recirculation spray is provided by the
RHR pumps, which draw suction from the containment sump. Spray recirculation phase
operation of the RHR pumps is discussed in Section 6.3.2.4.3.1 and Section 5.5.6.
6.2.2.3.8.2 Containment Fan Cooler System 6.2.2.3.8.2.1 Containment Fan Coolers In addition to limiting maximum containment pressure and temperature, the containment fan coolers provide mixing of the containment atmosphere for iodine removal and
control of hydrogen buildup as discussed in Sections 6.2.3 and 6.2.5, respectively.
The heat removal capability of the containment fan cooler cooling coils is 81 x 10 6 Btu/hr per fan cooler unit at saturation conditions (271
°F, 47 psig), with 2000 gpm cooling water supply at 125°F (refer to Table 6.2-26).
The design internal pressure of each coil is 200 psig and the coils can withstand an
external pressure of 47 psig at a temperature of 271°F without damage.
Each cooling coil assembly has a top and bottom horizontal coil casing made of
galvanized steel. The safety function of the cooling coil casings is to fill the air gap
between each stacked cooling coil assembly and direct airflow through the cooling fins
to ensure that adequate heat transfer occurs within the containment fan cooler units.
Each fan can provide a minimum flowrate of 47,000 cfm (in low speed) when operating against the system resistance of approximately 3-3/4 inches of water existing during the
accident condition.
As discussed in Section 6.2.2.2.1.2.3, each CFCS unit is comprised of four, pre-
fabricated modular units. As a result of the modifications made to the fan cooler
enclosure assemblies, only Module 3 serves to direct the airflow into the coils for normal
and post-LOCA operations. Modules 1 and 2 are no longer required for service.
Module 1 contains the locked-open accident flow inlet dampers. Locking the accident
inlet damper open prevents module overpressurization during a LOCA. Module 2
contains the locked-closed accident flow outlet dampers. Modules 1 and 2 are open to
one another (i.e., Module 1 and Module 2 do not have a division or dampers separating DCPP UNITS 1 &
2 FSAR UPDATE 6.2-36 Revision 23 December 2016 the two). Module 3 contains the locked-ope n normal flow inlet dampers and the pressure-relief damper.
The AARD is designed to withstand the dynamic loads associated with a 7 psi pressure differential.
The pressure relief dampers limit the pressure differential across the fan cooler
enclosure walls in the event of a LOCA or MSLB, and thus maintain the structural
integrity of the fan cooler units during the pressure transient.
The fan cooler motor heat exchanger has sufficient capacity to remove all motor assembly heat losses and external heat loads under all operating conditions, while limiting the maximum thermal environment consistent with motor design.
In the event of a LOCA or MSLB, two of the five fan coolers are required to operate.
The units are placed in post-accident mode of operation either by a safety injection signal or manually from the control room.
As shown in Table 6.2-26, during a LOCA or MSLB, each fan cooler motor will
automatically trip. The motors will then restar t at low speed (600 rpm) and provide a heat removal capacity of 81 x 10 6 Btu/hr at containment design conditions. The cooler heat removal capacity is based on 125°F cooling water and other parameters as
specified in Table 6.2-26.
A high degree of mechanical reliability is incorporated in the containment ventilation
system. The fan cooler system regularly de monstrates its availability because it is used during normal plant operation to control temperature inside the containment (refer to
Section 9.4.5). In the event of a f ailure of offsite electric power concurrent with a LOCA or MSLB, the fan cooler units and the CCW pumps supplying cooling water to these
coolers are started automatically and supplied with power from the standby power
source.
Immediately following a LOCA, the peak steam-air mixture entering the cooling coils is
at approximately 271°F with a density of 0.175 pounds per cubic foot. Part of the water
vapor condenses on the cooling coils. The air-side pressure drop is not appreciably
affected by the condensate on the cooling coils. The air leaving the coils is saturated at
a temperature somewhat below 271°F.
The steam-air mixture remains in this condition as it flows into the fan. At this point it
picks up some sensible heat from the fan and fan motor before entering the distribution
header and the dry-bulb temperature rises slightly above 271°F and the relative
humidity drops to slightly below 100 percent.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-37 Revision 23 December 2016
6.2.2.3.8.2.2 Flow Distribution and Flow Characteristics
The location of the distribution ductwork outlets, together with the location of the fan cooler unit inlets, ensures that the air will be directed to all areas requiring ventilation
before returning to the units.
In addition to ventilating areas inside the periphery of the polar crane wall, the
distribution system also includes branch ducts for ventilating the upper portion of the
containment. These ducts extend upward a long the containment wall to the dome area, although the volume control dampers have been closed and no longer supply air flow.
Refer to Section 6.2.2.3.1 for a discussion on the PG&E design classification of the
ductwork.
The air discharged inside the periphery of the polar crane wall will circulate and rise
above the operating floor though openings around the steam generators where it will
mix with air displaced from the dome area. This mixture will return to the fan cooler
inlets located on the operating floor. The temperature of this air will essentially be the design ambient for the containment (refer to Section 9.4.5).
In the accident mode of operation, the recirculation rate with five coolers operating is
approximately 5.4 containment volumes per hour.
6.2.2.3.8.2.3 Cooling Water for the Fan Cooler Units
The cooling water requirements for the fan cooler units during a LOCA or MSLB and the recovery are supplied by the CCW system, which is described in Section 9.2.2 and in
Table 6.2-26.
The CFCS removes sufficient heat from the containment, following the initial LOCA or MLSB containment pressure transients, to keep the containment pressure from exceeding the design pressure. The fans and cooling coils continue to remove heat
after the LOCA or MSLB and reduce the containment pressure close to atmospheric
within the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
In addition, the following objectives are met to provide the ESF functions:
(1) Each of the five fan cooler units is capable of transferring heat from the containment atmosphere, at the design ba sis rate for post-accident design
conditions (refer to Table 6.2-26).
(2) In removing heat at the design basis rate, the coils are capable of discharging the resulting condensate without impairing their flow
capacities and without raising the cooling water exit temperature to the
boiling point. Since condensation of water from the air-steam mixture is DCPP UNITS 1 &
2 FSAR UPDATE 6.2-38 Revision 23 December 2016 the principal mechanism for removal of heat from the post-accident containment atmosphere by the cooling coil s, the coil fins will operate as wetted surfaces under these conditions. Entrained water droplets added
to the air-steam mixture will therefore have essentially no effect on the
heat removal capability of the coils.
The post-accident heat removal capability of the fan coolers is demonstrated by the
Westinghouse computer program HECO (Reference 36).
6.2.2.3.9 General Design Criterion 38, 1967 - Reliability and Testability of Engineered Safety Features The CFCS and CSS are designed with high reliability and testability. This is demonstrated through testing included in the surveillance test procedures for the plant delineated in the Technical Specifications as discussed in Sections 6.2.2.3.8.2, 6.2.2.3.15, 6.2.2.3.17 through 6.2.2.3.23, 6.2.
2.3.26, 6.2.2.3.27, and 6.2.2.3.28.
6.2.2.3.10 General Design Criterion 40, 1967 - Missile Protection All of the fan coolers, the distribution ductwork to the annulus ring (inclusive), and cooling water piping are located outside the missile shield wall. This arrangement
provides protection from missiles for all system components. Portions of the branch
ductwork from the annulus penetrate and extend past the missile shield wall. However, as described in Section 6.2.2.3.12, they are not required for accident mitigation.
The CSS is protected from missiles, pipe whi p, or jet impingement from the rupture of any nearby high-energy lines (refer to Sections 3.5 and 3.6). A vulnerability of the
system to this is with the portion of the CSS piping located outside of containment in the
containment penetration area (GE/GW area) which could potentially result in missiles as
a result of a MSLB; however, a MSLB outside of containment does not require operation
of the CSS.
6.2.2.3.11 General Design Criterion 41, 1967 - Engineered Safety Features Performance Capability
The CFCS and CSS, including required aux iliary systems, are both designed to tolerate a single active failure during the spray injection phase following a LOCA or MSLB
without loss of protective function. Therefore, the CFCS and CSS provide sufficient
performance capability to accommodate partial loss of installed capacity and still fulfill
the required functions for containment heat removal, airborne fission product removal, and CFCS mixing for hydrogen control.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-39 Revision 23 December 2016 6.2.2.3.12 General Design Criterion 42, 1967 - Engineered Safety Features Components Capability The CFCS and CSS component design pressure and temperature conditions in Table 6.2-26 are specified as the most severe conditions to which each component is
exposed during either normal or post-LOCA operation.
During a LOCA, the primary objective of the CFCS is heat removal for containment pressure reduction. The exact distribution of the recirculation flow through the ductwork
is not critical. In the event of breakage/damage to PG&E Design Class II ductwork
branches, no significant reduction in CFCS performance is anticipated because
ductwork damage will not reduce the total heat removal capability of the operating fan
coolers. Therefore, the CFCS and CSS are designed so that the capabilities to perform the required functions are not impaired by the effects of a LOCA.
6.2.2.3.13 General Design Criterion 49, 1967 - Containment Design Basis
The CFCS and CSS ensure that the containment design pressure and temperature are maintained below the design conditions follo wing limiting LOCA events. Refer to
Appendix 6.2D for the containment integrity analysis.
The CSS penetrations are designed to withstand the pressures and temperatures that
could result from a LOCA without exceedi ng the design leakage rates. Refer to Section 3.8.2.1.1.3 for additional details.
6.2.2.3.14 General Design Criterion 52, 1967 - Containment Heat Removal Systems Adequate heat removal capability for the containment atmosphere is provided by two diverse and separate ESFs, the CSS and th e CFCS, capable of maintaining the containment below the maximum design pressure and temperature following a LOCA or
MSLB. During the recirculation phase, the CFCS alone is capable of maintaining the
containment below its maximum design pressure and temperature. These two systems
are designed to work in conjunction with one another to meet the single failure criteria
as discussed in Section 6.2.2.3.7.
6.2.2.3.15 General Design Criterion 54, 1 971 - Piping Systems Penetrating Containment The CSS containment isolation valves are periodically tested for operability and leakage. Testing of the components required for the containment isolation system (CIS) is discussed in Section 6.2.4. Test connections are provided in the penetration and in
the piping to verify valve leakage and penetration leakage are within prescribed limits.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-40 Revision 23 December 2016 6.2.2.3.16 General Design Criterion 56, 1971 - Primary Containment Isolation Valves The CSS containment penetrations that are part of the CIS include the containment spray pump discharge lines which comply with the requirements of GDC 56, 1971, as described in Section 6.2.4 and Table 6.2-39.
6.2.2.3.17 General Design Criterion 58, 1967 - Inspection of Containment Pressure-Reducing Systems Access is available for visual inspection of the fan cooler components, including fans, cooling coils, enclosure dampers, and ductwork. Since these units are in use during
power operation, continuous checks of their status are available.
The containment fan cooler cooling coils are designed to be easily removed for
inspection and maintenance.
Refer to Section 6.2.2.3.19 and 6.2.2.3.27 for a discussion on testing and inspection of
the CSS.
Therefore, provisions are available to facilitate periodic inspection of all important
components of the CFCS and CSS.
6.2.2.3.18 General Design Criterion 59, 1967 - Testing of Containment Pressure-Reducing Systems Provisions are available for periodic testing of active CSS components to ensure operability and required functional performance is maintained (refer to Section
6.2.2.3.19).
The fan cooling units are used during normal operatio n (refer to Section 9.4.5). The fans not in use can be started from the contr ol room to verify readiness. A test signal is used to demonstrate proper fan starting.
6.2.2.3.18.1 Anti-Reverse Rotation Device Tests The air is discharged through the ducting to the annular ring where the air is distributed
to various compartments and areas. The return air to the fan coolers is taken at
elevation 140 feet where the containment fan coolers are located. After a LOCA, the
pressure rise at the upper elevation is relatively slow and, as a result, the pressure
difference that is expected between the inlet and outlet of the containment fan cooler
unit is extremely small. A value of 7 psi pressure difference was chosen as a
conservative design limit.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-41 Revision 23 December 2016 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED
To demonstrate the adequacy of the ARRD design, dynamic tests were performed to demonstrate that the device will withstand the dynamic loads associated with a 7 psi pressure differential. Static tests were performed to show that the subsequent transient differential pressure can react on the ARRD without failure. The results of these tests and subsequent stress analyses are summarized below:
(1) The ARRD was statically loaded to 3828 ft-Ib driven torque and 2400 ft-Ib reverse rotation torque with no failure. Spin testing was performed to
validate retraction of the pawls. T he ARRD showed no permanent
deformation.
(2) The ARRD was tested to validate i ts ability to react to reverse rotation of the fan shaft under dynamic loading conditions. The dynamic load test conditions include the following: 317 f t-lb of torque, 3 degrees total rotation, 0.391 radians/second angular velocity at end of rotation.
From these results, it was concluded that the ARRD will withstand the load imposed by a 7 psi pressure differential and the design of the ARRD is adequate for the intended use.
6.2.2.3.18.2 Containment Fan Cooler Cooling Coil Test Summary HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED
Plate-finned cooling coils are an integral part of the CFCS. These heat exchangers remove sensible heat during normal operati on, but become condensers in the post-accident environment. Because of limited experimental information concerning the performance of plate-finned cooling coils oper ating in a condensing environment in the presence of a noncondensible (air), a demonstration test was undertaken.
The test method was to subject a scaled coil to a parametric test. These parameters were (a) containment pressure (with correspondi ng steam density and temperature), (b) air flowrate, (c) cooling water fl owrate, (d) cooling water temperature, and (e) entrained water content. Each parametric test condition was then used as input to
the HECO computer program used in coil sel ections. The results of the test and the computer program predictions were compared.
In all cases, the measured heat transfer rate was greater than that predicted by the
HECO code. The range of parameter variations was selected to be consistent with the design points of the containment fan cooling coils contained in actual plants. It is
apparent that for this specific type of heat exchanger, functioning in the range of environments tested, no moisture separator is needed to protect the coils from
excessive waterlogging due to entrained spray droplets.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-42 Revision 23 December 2016 The extension of the test to full-size units is m erely an increase in component size and total flow quantities, but not a change in controlling parameters. It is concluded that the test demonstrates that the computer code used to select cooling coil design is valid in
defining the heat removal rates of plate-finned tube cooling coil assemb lies of the CFCS. Therefore, these tests demonstrate that fan cooler designs, which are selected
by this computer program, will perform as required in the post-accident containment environment.
6.2.2.3.19 General Design Criterion 60, 1967 - Testing of Containment Spray Systems The CSS is designed so that component surveillance can be performed periodically to
demonstrate system readiness. Containme nt spray valve alignment is periodically verified.
The containment spray pumps are tested individually by manually shutting the spray
header isolation valves, manually opening the RWST test return line isolation valves, and manually starting the pumps. Pump differential pressure can be used to verify
pump performance.
Periodic testing of the containment spray nozzles, as required by Technical
Specifications, will ensure that nozzles are unobstructed. Each containment spray
header can be tested individually by connecting an air source to the normally capped
flange connection on the spray pump discharge header, shutting the manual spray
header isolation valve, and opening the air test line isolation valve and the motor-operated spray header isolation valve. Individual nozz les can then be checked for proper performance by streamers, which indicate unobstructed air flow.
6.2.2.3.20 General Design Criterion 61, 1967 - Testing of Operation Sequence of Containment Pressure Reducing Systems The CFCS is periodically tested to verify the CFCUs actuate properly upon receipt of an actual or simulated actuation signal and transfer to the standby power supply in the
required timeframe.
The aim of the periodic CSS testing is to:
(1) Verify the proper sequencing of valves and pumps on initiation of the containment spray signal and demonstrate the proper operation of
remotely operated valves. A spray test interlock prevents accidental
actuation of containment spray during testing.
(2) Verify the operation of the spray pumps; each pump will be run at minimum flow and the flow directed through the normal path back to the
RWST.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-43 Revision 23 December 2016 Testing of the S and P actuation signals is addressed in Section 7.3.4.1.5.2.
Component actuation logic, transfer to the standby power supply, and component
capability are governed and tested in accordance with the Technical Specifications.
6.2.2.3.21 General Design Criterion 62, 1967 - Inspection of Air Cleanup Systems
During periodic tests, the CFCS and CSS equipment are inspected visually. Leaking seals, packing, or flanges are corrected to eliminate the leak. Valves and pumps are operated and inspected after every maintenance activity to ensure proper operation.
Visual inspection of major components of th e CFCS is regularly performed. The major components inspected include access doors to the fan cooler units, the fan cooler
housings, the CFCS duct work, cooling coils, and dampers. Therefore, the CFCS and
CSS are designed with provisions to allow for physical inspection of all critical components.
6.2.2.3.22 General Design Criterion 63, 1967 - Testing of Air Cleanup Systems Components Periodic testing of the CFCS is performed in accordance with the Technical Specifications. The testing performed on the CFCS includes operation of the fans and
ensuring that equipment in the CFCS actuates upon receipt of actual or simulated
actuation signals.
Provisions are available for periodic testing of active CSS components to ensure
operability and required functional performance is maintained (refer to Section
6.2.2.3.19).
Therefore, provisions are provided for testing of the active components of the CFCS and CSS. 6.2.2.3.23 General Design Criterion 64, 1967 - Testing of Air Cleanup Systems The CSS, in conjunction with the SAS, serves as the air cleanup system. System water flow can be tested through the test lines permitting the test of the system up to the
containment isolation valves. Periodically, a sample of the sodium hydroxide solution is
tested to assure proper concentration. Testing of the sodium hydroxide additive
solution is discussed in Section 6.2.3.3.14. The fan coolers are normally in use and, therefore, the readiness of the system is verified.
6.2.2.3.24 General Design Criterion 65, 1967 - Testing of Operational Sequence of Air Cleanup Systems The testing of the operational sequences of the CFCS and CSS are performed in accordance with the Technical Specifications.
The surveillance tests include verifying that the containment spray pumps, automatic valves, and fan cooler units actuate
properly upon receipt of an actual or simulated actuation signal and transfer to the DCPP UNITS 1 &
2 FSAR UPDATE 6.2-44 Revision 23 December 2016 standby power supply in the required timeframe; therefore, provisions for testing of the operational sequence of the CFCS and CSS is provided.
6.2.2.3.25 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment As discussed in Section 6.2.3, the functional performance requirements of the CFCS and CSS, in conjunction with the SAS, are to ensure that the offsite radiological
exposures resulting from a LOCA are within the limits of 10 CFR Part 100. Refer to
Section 15.5 for a discussion of offsite radiological exposures resulting from a LOCA.
6.2.2.3.26 10 CFR 50.49 - Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants The CFCS and CSS meet the requirements of 10 CFR 50.49 as described in the DCPP EQ Program. CFCS and CSS 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. 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 CFCU motors, valves, pressure sensors, and flow transmitters and are listed on the EQ Master List. The
containment spray pump motors are not located within a harsh environment area and
are therefore not environmentally qualified.
6.2.2.3.26.1 Containment Fan Cooler System
The containment fan cooler motor insulation is Class F (National Electrical Manufacturers Association rated hot spot temperature of 155°C). At DBA conditions, the motor insulation hot spot temperature is not expected to exceed 122°C. The motors have 2300-V insulation for 460-V service that provides additional insulation margin. The insulation is impregnated and coated to give a homogeneous insulation system that is highly moisture resistant. Internal leads and terminal box-motor interconnections are given special design consideration to ensure that the level of insulation exceeds that of the service voltage for the motor. The motor ball bearings are lubricated with high-temperature grease.
The motor heat exchanger housings enclose the major functional element of the motor and limit exposure to the environment that would exist in the containment under post-accident conditions.
The containment fan cooler motor heat exchanger is designed to maintain the environment of the fan cooler motor within an acceptable range during normal and post-accident operation.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-45 Revision 23 December 2016 The design of the motor and heat exchan ger housing seals out the post-accident environment to minimize the amount of mois ture entering the motor windings. In addition, any moisture entering the motor housing will condense on the motor heat
exchanger. The chief attribute of this design approach is that it ensures that the internal motor parts are maintained in a "usual service condition" despite the hostile
environment outside the motor.
The motors are designed for Class F temperature in normal operation, which is
consistent with the 40-year plant life requirement. During post-accident operation, the motor heat exchanger keeps winding temperatures below the 155°C insulation hot spot
temperature rating.
Qualification of the fan cooler motors is described in PG&E EQ File IH-05 (Reference
39). The bases for motor temperature rise calculations are complete engineering tests of typical motors, which provide winding temperature rises and heat losses. These data are used to validate equations that predict temperature rises and heat losses for various
design parameters and service conditions. A number of machines are tested to permit
computer interpolation of each loss curve.
The qualification (i.e., tested) hot spot winding temperatures are 105°C (Normal), 122°C (DBA), and 97°C (Post DBA). The calculated total (i.e., Maximum) winding temperature values are 102°C (Normal), 111°C (DBA), and 92°C (Post DBA).
These results confirm the ability of the motor heat exchanger to maintain winding hot-spot temperatures below the qualification level temperatures established in WCAP-7829 (Reference 4 of EQ File IH-05).
HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED.
Tests indicate that the insulation system will survive direct exposure to the postulated
post-accident environment. Hence, the heat exchanger system used to cool the
windings provides an additional margin of safety. In addition, it should be noted that at the time of the postulated incident, the fan motor would be started if not already running, and its internal temperature, being higher than the ambient, would tend to drive any
moisture present out of the windings.
The heat exchanger was designed using a very conservative fouling factor. However, if surface fouling reduces the capability of the heat exchanger by one-half, the motor would still have a normal life expectancy, even under postulated accident conditions.
To prove the effectiveness of the heat exchanger in inhibiting large quantities of the
steam-air mixture from imp inging on the winding an d bearings, several full-scale motor tests were performed at representative accident conditions. The tests exposed the
motors to a steam-air mixture as well as boric acid and alkaline spray at 80 psig and saturated temperature conditions.
Insulation resistance, winding and bearing
temperatures, relative humidity, voltage and current, as well as heat exchanger water temperature and flow, were recorded periodically during the test. Following the test, the motors were disassembled and inspected and tested to further ensure that the units had performed as designed. In all tests, the motor unit performed satisfactorily.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-46 Revision 23 December 2016 The bearings are designed to perform in the ambient temperature conditions resulting from the postulated incident. It sh ould be noted, however, that the interior bearing housing details are cooled by the heat exchanger, thus providing an extra margin of
assurance. In addition, separate tests were performed on bearings mounted within a test rig. These bearings were directly exposed to the immediate accident environment including temperature, pressure, steam-air mixture, and chemic al sprays. The bearing ran continuously for 22 months without failure.
In all tests, bearings were lubricated with fully irradiated grease prior to testing.
To further ensure motor insulation effectiveness (thermalastic epoxy), a separate motor
test was conducted by Westinghouse in accordance with IEEE-334-1971 (Reference
- 47) without the heat exchanger attached to the motor. The test was completed satisfactorily.
6.2.2.3.26.2 Motor Unit Testing HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED.
Tests were conducted (Reference 39) to demonstrate the effectiveness of a heat exchanger assembly in isolating motor windings from the steam and chemistry of the post-LOCA environment. Additional tests were conducted in 1971 to comply with provisions of IEEE 334-1971. These tests also qualified design features not included in
the original motor. Steam exposure tests per IEEE 334-1971 were performed on the same motor with and without the heat exchanger to qualify it for both types of
application.
Objectives given particular attention in the current tests to meet proposals of IEEE 334-
1971 included:
(1) Aging of all samples to full service life prior to exposure to simulated DBA conditions (2) Vibration of thermally aged mo dels prior to steam exposure (3) Change of facilities to provide fast pressure transients to simulate accident conditions during the initial transient (4) Performance of five pressure transients and exposure to a saturated steam environment for 7 days on the prototype in accordance with the
DBA simulation model (5) Comparisons between insulatio n samples subjected to combined environment and irradiation and those exposed sequentially to thermal aging, irradiation, moisture, and voltage (6) Irradiation of all lubricants to 2 x 10 8 rads before use DCPP UNITS 1 &
2 FSAR UPDATE 6.2-47 Revision 23 December 2016 (7) Destructive tests of statistically selected insulation samples to measure degradation caused by environments including irradiation, with various combinations and durations
The tests in this series qualified all motor m aterials and design features for the conditions and duration of the test.
6.2.2.3.26.3 Containment Fan Cooler Motor Insulation Irradiation Testing HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED.
The testing program on the effects of radiation on the WF-SAC "Thermalastic" Epoxy insulation system used in the fan cooler motors has been completed. In these tests, irradiation of form-wound motor coil sections was accomplished up to exposure levels exceeding that calculated for the design basis LOCA. Three coil samples received the
following treatment sequence: irradiation, vibration test, high-potential test, and
vibration test. Six of the nine coil samples received high-potential and breakdown
voltage tests. All coil samples passed the high-potential tests. The breakdown voltage levels of all coils were well in excess of those required by the design and clearly indicate that the fan cooler motor insulation system will perform satisfactorily following exposure to the radiation levels calculated for the DBA.
6.2.2.3.26.4 Containment Fan Cooler Motor Lubricant Irradiation Testing HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED.
This section summarizes the results of tes ts performed on samples of unirradiated and irradiated Westinghouse Style No. 773A773G05 (Chevron SRI) lubricant, which is used in the fan cooler fan bearing as well as in the motor bearing. The results of these tests
indicate that the shear stability or consistency of the grease is increased by irradiation levels anticipated in the containment following a DBA. The consistency of the grease
following irradiation remained wit hin the most commonl y recommended consistency for ball bearing application (NLGI No. 2).
The purpose of this test program was to establish the effect of irradiation on the bearing
lubrication used on both the fan cooler motor and fan bearings. The maximum
calculated 1-year integrated dose for the bearing lubricant, using the DBA with no credit
for fission product removal from the contai nment atmosphere other than by natural decay, is 1.5 x 10 8 rads. The fan and motor bearings would receive a lesser exposure due to self-shielding effects of the motor housing bearing seals and bearing pillow blocks.
Samples of the lubricant were placed in a vented 1.5-x 12 inch aluminum tube. The tube was then placed adjacent to a 34 kilo-curie Cobalt 60 source and irradiated for DCPP UNITS 1 &
2 FSAR UPDATE 6.2-48 Revision 23 December 2016 79 hours9.143519e-4 days <br />0.0219 hours <br />1.306217e-4 weeks <br />3.00595e-5 months <br />. Dosimetry measurements were made at various locations in the tube using Dupont light blue calibration paper 300 MS-C, No. CB-91639.
Following exposures to average levels of 1.2 x 10 8 , 1.5 x 10 8 , and 1.8 x 10 8 rads, the irradiated grease along with unirradiated grease taken from the same supply were subjected to the Micro-Cone Penetration Test using standard apparatus conforming to
ASTM D1403-56T.
The results of the penetration test indicate that as exposure increased, the grease underwent a change in thickness function to the point that, at 1.8 x 10 8 rads, sufficient change had taken place to cause the grease to increase in consistency to an NLGI No. 2 rating, as the grease was "worked" or sheared rather than decreased as in
the unirradiated grease. The most comm only used greases, for ball bearing applications such as those in the fan cooler, have consistencies ranging between
NLGI No. 1 and No. 3.
Based on the tests results from irradiation and ASTM Micro-Cone penetration measurements, the containment fan cooler bearing lubricant, Westinghouse Style No. 773A773G05 (Chevron SRI) undergoes no significant change in properties, as
measured in terms of consistency.
6.2.2.3.27 10 CFR 50.55a(f) - Inservice Testing The CFCS and CSS meet the inservice testing requirements as described in the DCPP IST Program Plan.
6.2.2.3.28 10 CFR 50.55a(g) - Inservice Inspection Design provisions have been made, to the extent practicable, to facilitate access for
periodic visual inspection of all important components of the CFCS. Testing of any
components, after maintenance or as a par t of a periodic inspection program, may be performed at any time, since the CFCS un its are in operation on an essentially continuous schedule during normal plant operation. The inservice inspection
requirements for the CFCS and CSS are contained in the DCPP ISI Program Plan.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-49 Revision 23 December 2016 6.2.2.3.29 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 CFCS and CSS post-accident instrumentation for meeting Regulatory Guide 1.97, Revision 3, requirements consist of containment isolation valve position indication, containment spray flow indication, containment pressure indication, heat removal by
containment fan heat removal system through: CFCU motor speed indication, CCW flow
indication, containment pressure indication, and containment atmosphere temperature
indication (refer to Table 7.5-6). The instrumentation described above is directly
monitored in the control room.
Based on the above discussion, the CFCS and CSS instrumentation satisfy the
requirements of Regulatory Guide 1.97, Revision 3, as described in Table 7.5-6.
6.2.2.3.30 NUREG-0737 (Items II.F.1, III.D.
1.1), November 1980 - Clarification of TMI Action Plan Requirements Item II.F.1 - Additional Accident Monitoring Instrumentation:
Position (4) - Instrumentation to continuous ly monitor and record the containment pressure is provided in the control room (refer to Sections 6.2.2.3.3 and 7.5.2.1.2). The wide-range containment pressure indicators satisfy the measurement and indication capability requirements of NUREG-0737, November 1980, Item II.F.1, Position (4).
Instrument ranges and accuracies are provided in Table 7.5-4.
Item III.D.1.1 - Integrity of Systems Outside Containment Likely to Contain Radioactive Material for Pressurized-Water Reactors and Boiling-Water Reactors:
The pressure-containing portions of the CSS are tested periodically to check for leakage. This testing includes the portions of the system that would circulate
radioactive water from the containment sump, if recirculation spray is required.
6.2.2.3.31 Generic Letter 89-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance The CSS PG&E Design Class I 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.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-50 Revision 23 December 2016 6.2.2.3.32 Generic Letter 97-04, October 1997
- Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps
The containment spray pumps have been evaluated in accordance with Generic Letter
97-04, October 1997 and adequate NPSH is shown to be available. Refer to Section
6.3 for the evaluation of the ECCS pumps.
6.2.2.4 Tests and Inspections Refer to Sections 6.2.2.3.15, 6.2.2.3.17 through 6.2.2.3.23, 6.2.2.3.26, 6.2.2.3.27, and 6.2.2.3.28 for details regarding tests and inspections of the CHRS.
6.2.2.4.1 Preoperational Testing HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED.
The aim of preoperational CSS testing was to:
(1) Demonstrate that the system is adequate to meet the design pressure conditions. Outside containment piping welds were subjected to
radiographic inspection and/or partial hydrotesting; inside the containment
the spray header welds were subjected to 100 percent radiographic
inspection.
(2) Demonstrate that the spray nozzles in the containment spray header are clear of obstructions by passing air through the test connections.
(3) Verify that the proper sequencing of valves and pumps occurs on initiation of the containment spray signal and demonstrate the proper operation of remotely operated valves.
(4) Verify the operation of the spray pumps; each pump is run at minimum flow and the flow directed through the normal path back to the RWST.
During this time, the minimum flow is adjusted to that required for routine testing.
Each fan cooler unit was tested after installation for proper flow and distribution through
the duct distribution system.
6.2.2.4.2 Periodic Testing Periodic testing of the CHRS is discussed in Sections 6.2.2.3.15, 6.2.2.3.18 through
6.2.2.3.20, 6.2.2.3.22, 6.2.2.3.23, and 6.2.2.3.26.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-51 Revision 23 December 2016 6.2.2.5 Instrumentation Applications Refer to Section 6.2.2.3.4 for the instrumentation requirements related to the CHRS.
6.2.2.6 Materials Design parameters and materials used in the construction of CHRS components are
listed in Table 6.2-26. Those parts of the system that may come in contact with borated
water or sodium hydroxide solution are made of stainless steel or a similarly
corrosion-resistant material.
6.2.3 CONTAINMENT AIR PURIFICATION AND CLEANUP SYSTEMS The containment air purification and cleanup systems are made up of the spray additive system (SAS), the containment spray system (CSS), and the containment fan cooler system (CFCS). The functional performance objectives of the containment air
purification and cleanup systems are:
(1) The SAS provides a chemical additive to the CSS to prevent the re-evolution of the iodine in the recirculated core cooling solution (also
known as sump water) following a loss-of-coolant-accident (LOCA);
(2) The CSS removes airborne iodine from the containment atmosphere following a LOCA;
(3) The CFCS provides mixing of the sprayed and unsprayed regions of the containment to improve airborne iodine removal. Mixing the
containment atmosphere maximizes the gas volume treated by the
(4) The CFCS and CSS reduce the containment ambient temperature and pressure following a LOCA or a main steam line break (MSLB). While
performing this cooling function, the containment heat removal system (CHRS) also helps limit offsite radiation levels by reducing the pressure
differential between containment and outside atmospheres, thus
reducing the driving force for leakage of fission products from the
containment atmosphere (refer to Section 6.2.2)
(5) The CFCS provides a mixed containment atmosphere for hydrogen control (refer to Section 6.2.5);
(6) The CFCS limits the containment ambient temperature during normal plant operating conditions (refer to Section 9.4.5).
The SAS boundary consists of all piping and valves between, and including, the
isolation valve from the refueling water stora ge tank (RWST), the spray additive tank DCPP UNITS 1 &
2 FSAR UPDATE 6.2-52 Revision 23 December 2016 and the spray eductors. The safety function of iodine removal occurs during operation of the containment spray system and while core coolant is retained in the containment
sump. The physical SSC design bases, testing and inspection requirements of the CSS
and CFCS are discussed in Section 6.2.2, th erefore the following section addresses the SAS SSCs only.
The two small charcoal filter units in the containment air purification system are not
classified as ESFs and are described in Se ction 9.4.5. These units are not necessary for cleanup during accident conditions.
6.2.3.1 Design Bases
6.2.3.1.1 General Design Criterion 2, 1967 - Performance Standards The SAS is designed to withstand the effects of, or is protected against, natural phenomena, such as earthquakes, winds and tornadoes, floods and tsunamis, and
other local site effects.
6.2.3.1.2 General Design Criterion 3, 1971 - Fire Protection The SAS is designed and located to minimize, consistent with other safety
requirements, the probability and effects of fires and explosions.
6.2.3.1.3 General Design Criterion 11, 1967 - Control Room The SAS is designed to support actions to maintain and control the safe operational status of the plant from the control room.
6.2.3.1.4 General Design Criterion 12, 1967 - Instrumentation and Control Systems Instrumentation and controls are provided as required to monitor and maintain SAS
variables within prescribed operating ranges.
6.2.3.1.5 General Design Criterion 19, 1971 - Control Room The SAS, in conjunction with the iodine removal function of the CSS and CFCS, is designed to support radiation protection to permit access and occupancy of the control
room under accident conditions without personnel receiving radiation in excess of 5 rem
whole body, or its equivalent to any part of the body, for the duration of the accident.
6.2.3.1.6 General Design Criterion 21, 1967 - Single Failure Definition The SAS is designed to tolerate a single failure during the period of recovery following
an accident without loss of its protective function, including multiple failures resulting
from a single event, which is treated as a single failure.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-53 Revision 23 December 2016 6.2.3.1.7 General Design Criterion 37, 1967 - Engineered Safety Features Basis for Design The SAS, in conjunction with the iodine removal function of the CSS, is designed to prevent the re-evolution of iodine in the core cooli ng solution and provide back-up to the safety function provided by the core design, the reactor coolant pressure boundary, and
their protection systems.
The CFCS is designed to ensure mixing of the sprayed and unsprayed regions of the containment atmosphere, to improve airborne fission product removal following any size
reactor coolant pressure boundary break.
6.2.3.1.8 General Design Criterion 38, 1967 - Reliability and Testability of Engineered Safety Features The SAS is designed to provide high functional reliability and ready testability.
6.2.3.1.9 General Design Criterion 40, 1967 - Missile Protection The SAS is designed to be protected against dynamic effects and missiles that might
result from plant equipment failures.
6.2.3.1.10 General Design Criterion 41, 1967 - Engineered Safety Features Performance Capability The SAS is designed to provide sufficient performance capability to accommodate a partial loss of installed capacity, such as a single failure of an active component, and
still perform its required safety function.
6.2.3.1.11 General Design Criterion 42, 1967 - Engineered Safety Features Components Capability In support of the iodine removal function, the CSS and CFCS are designed so that the
capability of each component and system to perform its required function is not impaired by the effects of a LOCA.
6.2.3.1.12 General Design Criterion 62, 1967 - Inspection of Air Cleanup Systems The SAS is designed to facilitate physical inspection of critical parts required for
operation.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-54 Revision 23 December 2016 6.2.3.1.13 General Design Criterion 63, 1967 - Testing of Air Cleanup Systems Components The SAS is designed so that active components can be tested periodically for
operability and required functional performance.
6.2.3.1.14 General Design Criterion 64, 1967 - Testing of Air Cleanup Systems The SAS is designed for in-situ periodic testing and surveillance.
6.2.3.1.15 General Design Criterion 65, 1967 - Testing of Operational Sequence of Air Cleanup Systems
The SAS is designed with the capability to test, under conditions as close to design as
practical, the full operational sequence that would bring the SAS into action.
6.2.3.1.16 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment The SAS, in conjunction with the iodine removal function of the CSS and CFCS, is designed with provisions for maintaining cont rol over the plants radioactive gaseous effluents.
6.2.3.1.17 10 CFR 50.55a(f) - Inservice Testing Requirements
American Society of Mechanical Engineers (ASME) code components within the SAS are tested to the requirements of 10 CFR 50.55a(f)(4) and 10 CFR 50.55a(f)(5) to the extent practical.
6.2.3.1.18 10 CFR 50.55a(g) - Inservice Inspection Requirements ASME code components within the SAS are inspected to the requirements of
10 CFR 50.55a(g)(4) and 10 CFR 50.55a(g)(5) to the extent practical.
6.2.3.1.19 Generic Letter 89-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance The SAS PG&E Design Class I motor-operated valves (MOVs) meet the requirements
of Generic Letter 89-10, June 1989, and associated Generic Letter 96-05, September
1996. 6.2.3.2 System Description 6.2.3.2.1 General Description
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-55 Revision 23 December 2016 The SAS is designed to add sodium hydroxide to the containment spray water to enhance the absorption of elemental iodine from the containment atmosphere and to
retain the iodine in the containment sump water in nonvolatile forms.
According to the known behavior of elemental iodine in highly dilute solutions, the
hydrolysis reaction, described by the relationships:
++IHIOOHI 2 (6.2-8) proceeds nearly to completion (Reference 4) at pH values between 8 and 9.5. The iodine form is highly soluble, and HIO readily undergoes additional reactions to form
iodate.
The overall iodate reaction is:
(6.2-9)
Values of the spray removal half-life of the molecular iodine in a typical containment are on the order of minutes, or less. This makes the spray system a very efficient fission product removal system in comparison to such alternatives as charcoal filtration
systems.
The SAS, as shown schematically in Figure 3.2-12, consists of the spray additive tank, eductors, valves, and connecting piping. The design parameters are presented in
Table 6.2-29. Applicable codes and standard are given in Table 6.2-30.
The CFCS ensures appropriate mixing of the containment atmosphere following a
design basis accident to maximize the containment gas volume treated by the
On actuation, approximately 4 percent of each spray pump discharge flow will be
diverted through each spray additive eductor to draw sodium hydroxide from the tank.
This sodium hydroxide solution will then mix with the liquid entering the suction line of the pumps to give a solution suitable for removal of iodine from the containment
atmosphere.
During the injection phase, the emergency core cooling pumps will inject borated water
drawn from the RWST into the reactor and containment. Since these flowpaths will not
inject sodium hydroxide, the ratio of the total volume injected by all pumps to the volume
injected by the spray pumps will determine the change in sodium hydroxide
concentration during the injection phase. The total volume of water in the sump
includes the total amount contained in the primary coolant and accumulators that could
be released to the containment recirculation sump at the start of the LOCA.
++HIO 3 5HI 3H 2 O 3I 2 DCPP UNITS 1 &
2 FSAR UPDATE 6.2-56 Revision 23 December 2016 6.2.3.2.2 Component Descriptions 6.2.3.2.2.1 Spray Additive Tank The stainless steel tank will contain sufficient 30 to 32 weight percent sodium hydroxide
solution to bring the containment sump fluid to a minimum pH of 8.0 on mixing with the
borated water from the RWST, the accumulators, and reactor coolant following a large
break LOCA. This will ensure continued iodine removal and retention effectiveness of
the containment sump water during the recirculation phase.
6.2.3.2.2.2 Spray Additive Eductors Sodium hydroxide will be added to the spray liquid by a liquid jet eductor, a device which uses kinetic energy of a pressurized liquid to entrain another liquid. The pressurized liquid in this case is the spray pump discharge used to entrain the sodium
hydroxide solution, which is then discharged back into the suction of the spray pumps.
On actuation, approximately 4 percent of each spray pump discharge flow is diverted
through each spray additive eductor to draw sodium hydroxide solution from the spray
additive tank. An eductor motive flowrate of 104 gpm and an eductor suction flowrate of
35 gpm results in a spray solution pH of greater than or equal to 9.5 (refer to design case in Table 6.2-36).
6.2.3.2.2.3 Spray Nozzles A description of the containment spray nozzles is provided in Section 6.2.2.
6.2.3.2.2.4 Piping and Valves The piping and valves for the SAS are designed for 240 psig at 200 degrees F.
6.2.3.2.2.5 Electrical Supply Details of the Class 1E power sources are discussed in Chapter 8.
6.2.3.3 Safety Evaluation
6.2.3.3.1 General Design Criterion 2, 1967 - Performance Standards
The SAS is located within the auxiliary building, which is a PG&E Design Class I structure (refer to Section 3.8). The auxil iary building is designed to withstand the effects of winds and tornadoes (refer to Section 3.3), floods and tsunamis (refer to
Section 3.4), external missiles (refer to Section 3.5), and earthquakes (refer to Section
3.7) therefore protecting the SAS, ensuring its design functions will be performed.
The SAS is designed to accommodate the DE within applicable code stress limits and to
withstand the DDE or HE without rupture or loss of function.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-57 Revision 23 December 2016 6.2.3.3.2 General Design Criterion 3, 1971 - Fire Protection The SAS is designed to the fire protection guidelines of Branch Technical Position APCSB 9.5-1 (refer to Appen dix 9.5B Table B-1).
6.2.3.3.3 General Design Criterion 11, 1967 - Control Room
Each of the system's MOVs can be operated individually by switches in the control
room. Valve position indication on the same control room panel is used to verify valve
operability.
The spray additive tank level instruments provide two alarms to announce when the solution in the tank has dropped below a level approaching the Technical Specification minimum requirements.
6.2.3.3.4 General Design Criterion 12, 1967 - Instrumentation and Control Systems Analog and logic channels employed for initiation of SAS operation are discussed in Section 7.3. All alarms will be annunciated in the control room.
The SAS will be actuated by a "P" signal initiated either manually from the control room
or on coincidence of two sets of two-out-of-four high-high containment pressure signals.
Coincidence of "S" and "P" signals will start the containment spray pumps and open the
discharge valves to the spray headers. The "P" signal alone will open the valves associated with the spray additive tank.
A locally mounted indicator on the nitrogen line monitors the spray additive tank
pressure while adding nitrogen and during periodic inspections.
A flow element is located in the discharge line from the spray additive tank. Flow indication is provided in the control room.
For the spray additive tank, two separate instruments are provided: one to supply
readout in the control room and the other to provide local indication.
6.2.3.3.5 General Design Criterion 19, 1971 - Control Room The CSS provides iodine removal from the containment atmosphere. The CFCS provides mixing of the containment atmosph ere to maximize iodine removal. The SAS increases pH in the containment recirculation sump water. The SAS, CSS and CFCS
work in conjunction to ensure that radiological exp osures for control room personnel are within 5 rem whole body, refer to Section 15.5 for radiologica l consequences of plant accidents.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-58 Revision 23 December 2016 6.2.3.3.6 General Design Criterion 21, 1967 - Single Failure Definition
A failure analysis was conducted on the active components, MOVs 8994A/B, of the SAS
to show that failure of any single active component will not prevent the design function
from being fulfilled. This analysis is summarized in Table 6.2-38.
MOV 8992 in the spray additive line, which is required only in the short term, is not
included in the single failure analysis of Table 6.2-38 because it is not an active component. It performs no active function, is normally open, receives a "P" signal to
ensure positive opening, and is designed to fail as is, i.e., in the open position.
In addition, during power operation, power is removed from the single inline SAS outlet MOV 8992 at the circuit breaker at the motor control center with the valve in the open position to prevent a single active failure.
The single failure analysis for the CSS is given in Section 6.2.2.
6.2.3.3.7 General Design Criterion 37, 1967 - Engineered Safety Features Basis for Design The design basis of the system is two-fold:
(1) Sufficient sodium hydroxide must be added to the containment spray water to ensure rapid absorption by the spray of elemental iodine present
in the containment atmosphere following a LOCA.
(2) During the injection phase operation of the spray pumps, a sufficient amount of sodium hydroxide must be carried to the containment sump
water via the containment spray to ensure retention of the iodine in the
sump solution.
Performance of the CSS as an iodine removal mechanism is conservatively evaluated
at the containment design temperature and pressure. Since this peak pressure
condition is expected to exist for a few minutes at most, and mass transfer parameters
and spray flowrate improve with decreasing pressure, an appreciable margin is added to the evaluation. The design case removal constant for the CSS (s) provided in Table 6.2-36 was calculated by applying the model derived in Section 6.2.3.3.7.3 at this back pressure condition to the sprayed portion of the containment volume.
The CSS, by virtue of the large contact surface area provided between the droplets and
the containment atmosphere, affords an excellent means of absorbing radioactive
iodine released as a consequence of a LOC A. Sodium hydroxide is added to the spray fluid to increase the absorption of iodine in the spray to the point where the rate of
absorption is largely limited by the transfer rate through the gas film surrounding the drops. Reference 5 describes in detail the analytical and experimental basis for the DCPP UNITS 1 &
2 FSAR UPDATE 6.2-59 Revision 23 December 2016 above containment atmosphere iodine removal mechanism. The approach used is summarized below.
The SAS is dependent upon the CSS for operation. The CSS can function with one
spray train operating (abnormal operating mode) or with both spray trains operating (normal operating mode). In addition, the operation of one or both ECCS trains affects
the rate of withdrawal of water from the RWST, the duration of the spray injection phase, and thus the amount of sodium hydroxide added to the containment.
Spray iodine removal performance has been evaluated for the design case, a
double-ended LOCA, assuming that:
(1) Only one-out-of-two spray pumps operate (one spray train operating)
(2) The ECCS operates at its maximum capacity (two ECCS trains operating)
(3) Borated water is retained in the RWST for the exclusive use of the spray during the first part of the ECCS recirculation phase
The second assumption maximizes ECCS flow from the RWST. Overall, the first two assumptions give the most conservative prediction of sodium hydroxide introduction into
the containment (minimum containment recirculation sump pH). The third assumption
ensures that sufficient sodium hydroxide solution is added to provide for a sump pH of
8.0 or greater when spray injection terminates. Figure 6.2-15 presents various resulting
sump pH versus time curves.
The variation of sump pH with time after the accident is shown for various cases in Figure 6.2-15 (Reference 51). At the time spray injection terminates, the sump water
has reached an equilibrium pH of at least 8.0.
The sump pH will remain at the same pH after the spray injection phase because no additional water is added to the sump.
Any reevolution of dissolved iodine from the sump to the containment atmosphere depends upon the concentration gradient between the liquid and vapor phases. The
equilibrium between these iodine concentrations is given by the partition coefficient, H, and is a function of concentration, pH, and temperature. The partition coefficient at pH
8.0 exceeds the value of approximately 4 x 10 3 required to maintain a decontamination factor of 100 in the containment atmosphere for sump temperature above 120° F, and
thus a containment atmosphere decontamination factor of 100 or greater can be
expected. Figure 6.2-16 presents equilibriu m elemental iodine partition coefficients in the containment at various temperatures for t he minimum sump pH case. The equations given by Eggleton (Reference 8) were used to determine the partition
coefficients. Although the iodate reaction is expected to contribute significantly to the
iodine partition at high sump pH values, it has been neglected in these calculations in
the interest of conservatism.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-60 Revision 23 December 2016 Approximately 17 percent of the containment free air volume is not reached by the spray. The values listed below were used to estimate the total unsprayed volume in the containment.
(1) Containment radius 70 ft Height between operating deck and spring line 91 ft Approximate deck area covered only by grating 2,990 ft 2 Height between elevation 91 feet and deck 49 ft Average fall height below deck through grating 12 ft (2) These data result in the following volumes:
Volume in dome 717,000 ft 3 Volume in cylinder above deck 1,400,000 ft 3 Occupied volume above deck -95,000 ft 3 Sprayed volume below deck 36,000 ft 3 Sprayed refueling cavity volume 45,000 ft 3 Total sprayed volume 2,103,000 ft 3 Total free volume 2,550,000 ft 3 Total unsprayed volume 447,000 ft 3 Percent unsprayed volume
~17 %
The calculations of thyroid exposures in Section 15.5.17.2.4 were based on the
assumption of uniform mixing in the full free v olume of the containment. As a result of the circulation of air from the unsprayed portions of the containment free volume to the sprayed areas by the CFCS, good mixing is provided. As shown in Table 6.2-26, the fan cooler unit capacity is 47,000 cfm. If a simplified two-volume model were used, in combination with assumptions that some iodine was available for leakage from the
lower containment section (unsprayed), some reduction in the effective calculated spray
removal coefficient would result.
HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED In order to evaluate these and other possible combinations of degraded performance of the spray system, a sensitivity study was performed to determine the effect of a reduced
removal coefficient on thyroid exposures. The results of this study are presented in
Figures 15.5-6, 15.5-7, and 15.5-8. As shown in these figures, both the short-term and
the long-term thyroid exposures are very insensitive to reduced spray removal
coefficient down to values as low as 10 hr
-1.
6.2.3.3.7.1 Drop Size Distribution The drop size distribution used in the analytical model is based on data obtained from
measurements of the actual size distribution from the Spraco 1713A nozzle for the
range of pressure drops encountered during operation of the spray system. A complete DCPP UNITS 1 &
2 FSAR UPDATE 6.2-61 Revision 23 December 2016 analysis of the expected drop size distributions, including a statistical analysis is contained in References 5, 6, and 33. The parameters used in applying these distributions to the calculation of the iodine removal coefficient for the DCPP units are given in Tables 6.2-29, 6.2-36, and 6.2-37.
6.2.3.3.7.2 Condensation As the spray solution enters the high-temperature containment atmosphere, steam will
condense on the spray drops. The amount of condensation is calculated by an enthalpy
balance on the drop:
mh + m c h g= m'h f (6.2-10)
where:
m and m' = mass of the drop before and after condensation
m c = mass of condensate, lb h = initial enthalpy of the drop, Btu/lb
h g and h f = saturation enthalpy of water vapor and liquid, respectively, Btu/lb
The increase in each drop diameter in the distribution is, therefore, given by:
(6.2-11)
where:
v f = specific volume of liquid at saturation, ft 3/lb v = specific volume of the drop before condensation, ft 3/lb h fg = latent heat of evaporation, Btu/lb h g = enthalpy of steam at saturation, Btu/lb d = drop diameter before condensation, cm d' = drop diameter after condensation, cm
The increase in drop size due to condensation is expected to be complete in a few feet
of fall for the majority of drop sizes in the distribution. More detailed calculations by Parsly show that even for the largest drops in the distribution, thermal equilibrium is
reached in less than half the available drop fall height.
6.2.3.3.7.3 Mass Transfer Model The basic equation for the iodine concentration in the containment atmosphere is
derived from a material balance of the elemental iodine in the containment. The iodine
removal by the spray system may be expressed by:
) h h h ( )v ( )d d ( fg g'3 =v f DCPP UNITS 1 &
2 FSAR UPDATE 6.2-62 Revision 23 December 2016
)C(HCEF dt dC V L1 g g c= (6.2-12) where:
V c = containment free volume, cc C g = iodine concentration in the containment atmosphere, gm/cc H = iodine partition coefficient, (gm/liter of liquid)/(gm/liter of gas)
F = spray flowrate, cc/sec
The resulting change in the drop size distribution is taken into consideration in the mass transfer calculations described below.
The variable E is the absorption efficiency, which may also be described as the
fractional approach to saturation:
L1 L L1 L2CCCC E*= (6.2-13) where:
C L1 = iodine concentration in the liquid entering the dispersed phase, gm/cc C L2 = iodine concentration in the liquid leaving the dispersed phase, gm/cc C L*= equilibrium iodine concentration in the liquid, gm/cc This absorption efficiency is calculated from the time-dependent mass transfer model
suggested by L. F. Parsly (Reference 7).
The absorption efficiency calculated is a function of drop size, and the removal constant s , in reciprocal hours, for the entire spray is, therefore, obtained by an appropriate summation over all drop size groups:
cii n1i s VHFE== (6.2-14) A further discussion of drop size distribution, drop trajectories, drop coalescence and
mass transfer modeling is presented in References 5, 6, and 33.
6.2.3.3.7.4 Experimental Verification of Models HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED
The ability of the model described to give conservative estimates of actual spray performance was demonstrated in test runs m ade at Oak Ridge National Laboratory (ORNL) and Battelle Pacific Northwest Laboratory. The results of these tests DCPP UNITS 1 &
2 FSAR UPDATE 6.2-63 Revision 23 December 2016 (Reference 5), shown in Figure 6.2-14 for Run A6, verified that the spray removal model used is conservative in all cases.
6.2.3.3.8 General Design Criterion 38, 1967 - Reliability and Testability of Engineered Safety Features
The SAS testing and inspections are enveloped by the inservice inspection (ISI) and inservice testing (IST) programs. Pressure containing p ortions of the CSS and SAS are inspected in accordance with ASME BPVC,Section XI, as required by the Technical
Specifications and the Inservice Inspection Program Plan (Reference 38) (refer to
Section 6.2.3.3.17 for inservice testing and Section 6.2.3.3.18 for inservice inspection).
6.2.3.3.9 General Design Criterion 40, 1967 - Missile Protection The SAS is protected from internal missiles, pipe whip, and jet impingement from the rupture of any nearby high-energy line (refer to Sections 3.5.1.2 and 3.6.2.2.2).
6.2.3.3.10 General Design Criterion 41, 1967 - Engineered Safety Features Performance Capability
The SAS is designed to tolerate a single active failure without loss of protective function
and as such is designed with the redundancy to accommodate a partial loss of installed
capacity. The SAS is in use only during the injection period following a LOCA (refer to
Section 6.2.3.3.6).
6.2.3.3.11 General Design Criterion 42, 1967 - Engineered Safety Features Components Capability The CSS and CFCS component design pressure and temperature conditions in Table 6.2-26 are specified as the most severe conditions to which each component is exposed
during either normal or post-LOCA operation while maintaining the capability of the
system to provide its required design function in post-LOCA conditions (refer to Section
6.2.2). 6.2.3.3.12 General Design Criterion 62, 1967 - Inspection of Air Cleanup Systems During periodic tests, the equipment is inspe cted visually for leaks. Leaking seals, packing, or flanges are corrected to eliminate the leak. Valves and pumps are operated
and inspected after every maintenance to ensure proper operation.
All critical parts of the SAS are inspected in accordance with the plant ISI Program and
in compliance with 10 CFR 50.55a(g). Refer to Section 6.2.3.3.18 for further discussion.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-64 Revision 23 December 2016
6.2.3.3.13 General Design Criterion 63, 1967 - Testing of Air Cleanup Systems Components
Routine periodic testing of the SAS components and all necessary support systems at
power, under the conditions defined in the Technical Specifications, is performed.
Each of the SAS MOVs are tested while the pumps are shut down. During eductor
suction valve operation, the normally open spray additive tank valve is closed. Relief
valves and vacuum breakers on the sodium hydroxide tank are set and tested prior to
installation and periodically thereafter.
The purpose of the motor-operated double-disk gate valve in the spray additive line, shown in Figure 3.2-12, is to permit periodic testing of the two parallel MOVs
downstream from that valve.
6.2.3.3.14 General Design Criterion 64, 1967 - Testing of Air Cleanup Systems The concentration of the sodium hydroxide additive s olution is established at the time of initial tank fill, and then periodically checked by titration of tank samples taken from the local sample connection.
The testing of the CSS and CFCS is discussed in Section 6.2.2.
6.2.3.3.15 General Design Criterion 65, 1967 - Testing of Operational Sequence of Air Cleanup Systems Provisions are made in the circuitry of the various system alarms to test the proper operations of the alarm circuitry with a test signa l input. These circuits include the sodium hydroxide tank low-level alarms, the RWST low-level alarms, and the containment high-pressure alarms.
The SAS is tested to verify operational sequence under conditions as close to design as practical. For testing of the operational sequence required to actuate the SAS, refer to Section 7.3.4.1.5.2.
6.2.3.3.16 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment
The functional performance requirement for the SAS, in conjunction with the iodine
removal functions of the CSS and CFCS, is to ensure that offsite radiological exposures
resulting from a LOCA are within the limits of 10 CFR Part 100, refer to Section 15.5.
The use of the spray removal constant in the radiological release calculations for the LOCA is described in Section 15.5.17.2.4.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-65 Revision 23 December 2016
6.2.3.3.17 10 CFR 50.55a(f) - Inservice Testing Requirements
The IST requirements for SAS ASME code class valves are contained within the IST Program Plan.
6.2.3.3.18 10 CFR 50.55a(g) - Inservice Inspection Requirements Pressure containing portions of the CSS an d SAS are inspected in accordance with ASME BPVC,Section XI, as required by the Technical Specifications and the Inservice Inspection Program Plan (Reference 38).
6.2.3.3.19 Generic Letter 89-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance The SAS 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
6.2.3.4 Tests and Inspections HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED
The CSS was tested functionally in accordance with written procedures, as outlined in Chapter 14.
Spray pump delivered flow and head data were recorded to verify that the containment
spray pumps meet design criteria.
Spray additive eductor performance data were provided by the manufacturer based on actual tests of a similar eductor. These tests were conducted using a 1.3 specific gravity solution to verify eductor design perfo rmance. Additional manufacturer's tests were run using water so that comparative performance data were available for the two different additive solutions at eductor design conditions. Eductor performance was
checked subsequent to installation into the system. Spray additive flowrates were measured, with resulting rates in the range 31.5 to 38.5 gpm (35 gpm
+/- 10 percent) considered acceptable.
Each containment spray header was tested individua lly by connecting a source of air to the normally capped flange connection on the spray pump discharge header, shutting
the manual spray header isolation valve and opening the air test line isolation valve and the motor-operated spray header isolation valve.
Individual nozzles were checked for proper performance by streamers, which indicated unobstructed air flow.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-66 Revision 23 December 2016 The containment sump recirculation mode w as tested initially as part of a preoperational flow test under ambient conditions of the safety injection system (SIS). The purpose of the test was to demonstrate the capabi lity of appropriate subsystems to deliver fluid from the containment sump into the reactor coolant system (RCS) in the required time.
6.2.3.5 Instrumentation Applications Refer to Section 6.2.3.3.4 for instrumentation important to the operation of the SAS.
6.2.3.6 Materials The SAS design parameters and materials of construction are listed in Table 6.2-29.
Code compliance is shown in Table 6.2-30.
Parts of the system in contact with borated water, the sodium hydroxide spray additive, or mixture of the two are stainless steel or an equivalent corrosion-resistant material.
6.2.4 CONTAINMENT ISOLATION SYSTEM The containment isolation system (CIS) prev ents excessive radioactivity from passing through the containment to the atmosphere in the event of a LOCA. This is accomplished by automatically sealing the various lines through the containment walls.
6.2.4.1 Design Bases The CIS is designed to meet the containment isolation requirements of the 1971
General Design Criteria (GDC) except where specifically indicated. When deviations are noted, these cases do not meet the 1971 GDC because of commitment to design and construction prior to issuance of these criteria. Such cases do comply with the
1967 GDC, however.
6.2.4.1.1 General Design Criterion 2, 1967 - Performance Standards The components that make up the CIS are designe d to withstand the effects of, or are protected against, natural phenomena, such as earthquakes, flooding, tornados, winds, and other local site effects.
6.2.4.1.2 General Design Criterion 10, 1967 - Containment The CIS is designed, together with other engineered safety features as may be
necessary, to retain the functional capability of the containment to protect the public.
6.2.4.1.3 General Design Criterion 11, 1967 - Control Room The CIS is designed with indication and controls located in the control room as
necessary to shut down and maintain safe control of the facility.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-67 Revision 23 December 2016 6.2.4.1.4 General Design Criterion 12, 1967 - Instrumentation and Control Systems The CIS is designed with instrumentation an d controls as required to monitor and maintain variables within prescrib ed operating ranges.
6.2.4.1.5 General Design Criterion 21, 1967 - Single Failure Criterion The CIS is designed to tolerate a single failure during the period of recovery following
an accident without loss of its protective function, including multiple failures resulting
from a single event, which is treated as a single failure.
6.2.4.1.6 General Design Criterion 40, 1967 - Missile Protection The CIS is designed with protection against dynamic effects and missiles that might
result from plant equipment failures.
6.2.4.1.7 General Design Criterion 53, 1967 - Containment Isolation Valves The CIS is designed such that the component cooling water (CCW) penetration to the
excess letdown heat exchanger is provided with redundant valving and associated apparatus. The CCW penetration to the excess letdown heat exchanger does not meet GDC 57, 1971 because of commitments to design and construction made prior to the
issuance of the 1971 GDC. This penetration does comply with GDC 53, 1967.
6.2.4.1.8 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment
The CIS is designed such that the piping systems that penetrate containment are
provided with leak detection, isolation, and containme nt capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of
isolating these piping systems. These piping systems are designed with a capability to test periodically that operability of the isolation valves and associated apparatus and to
determine if valve leakage is within acceptable limits.
6.2.4.1.9 General Design Criterion 55, 1971 - Reactor Coolant Pressure Boundary Penetrating Containment The CIS is designed such that each line that is part of the reactor coolant pressure
boundary that penetrates containment is provided with containment isolation valves as
follows, unless otherwise demonstrated that the containment isolation provisions for a
specific class of lines are acceptable on another defined basis:
(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or DCPP UNITS 1 &
2 FSAR UPDATE 6.2-68 Revision 23 December 2016 (2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve is not used as the
automatic isolation valve outside containment; or
(4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve is not used as the automatic
isolation valve outside containment.
Isolation valves outside containment are located as close to the containment as
practical and upon loss of actuating power, automatic isolation valves are designed to
take the position that provides greater safety.
6.2.4.1.10 General Design Criterion 56, 1971 - Primary Containment Isolation The CIS is designed such that each line that connects directly to the containment
atmosphere and penetrates containment is provided with containment isolation valves
as follows, unless it is demonstrated that the containment isolation provisions for a
specific class of lines are acceptable on another defined basis:
(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or
(2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or
(3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve is not used as the
automatic isolation valve outside containment; or
(4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve is not used as the automatic
isolation valve outside containment.
Isolation valves outside containment are located as close to the containment as
practical and upon loss of actuating power, automatic isolation valves are designed to
take the position that provides greater safety.
6.2.4.1.11 General Design Criterion 57, 1971 - Closed System Isolation Valves The CIS is designed such that each line that penetrates containment and is neither part
of the reactor coolant pressure boundary nor connected directly to the containment
atmosphere has at least one containment isolation valve which is either automatic or DCPP UNITS 1 &
2 FSAR UPDATE 6.2-69 Revision 23 December 2016 locked closed or capable of remote-manual operation. This valve is outside containment and located as close to the containment as practical. A simple check valve
is not used as the automatic isolation valve.
Refer to Table 6.2-39 for exceptions to GDC 57, 1971.
6.2.4.1.12 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment The containment is designed as a barrier to maintain control over plant radioactive effluents, whether gaseous, liquid, or solid to meeting the radiological limits of 10 CFR Part 100. Appropriate holdup capacity is provided for retention of gaseous effluents, particularly where unfavorable environmental conditions can be expected to require operational limitations upon the release of radioactive effluents to the environment.
6.2.4.1.13 10 CFR Part 50, Appendix J, Option B - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors
The CIS is designed to enable implementation of a performance-based containment
leakage rate testing program for Type C loc al leak rate tests with approved exemptions.
6.2.4.1.14 Regulatory Guide 1.163, September 1995 - Performance-Based Containment Leak-Test Program The CIS is designed to allow the use of a performance-based leak-test program, including the leakage-rate test me thods, procedures, and analyses as required by Regulatory Guide 1.163, September 1995.
6.2.4.1.15 NUREG-0737 (Item II.E.4.2), November 198 0 - Clarification of TMI Action Plan Requirements Item II.E.4.2 - Dependabil ity of Containment Isolation:
Position (1) - The CIS is designed with diverse parameters sensed for the initiation of
containment isolation.
Position (2) - The CIS process penetrations are classified as nonessential, essential, and safety system process lines for the determination of those penetrations isolated by
a containment isolation signal.
Position (3) - The CIS nonessential systems use either manually sealed closed valves or are automatically isolated on a Phase A containment isolation signal. Additionally, essential systems are automatically isolated on a Phase B isolation signal.
Position (4) - The CIS is designed so that re-setting of a containment isolation signal will not result in the automatic re-opening of any containment isolation valves. Ganged re-DCPP UNITS 1 &
2 FSAR UPDATE 6.2-70 Revision 23 December 2016 opening cannot result from a single operator action after the containment isolation signal has been reset.
Position (5) - The CIS is designed so that the containment setpoint pressure that
initiates containment isolation for nonessential penetrations is set to the minimum
compatible with normal operating conditions with additional margin to allow for a small
pressure transient.
Position (6) - The DCPP purge system valves satisfy Branch Technical Position CSB 6-
4, September 1975.
Position (7) - The containment vent and purge isolation valves close on a high radiation
signal.
6.2.4.1.16 Generic Letter 89-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance In the CIS, PG&E Design Class I position-changeable motor-operated valves (MOVs) meet the requirements of Generic Letter 89-10, June 1989, and associated Generic
Letter 96-05, September 1996.
6.2.4.1.17 Generic Letter 96-06, September 1996 - Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions
The CIS is designed to prevent thermally induced overpressurization of isolated water-filled piping sections in containment during design-basis accidents.
6.2.4.2 System Description The CIS includes the mechanical and instrumentation fluid penetrations and associated
valves and isolation devices. These penetrations are identified in Figure 6.2-19 and
Table 6.2-39. The CIS design uses the following premises: (1) An automatic containment isolation barrier is provided by a closed system, a trip valve, or a check valve.
(2) A closed system meets the following requirements:
a) Inside the containment:
- 1. No mass transfer with either the RCS or the reactor containment interior
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-71 Revision 23 December 2016
- 3. Must withstand an external pressure and temperature that is greater than containment design pressure and temperature
- 4. Must withstand accident transient and environmental parameters
- 5. Must be protected against missiles and high-energy jets b) Outside the containment:
- 1. Does not communicate with the atmosphere outside the containment
- 3. Internal design pressure and temperature must be greater than containment design pressure and temperature
- 4. Must be protected against missiles and high-energy jets (3) A trip valve is a motor-, air-operated, or solenoid valve that moves to a preferred position upon a containment isolation signal. It can
additionally be opened or closed manually from a remote location.
(4) Lines that must remain in service subsequent to certain accidents have, as a minimum, one manual isolation valve.
(5) Lines 1 inch nominal pipe size and larger that penetrate the containment and are connected to the RCS have at least two valves
inside the containment. The valves are normally closed or have
automatic closure. For incoming lines, check valves are permitted and
are considered as an automatic barrier inside containment.
(6) All isolation valves (automatic and manual) and associated equipment are PG&E Design Class I. The is olation section piping is PG&E Design Class I, Quality/Code Class II.
6.2.4.2.1 Containment Penetration Piping Isolation Grouping The definitions used in the design bases and the physical configuration of various
systems that penetrate the containment divide themselves naturally into five groups.
These groups were used to determine the necessary valves on lines penetrating the containment. The lines and valves are shown graphically in Figure 6.2-17. PG&E has
named and defined each group as shown below. The groupings, A through E, do not
correspond to the piping code classes described in Section 3.2.2.3.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-72 Revision 23 December 2016 6.2.4.2.1.1 Group A Piping Group A piping complies with the requirements of either GDC 55, 1971 or GDC 56, 1971. Outside the containment this piping either connects directly to the atmosphere or is considered open, even though it may be physically closed. Inside the containment, it
is either part of the reactor coolant pressure boundary (RCPB), opens directly to the
containment atmosphere, or is considered open, even though it may be physically
closed. In this group, the following minimum requirements apply:
(1) Incoming Lines: One trip valve inside the containment and one trip valve outside the containment, or one check valve inside the containment and
one trip valve outside the containment
(2) Outgoing Lines: One trip valve inside the containment and one trip valve outside the containment 6.2.4.2.1.2 Group B Piping Group B piping complies with the requirements of either GDC 55, 1971 or GDC 56, 1971. Outside the containment, this piping operates in a closed system (physically
closed and PG&E Design Class I), and inside the containment it is either part of the RCPB or connects directly to the containment atmosphere. For this group the following
minimum requirements apply:
(1) Incoming Lines: One check valve inside the containment and a closed system outside the containment (2) Outgoing Lines: One trip valve inside the containment and a closed system outside the containment 6.2.4.2.1.3 Group C Piping Group C piping complies with the requirements of GDC 57, 1971 which states that
isolation valves in closed systems must be outside the containment and no simple
check valve may be used. Outside the containment, this piping connects with systems
that are either opened or closed. Inside the containment, both types of systems are separated from the RCPB and from the containment atmosphere by a membrane
barrier. For this group, the following minimum requirements apply:
(1) Incoming Lines: One t rip valve outside containment (2) Outgoing Lines: One trip valve outside containment
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-73 Revision 23 December 2016 6.2.4.2.1.4 Group D Piping Group D piping complies with the requirements of the applicable GDC 55, 1971; GDC
56, 1971; or GDC 57, 1971 to the extent that valves are provided in the proper
locations. These lines must, however, remain in service following an accident and, therefore, the valves do not isolate automatically, but trip to the required position.
Piping for the ESFs and supporting systems are included in this class. For this group, the following minimum requirements apply:
(1) Incoming Lines: One local or remote-manual valve outside the containment (2) Outgoing Lines: One local or remote-manual valve outside the containment 6.2.4.2.1.5 Group E Piping Group E piping complies with the requirements of either GDC 55, 1971; GDC 56, 1971;
or GDC 57, 1971. This piping is characterized by sealed closed valves and is used for
intermittent service not related to system functions.
For this group, the following
minimum requirements apply:
(1) Incoming Lines: Sealed closed manual valve outside the containment and a sealed closed manual valve or a check valve inside the containment (2) Outgoing Lines: Sealed closed manual valve outside the containment and a sealed closed manual valve inside the containment (3) No-flow Lines: Diaphragm or sealed closed valve outside containment and diaphragm inside containment 6.2.4.2.2 Piping Systems With some exceptions, such as those noted in Table 6.2-39, piping systems penetrating
the containment conform to GDC 54, 1971; GDC 55, 1971;
GDC 56, 1971; and GDC 57, 1971. The number and location of isolation valves are
shown graphically in Figure 6.2-19; the legend for the diagrams is given in Figure 6.2-
- 18. The criteria to which each penetration complies and a description of the isolation
valves are given in Table 6.2-39. To the extent indicated in Table 6.2-39 and
Figure 6.2-19, piping penetrations are designed with the capability of leak detection and
periodic testing of the isolation valve operability.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-74 Revision 23 December 2016 6.2.4.3 System Design All piping, valves, and connected equipment necessary to maintain the containment
isolation boundary are designed to withstand post-accident conditions with respect to pressure, temperature, and atmospheric conditions at which they are required to
maintain that boundary.
Analyses were made to ensure the integrity of the isolation valve system and
connecting lines due to the forces resulting from inadvertent closure of isolation valves
under operating conditions. Potential maximum forces and moments have been
calculated, and valves, piping systems, and p iping configurations have been designed to withstand these forces. Flued heads have also been designed to withstand these
forces. Where required, snubbers and pipe restraints have been installed to absorb
forces and prevent pipe ruptures caused by the inadvertent closure of an isolation valve.
6.2.4.3.1 Valve Positioning Valves that operate as part of the SIS are designated by the letter "S" in the penetration
diagrams in Figure 6.2-19.
Specific administrative procedures govern the positioning of all of the containment
isolation valves (except check valves) during normal operation, shutdown, and accident
conditions. The positioning of all of the valves required to maintain the penetration
pressure boundary and containment integrity, as well as any flanged closures, is
governed by the administrative procedures.
The main steam lines each have a check valve in series with the isolation valve to
prevent reverse flow of steam in the event of the rupture of a steam line inside the containment. Instrumentation and logic circ uits are provided to detect a ruptured steam line and to close the automatic trip isolation valves on the steam lines.
6.2.4.3.2 Systems Data Table 6.2-39 lists the CIS penetrations and the valves and closed systems employed for
containment isolation. This table lists the number and types of isolation valves, valve
positions during normal operation, shutdown, accident conditions, and primary and
secondary modes of actuation, as well as their functional classification in accordance
with the definitions in Section 6.2.4.2.1 above. The containment isolation valves that
are normally closed to maintain the containment isolation and do not perform an active
function following an accident are administratively sealed closed. Associated isolation
devices required to maintain containment leakage integrity of penetrations and closed
systems such as vent, drain, test, instrumentation and branch line valves, blind flanges, caps or other passive devices are not shown on Table 6.2-39, but are administratively
controlled to be in their proper isolation configuration by plant procedures.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-75 Revision 23 December 2016 Supplementary information regarding the listing in the table is discussed in the following paragraphs:
HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED (1) Leakage Characteristics at Accident Pressures All valves for containment isolation were specified, constructed, and tested
to the maximum allowable leakag e rates as shown in the following (Type C leakage limits are administratively assigned in the Containment
Leakage Rate Testing Program):
Valve Type Seat Leakage (a) Ball <0.3 Globe and gate 3 Check(b) 3 Diaphragm (Saunders Patent) Negligible Butterfly (rubber-seated) Negligible Notes:
(a) Leakage is expressed in units of cubic centimeters of water per hour
per inch of nominal pipe size at valve design conditions (b) The main steam isolation valve is a Schutte-Koerting reverse check which does not meet this seat leakage criterion.
Maximum allowable stem leakage for open backseated valves was specified as one cubic centimeter of water per hour per inch of stem
diameter at design conditions.
(2) Control System Type Containment isolation valves are provided with actuation and control
features appropriate to the valve type. For example, air-operated globe
and diaphragm valves are generally equipped with air diaphragm
operators, spring loaded to fail-closed on loss of air or electrical signal.
Motor-operated gate valves can be supplied from Class 1E Standby
Power Supply as well as their normal power source. Manual and check
valves do not require actuation or control systems. Valve and operator
types are listed in Table 6.2-39.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-76 Revision 23 December 2016 (3) Signal to Operate the Valve All remote-manual containment isolation valves are opened and closed normally from the control room or from local control panels (e.g., sampling
system valves are operated from a panel in the post-accident sample
room). (4) Power Source Required to Actuate or Operate the Valve Remote-manual containment isolation valves are actuated by compressed
gas or electrical power (refer to Table 6.2-39).
(5) Time Necessary to Close the Valve Standard closing times normally available are adequate for the sizes of
containment isolation valves used. Valves equipped with air diaphragm operators generally close in approximately 2 seconds; 10 seconds is
typical of the closing time available in large motor-operated gate valves.
(6) Normal and Failed Positions of the Valves Normal and failed positions of the valves are indicated in Table 6.2-39.
Diagrams for each penetration, showing all valves, barriers, missile
shielding, and leakage test connections are shown in Figure 6.2-19. The
parts of the piping systems that are PG&E Design Class I are also
indicated in this figure. The conditions requiring containment isolation are listed in Table 6.2-40.
6.2.4.4 Safety Evaluation
6.2.4.4.1 General Design Criterion 2, 1967 - Performance Standards
The CIS is designed such that many of the CIS components are contained in the
auxiliary building and containment structure. These structures are PG&E Design Class I (refer to Section 3.8). These buildin gs or applicable portions thereof are designed to withstand the effects of winds and tornados (refer 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, and to protect CIS components, ensuring
their design function will be performed.
Portions of the containment isolation system are not contained within a building and are
exposed directly to potential wind and tornado loads and have been evaluated. Loss or failure of this equipment does not compromise the capability of maintaining containment integrity (refer to Section 3.3.2).
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-77 Revision 23 December 2016 All valves, piping, and equipment that are considered to be isolation barriers are designed to PG&E Design Class I requirements and are protected against potential missiles and water jets.
6.2.4.4.2 General Design Criterion 10, 1967 - Containment
6.2.4.4.2.1 Piping Penetrations
The CIS limits radioactivity passing from the containment to the atmosphere in the event of a design basis accident. This is accomplished as described in Section 6.2.4.2 and
Table 6.2-39.
The CIS provides a minimum of two barriers to prevent leakage of radioactivity to the outside environment. Either barrier is sufficient to keep leakage within the allowable limits.
6.2.4.4.2.2 Instrument Lines Instrument lines penetrating containment meet the intent of Safety Guide 11, March 1971 (Reference 48), by providing a double barrier (one inside and one outside)
between containment and the outside atmosphere. It provides double barrier isolation
without operator action and without sacrificing any reliability with regard to its
engineered safety functions (refer to Figure 6.2-19 and Table 6.2-39 for the penetration configurations).
The containment pressure instrumentation penetration lines (refer to Sheet 15 of Figure 6.2-19 and Table 6.2-39) consist of a sealed, fluid-filled system with a sealed bellows sensor connected to the diaphragm of the pressure transmitter by a sealed fluid-filled tube. The bellows and tubing inside containment and transmitter diaphragm and tubing outside containment are protected from postulated missile and HELB pipe whip/jet
impingement effects by their location, which provides separation and shielding from these hazards. Isolation valving is not essential to meet the intent of Safety Guide 11, March 1971.
The abandoned-in-place deadweight pressure calibrator penetration line (refer to Sheet
22 of Figure 6.2-19 and Table 6.2-39) meets the intent of Safety Guide 11, March 1971, by the use of a diaphragm on both Units plus one sealed-closed valve on Unit 1 and an
instrument cap on Unit 2. 1-PT-458A and 2-PT-458A have been abandoned-in-place, resulting in the isolation valves being closed. Therefore, the inboard barrier is a closed
valve with the PT-458A diaphragm as a back-up. The calibrator and tube are filled with
distilled water that is separated from the reactor coolant by the diaphragm in the pressure sensor. This diaphragm is designed to withstand full RCS pressure from
either side.
The spare instrument test line penetrations used to measure containment pressure
during the integrated leakrate test meet the intent of Safety Guide 11, March 1971 by DCPP UNITS 1 &
2 FSAR UPDATE 6.2-78 Revision 23 December 2016 providing a double barrier, one inside and o ne outside containment (refer to Sheet 11 of Figure 6.2-19 and Table 6.2-39).
Reactor vessel level instrumentation penetration lines (refer to Sheet 25 of Figure
6.2-19 and Table 6.2-39) consist of a sealed fluid-filled system, with a sealed bellows sensor inside containment connected to one side of a differential pressure unit (DPU)
outside of containment by a sealed fluid-filled system. The sensor and the DPU are
capable of withstanding full RCS pressure. I solation valving is not essential to meet the intent of Safety Guide 11, March 1971.
6.2.4.4.3 General Design Criterion 11, 1967 - Control Room The CIS is designed with indication and controls located in the control room as necessary to shut down and maintain safe control of the facility.
Each automatic isolation valve is provided with a manual switch for operation. The
position of each automatic isolation valve and remote-manual valve is displayed in the
control room. All automatic isolation valves are operable from the control room. All
remote-manual containment isolation valves are opened and closed normally from the
control room or from local control panels (e.g., sampling system valves are operated
from a panel in the sampling room). Position indicators are provided for each valve
near its manual control switch. Control room indication, valve actuators, trip signals, and
valve positions are listed in Table 6.2-39.
6.2.4.4.4 General Design Criterion 12, 1967 - Instrumentation and Control Systems The CIS is designed with automatic and remote-manual containment isolation valves that are provided with instrumentation and controls as required to monitor and maintain
variables within prescribed operating ranges.
Governing conditions regarding closure of isolation valves and the instrumentation and controls for the system are described in
Sections 6.2.4.4.3 and in 7.3.2.4.3, and in Table 6.2-39.
6.2.4.4.5 General Design Criterion 21, 1967 - Single Failure Criterion The CIS is designed with penetration configurations to ensure that a single failure will not prevent the CIS from performing its design function. Refer to Section 6.2.4.2.1 for a description of the PG&E containment penetration piping isolation grouping.
No manual operation is required for Phase A and Phase B containment isolation
although isolation can be accomplished manually.
Each remote-manual and automatic isolation valve is designed to close or go to a preferred position on a loss of power or air
or nitrogen supply, except for motor-operated valves, which fail as-is.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-79 Revision 23 December 2016 Also, a single failure in the instrumentation and control circuits will not prevent isolation.
The instrumentation and control circuits are redundant in the sense that a single failure
cannot prevent containment isolation.
Control circuits are designed to close the air and solenoid operated isolation valves on a de-energized state. No power is therefore required to isolate the containment. The
exception to this is steam line isolation, which requires the energization of one of two
mutually redundant circuits.
6.2.4.4.6 General Design Criterion 40, 1967 - Missile Protection The CIS is designed with adequate protection for containment isolation, including piping, valves, and vessels, against dynamic effects and missiles that might result from plant equipment failures, including a LOCA.
No valve is considered to be an isolation valve if it is not missile-protected. Isolation
valves, actuators, and control devices required inside the containment are located
between the crane wall or some other missile shield and the outside containment wall.
Isolation valves, actuators, and control devices outside the containment are located
outside the path of potential missiles or are provided with missile protection. Piping or
vessels that provide one of the isolation barriers outside the containment are similarly
protected. The missile barrier for each isol ation valve is shown schematically on the penetration diagrams (Figure 6.2-19). Refer to Section 3.5 for additional information on
missile protection. Refer to the individual system sections for a discussion regarding
missile protection.
6.2.4.4.7 General Design Criterion 53, 1967 - Containment Isolation Valves
The CIS is designed such that penetrations that require closure for containment function are provided with at least two barriers. The CIS is designed to meet either GDC 55, 1971; GDC 56, 1971; or GDC 57, 1971 requirements with exceptions. The CCW
penetration to the excess letdow n heat exchanger is an exception that does not meet the 1971 GDC because of commitments to d esign and construction made prior to the issuance of the 1971 GDC. The CCW penetration to the excess letdown heat exchanger does comply with GDC 53, 1967. Refer to Table 6.2-39 for penetration and
configuration details with regards to GDC 53, 1967.
6.2.4.4.8 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment The CIS design provides for a double barrier at the containment penetration in those fluid systems that are not required to function following a design basis event. The capability for periodic leakage testing is discussed in Section 6.2.4.4.13. Those
automatic isolation valves that do not restrict normal plant operation are periodically
tested to ensure operability. Refer to Section 6.2.4.2.1 and Table 6.2-39 for penetration
and configuration details with regards to GDC 54, 1971.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-80 Revision 23 December 2016 Automatic Phase A and Phase B valves and sealed closed containment isolation valves are periodically tested for leak-tightness as described in Table 6.2-39.
6.2.4.4.9 General Design Criterion 55, 1971 - Reactor Coolant Pressure Boundary Penetrating Containment The CIS is designed such that for each line that is part of the reactor coolant pressure
boundary that penetrates containment is provided with containment isolation valves 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.
6.2.4.4.10 General Design Criterion 56, 1971 - Primary Containment Isolation
The CIS is designed such that each line that connects directly to the containment atmosphere and penetrates containment is provided with containment isolation valves in
compliance with GDC 56, 1971. Refer to Section 6.2.4.2.1 and Table 6.2-39 for
penetration and configuration details with regards to GDC 56, 1971.
6.2.4.4.11 General Design Criterion 57, 1971
- Closed System Isolation Valves The CIS is designed such that each line that penetrates containment and is neither part of the reactor coolant pressure boundary nor connected directly to the containment
atmosphere is provided with at least one containment isolation valve in compliance with
GDC 57, 1971. Refer to section 6.2.4.2.1 and Table 6.2-39 for penetration and
configuration details with regards to GDC 57, 1971.
6.2.4.4.12 General Design Criterion, 70, 1967 - Control of Releases of Radioactivity to the Environment The CIS, in conjunction with the containment (refer to Section 6.2.1.3.6), is designed to be a barrier to maintain control over plant radioactive effluents, whether gaseous, liquid, or solid. The CIS is designed to withstand the effects of a LOCA (refer to Section 6.2.4.4.2), ensuring that the offsite radiological exposures resulting from a LOCA are within the limits of 10 CFR Part 100 (refer to Section 15.5).
6.2.4.4.13 10 CFR Part 50, Appendix J, Option B - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors Testing of containment penetrations (Type C testing) is performed in accordance with the Technical Specifications 5.5.16, Containment Leakage Rate Testing Program, as
required by 10 CFR 50.54(o), and 10 CFR Part 50, Appendix J, Option B, as modified by approved exemptions. As a g eneral requirement, all containment isolation valves will
be tested periodically with a gas to determine leaktightness. Refer to Section 6.2.1 for
information regarding Type A (ILRT) and Type B (air locks, electrical penetrations, hatches, etc.) testing.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-81 Revision 23 December 2016 Exceptions to this requirement are those valves not required to be testable by Appendix J to 10 CFR Part 50, and certain valves that cannot be isolated for air testing. These
include the first of double check valves to the RCS and valves for which such testing
would require draining significant portions of the RHR system or the SIS. Even if these
systems were drained, the presence of other valves associated with the systems would
make it impractical to determine the source of any measured leakage. Where a
quantitative leakage test is necessary, provisions are made for each valve to measure
the inflow of the pressurizing medium, collect and measure leakage, or calculate the
leakage from the rate of pressure drop. The test pressure on the valve will be at a differential pressure of not less than the peak calculated containment internal pressure
related to the design basis LOCA (P a). The P a value specified in Technical Specification 5.5.16 bounds the calculated LOCA containment integrity results in Section 6.2D.3.2.6.
Check valves and single-disk gate valves wil l have the test pressure applied to the inboard side of the valve. Exceptions are the three RHR injection lines. The valves in
these lines will be tested from the outboard side, as there is no practical method to test
from the inboard side. Diaphragm valves may be tested on either side since their
leakage characteristics are the same in either direction. Double-disk gate valves may
be tested by applying the test pressure between the disks. Globe valves may be tested
by pressurizing either the inboard side or under the seat.
Piping systems are provided with test vents (TV) and test connections (TC) or have other provisions to allow periodic leakage testing of the containment isolation valves, as
required. Locations of TC and TV are shown on the penetration diagram (refer to Figure
6.2-19). In most cases, equipment vents or drains can be used as TC or TV.
6.2.4.4.14 Regulatory Guide 1.163, September 1995 - Performance-Based Containment Leak-Test Program The DCPP Containment Leakage Rate Testing Program is performed in accordance with the Technical Specifications. The DCPP Containment Leakage Rate Testing
Program utilizes a performance-based approach, consistent with Regulatory Guide
1.163, September 1995, to comply with the requirements of 10 CFR Part 50, Appendix
J, Option B (refer to Section 6.2.4.4.13).
6.2.4.4.15 NUREG-0737 (Item II.E.4.2), November 198 0 - Clarification of TMI Action Plan Requirements Item II.E.4.2 - Containment Isolation Dependability:
Position (1) - The automatically tripped isolation valves are actuated to the closed
position by one of two separate containment isolation signals.
Immediate isolation of the containment is accomplished automatically. There are two
automatic phases of containment isolation at DCPP.
Phase A isolates all nonessential DCPP UNITS 1 &
2 FSAR UPDATE 6.2-82 Revision 23 December 2016 process lines but does not affect safety injection, containment spray, component cooling water supplied to the reactor coolant pumps and containment fan coolers, and steam
and auxiliary feedwater lines. Phase B isolates all process lines except safety injection, containment spray, auxiliary feedwater, and the containment fan coolers component
cooling water system. Valves which close automatically upon receipt of a Phase A isolation signal are designated by the letter "T" in the penetration diagrams (Figure 6.2-19). The letter "P" is used to designate those valves that close automatically upon
receipt of a Phase B isolation signal.
Phase A isolation is initiated by high containment pressure, low pressurizer pressure, low steamline pressure, or manual initiation.
Phase B isolation is initiated by high-high containment pressure or manual initiation.
Section II.6 of SRP 6.2.4 establishes the DCP P licensing basis, but the design basis preceded the issuance of NUREG-0578, July 1979, and subsequently NUREG-0737, May 1980. The DCPP Phase A and Phase B containment isolation signals comply with Section II.6 of Standard Review Plan (SRP) 6.2.4, 1975, for diversity of parameters that
initiate containment isolation.Section II.6 of SRP 6.2.4 does not constitute the DCPP
design basis because the definitions of essential, nonessential, and safety system
process lines, as well as those parameters that initiate a Phase A or Phase B
containment isolation signal, were established prior to the issuance of NUREG-0578, July 1979, and subsequently NUREG-0737, May 1980.
Position (2) - Three levels of containment process penetrations have been defined for
the DCPP:
(1) "Nonessential" process lines are defined as those that do not increase the potential for damage for in-containment equipment when isolated.
These are isolated on Phase A isolation.
(2) "Essential" process lines are those providing cooling water and seal water flow through the reactor coolant pumps. These services should not be
interrupted while the reactor coolant pumps are operating unless
absolutely necessary. These are isolated on Phase B isolation.
(3) Safety system process lines are those required to perform the function of the ESF system.
Table 6.2-39 identifies nonessential, essential, and safety systems penetrating containment.
Position (3) - All nonessential systems use either manually sealed closed valves or else the valves are automatically isolated on a Phase A containment isolation signal.
Additionally, all essential systems (defined in Position (2) above) are automatically
isolated on a Phase B containment isolation signal.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-83 Revision 23 December 2016 Position (4) - Resetting the isolation signals will not result in the automatic reopening of containment isolation valves. Reopening of containment isolation valves requires
deliberate action and ganged reopening will not result from a single operator action after the signal has been reset.
Position (5) - Table 6.2-40 shows operating conditions that make containment isolation
mandatory. Setpoints are specified in the Technical Specifications.
Position (6) - The DCPP containment purge system valves and vacuum/overpressure
relief valves satisfy the operability criteria set forth in BTP CSB 6-4, 1975. The opening
of the 12 inch vacuum/overpressure relief valves is restricted to no more than 50
degrees.
Position (7) - The containment purge and vent isolation valves are closed automatically
by any one of the following:
(1) Phase A containment isolation signal (2) High gaseous or air particulate radioactivity in containment (3) High radiation at the plant vent 6.2.4.4.16 Generic Letter 89-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance The CIS 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
6.2.4.4.17 Generic Letter 96-06, September 1996 - Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions Containment isolation valves LWS-FCV-253 and LWS-FCV-500 (Penetrations 50 and 49, respectively) have both been modified through the addition of a pressure relief hole to prevent overpressurization of the associated isolated section of piping during design-basis accidents to ensure containment integrity is maintained. All other piping
penetrations that are susceptible to overpressurization either have valves whose design
prevents overpressurization (air-operated diaphragm valves, solenoid valves, or air-
operated globe valves) or have been drained to prevent overpressurization and
thereafter maintained drained as appropriate.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-84 Revision 23 December 2016 6.2.4.5 Tests and Inspections The CIS design provides such functional relia bility and ready testing facilities as are necessary to avoid undue risk to the health and safety of the public. CIS periodic tests
and inspections are provided to ensure a continuous state of readiness to perform its
safety function.
Containment isolation signal actuation channels are de signed with sufficient redundancy to provide the capability for channel testing and calibration during power operation
without tripping the system (refer to Section 7.3.3.4 for more details).
The pneumatic-operated isolation valves close on loss of control power or compressed
gas. Isolation valves will be periodically tested for operability.
For additional details regarding periodic testing and inspection of valves, refer to the Technical Specifications.
6.2.4.6 Materials Materials selection for the penetration lines and isolation valves of the CIS depends on
the particular application and function of the systems involved. Further information is
provided in the sections describing the individual systems of interest, and in Section 3.8 where details of penetration designs are presented.
6.2.5 COMBUSTIBLE GAS CONTROL IN CONTAINMENT
Following a loss-of-coolant accident (LOCA), hydrogen may be produced inside the reactor containment by radiolysis of the core and sump solutions, by corrosion of aluminum and zinc, by reaction of the zirconium in fuel cladding with water, and by
release of the hydrogen contained in the reactor coolant system.
The original design of the containment structure utilized the containment hydrogen
purge system (CHPS) which includes dedicated containment penetrations for the
purposes of purging hydrogen.
Following the TMI-2 incident and issuance of NUREG-0737, November, 1980, PG&E committed to provide an electrical hydrogen recombiner system (EHRS) inside the
containment structure as the primary mean s of hydrogen buildup mitigation. In addition, PG&E committed to providing the installed ca pability to connect external hydrogen recombiners outside the containment structure, cross tying the CHPS supply and
exhaust piping into a closed loop.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-85 Revision 23 December 2016 In 2003, revisions to 10 CFR 50.44 removed the definition of a design basis LOCA
hydrogen release, and eliminated the requirements for hydrogen control systems to
mitigate such a release. The requirements remaining include the need to maintain a
mixed atmosphere to prevent coalescence of high concentration local hydrogen
collections that may provide a flammability risk and the need to provide equipment for
monitoring hydrogen concentrations in the containment atmosphere.
To fulfill the requirements of 10 CFR 50.44, PG&E credits the use of the containment
fan cooler system (CFCS) as the means of c ontainment atmosphere mixing. Refer to Section 6.2.2 for discussion of other CFCS design bases.
HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED The research and development work discussed in more detail in Section 6.2.5.3 substantially reduced the uncertainties in both the expected rates of hydrogen
accumulation and the potential exposures that would result from hydrogen control by venting as follows:
(1) Research on the corrosion of aluminum and associated hydrogen production rates by Westinghouse (References 12 - 15 and 42) has
reduced uncertainties on corrosion rates in the expected post-accident environment and allowed a reduction in the expected corrosion rate from
the 42 mg/dm 2/hr used in the Unit 2 PSAR to the value shown in Figure 6.2-24.
(2) The amounts of aluminum used in the as-built plant have been minimized through materials design specifications. Zinc is another significant
contributor. The uncertainties in the amounts of hydrogen produced from
both have been reduced by itemized accoun ting (refer to Table 6.2-42).
(3) The amounts of hydrogen expected to be produced by the zirconium-water reaction have been reduced by the more stringent limits established on ECCS performance.
(4) Research by the Atomic Energy Commission (AEC) and its contractors (a partial compilation is included in Reference 16) in the context of
emergency core cooling system (ECCS) studies has substantially reduced uncertainties in the extent of zirconiu m-water reactions following a LOCA.
(5) Reevaluation of energy generation rates has allowed reduction of hydrogen generation rate from sump radiolysis.
(6) Research on hydrogen yield in the core and sumps by Westinghouse (References 13-15) has reduced uncertainties in these constants.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-86 Revision 23 December 2016 (7) Refined analysis of the distribution of fission product decay energy (Reference 17) has resulted in more precise values for the fractions of beta and gamma energies absorbed by water.
(8) Additional meteorological data and analysis (References 18-20) conducted by PG&E as a part of the 2-year site program has established
high probabilities of conditio ns favorable for controlled venting.
(9) Development of the general purpo se EMERALD (Reference 21) computer program for the calculation of doses following accidents has resulted in
more accurate estimates of potential exposures, and permitted additional
sensitivity studies of the influence of various parameters on potential
exposures.
6.2.5.1 Design Bases 6.2.5.1.1 General Design Criterion 2, 1967 - Performance Standards The CHPS, the EHRS and the hydrogen monitoring system (containment penetrations and containment isolation valves only) are designe d to withstand the effects of, or are protected against, natural phenomena such as earthquakes, winds and tornadoes, floods and tsunamis, and other local site effects.
6.2.5.1.2 General Design Criterion 3, 1971 - Fire Protection The CHPS, the EHRS and the hydrogen monitoring system SSCs are designed and
located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions.
6.2.5.1.3 General Design Criterion 11, 1967 - Control Room The CHPS and hydrogen monitoring system are designed to support actions to maintain and control the safe operational status of th e plant from the control room.
6.2.5.1.4 General Design Criterion 12, 1967 - Instrumentation and Control Systems Instrumentation and controls are provided as required to monitor and maintain hydrogen concentrations within prescribed operating ranges.
6.2.5.1.5 General Design Criterion 17, 1967 - Monitoring Radioactivity Releases Means are provided for monitoring the effluent discharge path of the CHPS for radioactivity that could be released.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-87 Revision 23 December 2016 6.2.5.1.6 General Design Criterion 37, 1967 - Engineered Safety Features Basis for Design Means are provided in the containment to back up the safety function provided by the core design, the reactor coolant pressure boundary, and their protection systems. The
CFCS is designed to ensure containment atmosphere mixing as a result of any size
reactor coolant pressure boundary break to prevent the coalescence of local hydrogen
concentrations within containment.
6.2.5.1.7 General Design Criterion 41, 1967 - Engineered Safety Features Performance Capability The CFCS is designed to provide sufficient performance capability to accommodate partial loss of installed capacity and still fulfill containment atmosphere mixing.
6.2.5.1.8 General Design Criterion 42, 1967 - Engineered Safety Features Components Capability The CFCS is designed so that the capability of each component and system to perform its required function is not impaired by the effects of a LOCA.
6.2.5.1.9 General Design Criterion 49, 1967 - Containment Design Basis The containment combustible gas control sy stems are 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 LOCA, including a considerable margin for effec ts from metal-water or other chemical reactions that could occur as a consequence of failure of emergency core cooling systems.
6.2.5.1.10 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment The piping that is part of the CHPS and hydrogen monitoring system that penetrate 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.
6.2.5.1.11 General Design Criterion 56, 1971 - Primary Containment Isolation Valves The CHPS and hydrogen monitoring system contain valves in piping that penetrate containment and connect directly to the containment atmosphere. Remote manual
isolation valves are provided outside containment and automatic (check) valves are
provided inside containment to ensure containment integrity is maintained.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-88 Revision 23 December 2016 6.2.5.1.12 10 CFR 50.44 - Combustible Gas Control for Nuclear Power Reactors The CFCS ensures a mixed atmosphere is maintained within containment to prevent high localized concentrations of hydrogen gas accumulation.
The hydrogen monitoring system is designed to be functional, reliable, and capable of
continuously measuring the concentration of hydrogen in the containment atmosphere.
6.2.5.1.13 10 CFR 50.49 - Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants CHPS, EHRS and the hydrogen monitoring system components that require environmental qualification (EQ) are qualifie d to the requirements of 10 CFR 50.49.
6.2.5.1.14 10 CFR 50.55a(f) - Inservice Testing Requirements American Society of Mechanical Engineers (ASME) code components of the CHPS and hydrogen monitoring system are tested to the requirements of 10 CFR 50.55a(f)(4) and
10 CFR 50.55a(f)(5) to the extent practical.
6.2.5.1.15 10 CFR 50.55a(g) - Inservice Inspection Requirements ASME code components of the CHPS and hydrogen monitoring system are inspected to the requirements of 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(g)(5) to the extent practical.
6.2.5.1.16 Regulatory Guide 1.7, Revision 2, November 1978 - Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident The CFCS is designed to provide a mixed atmosphere in containment and thus control combustible gas concentrations without relying on purging of the containment atmosphere following a LOCA. The CFCS meets the design, quality assurance, redundancy, energy source, and instrumentation requir ements for an engineered safety feature. The hydrogen monitoring system provides a means to measure the hydrogen
concentration in the containment.
6.2.5.1.17 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 hydrogen monitoring system is designed to provide continuous indication in the
control room of hydrogen concentration in the containment atmosphere following a
beyond design basis accident and meets th e design provisions of Regulatory Guide 1.97, Revision 3, May 1983 including qualification, redundancy and testability.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-89 Revision 23 December 2016 6.2.5.1.18 NUREG-0737 (Items II.E.4.1, II.F.1), November 1980 -
Clarification of TMI Action Plan Requirements Item II.E.4.1 - Dedicated Hydrogen Penetrations:
The CHPS for post-accident combustible gas control of the containment atmosphere is
provided with dedicated containment penetrations separate from other containment
venting systems.
Item II.F.1 - Additional Accident Monitoring Instrumentation:
Position (6) - The display instrumentation is designed to include the containment
hydrogen monitors. Indication of hydrogen concentration in the containment
atmosphere is provided in the control room.
6.2.5.1.19 Generic Letter 89-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance The CHPS motor-operated valves (M OVs) meet the requirements of Generic Letter 89-10, June 1989, and associated Generic Letter 96-05, September 1996.
6.2.5.2 System Description 6.2.5.2.1 Containment Fan Cooler System The CFCS provides mixing of the containme nt atmosphere following a design basis accident to ensure that hydrogen concentrations are within prescribed limits. Refer to Sections 6.2.2 and 9.4.5 for further details on the CFCS.
6.2.5.2.2 Electric Hydrogen Recombiner System Internal EHRSs are installed in both units of the DCPP as a means of controlling
containment atmosphere hydrogen concentration following a LOCA.
The EHRSs are natural convection, flameless, thermal reactor-type hydrogen-oxygen
recombiners. In their basic operation, they heat a continuous stream of air-hydrogen
mixture to a temperature sufficient for spontaneous recombination of the hydrogen with
the oxygen in the air to form water vapor.
The system for each unit consists of two independent recombiners, each of which
contains the electric heater banks, a power supply panel that contains the equipment for
powering the heaters, and a power control panel to the heaters. The recombiners are
located inside the containment building; the power supply and control panels are located outside this building. The EHRS units are completely enclosed and the
internals are protected against impingement from containment spray.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-90 Revision 23 December 2016 Each recombiner consists of an inlet preheater section, a heater-recombination section, and a mixing chamber (refer to Figure 6.2-23).
Air and the hydrogen are drawn into the unit by natural convection via the inlet louvers and pass through the preheater section, which consists of a shroud placed around the central heaters to take advantage of heat
conduction through the walls. In this area, the temperature of the inlet air is raised.
This rise in temperature accomplishes the dual function of increasing system efficiency
and evaporating any moisture droplets that may be entrained in the air. The warmed air
then passes through the flow orifice that has been specifically sized to regulate air flow
through the unit. After passing through the orifice plate, the air flows vertically upward through the heater section, where its temperature is raised to the range of 1150
°F to 1400°F, causing the recombination of H 2 and O 2 to occur. The recombination temperature is approximately 1135°F.
Next, the air rises from the top of the heater section and flows into the mixing chamber, which is at the top of the unit. Here, the hot air is mixed with the cooler containment air
and then discharged back into the containment at a lower temperature. The cooler
containment air enters the mixing chamber through the lower part of the upper louvers
located on three sides of the unit.
The major structural components are manuf actured from stainless steel and Incoloy-800. The heater sheathing is also Incoloy-800. Each bank is constructed of Incoloy-
800 sheathed tubular elements mounted in a heavy-gauge steel flange with holes for
mounting into the recombiner heater frame.
There are four banks of heaters in each recombiner. Each bank contains 60 individual, U-type heating elements connected in series-parallel arrangements as required to obtain the power rating for each bank. The internal connections are wired to special terminal blocks located outside the heater flange. Each bank is sized for a specific
power rating.
The power supply panel contains all the necessary electrical equipment to provide the power required by the heaters in the recombiner.
The panel consists of an isolation transformer, silicon-controlled rectifier module, an
auxiliary control power transformer, and a main line contactor. The control panel contains all the control and monitoring equipment required for operating the recombiner
and is easily accessible to the plant operators.
Thermocouples are provided for convenience in testing and periodic checkout; they are
not considered necessary, however, to ensure proper operation of the recombiner.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-91 Revision 23 December 2016 The EHRS design characteristics conservatively bound the design conditions following a LOCA and are as follows:
Normal Operating Conditions Post-LOCA Operating Conditions Temperature, °F 120 288 (max) Pressure, psia 15 77 (max) Pressure transient, psia N/A 77 (max in 10 sec)
Relative humidity, %
0-100 100 Radiation, rads/hr 5 3.3 x 10 5 Radiation-total dose, rads NA 2 x 10 8 (max) Spray solution, ppm B/pH NA/NA 2550/9-10.5 Design life, yr 40 NA
Recombiner capacity, scfm of containment gas at 1 atm NA 100 (min)
From the results shown in Figures 6.2-26 through 6.2-29, a 100 scfm hydrogen recombiner, started when the bulk containment hydrogen concentration reaches
3.5 percent by volume (after 3 days), or earlier, will ensure that the bulk containment
hydrogen concentration will not reach the lower flammability limit of 4 percent by
volume. The licensing limit of 4.0 percent by volume is assured by operating
procedures that direct operators to initiate recombiner operation at hydrogen
concentrations as low as 0.5 percent by volume. Thus, neither hydrogen burning nor
detonation will occur.
6.2.5.2.3 Containment Hydrogen Purge System The CHPS is designed for either intermittent or continuous flow operation. While the hydrogen recombiner system is the primary system for post-LOCA containment
hydrogen control, in the event that hydrogen concentration were to reach the control
limit of 3.5 percent, the hydrogen purge system may be placed into operation under
strict administrative controls.
The CHPS is a PG&E Design Class I system consisting of two diverse purge routes and
two redundant supply routes. The system is available for control of hydrogen and
includes provisions for post-accident installation of portable recombiners. The basic
features of the CHPS are shown in Figures 6.2-20, 6.2-21, and 6.2-22.
Each purge stream leaves the containment through a motor-operated isolation valve, which opens remotely during venting. One purge stream is routed through a manual valve, roughing filter, HEPA filter, charcoal, HEPA after-filter, a blower, a hand control
valve, a flow measuring device, plant vent radiation monitor, and the associated plant
vent radiation monitoring systems. The second purge stream is routed through a flow
measuring device, a hand-controlled valve, and the containment excess pressure relief
line to the auxiliary bu ilding ventilation system carbon filter plenum. This purge stream DCPP UNITS 1 &
2 FSAR UPDATE 6.2-92 Revision 23 December 2016 then follows the auxiliary building ventilation exhaust flowpath through the roughing filter, HEPA filter, carbon adsorber, exhaust fan, plant vent and the associated plant vent radiation monitoring system. The purge stream can be operated independently of
the supply stream.
The supply stream is drawn through a roughing filter and a blower and routed through a
hand control valve, a flow measuring device, a gate valve, and isolation check valves.
The supply stream can be operated independently of the purge stream.
The supply stream entrance and the purge stream exit are widely separated to prevent
short circuiting. The containment fan cooler units ensure complete mixing of the post-
accident containment atmosphere.
All CHPS motor-operated valves inside cont ainment are supplied with Class 1E power.
Prior to initiation of hydrogen purge, the containment radiation monitoring system is
used to monitor, either continuously or intermittently, the radioactivity in containment.
The plant vent radiation monitors measure the radioactivity in the purge stream. Refer
to Section 9.4.2 for information on the plant vent.
The supply stream isolation valves and blower are operated manually.
The supply stream provides for immediate dilution, and the hydrogen concentration
decreases. The purge stream is initiated after the supply stream begins diluting
The containment hydrogen purge system is provided with charcoal filters to minimize the release of radioactive iodine. The filters are sized in accordance with activity loading specifications associated with ESFs.
The analysis of potential radiation exposures that could result from venting for hydrogen control is contained in the LOCA analysis in Section 15.5.17.2.9. The estimated
incremental exposures resulting from containment venting are a negligible addition to those estimated for containment leakage.
The operators are provided with current data on containment hydrogen concentration, containment activity levels, wind direction, and wind speed to determine optimum purge
schedules.
HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED
In the Preliminary Safety Analysis Report (PSAR) (Reference 9) for DCPP Unit 2, an analysis of the expected hydrogen production concluded that any hydrogen
accumulation could be controlled by containment venting, with radiological exposures below the annual limits specified in 10 CFR Part 20. Using more conservative parameters, the AEC regulatory staff calculated that the lower flammability limit would DCPP UNITS 1 &
2 FSAR UPDATE 6.2-93 Revision 23 December 2016 be attained in less than 40 days (Reference 10). The staff concluded that the purging operation could result in offsite activity concentration levels that exceed 10 CFR Part 20
limits. Additional capability for filtering of containment effluent, including charcoal beds, would, however, reduce the I-131 concentration level.
Assuming 90 percent filter
efficiency for iodine removal, the staff estima ted (Reference 11) that doses at the site boundary would be about 0.8 rem whole body and about 8.5 rem to the thyroid if the entire contents of the containment were vented over a 30-day period. Estimated
exposures were, therefore, less than 10 percent of the guideline levels established in 10 CFR Part 100.
6.2.5.2.4 Hydrogen Monitoring System The hydrogen monitors are located outside containment in the containment hydrogen
monitor panels. The sample flows pass from the containment through inner and outer
solenoid containment isolation valves to the monitors. The sample flows return to the containment through outer solenoid containment isolation valves and inner containment isolation check valves. Refer to Figure 6.2-22 for schematic configuration of the
system.
Refer to Section 6.2.5.5.1 for additional information on the hydrogen monitors.
6.2.5.3 Safety Evaluation 6.2.5.3.1 General Design Criterion 2, 1967 - Performance Standards The auxiliary building and containment structure, which contain the EHRS, CHPS and hydrogen monitoring system SSCs are PG&E Design Class I (refer to Section 3.8).
These buildings or applicable portions thereof are designed to withstand the effects of
winds and tornadoes (refer 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, and to protect EHRS, CHPS and hydrogen monitoring system
SSCs ensuring their design functions will be performed.
The EHRS and CHPS are PG&E Design Cla ss I and are maintained as such to withstand additional forces that might be imposed by natural phenomena such as winds and tornadoes (refer to Section 3.3), floods and tsunamis (refer to Section 3.4), external
missiles (refer to Section 3.5) and earthquakes (refer to Section 3.7).
6.2.5.3.2 General Design Criterion 3, 1971 - Fire Protection The CHPS, EHRS and hydrogen monitoring system are 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 6.2-94 Revision 23 December 2016 6.2.5.3.3 General Design Criterion 11, 1967 - Control Room Two redundant hydrogen monitors are installed in each unit to provide continuous
indication and recording in the control room of containment hydrogen concentration.
Containment isolation valve status is shown on the main control board as indicated in
Table 6.2-39. For the CHPS, annunciation is provided to alarm on high radioactivity, high flowrate, and fan failure.
Refer to Section 6.2.5.5 for additional discussion of instrumentation and controls.
6.2.5.3.4 General Design Criterion 12, 1967 - Instrumentation and Control Systems The EHRS does not require any instrumentation inside the containment for proper
operation after a LOCA. Proper recombiner operation after an accident is ensured by
measuring the amount of electric power to the recombiner from the control panel
located outside containment accessible to operators following an accident. The
temperature readout device is a monitoring unit, not a control unit. For additional
information refer to Section 6.2.5.2.2.
For the CHPS, the supply and the purge streams may be adjusted to regulate the
flowrates to values required to maintain hydrogen level in the containment at or below
the 4.0 percent limit. Instrumentation is provided to monitor flowrate and hydrogen concentration. The containment radiation monitoring system and the plant vent
radiation monitors are used to monitor the radioactivity in containment and the hydrogen purge line.
Refer to Section 6.2.5.5 for additional discussion of instrumentation and controls.
6.2.5.3.5 General Design Criterion 17, 1967 - Monitoring Radioactivity Releases The CHPS is vented through the plant vent which is equipped with a radiation
monitoring system. Refer to Sections 9.4.5.3.5 and 11.4 for further details on the plant vent and radiation monitoring system.
6.2.5.3.6 General Design Criterion 37, 1967 - Engineered Safety Features Basis for Design The basis for design is centralized upon the r equirements set forth in 10 CFR 50.44 in which the containment structure must have the capabi lity for ensuring a mixed atmosphere. A mixed atmosphere is obtained through use of the CFCS in which post-
accident hydrogen produced is circulated throughout the containment structure in
conjunction with other gases to prevent local hydrogen concentrations from reaching a
lower flammability limit of 4 percent hydrogen by volume. The method of analysis used DCPP UNITS 1 &
2 FSAR UPDATE 6.2-95 Revision 23 December 2016 to determine the post-accident production rate of hydrogen used in the basis of the system design is given below.
6.2.5.3.6.1 Analysis of Hydrogen Generation and Accumulation The quantity of zirconium, which reacts with the core cooling solution, depends on the
performance of the ECCS.
The criteria for ECCS evaluation (10 CFR 50.46) requires that the zirconium-water
reaction be limited to 1 percent by weight of the total quantity of zirconium in the core.
ECCS calculations have shown the zirconiu m-water reaction to be less than 1 percent.
Aluminum inside the containment is not used in safety-related components that are in
contact with the recirculating core cooling fluid. It is more reactive with the containment
spray alkaline borate solution than other plant materials such as galvanized steel, copper, and copper nickel alloys.
The zirconium-water reaction and aluminum and zinc corrosion with containment spray are chemical reactions, which are essentially independe nt of the radiation field inside the containment following a LOCA. Radiolytic decomposition of water is dependent on
the radiation field intensity. The radiation field inside the containment is calculated for the maximum credible accident for which the fission product activities are given in
TID-14844 (Reference 24).
The hydrogen generation is calculated using the NRC model discussed in Regulatory
Guide 1.7, Revision 2, November 1978 (Reference 22), Standard Review Plan 6.2.5 (Reference 40) and Branch Technical Position CSB 6-2 (Reference 41).
6.2.5.3.6.1.1 Hydrogen Generation from the Zirconium-Water Reaction The zirconium-water reaction is described by the chemical equation:
Zr + 2H 2 O Zr O 2 + 2H 2 + Heat Hydrogen generation due to this reaction will be completed during the first day following the LOCA. The NRC model assumes a 5 percent (5 times the maximum allowable value defined by 10 CFR Part 50, Appendix K (ECCS)) zirconium-water reaction. The
hydrogen generated is assumed to be released immediately to the containment
atmosphere.
6.2.5.3.6.1.2 Hydrogen Available from the Reactor Coolant System
The quantity of hydrogen in the RCS during normal operation includes hydrogen from
the pressurizer gas space and hydrogen dissolved in the reactor coolant. The
pressurizer gas space hydrogen is based on:
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-96 Revision 23 December 2016 (1) A maximum allowable coolant hydrogen concentration of 60 cc(stp)/kg of coolant (stp denotes standard temperature and pressure)
(2) Control banks of pressurizer heaters will modulate to control pressurizer heat losses to maintain constant pressurizer temperature and pressure (3) Minimum bypass spray rate of 2.0 gpm (4) Normal liquid level of the pressurizer (60 percent)
(5) Pressurizer power-operated relief valves closed The hydrogen from the reactor coolant and the pressurizer vapor space is available for
release to the containment immediately following a LOCA.
6.2.5.3.6.1.3 Hydrogen Generation from the Corrosion of Plant Materials
Oxidation of metals in aqueous solution generates hydrogen gas as one of the corrosion
products. Extensive corrosion testing has been conducted to determine the behavior of the various metals used within the containment.
Metals tested include zirconium alloys, Inconel, aluminum alloys, cupronickel alloys, carbon steel, galvanized carbon steel, and copper.
Tests conducted at Oak Ridge National Laboratory (ORNL) (References 25 and 26) and
Westinghouse (Reference 42) have verified the compatibility of the various materials
with alkaline borate solution and have shown that aluminum and zinc will corrode at a rate that will significantly add to the hydrogen accumulation in the containment
atmosphere.
The corrosion of aluminum may be described by the overall reaction:
2 Al + 3 H 2 O Al 2 O 3 + 3 H 2 Three moles of hydrogen are produced for every two moles of aluminum oxidized.
The corrosion of zinc may be described in the overall reaction:
Zn + 2H 2 O Zn(OH)2 + H 2 One mole of hydrogen is produced for each mole of zinc oxidized.
The time-temperature cycle (refer to Table 6.2-41) considered in the calculation of
aluminum and zinc corrosion is a step-wise representation of the postulated post-
accident containment temperature transient. The corrosion rates at the various steps
were determined from the aluminum and zinc corrosion rate design curves (References
42 and 55) shown in Figure 6.2-24, which include the effects of temperature and spray DCPP UNITS 1 &
2 FSAR UPDATE 6.2-97 Revision 23 December 2016 solution conditions. For conservative estimation, no credit was taken for protective shielding effects of insulation or enclosures from the spray, and complete and
continuous immersion was assumed.
The calculations were performed by Westinghouse using the methodology of
Regulatory Guide 1.7, Revision 2, November 1978, but using the corrosion rates given
in Figure 6.2-24 and the containment time-temperature values given in Table 6.2-41.
For this hydrogen generation reanalysis the aluminum and zinc inventories inside
containment are as shown in Table 6.2-42.
6.2.5.3.6.1.4 Hydrogen Generation from the Radiolysis of Core and Sump Water Water radiolysis is a complex process involving reactions of numerous intermediates.
However, the overall radiolytic process may be described by the reaction:
2 2 22/1 OHOH+ Table 6.2-43 presents the total decay energy ( + ) of a reactor core. It assumes full power operation with extended fuel cycles prior to the accident. For the maximum credible accident case, the contained decay energy in the core accounts for the
assumed TID-14844 release of 50 percent halogens and 1 percent other fission
products. In the TID-14844 model, the noble gases are assumed to escape to the
containment vapor space.
HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED
The yield of hydrogen from radiolytically decomposed solution has been studied extensively by Westinghouse and ORNL. The results of static capsule tests conducted by Westinghouse indicate hydrogen yields much lower than 0.44 molecules per 100 eV for core radiolysis.
There are, however, differences between the static capsule tests and the dynamic
condition in core, where the cooling fluid is continuously flowing. The flow is assumed to disturb the steady state conditions that are observed in static capsule tests, and while
the occurrence of back reactions is still significant, the overall net yield of hydrogen is
somewhat higher in th e flowing system.
Westinghouse studies of radiolysis in dynamic systems (Reference 15) show 0.44
molecules per 100 eV to be a maximum yield for high solution flowrates through a gamma radiation field. Work by ORNL (References 25 and 26), Zittel (Reference 28), and Allen (Reference 29) confirm this value.
Analysis, based on Regulatory Guide 1.7, Revision 2, November 1978, is conservative because it assumes a hydrogen yield value of 0.5 molecules per 100 eV. It also DCPP UNITS 1 &
2 FSAR UPDATE 6.2-98 Revision 23 December 2016 assumes that 10 percent of the gamma energy, produced from fission products in the fuel rods, is absorbed by the solution in the region of the core, and the noble gases
escape to the containment vapor space.
Another potential source of hydrogen assumed for the post-accident period arises from
water in the reactor containment sump being subjected to radiolytic decomposition by
fission products. An assessment must therefore be made of the decay energy
deposited in the solution and the radiolytic hydrogen yield, much in the same manner as
for core radiolysis.
The energy deposited in solution is computed using the following basis:
(1) For the maximum credible accident, a TID-14844 release model (Reference 24) is assumed where 50 percent of the total core halogens
and 1 percent of all other fission products, excluding noble gases, are
released from the core to the sump solution.
(2) The quantity of fission product release considers a reactor operating with extended fuel cycles prior to the accident.
(3) The total decay energy from the released fission products, both beta and gamma, is assumed to be fully absorbed in the solution.
The calculation of the fission product decay energy deposited in the sump solution
considers the decay of halogens and the decay of the remaining 1 percent of fission
products. The energy release rates and integrated energy release for various times after a LOCA are listed in Table 6.2-44.
HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED
The yield of hydrogen from sump solution radiolysis is most nearly represented by the static capsule tests performed by Westinghouse and ORNL with an alkaline sodium
borate solution. The differences between these tests and the actual conditions for the
sump solution, however, are important and render the capsule tests conservative in
their predictions of radiolytic hydrogen yields.
In this assessment, the sump solution will have considerable depth, which inhibits the
ready diffusion of hydrogen from solution, as compared to the case with shallow-depth
capsule tests. This retention of hydrogen in solution will have a significant effect in
reducing the hydrogen yields to the containm ent atmosphere. The buildup of hydrogen in solution will enhance the back reaction to form water and lower the net hydrogen yields, in the same manner as a reduction in the gas to liquid volume ratio will reduce
the yield. This is illustrated by the data presented in Figure 6.2-25 for capsule tests with
various gas to liquid volume ratios.
The data show a significant reduction in the net hydrogen yield from the primary maximum yield of 0.44 molecules per 100 eV. Even at DCPP UNITS 1 &
2 FSAR UPDATE 6.2-99 Revision 23 December 2016 the very highest ratios, where capsule solution depths are very low, the yield is less than 0.30, with the highest scatter data point at 0.39 molecules per 100 eV.
Taking these data into account, a reduced hydrogen yield is a reasonable assumption for the case of sump radiolysis. The ex pected yield is on the order of 0.1 molecules per 100 eV or less. Regulatory Guide 1.7, Revision 2, November, 1978 does not, however, allow credit for the reduced hydrogen yields and a yield value of 0.5 molecules per 100 eV is used in the analyses.
All containment volumes are connected by large vent areas to promote good air
circulation. Hydrogen will diffuse very rapidly giving an even distribution under the
conditions existing in the containment struct ure. In addition, thermal mixing effects, heating of air above the hot sump water, and possible steam released from the RCS will
move the hydrogen-laden air from the points of generation toward the cool external
walls. Although hydrogen is lighter than air, it will not concentrate significantly in high areas because of the high diffusion rate, the open design of the containment, and the fan cooler air mixing.
The ability of hydrogen to diffuse rapidly into all volumes is inferred from a CSE
experiment (Reference 23). These tests showed very good mixing in the main chamber
and a rapid interchange by diffusion and mixing with the atmosphere of other chambers
that had limited communication. The diffusivity of hydrogen is approximately 10 times
that of iodine, so a more uniform mixture is expected for hydrogen. Also, higher
concentration provides greater concentration gradients for better diffusion than indicated
by the CSE tests.
Table 6.2-45 summarizes the calculated hydrogen production and accumulation data.
6.2.5.3.6.1.5 Results of the Hydrogen Generation and Accumulation Analyses The results of the hydrogen generation and accumulation analyses are presented in
Figures 6.2-26 through 6.2-29.
6.2.5.3.7 General Design Criterion 41, 1967 - Engineered Safety Features Performance Capability The CFCS, including required au xiliary systems, is designed to tolerate a single active failure following a LOCA without loss of protective function. Refer to Section 6.2.2.3.2
for further discussion.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-100 Revision 23 December 2016 6.2.5.3.8 General Design Criterion 42, 1967 - Engineered Safety Features Components Capability The component design pressure and temperature conditions in Table 6.2-26 are specified as the most severe conditions to which each CFCS component is exposed
during either normal or post-LOCA operation allowing the system to provide a mixed
atmosphere in post-LOCA conditions. Refer to Section 6.2.2.3.1 for further discussion.
6.2.5.3.9 General Design Criterion 49, 1967 - Containment Design Basis The CFCS ensures a mixed atmosphere within the containment structure to prevent the
ignition of high concentration hydrogen collec tions that may produce quick and large pressure and temperature rises.
The containment penetrations, including the CHPS and hydrogen monitoring system piping and valves required for containment isolation, are designed and analyzed to withstand the pressures an d temperatures that could res ult from a L OCA without exceeding design leakage rates.
Refer to Section 3.8.2.1.1.3 for additional details.
6.2.5.3.10 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment The CHPS and hydrogen monitoring system valves required for containment isolation
are periodically tested. Testing of the components required for the containment
isolation system (CIS) is discusse d in Section 6.2.4.
6.2.5.3.11 General Design Criterion 56, 1971 - Primary Containment Isolation Valves The CHPS and containment hydrogen monitoring system containment penetrations
comply with the requirements of GDC 56, 1971, as described in Section 6.2.4 and Table
6.2-39.
6.2.5.3.12 10 CFR 50.44 - Combustible Gas Control for Nuclear Power Reactors The CFCS is credited as the means for providing a mixed atmosphere in the
containment structure in accordance with 10 CFR 50.44. Refer to Section 6.2.2.2 for
discussion of the CFCS.
The hydrogen monitoring system is provided to continuously measure hydrogen
concentrations in the containment structure follo wing a significant beyond design basis accident for accident management, including emergency planning. Refer to
Section 6.2.5.5 for further details on the hydrogen monitoring instrumentation.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-101 Revision 23 December 2016 6.2.5.3.13 10 CFR 50.49 - Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants CHPS, EHRS and hydrogen monitoring system components 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. Section 3.11 describes the DCPP EQ Program and the requirements for the environmental design of electrical and related mechanical equipment. The affected
equipment, including the CFCS motors, containment isolation solenoid valves, containment isolation valve motor actuators, and the recombiners, are listed on the EQ
master equipment list.
6.2.5.3.14 10 CFR 50.55a(f) - Inservice Testing Requirements Periodic inservice testing (IST) of all containment isolation valves, in the system is
performed. The IST requirements are contained within the IST Program Plan and
comply with the ASME code for Operation and Maintenance of Nuclear Power Plants.
Refer to Section 6.2.2.4 for a discussion of testing of the CFCS.
6.2.5.3.15 10 CFR 50.55a(g) - Inservice Inspection Requirements The inservice inspection (ISI) requirements for the CHPS and containment hydrogen
monitoring penetrations are contained within the ISI Program Plan and comply with the
ASME BPVC,Section XI. Refer to Section 6.2.2.4 for a discussion of inspection of the CFCS.
6.2.5.3.16 Regulatory Guide 1.7, Revision 2, November 1978 - Control of Combustible Gas Concentrations in Containment Following a Loss-of-
Coolant Accident
The CFCS, serving as the credited means for containment atmosphere mixing in
accordance with 10 CFR 50.44, is designed and constructed to PG&E Design Class I
standards.
The hydrogen monitoring system includes two hydrogen monitors to measure the
hydrogen concentration in the containment.
The hydrogen monitoring system was origina lly designed and constructed as Category
1, as defined by Regulatory Guide 1.97, Revision 2, December 1980 however has been
reclassified as Category 3, as defined by Regulatory Guide 1.97, Revision 3, May 1983
as a result of the rulemaking revision to 10 CFR 50.44 (refer to Section 6.2.5.3.17).
The EHRS is available to control containment hydrogen concentration following a LOCA to at or below 4.0 percent by volume without relying on the CHPS. Each of the two
redundant recombiners is capable of providing the required removal capacity. The
CHPS is also available.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-102 Revision 23 December 2016 To ensure that the lower flammability limit (4 percent) will not be exceeded, the internal
electric hydrogen recombiners will be started at or below 3.5 percent by volume.
The EHRS and CHPS systems meet PG&E Design Class I design and construction
standards.
The EHRS provides 100 percent redundancy since each recombiner and its associated
power supply and control panel are capable of providing the required hydrogen removal capacity. The second unit, including its associated power supply and control panels, is
normally on standby following a postulated LOCA.
The CHPS is provided as a means to carry out controlled purging of the containment
atmosphere. While it is intended to serve only as an additional option for containment
hydrogen control, and may not be acceptable for operation following a LOCA, the CHPS otherwise satisfies the design requirements for an ESF system. Two redundant
systems complete with separate lines, blowers, and filters have been provided. These
lines, valves, instrumentation, and blowers are PG&E Design Class I. Each blower and
its associated controls are powered by independent electrical power supplies.
The basic data and analytical models and assumptions used for determining hydrogen production and accumulation shall be based on Regulatory Guide 1.7, Revision 2, November 1978 (Reference 22). The parameter values listed in Table 1 of Regulatory
Guide 1.7, Revision 2, November 1978 were used in calculating the hydrogen gas concentration in containment. Refer to Section 6.2.5.3.6 for further discussion of the analysis. In addition, refer to Section 15.5.17.2.9 for analysis of potential radiation exposures that could result from venting hydrogen.
The use of corrodible materials that yield hydrogen due to corrosion from the
emergency cooling or containment spray solutions has been controlled in the
containment to minimize hydrogen production.
6.2.5.3.17 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
Two Category 3 hydrogen monitors per unit monitor post-accident hydrogen concentration (refer to Figure 6.2-22) and indication is provided in the control room for
Regulatory Guide 1.97, Revision 3, May 1983 monitoring. Although designed to
Category 1, the monitors have been reclassified to meet the requirements of Category 3
as described in Section 6.2.5.3.16. Their characteristics are described in Section
6.2.5.5 and Table 7.5-6.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-103 Revision 23 December 2016 6.2.5.3.18 NUREG-0737 (Items II.E.4.1, II.F.1), November 1980 -
Clarification of TMI Action Plan Requirements Item II.E.4.1 - Dedicated Hydrogen Penetrations:
Dedicated penetrations are used for the purge system. Refer to Section 6.2.5.2.2 for a description of the CHPS and Sections 6.2.5.3.10 and 6.2.5.3.11 for discussions of GDC
54, 1971 and GDC 56, 1971, respectively.
Item II.F.1 - Additional Accident Monitoring Instrumentation:
Position (6) - The display instrumentation for containment hydrogen monitoring is
described in Section 6.2.5.5.1.
6.2.5.3.19 Generic Letter 89-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance The CHPS MOVs are subject to the require ments 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.
6.2.5.4 Tests and Inspections Tests and inspections for the containment hydrogen control equipment are discussed in
Sections 6.2.2, 6.2.5.3.14, and 6.2.5.3.15. The valves associated with containment
isolation will be tested as described in Sections 6.2.4.4.13 and 6.2.4.5. Tests and inspections of filters and fans are described in Section 9.4.
HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED
All PG&E Design Class I components of the combustible gas control systems were designed, fabricated, installed, and tested under quality assurance requirements in accordance with 10 CFR Part 50, Appendix B, as described in Chapter 17.
Nondestructive examination has been performed on the components of the systems in
accordance with the requirements of the applicable codes as described in Section 3.2.
The systems have been tested in accordance with the procedures outlined in Chapter
- 14. During preoperational startup testing of the plant, a functional test using predetermined sample hydrogen gas m ixtures was performed to verify hydrogen analyzer operation.
6.2.5.5 Instrumentation Applications Because of the possibility of hydrogen release to the containment atmosphere following
a LOCA, means to monitor and control the post-accident concentration of hydrogen in
the containment are necessary.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-104 Revision 23 December 2016 6.2.5.5.1 Hydrogen Monitoring System Two redundant hydrogen monitors are installed in each unit to provide continuous
indication and recording in the control room of containment hydrogen concentration.
The monitors are Instrument Class II, Type C, Regulatory Guide 1.97 Category 3 and
each has its own dedicated containment penetration and isolation valves to meet single failure criteria (refer to Figure 6.2-22 for syst em layout). The monitoring system is capable of sampling and measuring the hydrogen concentration inside the containment to diagnose the course of beyond-design-basis accidents. The system has a range of 0-10 percent by volume. The normal system configuration is with the hydrogen
monitors off-line and their respective containment isolation valves closed.
Refer to Sections 6.2.5.3.3, 6.2.5.3.16, and 6.2.5.3.17 for additional discussion on the
hydrogen monitors.
6.2.5.5.2 Containment Hydrogen Purge System Instrumentation is provided to monitor the flowrate and the amount of radioactivity
released by the purging operation.
The containment radiation monitoring system and the plant vent radiation monitors are
used to monitor the radioactivity in containment and the hydrogen purge line. Refer to
Sections 9.4.2 and 11.4 for information regarding the plant vent.
A manual sample point is provided on each exhaust line to obtain a grab sample for laboratory analysis.
Flow indicators are provided for each CHP S exhaust line. The indicators are PG&E Design Class I. The range is 500 to 4000 feet per minute (corresponding to flowrates of approximately 45 to 350 cfm).
Containment isolation valves status is shown on the main control board as indicated in
Table 6.2-39. Annunciation is provided to ala rm on high radioactivity, high flowrate, and fan failure.
Refer to Sections 6.2.5.3.3 and 6.2.5.3.4 for additional discussion on the CHPS instrumentation.
6.2.5.6 Materials Materials of construction of components are indicated in Section 6.2.5.2.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-105 Revision 23 December 2016 6.
2.6 REFERENCES
- 1. Deleted in Revision 18.
- 2. Deleted in Revision 18.
- 3. Deleted in Revision 18.
- 4. M. A. Styrikovich, et al, Atomnoya Energya, Vol. 17, No. 1, July 1964, pp. 45-49, (Translation in UD 621.039.562.5).
- 5. R. M. Kemper, Iodine Removal by Spray in the Diablo Canyon Station Containment, WCAP-7977, September 1973.
- 6. W. F. Pasedag and J. L. Gallagher, "Drop Size Distribution and Spray Effectiveness," Nuclear Technology, 10, 1971, p. 412.
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- 14. W. D. Fletcher, et al, "Post-LOCA Hydrogen Generation in PWR Containments," Journal of the American Nuclear Society, June 1970.
- 15. W. D. Fletcher, et al, "Post-LOCA Hydrogen Generation in PWR Containments", Nuclear Technology 10, 1971, pp. 420-427.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-106 Revision 23 December 2016
- 16. Draft Environmental Statement Concerning Proposed Rule-Making Action-Acceptance Criteria for Emergency Core Cooling Systems for Light-Water-Cooled Nuclear Power Reactor, USA EC Regulatory Staff, December 1972.
- 17. J. Sejvar, Distribution of Fission Product Decay Energy in PWR Cores, Westinghouse Electric Corporation, WCAP-7816, December 1971.
- 18. M. L. Mooney and H. E. Cramer, Meteorological Study of the Diablo Canyon Nuclear Power Plant Site, Pacific Gas and Electric Company, May 1970.
- 19. M. L. Mooney, Supplement No. 1 to Meteorology Report for Diablo Canyon Site, May 1971.
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- 21. W. K. Brunot, EMERALD - A Program for the Calculation of Activity Releases and Potential Doses from a Pressurized Water Reactor Plant, Pacific Gas and Electric Company, October 1971.
- 22. Regulatory Guide 1.7, Rev. 2, Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident, USNRC, November 1978.
- 23. H. E. Zittel, Radiation and Thermal Stability of Spray Solutions, ORNL-NSRD Program Bi-Monthly Report for May-June, 1969. ORNL-TM-2663, September, 1969.
- 24. J. J. DiNunno, et al, Calculation of Distance Factors for Power and Test Reactor Sites, AEC Report Number TID-14844, March 23, 1962.
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- 29. A. O. Allen, The Radiation Chemistry of Water and Aqueous Solutions, Princeton, N. J., Van Nostrand, 1961.
- 30. Deleted in Revision 18.
- 31. Deleted in Revision 18.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-107 Revision 23 December 2016
- 32. Deleted in Revision 18.
- 34. Deleted in Revision 18.
- 35. Deleted in Revision 18.
- 36. G. A. Israelson, J. R. v an Searen, W. C. Boettinger, Reactor Containment Fan Cooler Cooling Test Coil, WCAP-7336-L, 1969.
- 37. Deleted in Revision 18.
- 38. Diablo Canyon Power Plant - Inservice Inspection Program Plan - The Third 10-Year Inspection Interval, Pacific Gas and Electric Company.
- 39. EQ File IH-05, Containment Fan Cooler Motor, Pacific Gas and Electric Company. 40. Standard Review Plan 6.2.5, Combustible Gas Control.
- 41. Branch Technical Position CSB 6-2, Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident.
- 42. D. D Whyte., R. C. Burchell, Corrosion Study for Determining Hydrogen Generating From Aluminum and Zinc During Postaccident Conditions, WCAP-8776, April 1976.
- 43. Deleted in Revision 18.
- 44. Deleted in Revision 18.
- 45. Deleted in Revision 23.
- 46. Technical Specifications, Diablo Canyon Power Plant Units 1 and 2, Appendix A to License Nos. DPR-80 and DPR-82, as amended.
- 47. IEEE-Std-334, Guide for Type Tests of Clas s I Motors Installed Inside the Containment of Nuclear Power Generating Stations, 1971.
- 48. AEC Safety Guide 11, Instrument Lines Penetrating Primary Reactor Containment, March 10, 1971.
- 49. Deleted in Revision 18.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-108 Revision 23 December 2016
- 50. Westinghouse Letter PGE-91-533, Safety Evaluation for Containment Spray Flow Reduction, February 7, 1991.
- 51. Westinghouse Letter PGE-89-673, RWST Setpoint Evaluation, July 24, 1989.
- 52. Deleted in Revision 18.
- 53. Deleted in Revision 18.
- 54. Deleted in Revision 18.
- 55. J. C. Griess, A. L. Bacarella, Design Considerations of Reactor Containment Spray Systems - Part III. The Cor rosion of Materials in Spray Solutions, ORNL-TM-2412, Part III, December 1969.
- 56. Deleted in Revision 18.
- 57. Deleted in Revision 18.
- 58. Deleted in Revision 18.
- 59. Westinghouse Mass and Energy Release Data for Containment Design, WCAP-8264-P-A, Rev. 1, (Proprietary), WCAP-8312-A, August 1975.
- 60. Ice Condenser Containment Pressure Transient Analysis Methods, WCAP-8077 (Proprietary), WCAP-8078, March 1973.
- 61. Letter from Sheri R. Peterson (NRC - Office Nuclear Reactor Regulation) to Mr.
Gregory M. Rueger, Nuclear Power Generation, Pacific Gas and Electric Company, Leak-Before-Break Evaluation of Reactor Coolant System Piping for Diablo Canyon Nuclear Power Plant, Unit No. 1 (TAC NO. M83283) and Unit No.2 (TAC NO. M83284), March 2, 1993.
- 62. Cloud, R.L., et al, Evaluation of the Reactor Coolant System for Postulated Loss-of-Coolant Accidents for the Diablo Ca nyon Nuclear Plant, WCAP-9241 (Proprietary), December 1977.
- 63. Appendix E, "Acceptable Simpl ified Methods for Calculating Mass and Energy Release Rates for Compartment Pressurization Evaluation," to ANSI/ANS-58.2-1980, Design Basis for Protection of Light Water Nuclear Power Plants Against Effects of Postulated Pipe Rupture, approved December 31, 1980 by the American National Standards Institute, INC.
- 64. F.J. Moody, "Maximum Flow Rate of a Single Component, Two Phase Mixture," Journal of Heat Transfer, Transactions of the ASME, Series C, vol. 87, February 1965.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2-109 Revision 23 December 2016 6.2.7 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 6.3-1 Revision 23 December 2016 6.3 EMERGENCY CORE COOLING SYSTEM The emergency core cooling system (ECCS) is comprised of the refueling water storage
tank (RWST) and piping and components of the residual heat removal (RHR) system, safety injection (SI) system, and chemical and volume control system (CVCS).
The primary function of the ECCS is to deliver borated cooling water to the reactor core
in the event of a loss-of-coolant accident (LOCA). This limits the fuel cladding
temperature and thereby ensures that the core will remain intact and in place, with its
essential heat transfer geometry preserved. This protection is afforded for:
- All pipe break sizes up to and including the hypothetical circumferential rupture of a reactor coolant loop
- A loss of coolant associated with a rod ejection accident
The following normal operation functions are not covered in this section:
- The RHR system pumps, heat exchangers, valves, and associated piping are normally used during the latter stages of normal reactor cooldown and during
refueling operations (refer to Section 5.5.6).
- The centrifugal charging pumps (CCPs) are normally aligned for charging and letdown service; along with providing seal water to the reactor coolant pumps (refer to Section 9.3.4).
- The RWST is used to fill the refueling canal for refueling operations (refer to Section 9.1.4).
The CCPs, SI pumps, and RHR pumps are commonly referred to as "high-head
pumps," intermediate-head pumps, and "low-head pumps, respectively. The term "high-head injection" is used to denote CCP and SI pump injection; while the term "low-
head injection" refers to RHR pump injection.
6.3.1 Design Bases 6.3.1.1 General Design Criterion 2, 1967 - Performance Standards The ECCS is designed to withstand the effects of, or be protected against, natural phenomena, such as earthquakes, flooding, tornadoes, winds, and other local site
effects. 6.3.1.2 General Design Criterion 3, 1971 - Fire Protection The ECCS is designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-2 Revision 23 December 2016 6.3.1.3 General Design Criterion 11, 1967 - Control Room
The ECCS is designed to support actions to maintain and control the safe operational status of the plant from the control room.
6.3.1.4 General Design Criterion 12, 1967 - Instrumentation and Control Systems Instrumentation and controls are provided, as required, to monitor and maintain ECCS variables within prescribed operating ranges.
6.3.1.5 General Design Criterion 21, 1967 - Single Failure Definition
The ECCS is designed to tolerate a single fa ilure during the period of recovery following an accident without loss of its protective function, including multiple failures resulting from a single event, which is treated as a single failure.
6.3.1.6 General Design Criterion 37, 1967 - Engineered Safety Features Basis for Design The ECCS is designed to provide back-up to the safety provided by the core design, the reactor coolant pressure boundary (RCPB), and their protection systems. The ECCS is designed to cope with any size RCPB break, up to and including the circumferential
rupture of any pipe in that boundary, assuming unobstructed discharge from both ends.
6.3.1.7 General Design Criterion 38, 1967 - Reliability and Testability of Engineered Safety Features
The ECCS is designed to provide high functional reliability and ready testability.
6.3.1.8 General Design Criterion 40, 1967 - Missile Protection
The ECCS is designed to be protected against dynamic effects and missiles that might result from plant equipment failures.
6.3.1.9 General Design Criterion 41, 1967 - Engineered Safety Features Performance Capability The ECCS is designed to provide sufficient performance capability to accommodate a partial loss of installed capacity, such as a single failure of an active component, and still perform its required safety function.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-3 Revision 23 December 2016 6.3.1.10 General Design Criterion 42, 1967 - Engineered Safety Features Components Capability The ECCS is designed so that the capability of each component and system to perform its required function is not impaired by the effects of a LOCA.
6.3.1.11 General Design Criterion 43, 1967 - Accident Aggravation Prevention
The ECCS is designed so that any action of the engineered safety features (ESF) which might accentuate the adverse after effects of the loss of normal cooling is avoided.
6.3.1.12 General Design Criterion 44, 1967 - Emergency Core Cooling Systems Capability The ECCS is designed to provide the capability of accomplishing abundant emergency core cooling with two systems of different design principles. Each ECCS and the core are designed to prevent fuel and clad damage that would interfere with the emergency core cooling function and to limit the clad metal-water reaction to negligible amounts for
all sizes of breaks in the RCPB, including the double-ended rupture of the largest pipe.
The performance of each ECCS is evaluated conservatively in each area of uncertainty.
The systems do not share active components and do not share other features or
components unless it is demonstrated that (a) the capability of the shared feature or
component to perform its required function is readily ascertained during reactor
operation, (b) failure of the shared feature or component does not initiate a LOCA, and (c) capability of the shared feature or compo nent to perform its required function is not impaired by the effects of a LOCA and is not lost during the entire period this function is required following the accident.
6.3.1.13 General Design Criterion 45, 1967 - Inspection of Emergency Core Cooling Systems The ECCS is designed to facilitate physical inspection of all critical parts, including reactor vessel internals and water injection nozzles.
6.3.1.14 General Design Criterion 46, 1967 - Testing of Emergency Core Cooling Systems Components The ECCS is designed so that active components, such as pumps and valves, can be tested periodically for operability and required functional performance.
6.3.1.15 General Design Criterion 47, 1967 - Testing of Emergency Core Cooling Systems The ECCS is designed to provide the capabi lity to periodically test the delivery capability at a location as close to the core as is practical.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-4 Revision 23 December 2016 6.3.1.16 General Design Criterion 48, 1967 - Testing of Operational Sequence of Emergency Core Cooling Systems The ECCS is designed to provide the capability to test, under conditions as close to design as practical, the full operational sequence that would bring the ECCS into action, including the transfer to alternate power sources.
6.3.1.17 General Design Criterion 49, 1967 - Containment Design Basis
The emergency core cooling systems 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 LOCA, including a considerable margin for effec ts from metal-water or other chemical reactions that could occur as a consequence of failure of emergency core cooling systems. 6.3.1.18 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment The piping that is part of the ECCS that penetrates containment is provided with leak detection, isolation, redundancy, reliability, and performance capabilities which reflect
the importance to safety of isolating this syst em. 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.
6.3.1.19 General Design Criterion 55, 1971 - Reactor Coolant Pressure Boundary Penetrating Containment Each ECCS line that penetrates the containment is provided with containment isolation valves. 6.3.1.20 General Design Criterion 56, 1971 - Primary Containment Isolation
The ECSS contains valving in piping that penetrates containment and that is connected directly to the containment atmosphere. Remote manual isolation valves are provided
outside containment and either automatic (check) valves are provided inside
containment, or the system outside containment is considered a closed system, to
ensure containment integrity is maintained.
6.3.1.21 Emergency Core Cooling System Safety Function Requirements (1) Leakage Provisions and Flooding Protection The ECCS includes provisions to detect and accommodate leakage during post-LOCA operation and mitigate any consequential flooding.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-5 Revision 23 December 2016 6.3.1.22 10 CFR 50.46 - Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Plants The ECCS is designed so that its calculated cooling performance f ollowing postulated LOCAs conforms to the criteria set forth in 10 CFR 50.46. ECCS cooling performance is calculated in accordance with an acceptable evaluation model for a number of
postulated LOCAs of different sizes, locations, and other properties sufficient to provide
assurance that the most severe postulated LOCAs are calculated.
6.3.1.23 10 CFR 50.49 - Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants ECCS components that require environmental qualification (EQ) are qualified to the requirements of 10 CFR 50.49.
6.3.1.24 10 CFR 50.55a(f) - Inservice Testing Requirements ECCS ASME Code components are tested to the requirements of 10 CFR 50.55a(f)(4) and a(f)(5) to the extent practical.
6.3.1.25 10 CFR 50.55a(g) - Inservice Inspection Requirements ECCS ASME Code components are inspected to the requirements of 10 CFR 50.55a(g)(4) and a(g)(5) to the extent practical.
6.3.1.26 Safety Guide 1, November 1970 - Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System
Pumps The ECCS is designed such that adequate net positive suction head (NPSH) is provided to system pumps assuming maximum expected temperatures of pumped fluids and no increase in containment pressure fro m that present prior to postulated LOCAs.
6.3.1.27 Regulatory Guide 1.79, June 1974 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors A series of comprehensive preoperational te sts were performed on the ECCS in accordance with Regulatory Guide 1.79, June 1974, with noted exceptions, to assure the ECCS will accomplish its intended function when required.
6.3.1.28 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 ECCS provides instrumentation to monitor RHR system flow, SI pump discharge flow, CCP 1 and 2 injection header flow, RHR heat exchanger outlet temperature, DCPP UNITS 1 &
2 FSAR UPDATE 6.3-6 Revision 23 December 2016 containment recirculation sump water level and temperature, accumulator tank level and pressure, accumulator isolation valve position, RWST level, containment isolation valve
position, and subcooling margin indication du ring and following an accident.
6.3.1.29 NUREG-0737 (Items I.C.1, I.D.2, II.B.2, II.F.1, I I.F.2, II.K.3.30, II.K.3.31, III.D.1.1), November 1980 -
Clarification of TMI Action Plan Requirements Item I.C.1 - Guidance for the Evaluation and Development of Procedures for Transients and Accidents: NUREG-0737, Supplement 1, January 1983 provides the requirements for I.C.1 as follows:
Section 7.1(b) - Transients and accidents were reanalyzed for the purposes of
preparing technical guidelines an d upgrading emergency operating procedures.
Item I.D.2 - Plant Safety Parameter Display Console: NUREG-0737, Supplement 1, January 1983 provides the requirements for I.D.2 as follows:
Section 4.1(f)(v), Containment Conditions: The ECCS provides instrumentation for
control room personnel to monitor containment recirculation sump water level during
and following an accident. Additional monitors are provided in the technical support center (TSC) and emergency operations facility (EOF).
Item II.B.2 - Design Review of Plant Shielding and Environmental Qualification of Equipment for Space/Systems Which May Be Used in Postaccident Operations: Plant
shielding provides adequate access to, and occupancy of, the switchgear rooms for the purpose of restoring power to normally de-energized ECCS valves.
Item II.F.1 - Additional Accident Monitoring Instrumentation:
Position (5) - The ECCS provides continuous instrumentation to monitor containment
recirculation sump water level in the control room.
Item II.F.2 - Instrumentation for Detection of Inadequate Core Cooling: ECCS instrumentation provides an unambiguous indication of inadequate core cooling by
indicating the existence of inadequate core cooling caused by various phenomena and does not erroneously indicate inadequate core cooling due to the presence of an
unrelated phenomenon. 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 complete core uncovery.
Item II.K.3.30 - Revised Small-Break Loss-Of-Coolant-Accident Methods to Show
Compliance with 10 CFR Part 50, Appendix K: The analysis method for small-break
loss-of-coolant accidents (SBLOCAs) was revised for compliance with 10 CFR Part 50, Appendix K. The revision accounts for comparisons with experimental data, including data from the loss of fluid test and semiscale test facilities.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-7 Revision 23 December 2016 Item II.K.3.31 - Plant-Specific Calculations to Show Compliance with 10 CFR Part 50.46: Plant-specific calculations for SBLOCA, using a Nuclear Regulatory Commission (NRC) approved model, show compliance with 10 CFR 50.46.
Item III.D.1.1 - Integrity of Systems Outside Containment Likely to Contain Radioactive
Material for Pressurized-Water Reactors and Boiling-Water Reactors: Appropriate
portions of the ECCS are periodically pressure leak tested and visually inspected for leakage into the building environment.
6.3.1.30 Generic Letter 89-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance ECCS PG&E Design Class I and position changeable motor-operated valves (MOVs) meet the requirements of Generic Letter 89-10, June 1989, and associated Generic
Letter 96-05, September 1996, Periodic Verification of Design-Basis Capability of
Safety-Related Motor-Operated Valves.
6.3.1.31 Generic Letter 95-07, August 1995 - Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves ECCS PG&E Design Class I, power-operated gate valves meet the requirements of Generic Letter 95-07, August 1995.
6.3.1.32 Generic Letter 96-06, September 1996 - Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions
ECCS piping has been evaluated for the issue of thermal overpressurization of isolated piping sections that could affect c ontainment integrity during accident conditions, as described in Generic Letter 96-06, September 1996.
6.3.1.33 Generic Letter 97-04, October 1997
- Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment
Heat Removal Pumps The ECCS has been evaluated to assure adequate NPSH is available to ECCS pumps under all design basis accident scenarios.
6.3.1.34 Generic Letter 98-04, July 1998 - Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System After a Loss-of-Coolant Accident Because of Construction and
Protective Coating Deficiencies and Foreign Material in Containment The ECCS and containment recirculation sump have been evaluated to assure construction and protective coating deficiencies or foreign material in containment will DCPP UNITS 1 &
2 FSAR UPDATE 6.3-8 Revision 23 December 2016 not cause degradation of the ECCS, as required by Generic Letter 98-04, July 1998 and discussed in associated NRC Bulletin 93-02, May 1993, Debris Plugging of Emergency Core Cooling Suction Strainers.
6.3.1.35 Generic Letter 2004-02, September 2004 - Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents
at Pressurized-Water Reactors The ECCS and containment recirculation sump have been evaluated to assure that potential debris blockage due to a design b asis accident will not impact the ECCS safety-related functions, as required by Generic Letter 2004-02, September 2004.
6.3.1.36 Generic Letter 2008-01, January 2008 - Managing Gas Accumulation in Emergency Core Cooling, Decay H eat Removal, and Containment Spray
Systems The ECCS is designed, periodically inspected, and procedurally managed to assure that gas accumulation in ECCS piping will not impact the ECCS safety-related functions, as required by Generic Letter 2008-01, January 2008.
6.3.1.37 IE Bulletin 79-06A (Position 8), April 1979 - Review of Operational Errors and System Misalignments Id entified During the Three Mile Island Incident Position (8) (subsequently NUREG-0737, November 1980, Item II.K.1.5, Safety Related Valve Position):
ECCS PG&E Design Class I valve positions, positioning requirements, and positive
controls have been assured such that the valves remain positioned (open or closed) in a
manner to ensure the proper operation of the ESF to satisfy Position (8) of IE Bulletin 79-06A, April 1979.
6.3.1.38 IE Bulletin 80-18, July 1980 - Maintenance of Adequate Minimum Flow Thru Centrifugal Charging Pumps Following Secondary Side High
Energy Line Rupture The availability of adequate ECCS CCP minimum flow has been ensured under all conditions in order to protect the pumps during the possible deadheading conditions described in IE Bulletin 80-18, July 1980.
6.3.1.39 NRC Bulletin 88-04, May 1988
- Potential Safety-Related Pump Loss The ECCS is designed such that PG&E Desi gn Class I pumps that share a common minimum flow recirculation line will not be susceptible to the pump-to-pump interaction described in NRC Bulletin 88-04, May 1988.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-9 Revision 23 December 2016 6.3.1.40 NRC Bulletin 88-08, June 1988 - Thermal Stresses in Piping Connected to Reactor Coolant Systems Unisolable ECCS pipi ng sections connected to the RCS, which have the potential to be subjected to unacceptable thermal stresses due to temperature stratifications induced by leaking valves, have been identified. Means have been provided to ensure that the
pressure upstream from block valves, which might leak, is monitored and controlled.
6.3.1.41 NRC Bulletin 2003-01, June 2003 - Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors Modifications have been implemented to prevent potential blockage of drainage paths to the containment recirculation sump due to post-accident debris.
6.3.1.42 Branch Technical Position EICSB 18, November 1975 - Application of the Single Failure Criterion to Manually-Controlled Electrically-Operated Valves Certain ECCS manually-controlled, electrically-operated valves have their electric power removed during normal operation to satisfy the single failure criterion, as discussed in
Branch Technical Position EICSB 18, November 1975. Continuous, redundant position indication for these valves is provided in the control room.
6.3.2 System Description The ECCS is designed to cool the reactor core as well as to provide additional shutdown capability following initiation of the followi ng accident conditions:
(1) A pipe break or spurious valve lifting in the reactor coolant system (RCS) that causes a discharge larger than that which can be made up by the
normal makeup system, up to and including the circumferential rupture of
the largest pipe in the RCS (refer to Sections 15.3.1 and 15.4.1 for a
discussion of these accidents.)
(2) Rupture of a control rod drive mechanism (CRDM) causing a rod cluster control assembly (RCCA) ejection accident (refer to Section 15.4.6)
(3) A pipe break or spurious valve lifting in the steam system, up to and including the instantaneous circumferential rupture of the largest pipe in
the steam system (refer to Sections 15.2.14, 15.3.2, and 15.4.2)
(4) A steam generator tube rupture (SGTR) (refer to Section 15.4.3)
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-10 Revision 23 December 2016 6.3.2.1 Range of Coolant Ruptures and Leaks Sections 15.3.1 and 15.4.1 provide discussion on the ranges of coolant ruptures and
leaks evaluated for ECCS performance.
6.3.2.2 Fission Product Decay Heat The primary function of the ECCS following a LOCA is to remove the stored and fission
product decay heat from the reactor core to prevent fuel rod damage to the extent that such damage may impair effective core cooling. The acceptance criteria for the
accidents, as well as their analyses, are provided in Sections 15.3.1 and 15.4.1.
6.3.2.3 Reactivity Required for Cold Shutdown The ECCS provides shutdown capability for the accidents listed in Section 6.3.2 by means of shutdown chemical (boron) injection. The most critical accident for shutdown
capability is the main steam line break (MSLB) and for this accident the ECCS meets
the criteria defined in Sections 15.3.2 and 15.4.2.
6.3.2.4 Equipment and Component Descriptions The major components of the ECCS are described in the following sections. Pertinent design and operating parameters for ECCS components are given in Table 6.3-1.
6.3.2.4.1 Accumulators
The accumulators are pressure vessels partially filled with borated water and pressurized with nitrogen gas. During normal operation each accumulator is isolated
from the RCS by two check valves in series. Should the RCS pressure fall below the
accumulator pressure (refer to Table 6.3-1), the check valves open and borated water is
forced into the RCS. One accumulator is attached to each of the cold legs of the RCS.
Mechanical operation of the swing-disk check valves is the only action required to open
the injection path from the accumulators to the core via the cold leg. Sections
6.3.3.4.5.1 and 6.3.3.5.5 describe the accumulator MOVs and their position indicators.
Connections are provided to remotely adjust the level and boron concentration of the
borated water in each accumulator during normal plant operation, as required.
Accumulator water level may be adjusted either by draining to the reactor coolant drain
tank and then to the liquid holdup tank (LHUT), or by pumping borated water from the
RWST to the accumulator. Samples of the solution in the accumulators are taken periodically to check boron concentration.
Accumulator pressure is provided by a supply of nitrogen gas and can be adjusted as
required during normal plant operation. The accumulators are, however, normally
isolated from this nitrogen supply. Gas relief valves on the accumulators protect them
from pressures in excess of design pressure.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-11 Revision 23 December 2016 The accumulators are located within the containment but outside of the secondary
shield wall, which protects them from missiles. Since the accumulators are located
within the containment, a release of the nitrogen gas from the accumulators would
cause an increase in normal containment pressure. Containment pressure increase
following release of the gas from all accumulators has been calculated and is well below the containment pressure setpoint for ECCS actuation.
Release of accumulator gas is detected by the accumulator pressure indicators and
alarms. Thus, the operator can take action promptly as required to maintain plant
operation within the requirements of the Technical Specification (Reference 10)
covering accumulator operability.
6.3.2.4.2 Refueling Water Storage Tank The content of the RWST is normally used to supply borated water to the refueling
canal for refueling operations. In addition to its usual service, this tank provides borated water to the ECCS pumps and the containment spray system (CSS) pumps following a
LOCA or MSLB. For RWST volume requirements, refer to Section 6.3.2.11 and Table 6.3-1. For ECCS operation requirements followin g a LOCA or MSLB, refer to Section 6.3.3.6.1.1.
During normal operation, the RWST is aligned to the suction of the SI pumps, RHR
pumps, and CSS pumps. The suction of CCP1 and CCP2 is automatically aligned to
the tank by the safety injection signal ("S" signal).
The water in this tank is borated to a minimum concentration of 2300 ppm boron that ensures reactor shutdown by at least 5 percent k/k when all RCCAs are inserted with the most reactive RCCA completely removed from its fuel assembly, and when the reactor is cooled down for refueling.
The RWST is vented directly to the atmosphere.
6.3.2.4.3 Pumps 6.3.2.4.3.1 Residual Heat Removal Pumps The RHR pumps are provided to deliver water from the RWST to the RCS should the
RCS pressure fall below their shutoff head during the injection phase. The pumps are
automatically started upon receipt of the "S" signal. Each RHR pump is a single-stage, vertical, centrifugal pump. It has an integral motor-pump shaft, driven by an induction motor, and is powered by the Class 1E 4.16-kV system. The unit has a self-contained
mechanical seal, which is cooled by component cooling water (CCW).
During the injection mode of ECCS operation, the RHR pumps draw water from the
RWST; during the recirculation mode, they draw water from the containment DCPP UNITS 1 &
2 FSAR UPDATE 6.3-12 Revision 23 December 2016 recirculation sump. The changeover from the injection mode to recirculation mode (refer to Section 6.3.3.6.1.1.2 and Table 6.3-5) is initiated by low level in the RWST, which results in an automatic trip of the RHR pumps. The operator then manually
changes system alignment to the recirculation mode after water is available in the containment recirculation sump and before water is exhausted from the RWST.
Adequate NPSH is always available to the RHR pumps in both the injection phase and the recirculation phase. Table 6.3-11 lists available and required NPSH. Phase I of the
preoperational system test (refer to Section 6.3.3.27.1.1) verified that the RHR pump performance was satisfactory for all required alignments.
A minimum flow recirculation line is provided for the pumps to recirculate fluid through
the RHR heat exchangers and return the co oled fluid to the pump suction should these pumps be started with their normal flowpaths blocked. Once flow to the RCS is
established, the recirculation line is automatically closed. This line prevents deadheading the pumps and permits pump testing during normal operation. The RHR
pumps are also discussed in Section 5.5.6.
6.3.2.4.3.2 Centrifugal Charging Pumps (CCP1 and CCP2)
When aligned for safety injection operation, CCP1 and CCP2 deliver water from the
RWST to the RCS at the prevailing RCS pressure. The pumps are automatically
started upon receipt of the "S" signal. CCP1 and CCP2 are multistage, diffuser design, barrel-type casing pumps with vertical suction and discharge nozzles.
The pump is driven by an induction motor, powered by the Class 1E 4.16-kV system.
The unit has a self-contained lubrication system cooled by CCW and a mechanical seal system that requires no external cooling.
A minimum flow bypass line is provided on each pump discharge to recirculate flow to
the pump suction after cooling in the seal water heat exchanger during normal
operation. Valves in minimum flow bypass lines are closed by the operator when the
ECCS is transferred to the recirculation mode of operation following a LOCA. During
normal plant operation, CCP1 or CCP2 may be in use. CCP1 and CCP2 may be tested
during normal operation through the use of the minimum flow bypass line.
6.3.2.4.3.3 Safety Injection Pumps SI pumps deliver water from the RWST to the RCS after the RCS pressure is reduced below the shutoff head of the pumps. The pumps are automatically started upon receipt
of the "S" signal. Each SI pump is a multistage, centrifugal pump. The pump is driven
directly by an induction motor, powered by the Class 1E 4.16-kV system. The unit has
a self-contained lubrication system and a mechanic al seal system that are cooled by CCW.
A minimum flow bypass line is provided on each pump discharge to recirculate flow to
the RWST in the event the pumps are started with the normal flowpaths blocked. This DCPP UNITS 1 &
2 FSAR UPDATE 6.3-13 Revision 23 December 2016 line also permits pump testing during normal operation. Two MOVs in series are provided in this line. These valves are closed by operator action during the switchover
to the ECCS recirculation mode.
6.3.2.4.4 Residual Heat Removal Heat Exchangers The RHR heat exchangers are conventional shell- and U-tube-type units. During
normal cooldown of the primary system, reactor coolant flows through the tube side
while CCW flows through the shell side. During the emergency core cooling
recirculation phase, water, from the contain ment recirculation sump, flows through the tube side. Further discussion of the RHR heat exchangers is found in Section 5.5.6.
Design characteristics are included in Table 5.5-10.
6.3.2.4.5 Valves Stroke times for MOVs used in the ECCS are given in Table 6.3-1. This table also lists
the leakage specification for the various types of valves used in the ECCS. Setpoints
and capacities for relief valves are given in Table 6.3-10.
Design features employed to minimize valve leakag e include the following:
(1) Globe valves, which during post-accident recirculation are normally closed, are installed with the recirculated fluid pressure under the seat to
prevent stem leakage of recirculated (radioactive) water.
(2) Relief valves are enclosed; i.e., t hey are provided with a closed bonnet and discharge to a closed system.
(3) Control and motor-operated valves (2-1/2 inches and above) in the RHR portion of the ECCS recirculation loop outside containment have
double-packed stuffing boxes and stem leakoff connections to the
equipment drain system. Valves in the other portions of the ECCS
recirculation loop outside containment have their leakoff connections
capped. 6.3.2.4.5.1 Motor-Operated Gate Valves The seating design of all motor-operated gate valves is of either the parallel-disk design
or the flexible wedge design. The seating surfaces are hard-faced to prevent galling
and reduce wear.
Where a gasket is employed for the body-to-bonnet joint, it is either a fully trapped, controlled compression, spiral-wound gasket with provisions for seal welding, or of the pressure-seal design. The valve stuffing boxes are designed with a lantern ring leakoff
connection with a minimum of a full set of packing below the lantern ring and a minimum DCPP UNITS 1 &
2 FSAR UPDATE 6.3-14 Revision 23 December 2016 of one-half of a set of packing above the lantern ring for valves that have leakoff connections piped to an equipment drain system.
The motor operator incorporates a "hammer blow" feature that allows the motor to attain
its full speed prior to being placed under load. MOVs are further discussed in Sections
6.3.3.30 and 6.3.3.31.
6.3.2.4.5.2 Accumulator Check Valves (Swing-Disk)
The accumulator check valves are designed with a low pressure drop configuration with
all operating parts contained within the body.
Design considerations and analyses that ensure that leakage across the check valves located in each accumulator injection line will not impair accumulator availability are as
follows:
(1) During normal operation the check valves are in the closed position with a nominal differential pressure across the disk of approximately 1650 psi.
Since the valves remain in this position except for testing or when called
upon to function, and are therefore not subject to abuse of flowing
operation or impact loads caused by sudden flow reversal and seating, they do not experience significant wear of the moving parts and are
expected to function with minimal leakage.
(2) When the RCS is being pressurized during the normal plant heatup operation, the check valves are tested for leakage as soon as there is a stable differential pressure of about 100 psi or more across the valve.
This test confirms the seating of the disk and whether or not there has
been an increase in the leakage since the last test. When this test is
completed, the discharge line motor-operated isolation valves are opened
and the RCS pressure increase is continued. There should be no
increase in leakage from this point on since increasing reactor coolant
pressure increases the seating force and decreases the probability of
leakage. HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO
BE REVISED (3) The experience derived from the check valves employed in the emergency injection systems indicates that the system is reliable and workable. This is substantiated by the satisfactory experience obtained from operation of
the Ginna and subsequent plants where the usage of check valves is
identical to this application.
(4) The accumulators can accept some inleakage from the RCS without affecting availability. Inleakage would require, however, that the DCPP UNITS 1 &
2 FSAR UPDATE 6.3-15 Revision 23 December 2016 accumulator water volume be adjusted in accordance with Technical Specification (Reference 10) requirements.
6.3.2.4.5.3 Relief Valves Relief valves are installed in various sections of the ECCS to protect the system from overpressure. Valves that normally see liquid service have their stem and spring
adjustment assemblies isolated from the system fluid by a bellows seal between the
valve disk and spindle. The closed bonnet provides an additional barrier for enclosure of the relief valves. Table 6.3-10 lists the system relief valves with their capacities and
setpoints. The accumulator relief valves are sized to pass nitrogen gas at a rate in
excess of the accumulator gas fill line delivery rate.
The relief valves will also pass water in excess of the maximum water fill rate, but this is not considered important, because the time required to fill the gas space gives the operator ample opportunity to
correct the situation.
6.3.2.4.5.4 Ball Valves Each main RHR line has an air-operated ball valve, which is normally open and is designed to fail in the open position, thus max imizing flow from this system to the RCS during ECCS operation. These ball valves at the discharge of each RHR system heat
exchanger along with the ball valve in the RHR system heat exchanger bypass line are adjusted during RHR system operation to meet the design plant cooldown
requirements.
6.3.2.4.6 Piping All piping joints are either welded, flanged, or threaded connections.
Weld connections for pipes sized 2-1/2 inches and larger are butt-welded.
Minimum piping and fitting wall thickness, as determined by ANSI B31.7-1969 with 1970
Addenda Code formula, are increased to account for the material specifications permissible tolerance on the nominal wall and an appropriate allowance for wall thinning
on the external radius during any pipe bendi ng operations in the shop fabrication of the subassemblies.
6.3.2.4.7 Heat Tracing With the lowering of the normal boric acid solution from 12 to 4 percent, for Cycle 5 and
later, the heat tracing on piping, valves, flanges, and instrumentation lines carrying boric
acid solution were downgraded to PG&E Design Class II. Refer to Section 9.3.4.2.9.30
for further discussion.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-16 Revision 23 December 2016 6.3.2.4.8 Containment Recirculation Sump and Strainer The containment recirculation sump and strainer is a large collecting reservoir designed
to provide an adequate supply of water with a minimum amount of particulate matter to
the SI system, CCP1 and CCP2, the RHR system, and the CSS, if recirculation spray is
used, during the recirculation mode of ECCS operation following a postulated LOCA.
The sump is located in the annulus area of the containment between the crane wall and
the containment liner at the 91 foot elevation.
It is approximately 50 feet from RCS piping and components that could become sources of debris. The two 14-inch suction
pipes are located on opposite sides of the sump for ECCS train separation. In the
sump, there is a strainer assembly, which includes a trash rack with integral debris curb
and two strainers (front and rear). Each strainer is designed with 3/32-inch nominal
diameter openings, which are sized for the smallest credible restriction in the ECCS
flowpath. Each strainer has a plenum. These plenums come together above the two
14-inch RHR pump suction lines and both plenums feed both RHR lines. The lower or
rear plenum assembly is water tight to allow collection of any potential back leakage
from the RHR system.
The strainer assemblies are designed to minimize blockage. The design provides
enough screen area to ensure, with maximum accident debris loads and RHR flowrates, the RHR system has sufficient NPSH margin. Since both plenums feed both RHR
trains, without complete blockage of both strainer assemblies, there is no condition that
would cause both RHR trains to fail.
The physical arrangement of the strainer for DCPP Unit 1 is shown in Figure 6.3-6 and the DCPP Unit 2 arrangement is shown in Figure 6.3-7.
Any debris or other matter that passes into the sump through the 3/32-inch maximum
hole size allowed for the strainers will pass through the SI system, CCP1 and CCP2, the
RHR system, and the CSS, if recirculation spray is used, without restriction and
eventually will be pumped back into the cont ainment. This is based on the maximum nominal-sized debris (i.e., the evaluation is based on a 1/8 inch diameter, therefore
conservative using a 3/32 inch diameter) that could potentially pass through the sump
strainers and pass through the ECCS throttle valves.
Safety Injection System Flow Path
Fluid from the containment recirculation strainer passes into the 14 inch RHR pump
suction piping. The flow passes through the RHR pumps and heat exchangers. The
flowpath continues from the exit of the RHR heat exchanger to the suction of the SI
pumps. The SI pumps discharge to either the cold or hot legs of the RCS loops. The
flowpath continues through the RCS cold or hot legs into the core, through the reactor
vessel, into the ruptured RCS loop, through the rupture into the containment and, finally, ends in the containment recirculation sump.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-17 Revision 23 December 2016 Centrifugal Charging Pump System (CCP1 and CCP2) Flow Path
Fluid from the containment recirculation strainer passes into the 14 inch RHR pump
suction piping. The flow passes through the RHR pumps and heat exchangers. The
flowpath continues from the exit of the RHR heat exchangers to the suction of CCP1
and CCP2. From these CCPs, it goes through the charging injection to the cold legs of
the RCS loops. The flowpath continues through the 27-1/2 inch piping of the RCS cold
legs, into the core, through the reactor vessel, into the ruptured RCS loop, through the
rupture into the containment and, finally, into the containment recirculation sump.
CCP3, which replaced the positive displacement pump, is not credited as part of the
centrifugal charging pump system (CCP1 and CCP2) described within this section and
Chapter 15.
Residual Heat Removal System Flow Path
Fluid from the containment recirculation strainer passes into the 14 inch RHR pump
suction piping. The flow passes through the RHR pumps and heat exchangers. The
flowpath continues from the exit of the RHR heat exchangers to either the cold or hot
legs of the RCS loops. The flowpath continues through the cold legs or hot legs of the
RCS loop piping, into the core, through the reactor vessel, into the ruptured RCS loop, through the rupture into the containment and, finally, into the containment recirculation
sump. During normal plant operation, when the RCS is hot and pressurized, there is no
direct connection between the RWST and the RCS. Also during normal plant shutdown, when the RCS is being cooled down and the RHR system begins to operate, the RHR
system is isolated from the RWST by an MOV in addition to a check valve.
Containment Spray System Flow Path
Fluid from the containment recirculation strainer passes into the 14 inch RHR pump
suction piping. The flow passes through the RHR pumps and heat exchangers. The
flowpath continues from the exit of the RHR heat exchangers to the spray headers and
out of the 3/8 inch spray nozzles into the containment. Finally, fluid in the containment
drains into the containment recirculation sump. The CSS is discussed in Section 6.2.2.
6.3.2.4.8.1 Sump Strainer Submergence During an SBLOCA
Since SBLOCA scenarios may result in a partially-submerged strainer, a set of flow straighteners were added to the strainer system to reduce the tendency to vortex and to
eliminate air entrainment as a concern. The performance of the DCPP strainers at water
levels representative of a SBLOCA will not result in conditions which would entrain excessive amounts of air into the suctions of the RHR pumps.
6.3.2.5 Motor-Operated Valves and Controls Each containment recirculation sump isolation valve is interlocked with its respective
pump suction/RWST isolation valve to the RHR system. The interlock is provided with DCPP UNITS 1 &
2 FSAR UPDATE 6.3-18 Revision 23 December 2016 redundant signals from each isolation valve. This interlock prevents opening the sump isolation valve when the RWST isolation valves are open and thus prevents dumping the RWST contents into the containment recirculation sump.
To preclude spurious movement of specific MOVs that could result in a loss of ECCS
function, electric power is removed from certain valves during normal operation. These
valves are listed in Table 6.3-12 and further discussion is provided in Section 6.3.3.42.
6.3.2.6 Schematic Piping and Instrumentation Diagrams Piping schematic diagrams of the ECCS are shown in Figures 3.2-8, 3.2-9, and 3.2-10.
6.3.2.7 ECCS Flow Diagrams Alignment of the major ECCS components during the injection and recirculation phases
is shown in Figures 6.3-4 and 6.3-5, respectively. Tables 15.3-2 and 15.3-3 summarize
the calculated times at which the major components perform their safety-related
functions for various accident conditions (tabulated in Table 15.1-2) that require ECCS
operations.
6.3.2.8 Applicable Codes and Classifications The codes and standards to which the indivi dual ECCS components are designed are listed in Table 6.3-2.
6.3.2.9 Materials Table 6.3-3 lists the materials used in ECCS components.
6.3.2.9.1 Material Specifications and Compatibility Materials employed for ECCS components are given in Table 6.3-3. Materials are
selected to meet the applicable material code requirements of Table 6.3-2 and the
following additional requirements:
(1) All parts of components in contact with borated water are fabricated of or clad with austenitic stainless steel or similar corrosion-resistant material.
(2) All parts of components in contact (internal) with sump solution during recirculation are fabricated of austenitic stainless steel or similar corrosion
resistant material.
(3) Valve seating surfaces are hard-faced with Stellite No. 6 or similar to prevent galling and to reduce wear.
(4) Valve stem materials were selected for their corrosion resistance, high tensile properties, and resistance to surface scoring by the packing.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-19 Revision 23 December 2016 The elevated temperature of the sump solution during recirculation is well within the
design temperature of all ECCS components. In addition, consideration has been given to the potential for corrosion of various types of metals exposed to the fluid conditions
prevalent immediately after the accident or during long-term recirculation operations.
Environmental testing of ECCS equipment inside the containment, which is required to
operate following a LOCA, is discussed in S ection 3.11. The results of the test program indicate that the equipment will operate satisfactorily during and following exposure to
the combined containment post-accident environmental temperature, pressure, chemistry, and radiation.
6.3.2.10 Design Pressures and Temperatures The component design pressure and temperature conditions as given in Table 6.3-1 are
specified as the most severe conditions to which each component is exposed to during
either normal plant operation or during ECCS operation.
6.3.2.11 Coolant Quantity The quantities and minimum boron concentration of coolant available to the ECCS are
summarized in Table 6.3-1.
The minimum volume that will be maintained in the RWST is 455,300 gallons (this
includes the usable and unusable volume). In the event of a LOCA, this volume also
provides a sufficient amount of borated water to meet the following requirements:
(1) Provide adequate coolant during the injection phase to meet ECCS design objectives (refer to Sections 15.3 and 15.4).
(2) Increase the boron concentration of reactor coolant and recirculation water to a point that ensures no return to criticality with the reactor at cold shutdown and all control rods, except the most reactive RCCA, inserted
into the core (refer to Sections 15.3.1 and 15.4.1).
(3) Fill the containment recirculation sump to support continued operation of the ECCS pumps at the time of transfer from the injection mode to the
recirculation mode of cooling. The sump strainer assemblies are required
to be fully submerged to prevent vortexing and air ingestion, during
changeover from the injection mode to the recirculation mode, for a large-
break loss-of-coolant accident (LBLOCA) (Reference 16). Refer to
Section 6.3.2.4.8.1 for discussion on partial submergence of the strainer assemblies during an SBLOCA.
(4) Fulfill spray requirements (refer to Section 6.2.2).
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-20 Revision 23 December 2016 6.3.2.12 Accumulator Availability Accumulator availability requirements during power operation, hot standby, and startup conditions are detailed in the Technica l Specifications (Reference 10).
6.3.2.13 Dependence on Other Systems Other systems that operate in conjunction with the ECCS are as follows:
(1) The CCW system (refer to Section 9.2.2) cools the RHR heat exchangers during the recirculation mode of operation. It also supplies cooling water
to CCP1 and CCP2, the SI pumps, and the RHR pumps during the
injection and recirculation modes of operation.
(2) The auxiliary salt water (ASW) system (refer to Section 9.2.1) provides cooling water to the CCW heat exchangers.
(3) The Preferred Power Supply (230-kV and 500-kV), Onsite Distribution System (120-Vac, 480-V, and 4.16-kV), and Standby Power Supply
provide normal and emergency power sources for the ECCS (refer to
Sections 8.2, and 8.3.1.1.3 through 8.3.1.1.6).
(4) The engineered safety features actuation system (ESFAS) (refer to Section 7.3) generates the initiation signal for emergency core cooling.
(5) The auxiliary feedwater (AFW) system (refer to Section 6.5) supplies feedwater to the steam generators.
(6) The auxiliary building ventilation system (ABVS) (refer to Section 9.4.2) removes heat from the pump compartments and provides for radioactivity
contamination control should some leakage occur in a compartment.
6.3.2.14 Lag Times The sequence and time-delays for actuation of ECCS components for the injection and
recirculation phases of emergency core cooling are given in Table 6.3-7. Alignment of
the major ECCS components during the injection and recirculation phases is shown in
Figures 6.3-4 and 6.3-5, respectively. Tables 15.3-2 and 15.3-3 summarize the
calculated times at which the major components perform the safety-related functions for
those various accident conditions (tabulated in Table 15.1-2) that require the ECCS.
The minimum active components will be capable of delivering full rated flow within a
specified time interval after process parameters reach the setpoints for the "S" signal.
Response of the system is automatic with appropriate allowances for delays in actuation
of circuitry and active components. The active portions of the system are actuated by the "S" signal. In analyses of system performance, delays in reaching the programmed DCPP UNITS 1 &
2 FSAR UPDATE 6.3-21 Revision 23 December 2016 trip points and in actuation of components are established on the basis that only Standby Power Supply is available. The starting sequence following a loss of offsite
power is discussed in detail in Chapter 8.
The ECCS is operational after an elapsed time not greater than 25 seconds, including
the time to bring the RHR pumps up to full speed.
The starting times for components of the ECCS are consistent with the delay times used
in the LOCA analyses for large and small breaks.
In the LOCA analysis presented in Sections 15.3 and 15.4, no credit is assumed for
partial flow prior to the establishment of full flow and no credit is assumed for the
availability of the Preferred Power Supply.
For smaller LOCAs, there is some additional delay before the process variables reach
their respective programmed trip setpoints since this is a function of the severity of the
transient imposed by the accident. This is allowed for in the analyses of the range of
LOCAs (refer to Tables 15.3-1, 15.4.1-1A, and 15.4.1-1B).
Accumulator injection occurs immediately when RCS pressure has decreased below the
operating pressure of the accumulator.
6.3.3 Safety Evaluation 6.3.3.1 General Design Criterion 2, 1967 - Performance Standards With the exception of the RWST, all ECCS components are housed 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 (refer 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 ECCS SSCs, ensuring their safety-related design functions will be performed.
The RWST, as discussed in Section 3.8, is a PG&E Design Class I outdoor water storage tank which is designed to withstand the ef fects of 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. The ability of the RWST to withstand the effects of winds
and tornadoes is addressed in Section 3.3. The loss of the RWST, which is not
contained within a building and is expose d directly to potential wind and tornado loads, has been evaluated. Loss of this equipment does not compromise the capability of
shutting down the plant safely (refer to Section 3.3.2.3). Leakage from the refueling water storage tanks due to tornado or missile-induced damage will not result in the
flooding of PG&E Design Class I equipment in the auxiliary building since essentially watertight cover plates are installed over the pipe entranceway from each tank into the
auxiliary building. The water will drain away from the building via the plant yard drainage system.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-22 Revision 23 December 2016 The ECCS is designed to perform its function of ensuring core cooling and providing shutdown capability following an accident under simultaneous DDE or 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. ECCS
components are designed to withstand the appropriate seismic loadings in accordance
with PG&E Design Class I criteria.
6.3.3.2 General Design Criterion 3, 1971 - Fire Protection
The ECCS is designed to the fire protection guidelines of Branch Technical Position APCSB 9.5.1 (refer to Appen dix 9.5B, Table B-1).
As described in Appendix 9.5B, Table B-1, no fire hazards exist that could adversely
affect the availability of the RWST. No com bustible materials are stored in proximity to the RWST, and hose stations and fire hydrants from the yard main provide fire
suppression capability, if required.
6.3.3.3 General Design Criterion 11, 1967 - Control Room No manual actions are required from the operator during the injection phase of ECCS
operation. Those manual actions from the control room required for changeover from the injection phase to the recirculation phase are described in Section 6.3.3.6.1.1.2 and
in Table 6.3-5.
Instrumentation, alarms, and controls are provided in the control room for operators to monitor and maintain ECCS parameters. Instrumentation and alarms for the ECCS is further discussed in Section 6.3.3.4.
6.3.3.4 General Design Criterion 12, 1967 - Instrumentation and Control Systems Varied instrumentation is available to assist the operator in assessing post-accident conditions. This instrumentation is listed below.
Instrumentation and associated analog and logic channels used to initiate ECCS operation are discussed in Section 7.3. This section describes the instrumentation
employed to monitor ECCS components during normal plant operation and ECCS post-
accident operation. All alarms are annunciated in the control room.
6.3.3.4.1 Temperature Indication 6.3.3.4.1.1 Residual Heat Exchanger Outlet Temperature The fluid temperature at the outlet of each RHR heat exchanger is recorded in the
control room.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-23 Revision 23 December 2016 6.3.3.4.1.2 ECCS Pump-motor Temperatures Temperature indicators are provided to monitor RHR pump, CCP1 and CCP2, and SI
pump motor/motor bearing temperatures. High temperatures activate alarms in the
control room.
6.3.3.4.1.3 Containment Recirculation Sump Water Temperature
The fluid temperature of the containment recirculation sump water is provided on the post-accident monitoring panel, PAM1.
6.3.3.4.2 Pressure Indication 6.3.3.4.2.1 Charging Injection Line Pressure The charging injection line pressure (PT-947 between valves 8801A & 8801B and 8803A & 8803B) shows that the CCPs are operating. The transmitters are outside the containment, with indicators on the control board.
6.3.3.4.2.2 Safety Injection Header Pressure SI pump discharge pressure for each pump shows that the SI pumps are operating.
The transmitters are outside the containment, with indicators on the control board.
6.3.3.4.2.3 Accumulator Pressure
Duplicate pressure channels are installed on each accumulator. Pressure indication in the control room and high- and low-pressure alarms are provided by each channel.
6.3.3.4.2.4 Residual Heat Removal Pump Discharge Pressure RHR pump discharge pressure for each pump is indicated in the control room. A
high-pressure alarm is actuated by each channel.
6.3.3.4.3 Flow Indication 6.3.3.4.3.1 CCP1 and CCP2 Injection Flow Injection flow through the common header to the reactor cold legs is indicated in the
control room.
6.3.3.4.3.2 Safety Injection Pump Header Flow Flow through the SI pump headers is indicated in the control room.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-24 Revision 23 December 2016 6.3.3.4.3.3 Residual Heat Removal Pump Injection Flow Flow through each RHR injection and recirculation he ader leading to the reactor cold or hot legs is indicated in the control room.
6.3.3.4.3.4 Test Line Flow Local indication of the leakage test line flow is provided to check for proper seating of
the accumulator check valves between the injection lines and the RCS.
6.3.3.4.3.5 Safety Injection Pump Minimum Flow A local flow indicator is installed in the SI pump minimum flow line.
6.3.3.4.3.6 Residual Heat Removal Pump Minimum Flow A flowmeter installed in each RHR pump discharge header provides control for the
valve located in the pump minimum flow line.
6.3.3.4.4 Level Indication 6.3.3.4.4.1 Refueling Water Storage Tank Level Three water level instrumentation channels are provided for the RWST. Each channel provides independent indication on the main control board, thus satisfying the requirements of paragraph 4.20 of IEEE 279-1971 (Reference 12). Two-out-of-three logic is provided for RHR pump trip and low-level alarm initiation. One channel provides low-low water level alarm initiation. One channel also provides a high water level alarm.
6.3.3.4.4.2 Accumulator Water Level Duplicate water level channels are provided for each accumulator. Both channels
provide indication in the control room and actuate high- and low-water level alarms.
6.3.3.4.4.3 Containment Recirculation Sump Water Level Two redundant containment wide-range water level channels are provided to measure
level from the bottom of the reactor cavity.
Wide-range recorders are located on the post-accident monitoring panel.
Two redundant narrow-range instruments measure level from the bottom of the
containment recirculation sump. The containment recirculation sump level
instrumentation consists of two level sensor elements, junction boxes, and connection
sleeves designed to operate in a post-accident environment inside containment. The
junction boxes and connection sleeves are located above any possible flooding level.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-25 Revision 23 December 2016 Level transmitters are located outside of containment and are not required to be environmentally qualified as they are not req uired for a postulated line break outside of containment. Narrow-range indicators are located on the main control board. The
containment recirculation sump level instrumentation is described in more detail in
Section 7.5.
6.3.3.4.5 Valve Position Indication Valve positions that are indicated on the control board are done so by a "normal off" system; i.e., should the valve not be in its proper position, a bright white light will give a
highly visible indication to the operator.
This indication is only active upon receipt of an "S" signal for those ECCS-related valves that are required to automatically change
position to align CCP1 and CCP2 for injection, as described in Section 6.3.3.6.1.1.1.
6.3.3.4.5.1 Accumulator Isolation Valve Position Indication The accumulator MOVs are provided with red (open) and green (closed) position
indicating lights located at the control switch for each valve. These lights are energized
from the Class 1E 120-Vac Instrument Power Supply System and actuated by the
associated valve motor operator limit switches.
A monitor light that is on when the valve is not fully open is provided in an array of
monitor lights that are all off when their respective valves are in proper position enabling
safety features operation. This light is energized from a separate Class 1E 120-Vac
circuit and actuated by a valve motor-operated limit switch.
An alarm annunciator point is activated by a valve motor operator limit switch or a stem travel limit switch whenever an accumulator valve is not fully open with the system at
pressure (the pressure at which the safety injection block is unblocked). The alarm is
reinstated once an hour. A separate annunciator point is used for each accumulator
valve. 6.3.3.4.6 Subcooling Meter Each subcooling meter provides continuous digital-type display of either the
temperature or pressure subcooling margin. The displays are a subset of the reactor
vessel level instrumentation system (RVLIS) and are located on PAM panels, PAM3
and PAM4, with low and low-low alarms being provided. A digital display of the
temperature margin is fed from train B and is available on the main control board. A
recorder fed from train A is located on PAM1. The subcooling meter displays, calculators, and inputs are described in Sections 5.6 and 7.5.
6.3.3.5 General Design Criterion 21, 1967 - Single Failure Definition To ensure that the ECCS will perform its intended function if any of the accidents listed
in Section 6.3.2 occurs, it is designed to tolerate a single active failure during the short-DCPP UNITS 1 &
2 FSAR UPDATE 6.3-26 Revision 23 December 2016 term immediately following an accident, or to tolerate a single active or passive failure during the long-term following an accident. This subject is detailed in Section 3.1 and
the following subsections.
6.3.3.5.1 Definition of Terms Definitions of terms used in this section are located in Section 3.1.1.1.
6.3.3.5.2 Active Failure Criteria The ECCS is designed to accept a single fa ilure following the incident without loss of its protective function. The system design will tolerate the failure of any single active
component in the ECCS itself or in the necessary associated service systems at any
time during the period of required system operations following the incident.
A single active failure analysis is presented in Table 6.3-13, and demonstrates that the
ECCS can sustain the failure of any single active component in either the short- or
long-term and still meet the level of performance for core cooling.
Since the operation of the active components of the ECCS following an MSLB is identical to that following a LOCA, the same analysis is applicable and the ECCS can
sustain the failure of any single active component and still meet the level of
performance for the addition of shutdown reactivity.
6.3.3.5.3 Passive Failure Criteria
The following philosophy provides for necessary redundancy in component and system arrangement to meet the single failure criterion, as it specifically applies to failure of passive components in the ECCS. Thus, for the long-term, the system design is based
on accepting either a passive or an active failure.
6.3.3.5.3.1 Redundancy of Flow Paths and Components for Long-Term Emergency Core Cooling In the design of the ECCS, Westinghouse utilizes the following criteria:
(1) During the long-term cooling peri od following a loss of coolant, the emergency core cooling flow paths are separable into two subsystems, either of which can provide minimum core cooling functions and return
spilled water from the floor of the containment back to the RCS.
(2) Either of the two subsystems can be isolated and removed for service in the event of a leak outside the containment.
(3) Adequate redundancy of check valves is provided to tolerate failure of a check valve during the long-term as a passive component.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-27 Revision 23 December 2016 (4) Should one of these two subsystems be isolated in this long-term period, the other subsystem remains operable.
(5) Provisions are also made in the design to detect leakage from components outside the containment and collect this leakage.
The two separate heat exchanger and pump flowpaths provide redundant capability of meeting the engineered safety function of the RHR system. The loss of one of these RHR system flowpaths would not negate the capability of the ECCS since the two flowpaths provide full redundancy for engineered safety requirements. The injection flow paths to loops one and two and to loops three and four are not redundant. Both of these flowpaths must be available with the heat exchanger discharge cross tie open to meet the engineered safety function of the RHR system.
Thus, for the long-term emergency core cooling function, adequate core cooling
capacity exists with one flow path removed from service whether isolated due to a leak, because of blocking of one flow path, or because failure in the containment results in a
spill of the delivery of one subsystem.
6.3.3.5.3.2 Subsequent Leakage from Co mponents in Engineered Safety Systems With respect to piping and mechanical equipment outside the containment, considering the provisions for visual inspection and leak detection, leaks will be detected before they
propagate to major proportions. A review of the equipment in the system indicates that
the largest sudden leak potential would be the sudden failure of a pump shaft seal.
Evaluation of leak rate assuming only the presence of a seal retention ring around the pump shaft showed flows less than 50 gpm would result. Piping leaks, valve packing leaks, or flange gasket leaks have been of a nature to build up slowly with time and are
considered less severe than the pump seal failure.
Larger leaks in the ECCS are prevented by the following:
(1) The system piping is located within a controlled area on the plant site.
(2) The piping system receives periodic pressure tests and is accessible for periodic visual inspection (refer to Sections 6.3.3.24 and 6.3.3.25).
(3) The piping is austenitic stainless steel that, due to its ductility, can withstand severe distortion without failure.
Based on this review, the design of the auxiliary building and related equipment is
based on handling of leaks up to a maximu m of 50 gpm. Means are also provided to detect and isolate such leaks in the emergency core cooling flow path within
30 minutes.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-28 Revision 23 December 2016 With these design ground rules, continued function of the ECCS will meet minimum core cooling requirements, and offsite doses resulting from the leak will be within
10 CFR 100.11 limits. Refer to Section 15.5 for discussion of radiological
consequences of plant accidents.
A single passive failure analysis is presented in Table 6.3-14. It demonstrates that the
ECCS can sustain a single passive failure du ring the long-term phase and still retain an intact flow path to the core to supply sufficient flow to maintain the core covered and
effect the removal of decay heat. The procedure followed to establish the alternate flow
path also isolates the component that failed.
6.3.3.5.4 Single Failures in Electrical and Control Circuitry ESF systems, including the ECCS, are designed to tolerate single failures in the
electrical and control circuitry as defined below. Each ECCS train is supplied from
separate Class 1E power sources. In addition, provisions are made in construction, layout, and installation that minimize the occurrence of electrical faults.
Single failures of switching components that are considered in the design include (a) a
single instance of contact failure, (b) a loss of control power, or (c) mechanical failure resulting in the sticking of a component in any position. Faults that require a particular
time sequence of abnormal connections or that involve selective combinations of
multiple contact closures are considered to involve more than one failure, and hence to
be incredible. Examples of a single wirin g failure considered in design include the single short and open circuit faults. Single short circuits considered are (a) a single
conductor shorted to ground or to a structure such as a cable tray, or (b) two conductors in the same cable shorted together. Faults that require several particular wires to be connected, or which require sustained application of power through a short circuit, are
not considered to be credible. Open circuit faults considered include (a) a single
conductor breaking, (b) a single connector being disconnected, or (c) a single field-run
cable severed.
The random shorting to the manual closing switch contacts or a hot short in the cable run to the closing coil have been identified as control system failure modes that could
lead to spurious movement of a passive MOV.
During the safety review of the operating license application for the DCPP, the Atomic Energy Commission (AEC, now NRC) regulatory staff adopted the position that failures
of the type discussed above that could lead to spurious movement of passive MOVs
must be considered in relation to satisfying the single failure criterion. The regulatory
staff's position, as stated in Branch Technical Position EICSB 18, November 1975, considers removal of electric power an acceptable means, under certain conditions, of
satisfying the single failure criterion. Refer to Section 6.3.3.42 for further discussion.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-29 Revision 23 December 2016 6.3.3.5.5 Control of Accumulator Motor-Operated Isolation Valves During power operation, electrical power is removed from the valves by opening the
480-V breaker, thereby preventing inadvertent closure due to an electrical short. As the
valve is provided with an automatic opening signa l whenever RCS pressure exceeds the unblocking pressure, P11, a manual override permits closing of the valves at pressures exceeding P11 (refer to Figure 7.3-33 for controls). Additionally, it will open
automatically upon receipt of the "S" signal should the valve be closed and the breaker
is racked in. This "S" signal overrides any bypass feature. Position indicators for these
valves are discussed in Section 6.3.3.4.5.1.
The accumulator MOVs are considered to be operating bypasses in the context of
IEEE 279-1971 (Reference 12), which requires that bypasses of a protective function be
removed automatically whenever permissive conditions are not met. In addition, as
these accumulator isolation valves fail to meet single failure criteria, removal of power to
the valves is required.
6.3.3.6 General Design Criterion 37, 1967 - Engineered Safety Features Basis for Design
6.3.3.6.1 Systems Operation The operation of the ECCS following a LOCA or MSLB is discussed below.
6.3.3.6.1.1 Operation After Loss of Primary Coolant
The operation of the ECCS following a LOCA can be divided into two distinct modes:
(1) The injection mode in which any reactivity increase following the postulated accident is terminated, initial cooling of the core is
accomplished, and coolant lost from the primary system is replenished (2) The recirculation mode in which long-term core cooling is provided during the accident recovery period
A discussion of these modes follows:
6.3.3.6.1.1.1 Injection Mode Af ter Loss of Primary Coolant As shown in Figure 6.3-4, the principal mechanical components of the ECCS that
provide core cooling immediately following a LOCA are the accumulators (one for each loop), the SI pumps, the CCPs (CCP1 and CCP2), the RHR pumps, and the associated valves, tanks, and piping.
For large pipe ruptures, (0.5 square feet equivalent and larger), the RCS would be
depressurized and voided of coolant rapidly. A high flowrate of emergency coolant is DCPP UNITS 1 &
2 FSAR UPDATE 6.3-30 Revision 23 December 2016 required to quickly cover the exposed fuel rods and limit possible core damage. This high flow is provided by the passive accumulators, followed by CCP1 and CCP2, SI pumps, and the RHR pumps discharging into the cold legs of the RCS.
During the injection mode, the RHR and SI pumps deliver into the accumulator injection
lines between the two check valves.
CCP1 and CCP2 deliver through the charging injection line directly into the cold legs.
Emergency cooling is provided for small ruptures primarily by high-head injection.
Small ruptures are those, with an equivalent diameter of 6 inches or less, that do not
immediately depressurize the RCS below the accumulator discharge pressure. CCP1
and CCP2 deliver borated water at the prevail ing RCS pressure to the cold legs of the RCS, from the RWST during the injection mode.
The SI pumps also take suction from the RWST and deliver borated water to the cold
legs of the RCS. The SI pumps begin to deliver water to the RCS after the pressure
has fallen below the pump shutoff head.
The RHR pumps take suction from the RWST and deliver borated water to the RCS.
These pumps begin to deliver water to the RCS only after the pressure has fallen below
the pump shutoff head.
The injection mode of emergency core cooling is initiated by the "S" signal. This signal is actuated by any of the following:
(1) Pressurizer low pressure (2) Containment high pressure (3) Low steamline pressure (4) Manual actuation
Operation of the ECCS during the injection mode is completely automatic. The "S" signal automatically initiates the following actions:
(1) Starts the diesel generators and, if all other sources of power are lost, aligns them to the Class 1E buses (refer to Section 8.3.1.1.6)
(2) Starts CCP1 and CCP2, the SI pumps, and the RHR pumps (3) Aligns CCP1 and CCP2 for injection by:
(a) Closing the valves in the charging pump discharge line to the normal charging line
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-31 Revision 23 December 2016 (b) Opening the valves in the charging pumps suction line from the RWST (c) Closing the valves in the charging pump normal suction line from the volume control tank (d) Opening the charging injection inlet and outlet line isolation valves (4) Provides an open signal to the accumulator isolation valves; and if they are energized and closed, they will open (refer to Section 6.3.3.5.5).
The injection mode continues until the low level is reached in the RWST, at which time
the RHR pumps are automatically tripped. The operator then manually changes system
alignment to the recirculation mode. The remaining water in the RWST provides a
reserve to be used by the CSS pumps to ensure that enough sodium hydroxide has
been added to the containment, utilizing the spray additive system (SAS), to maintain the pH of the recirculation fluid g reater than 8.0. Three channels of RWST instrumentation are used, providing three independent and redundant RWST level
indications that are displayed in the control ro om, to inform the operator of the water level in the tank at all times. Any two of these three channels will actuate the low-level
alarm and will automatically trip the RHR pumps on low RWST level. After the
changeover begins, the rate of possible demand on the tank is reduced, providing
increasing periods of operation for the pumps still drawing water from the tank, and
giving further assurance that water will remain in the tank after the completion of the
changeover.
6.3.3.6.1.1.2 Changeover from Injection Mode to Recirculation After Loss of Primary Coolant
Water level indication and alarms on the RWST and level indication in the containment
recirculation sump provide ample warning to terminate the injection mode while the
operating pumps still have adequate NPSH. Since the injection mode of operation
following a LOCA is terminated before the RWST is completely emptied, all pipes are
kept filled with water before recirculation is initiated. For some SBLOCAs, the RCS
pressure may remain above the shut off head of the RHR pumps that would be
providing only recirculation flow. For these scenarios, the operators will trip the RHR
pumps or initiate CCW flow to an RHR heat exchanger within 30 minutes. This ensures
that pump recirculation does not result in heating and pressurizing the RHR suction
piping.
Following receipt of the RWST low-level alarm and automatic tripping of the RHR
pumps, the remainder of the changeover sequence from injection mode to recirculation
mode is accomplished manually by the operator from the control room (except for restoring power to the RWST supply valves to the RHR and the SI pumps). The same
sequence (as delineated in Table 6.3-5) is fo llowed regardless of which power supply is available (Offsite Power System or Standby Power Supply). Controls for ECCS DCPP UNITS 1 &
2 FSAR UPDATE 6.3-32 Revision 23 December 2016 components are grouped together on the main control board. The component position lights verify when the function of a given switch has been completed. The total required
switchover time for the changeover from injection to recirculation is approximately
10 minutes, as shown in Table 6.3-5. The postulated single failure during the
changeover sequence is the failure of an RHR pump to trip on low RWST level. The
operator action requires approximately 5 minutes to locally open the breaker for an RHR
pump motor. The operator action is performed concurrently with the changeover
sequence and there is no increase in the total time for the changeover. The
changeover sequence can be completed, with the single failure, and the remaining
useable RWST volume exceeds the licensing basis of 32,500 gallons.
6.3.3.6.1.1.3 Recirculation Mode After Loss of Primary Coolant After the injection operation, water collected in the containment recirculation sump is cooled and returned to the RCS by the low-head/high-head recirculation flowpath. The
RCS can be supplied simultaneously from the RHR pumps and from a portion of the discharge from the RHR heat exchanger that is directed to CCP1 and CCP2 and SI
pumps that return the water to the RCS.
The latter mode of operation ensures flow in the event of a small rupture where the depressurization proceeds more slowly, so that
the RCS pressure is still in excess of the shutoff head of the RHR pumps at the onset of recirculation. Hot leg recirculation is implemented to ensure termination of boiling and prevent boric acid crystallization. Approximately 7.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> after LOCA inception, the operators will manually initiate hot leg recirculation and are expected to complete the switchover process within approximately 15 minutes. Some cold leg recirculation would be maintained with the CCPs (CCP1 and CCP2) after hot leg recirculation is initiated.
The RWST is protected from back flow of reactor coolant from the RCS. All connections to the RWST except those that are designed to return flow to the RWST
are provided with check valves to prevent back flow.
Redundancy in the external recirculation loop is provided by duplicate CCPs (CCP1 and
CCP2), SI pumps, and RHR pumps and heat exchangers. Inside the containment, the
high-pressure injection system is divided into two separate flow trains. For cold leg
recirculation, CCP1 and CCP2 deliver to all four cold legs and the SI and RHR pumps
deliver also to all four cold legs by separate flowpaths. For hot leg recirculation, each SI pump delivers through separate paths to two hot legs, while the RHR pumps deliver to
two of the four hot legs.
The sump isolation valves are located in steel-lined pressure-tight compartments. This
arrangement contains any leakage from an isolation valve stem or bonnet.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-33 Revision 23 December 2016 6.3.3.6.1.2 Operation After MSLB Following an MSLB, the ECCS is automatically actuated to deliver borated water from
the RWST to the RCS. The response of the ECCS following an MSLB is identical to its
response during the injection mode of operation follow ing a LOCA. The S signal initiates identical actions as described for the injection mode of the LOCA, even though
not all of these actions are required following an MSLB; e.g., the RHR pumps are not
required since the RCS pressure will remain above the pump shutoff head.
The delivery of the concentrated boric acid from the RWST provides negative reactivity
to counteract the increase in reactivity caused by the system cooldown. CCP1 and
CCP2 continue to deliver borated water from the RWST until enough water has been
added to the RCS to make up for the shrinkage due to cooldown. The SI pumps also
deliver borated water from the RWST for the interval when the RCS pressure is less than the shutoff head of the SI pumps.
After pressurizer water level has been restored, the injection is manually terminated.
The sequence of events following a postulated MSLB is described in Section 15.4.2.
6.3.3.6.2 ECCS Performance The following events were analyzed to ensure that the limits on core behavior following
an RCS pipe rupture are not exceeded when the ECCS operates with minimum design
equipment:
(1) Large pipe break analysis (2) Small line break analysis (3) ECCS recirculation mode cooling
The adequacy of flow delivered to the RCS by the ECCS with the operation of minimum
design equipment is demonstrated in Sections 15.3.1 and 15.4.1.
The design basis performance characteristic is derived from the specified performance
characteristic for each pump with a conservative estimate of system piping resistance, based on the final piping layout. The performance characteristic utilized in the accident
analyses includes a decrease in the design head for margin. When the initiating
incident is assumed to be the severance of an injection line, the injection curve utilized
in the analyses accounts for the loss of injection water through the broken line.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-34 Revision 23 December 2016 6.3.3.6.2.1 Large Pipe Break Analysis The large pipe break analysis is used to evaluate the initial core thermal transient for a
spectrum of pipe ruptures from a break size greater than 1.0 square foot up to the
double-ended rupture of the largest pipe in the RCS.
The injection flow from active components is required to control the cladding
temperature subsequent to accumulator injection, complete reactor vessel refill, and
eventually return the core to a subcooled state. The results indicate that the maximum
cladding temperature attained at any point in the core is such that the limits on core
behavior as specified in Section 15.4.1 are met.
6.3.3.6.2.2 Small Pipe Break Analysis The small pipe break analysis is used to evaluate the initial core thermal transient for a
spectrum of pipe ruptures, which bounds breaks corresponding to the smallest break
size, typically a 3/8 inch diameter opening (0.11 square inch), up to and including a break size of 1.0 square foot. For a break opening 3/8 inch or smaller, the makeup flow
rate from either CCP1 or CCP2 is adequate to allow time for an orderly plant shutdown
without automatic ECCS actuation.
The results of the small pipe break analysis indicate that the limits on core behavior are
adequately met, as shown in Section 15.3.1.
6.3.3.6.2.3 Recirculation Cooling
Core cooling during recirculation can be maintained by the flow from one RHR pump if RCS pressure is low. If RCS pressure remains high, either CCP1 or CCP2 and one SI
pump operating in series with one RHR pump provide the added head and flow needed
to maintain adequate cooling.
Heat removal from the recirculated sump water is accomplished via operation of one or
both of the RHR heat exchangers.
6.3.3.6.2.4 Required Operating Status of ECCS Components Normal operating status of ECCS components is given in Table 6.3-6.
ECCS components are available whenever the coolant energy is high and the reactor is critical. During low temperature physics tests, there is a negligible amount of stored
energy and low decay heat in the coolant; therefore, an accident comparable in severity
to accidents occurring at operating conditions is not possible in low temperature physics
tests, and ECCS components are not required.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-35 Revision 23 December 2016 6.3.3.6.2.5 Loss of Offsite Power The ECCS is designed to meet its minimum required level of functional performance
with onsite electrical power system operation (assuming offsite power is not available)
or with offsite electrical power system operation for any of the above abnormal
occurrences assuming a single failure as discussed in Sections 3.1 and 6.3.3.5.
Diesel generators supply power to ECCS components in the event that all sources of offsite power become unavailable.
The supply of emergency power to the ECCS components is arranged so that, as a
minimum, CCP1 or CCP2, one SI pump, and one RHR pump together with the
associated valves will automatically receive adequate po wer in the event that a loss of offsite power occurs simultaneously with any of the design basis accidents described in
Section 6.3.2. Adequate power is provided even if the single failure is the failure of an emergency diesel generator to start.
6.3.3.6.2.6 Range of Core Protection Core protection is afforded with the minimum ESF equipment, defined by consideration
of the single failure criteria as discussed in Sections 3.1 and 6.3.3.5. The minimum
design case will ensure that the entire break spectrum is accounted for and the core
cooling design bases of Section 6.3.2 are met. The analyses for this case are
presented in Sections 15.3 and 15.4.
For large RCS ruptures, the accumulators and the active high-head (CCP), intermediate-head (SI), and low-head (RHR) pumping components serve to complete the core refill. One RHR pump is required for long-term recirculation.
If the break is small (1.0 square foot or less), the accumulators, CCP1 or CCP2, and
one SI pump ensure adequate cooling during the injection mode. Long-term
recirculation requires operation of one RHR pump in conjunction with CCP1 or CCP2
and one SI pump and components of the auxiliary heat removal systems that are required to transfer heat from the ECCS (e.g., CCW system and ASW system). The LOCA analyses, presented in Sections 15.3 and 15.4, indicate that certain modifications (i.e., reduced component availability) to the normal operating status of the ECCS, as given in Table 6.3-6, are permissible without impairing the ability of the ECCS, to
provide adequate core cooling capacity.
6.3.3.6.2.7 ECCS Piping Failures The rupture of the portion of an injection line from the last check valve to the connection
of the line to the RCS can cause not only a loss of coolant but impair the injection as
well. To reduce the probability of an emergency core cooling line rupture causing a
LOCA, the check valves that isolate the ECCS from the RCS are installed adjacent to the reactor coolant piping.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-36 Revision 23 December 2016 For a small break, the reactor pressure maintains a relatively uniform back pressure in
all injection lines so that a significant flow imbalance does not occur. Also, system
resistances are balanced by adjustment of throttling valves in the injection lines prior to
plant operation. A rupture in an accumulator injection line is accounted for in the
analyses by assuming that for cold leg breaks the entire contents of the associated
accumulator are discharged through the break.
6.3.3.6.2.8 External Recirculation Loop Major ECCS components are shielded, as required, from their associated redundant
train to facilitate maintenance. During recirculation, following a LOCA, access to these
components is not credited.
Pressure relieving devices with setpoints below the shutoff head of the high-head ECCS pumps (CCP1 and CCP2), from portions of the ECCS located outside of containment
that might contain radioactivity, discharge to the pressurizer relief tank.
An analysis has been performed to evaluate the radiological effects of recirculation loop
leakage (refer to Section 15.5). A loop is assumed to include CCP1 or CCP2, an SI pump, an RHR pump, an RHR heat exchanger, and the associated piping. Thus two loops are provided, each of which is adequate for core cooling. In the analysis, maximum leakage was assumed as discussed in Section 6.3.3.5. Analyses indicate that the offsite dose resulting from such leakage is much less than the requirements of
Since redundant flowpaths are provided duri ng recirculation, a leaking component in one of the flowpaths may be isolated. This action curtails any further leakage.
Maximum potential leakage from components during normal operation is given in
Table 6.3-9.
Each pump compartment and heat exchanger compartment are provided with sufficient
drains to the RHR room sump to prevent compartment overflow due to the design leakage rate. Containment isolation valves can be remote manually closed before the
pump compartment can overflow.
This layout permits the detection of a leaking recirculation loop component by means of
a radiation monitor that samples the air exit ing the heat exchanger compartment.
Alarms in the control room will alert the operator when the activity exceeds a preset
level. Sump level alarms and operation of sump pumps will be indicated in the control
room as a backup for detection of water leaks.
Should a tube-to-shell leak develop in a RHR heat exchanger, the operator will be warned by a CCW high radiation alarm. For large leaks, the operator will also be
warned by a CCW surge tank high-level alarm.
In the event that the leak cannot be DCPP UNITS 1 &
2 FSAR UPDATE 6.3-37 Revision 23 December 2016 isolated before the CCW surge tank fills, the tank relief valve will lift and direct the excess water to the auxi liary building sump.
6.3.3.6.2.9 Evaluation of Shutdown Reactivity Capa bility Following an Abnormal Release of Steam from the Main Steam System Analyses are performed to ensure that the core limitations defined in Sections 15.2, 15.3, and 15.4 are met following a steam line rupture or a single active failure in the
main steam system.
6.3.3.6.2.9.1 Main Steam System Single Active Failure
Analyses of reactor behavior following any single active failure in the main steam
system that results in an uncontrolled release of steam are included in Section 15.2.
The analyses assume that a single valve (largest of the safety, relief, or bypass valves)
opens and fails to close, resulting in an uncontrolled cooldown of the RCS.
Results indicate that if the incident is initiated at the hot shutdown condition, which results in the worst reactivity transient, the limiting departure from nucleate boiling ratio (DNBR) values will be met. Thus, the ECCS provides adequate protection for this
incident.
6.3.3.6.2.9.2 Main Steam Line Break This accident is discussed in detail in Section 15.4. The limiting MSLB is a complete line severance.
The results of the analysis indicate that the design bas is criteria are met. Thus, the ECCS adequately fulfills its shutdown reactivity addition function.
Following a secondary side high-energy (steam or feedwater) line break, CCW is
supplied to the seal-water heat exchanger to provide co oling for CCP1 and CCP2 under miniflow conditions. This action prevents damage to the CCPs before safety injection
termination criteria are reached and CCP operation is terminated.
6.3.3.6.2.10 Alternate Analysis Methods The method of break analysis and the spectrum of breaks analyzed are described in
Sections 15.3.1 and 15.4.1.
6.3.3.6.2.11 Fuel Rod Perforations Results of the small pipe break and large pipe break analyses are presented in
Sections 15.3.1 and 15.4.1, respectively.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-38 Revision 23 December 2016 6.3.3.7 General Design Criterion 38, 1967 - Reliability and Testability of Engineered Safety Features ECCS reliability was considered in all aspect s of system evolution, from initial design to periodic testing of the components during plant operation.
The preoperational testing program ensures that the systems, as designed and
constructed, meet functional requirements.
The ECCS is designed with the ability for on-line testing of most components so that
availability and operational status can be readily confirmed. The integrity of the ECCS
is ensured through examination of critical components during routine inservice
inspection.
6.3.3.7.1 Provisions for Performance Testing Design features have been incorporated to ensure that the following testing can be
performed:
(1) Active components may be tested periodic ally for operability (e.g., pumps on miniflow, certain valves, etc.).
(2) An integrated system actuation test can be performed when the plant is cooled down and the RHR system is in operation. The ECCS will be
arranged so that no flow will be introduced into the RCS for this test.
(3) An initial flow test of the full operational sequence can be performed.
Specific design features that ensure this test capability are the following:
(1) Power sources are provided to permit actuation of individual ECCS active components.
(2) The SI pumps can be tested periodically during plant operation using the minimum flow recirculation lines provided.
(3) The RHR pumps are used every time the RHR system is put into operation. They can also be tested periodically when the plant is at power
using the miniflow recirculation lines.
(4) CCP1 and CCP2 are either normally in use for charging service or can be tested periodically on miniflow.
(5) Remotely operated valves can be exercised during routine plant maintenance.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-39 Revision 23 December 2016 (6) For each accumulator tank, level and pressure instrumentation is provided for continuous monitoring during plant operation.
(7) Pressure instrumentation and a flow indicator are provided in the SI pump header and in the RHR pump headers.
(8) An integrated system test can be performed when the plant is cooled down and the RHR system is in operation. This test does not introduce flow into the RCS but does demonstrate the operation of the valves, pump
circuit breakers, and automatic circuitry including diesel starting and the
automatic loading of ECCS components off the diesels (by simultaneously
simulating a loss of offsite power to the Class 1E electrical buses).
6.3.3.8 General Design Criterion 40, 1967 - Missile Protection The provisions taken to protect the ECCS from damage that might result from dynamic
effects associated with a postulated rupture of piping are discussed in Section 3.6. The provisions taken to protect the system from m issiles are discussed in Section 3.5. The ECCS design is such that physical protection is adequately provided against physical
hazards in areas through which the system is routed.
6.3.3.8.1 Safety Injection System Pump Missile Proneness The capability of SI pump-motor combination to generate an external missile has been evaluated in Section 3.5. The results of this evaluation showed that neither the SI pump nor the motor are capable of generating external missiles.
However, the flexible coupling be tween the pump and motor could conceivably become a missile in the unlikely event that it should fail to maintain its mechanical integrity, due
to a maximum overspeed condition caused by a large pressure head driving the pump
in reverse. Such failure would require the failure of two check valves in the open
position in conjunction with a rupture of the pipe on the suction side of the pump.
Despite the low probability of such a combination of failures, a shroud has been installed around the flexible coupling to eliminate all possibility of missiles being
generated in the unlikely event of gross coupling failure.
6.3.3.9 General Design Criterion 41, 1967 - Engineered Safety Features Performance Capability The ECCS is a two-train, fully redundant, standby ESF. The system was designed to withstand any single credible active failure during injection or active or passive failure during recirculation and maintain the performance objectives outlined in Section 6.3.2.
Two trains of pumps, heat exchangers, and flowpaths are provided for redundancy; only one train is required to satisfy performance requirements. Initiating signals for the DCPP UNITS 1 &
2 FSAR UPDATE 6.3-40 Revision 23 December 2016 ECCS are derived from independent sources as measured from process (e.g., low pressurizer pressure) or environmental variables (e.g., containment pressure).
Each train is physically separated and protected so that a single event cannot initiate a
common mode failure. Each ECCS train is supplied from separate Class 1E power
sources. The Class 1E power sources are discussed in Chapter 8.
6.3.3.9.1 Pump Characteristics Typical performance curves for RHR pumps, CCP1 and CCP2, and SI pumps are
shown in Figures 6.3-1, 6.3-2, and 6.3-3, respectively. The upper curves represent the
typical performance characteristics of the pump. The lower curves illustrate that
margins for the potential pump degradation have been considered in the analyses
described in Chapter 15.
6.3.3.9.2 Limits on System Parameters The specification of individual parameters as indicated in Table 6.3-1 includes due consideration of allowances for margin over and above the required performance value (e.g., pump flow and NPSH), and the most severe conditions to which the component
could be subjected (e.g., pressure, temperature, and flow).
This consideration ensures that the ECCS is capable of meeting its minimum required level of functional performance.
6.3.3.9.2.1 Coolant Storage Reserves A minimum RWST volume is provided to ensure that, after an RCS break, sufficient
water is injected and available within containment to permit recirculation cooling flow to
the core, to meet the NPSH requirement of the RHR pumps. This volume is less than that required to fill the refueling canal (to permit normal refueling operation). Thus, adequate emergency coolant storage volume for ECCS operation is provided. For
RWST operational volumes, refer to Section 6.3.2.11 and Table 6.3-1.
6.3.3.9.2.2 Limiting Conditions for Ma intenance During Operation Maintenance on an active component will be permitted if the remaining components
meet the minimum conditions for operation as well as the following conditions:
(1) The remaining equipment has been verified to be in operable condition, ready to function just before the initiation of the maintenance.
(2) A suitable time limit is placed on the total time span of successful maintenance that returns the components to an operable condition, ready
to function.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-41 Revision 23 December 2016 The design philosophy with respect to active components in the high-head/low-head injection system is to provide backup equipm ent so that maintenance is possible during operation without impairment of the system safety function (refer to Section 6.3.3.5).
Routine servicing and maintenance of equipment of this type that is not required more
frequently than on an outage basis would generally be scheduled for periods of refueling and maintenance outages. The Technical Specifications (Reference 10)
discuss in detail the applicable limiting cond itions for maintenance during operations.
6.3.3.10 General Design Criterion 42, 1967 - Engineered Safety Features Components Capability Instrumentation, motors, and pen etrations located inside the containment are selected to meet the most adverse accide nt conditions to which they may be subjected. These items are either protected from co ntainment accident c onditions or are designed to withstand, without failure, exposure to the worst combination of temperature, pressure, and humidity expected during the requi red operational period.
The ECCS pipes serv ing each loop are anchored at the missile barrier (i.e., the crane wall) in each loop area to restrict p otential accident damage to the portion of piping beyond this point. The anchorage is designed to withstand, without fai lure, the thrust force exerted by any branch line severed fr om the reactor coolant pipe and discharging fl uid to the atmosphere, and to withstand a bending moment equivalent to that producing failure of the piping under the actio n of a free-end discharge to atmos phere or motion of the broken reactor coolant pipe to which the ECCS pipe s are connected. Thi s prevents possible failure at any point up stream from the support point includi ng the branch line connection into the piping header.
6.3.3.11 General Design Criterion 43, 1967 - Accident Aggravation Prevention
As discussed in Section 15.4.
1.4, the introduction of ECCS suppli ed borated cooling water into the core do es not result in a net posi tive reactivi ty addition.
When water in the RWST at its minimum boron concentration is mixed with the contents
of the RCS, the resulting boron concentration ensures that the reactor will remain
subcritical in the cold condition with all control rods, except the most reactive RCCA, inserted into the core.
The boron concentration of the accumulator and the RWST is below the solubility limit
of boric acid at the respective temperatures.
Thermal stresses on the RCS are discussed in Section 5.2.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-42 Revision 23 December 2016 6.3.3.12 General Design Criterion 44, 1967 - Emergency Core Cooling Systems Capability By combining the use of passive accumula tors with two CCPs (CCP1 and CCP2), two SI pumps, and two RHR pumps, emer gency core cooling is provid ed even if there should be a failure of any single component in any system. The ECCS employs a passive system of accumulators that do not require any external signals or source of power for their operation to cope with the short-term cooling requirements of large react or coolant pipe breaks. Two independent and redunda nt high-pressure flow and pumpi ng systems, each capable of the required emergency cooling, are provided for small break protection and to keep the core submerged after the accumula tors have dischar ged following a large break. These systems are arranged so th at the single failure of any active compone nt does not interfere with meeting the short-term cooling requirements.
Borated water is inject ed into the RCS by accumulators, SI p umps, RHR pumps, and CCPs. Pump design includes consideration of fluid tempe rature and containment pressure in accordance with Safety Guide 1, November 1970 (refer to Sec tion 6.3.3.26).
The failure of any single acti ve component or the develop ment of excessive leakage during the long-term cooling pe riod does not interfere with the ability to meet necessary long-term cooling objectives with one of the systems.
The primary function of the ECCS is to deliver borated cooling water to the reactor core in the event of a LOCA. This limits the f uel cladding temperature and thereby ensures that the core will remain inta ct and in place, with its essential heat transfer geometr y preserved.
This protection is afforded for:
(1) All pipe break size s up to and including the hy pothetical circumferential rupture of a reactor coolant loop (2) A loss of coolant associated with a rod ejec tion accident
The basic criteria for LOCA evaluations are (a) no cladding melting will occur, (b) zirconium-water reactions will be limited to an insignificant amount, an d (c) the core geometry will remain es sentially in place and intact so that effectiv e cooling of th e core will not be impaired. The zirc onium-water reactions will be limited to an insignificant amount so that the accident:
(1) Does not interfer e with the emergency core cooli ng function to limit cladding temperatures (2) Does not produce hy drogen in an amount that, when burned, would cause the containment pressure to exceed the design value
For any rupture of a steam pipe and the assoc iated uncontrolled h eat removal from the core, the ECCS adds shutdown reactivity so that with a stu ck rod, no offsit e power, and minimum ESF, there is no con sequential damage to the pr imary system and the core DCPP UNITS 1 &
2 FSAR UPDATE 6.3-43 Revision 23 December 2016 remains in place and intact. With no stuc k rod, offsite pow er, and all equipment operating at design capacity, there is insigni ficant cladding rupture.
6.3.3.12.1 Use of Dual Function Components The ECCS contains components that have no other operating function as well as
components that are shared with other systems. Components in each category are as
follows:
(1) Components of the ECCS that perform no other function are:
(a) One accumulator for each reactor coolant loop that discharges borated water into its respective cold leg of the reactor coolant loop
piping (b) Associated piping, valves, and instrumentation (2) Components of the ECCS that also have a normal operating function are as follows:
(a) The RHR pumps and heat exchangers: These components are normally used during the latter stages of normal reactor cooldown
and when the reactor is held at cold shutdown for core decay heat
removal. However, during all other plant operating periods, they
are aligned to perform the low-head injection function.
(b) Two SI pumps that supply borated water for core cooling to the RCS (Note that the SI pumps are also used for SI accumulator fill
and makeup, but this is not a significant function.) During refueling, the pumps may be used for the boration flow path with all reactor
head bolts fully detensioned.
(c) CCP1 and CCP2: These pumps are normally aligned for charging service. The normal operation of the pumps as part of the CVCS is
discussed in Section 9.3.4.
(d) The RWST: This tank is used to fill the refueling canal for refueling operations (refer to Section 9.1). During all other plant operating
periods, it is aligned to the suction of the SI pumps and the RHR
pumps. CCP1 and CCP2 are automatically aligned to the suction
of the RWST upon receipt of the "S" signal.
An evaluation of all components required for operation of the ECCS demonstrates that
either:
(1) The component is not shared with other systems.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-44 Revision 23 December 2016 (2) If the component is shared with other systems, it is aligned during normal plant operation to perform its accident function, or, if not aligned to its accident function, two valves in parallel are provided to align the system
for injection, and two valves in series are provided to isolate portions of
the system not utilized for injection. These valves are automatically
actuated by the "S" signal.
Table 6.3-8 indicates the alignment of components during normal operation, and the
realignment required to perform the accident function.
6.3.3.13 General Design Criterion 45, 1967 - Inspection of Emergency Core Cooling Systems The components outsid e containment are acce ssible for leak tightnes s inspectio n during operation of the reactor. Periodic visual inspection of portions of the ECCS for leakage
is specified in the Technical Specifications (Reference 10). ECCS systems are
inspected in accordance with ASME BPVC Section XI-2001 through 2003 Addenda, as
stated in the Diablo Canyon Power Plant (DCPP) Inservice Inspection Program Plan (Reference 9). Details of the inspection program for the reactor vessel internals are included in Section 5.4.
6.3.3.14 General Design Criterion 46, 1967 - Testing of Emergency Core Cooling Systems Components Routine periodic testing of the ECCS components and all necessary support systems is
specified in the Technical Specifications (Reference 10). If such testing indicates a need for corrective maintenance, the redundancy of equipment in these systems
permits such maintenance without shutting down or reducing load under the conditions
established in the Technical Specifications (Reference 10).
Test connections are provided for periodic checks of the leakage of reactor coolant back
through the accumulator discharge line check valves and to ascertain that these valves
seat properly whenever the RCS pressure is raised. This test will be performed
following valve actuation due to automatic or manual action, or flow through the valve in
accordance with Technical Specifications (R eference 10). The SI test lines (and associated RCPB test valves) that are used f or ECCS check valve surveillance testing
are designed for testing in Modes 4 and 5 only. Per Section 6.3.3.6.2.2, maximum flow
rate through each RCPB test valve is limited such that, in the event of a downstream
pipe break during testing in Modes 4 or 5, th e makeup flow rate from either CCP1 or CCP2 is adequate to allow time for an orderly plant shutdown/cooldown without ECCS
actuation.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-45 Revision 23 December 2016 6.3.3.15 General Design Criterion 47, 1967 - Testing of Emergency Core Cooling Systems In addition to the Technical Specification (Reference 10) requirements, an ECCS subsystem is demonstrated operable during shutdown, following completion of
modifications to the ECCS subsystem that alter the subsystem flow characteristics by
performing a flow balance test to verify:
For CCP1 and CCP2, with a single pump running that:
(1) The sum of injection line flow rates, excluding the highest flow rate, is greater than or equal to 299 gpm, and (2) The total flow rate through all four injection lines is less than or equal to 461 gpm, and (3) The difference between the maximum and minimum injection line flow rates is less than or equal to 15.5 gpm, and (4) The total pump flow rate is less than or equal to 560 gpm.
For SI pumps, with a single pump running that:
(1) The sum of injection line flow rates, excluding the highest flow rate, is greater than or equal to 427 gpm, and (2) The total flow rate through all four injection lines is less than or equal to 650 gpm, and (3) The difference between the maximum and minimum injection line flow rates is less than or equal to 20.0 gpm, and (4) The total pump flow rate is less than or equal to 675 gpm.
The RHR subsystem is demonstrated operable during shutdown, following completion
of modifications to the RHR subsystem that alter the subsystem flow characteristics, by
performing a flow test and verifying a total flow rate greater than or equal to 3976 gpm with a single RHR pump running and delivering flow to all four cold legs.
6.3.3.16 General Design Criterion 48, 1967 - Testing of Operational Sequence of Emergency Core Cooling Systems As discussed in Section 6.3.3.7, a system actuation test can be performed when the plant is cooled down and the RHR system is in operation. Details of the testing of the
sensors and logic circuits associated with the generation of an "S" signal, together with the application of this signal to the operation of each active component, are given in
Section 7.3.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-46 Revision 23 December 2016 6.3.3.17 General Design Criterion 49, 1967 - Containment Design Basis
The ECCS containment pe netrations, including the system piping and valves required for containment isolation, are designed and analyzed to withstand the pressures and temperatures that could result from a LOCA without exceeding design leakage rates.
Refer to Section 3.8.2.1.
1.3 for additional details.
6.3.3.18 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment The ECCS isolation valves required for containment closure are periodically tested as part of the Inservice Testing (IST) Program Plan for operability in accordance with GDC 54, 1971. Refer to Section 6.3.3.14 for additional discussion on periodic testing. 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
containment isolation system (CIS) is discussed in Section 6.2.4.
6.3.3.19 General Design Criterion 55, 1971 - Reactor Coolant Pressure Boundary Penetrating Containment The ECCS is designed such that for each line that is part of the reactor coolant pressure boundary that penetrates containment is provided with containment isolation valves 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.
6.3.3.20 General Design Criterion 56, 1971 - Primary Containment Isolation The ECCS is designed such that each line that connects directly to the containment atmosphere and penetrates containment is provided with containment isolation valves in compliance with GDC 56, 1971. Refer to Section 6.2.4.2.1 and Table 6.2-39 for
penetration and configuration details with regards to GDC 56, 1971.
6.3.3.21 Emergency Core Cooling System Safety Function Requirements (1) Leakage Provisions and Flooding Protection In the event of a LOCA, fission products may be recirculated via the RHR system exterior to the containment. If an RHR pump seal should fail, water would spill onto the floor of the pump compartment. Each RHR pump is in a separate, shielded compartment that drains to a sump containing two pumps that can pump the spillage to the liquid radwaste system (LRS). Each sump pump is capable of removing the spillage that would result from the failure of one RHR pump seal.
If flooding occurred, overflow from one pum p compartment would drain through a 14 inch line to the pipe trench rather than flood the adjacent compartment. Added sump DCPP UNITS 1 &
2 FSAR UPDATE 6.3-47 Revision 23 December 2016 pump reliability is achieved by elevating the drive motors above the compartment overflow drain so that the pump motors would not be flooded. Gross leakage from the RHR system can be accommodated in the pump compartments, each of which has a capacity of 9450 gallons.
The RHR heat exchangers and pumps can also be isolated, in the event of gross leakage, through appropriate isolation valves. The isolation valves are operated manually by means of remote valve reach-rod operators located in a shielded valve gallery. Radiation levels in the vicinity of the recirculation loop are discussed in Chapter 12.
Recirculation loop component leakage is de tected by means of a radiation monitor that samples the air in the ventilation exhaust ducts from each compartment. Supplemental radiation monitoring is provided in the plant vent (refe r to Section 9.4.2.3.6). Alarms in the control room alert the operator when the activity exceeds a preset level, and the capability exists to detect small leaks within a short period of time. Operation of the sump pumps is a less sensitive indication of leakage. Recirculation loop components that are potential sources of leaks are described in Table 5.5-11. The table lists conservative estimates of the maximum leakage expected from each leak source during normal operation. However, the design basis for sizing auxiliary bu ilding sump pumps that will be required to dispose of this leakage employs a conservative value of 35 gpm, as described above.
The consequences of a leak through an RHR heat exchanger to the CCW system are discussed in Section 9.2.
6.3.3.22 10 CFR 50.46 - Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Plants
The ECCS calculated cooling performance is capable of demonstrating a high level of probability that the limits set forth in 10 CFR 50.46 are met. Sections 15.3.1 and 15.4.1 provide discussion of SBLOCA and LBLOCA analyses, respectively, including the
approved evaluation methodologies, range of coolant rupture and leak sizes evaluated, other properties sufficient to provide assurance that the most severe postulated LOCAs
are calculated, and demonstration that the limits set forth in 10 CFR 50.46 are met.
6.3.3.23 10 CFR 50.49 - Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants The ECCS 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. 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, valves, and
switches, and are listed on the EQ Master List.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-48 Revision 23 December 2016 6.3.3.23.1 Evaluation of the Capability to Withstand Post-Accident Environment A comprehensive testing program was undertaken to demonstrate that ECCS
components and associated instrumentation and electrical equipment that are located
inside the containment will operate for the required time period, under the combined
post-LOCA conditions of temperature, pressure, humidity, radiation, chemistry, and
seismic phenomena (Reference 6).
6.3.3.24 10 CFR 50.55a(f) - Inservice Testing Requirements The IST requirements for the ECCS are contained in the DCPP IST Program Plan.
6.3.3.25 10 CFR 50.55a(g) - Inservice Inspection Requirements ECCS systems are inspected in accordance with ASME BPVC Section XI-2001 through 2003 Addenda, as stated in the DCPP Inservice Inspection Program Plan (Reference
9).
6.3.3.26 Safety Guide 1, November 1970 - Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System
Pumps The ECCS is designed so that adequate NP SH is provided to system pumps. In addition to considering the static head and suction line pressure drop, the calculation of available NPSH in the recirculation mode for the RHR pumps assumes that the vapor pressure of the liquid in the sump equals containment pressure. This assumption
ensures that the actual available NPSH is always greater than the calculated NPSH.
The calculation of available NP SH during recirculation is as follows:
NPSH actual = (h)containment pressure
- (h)water vapor, partial pressure
+ (h)static head
- (h)loss NPSH calculated
= (h)static head
- (h)loss NPSH for ECCS pumps is further discussed in Section 6.3.3.33.
6.3.3.27 Regulatory Guide 1.79, June 1974 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors
The integrated system test for the ECCS was formulated using Regulatory Guide 1.79, June 1974 as a basis. The integrated system test was divided into three phases and
was performed after the five component tests, discussed in Section 6.3.4.2.1, had been
completed.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-49 Revision 23 December 2016 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED.
6.3.3.27.1 Integrated System Tests
6.3.3.27.1.1 System Test - Phase I
The Phase I test was conducted prior to hot functional tests and involved testing the ECCS at ambient conditions w ith the reactor vessel open.
A loss of offsite power was simulated prior to test initiation. The emergency diesel
generators supplied power to the ESF equi pment through Phase I of the test.
The ECCS was tested functionally by man ually initiating safety injection and monitoring components for correct system alignment, autostarts, and pump delivery rates.
Response time data were obtained for components being tested to demonstrate that they meet or exceed acceptance criteria as established in the test.
At the conclusion of the test, the RHR, SI pump, and CCP1/CCP2 were realigned to the recirculation mode to demonstrate the capability of the RHR pumps to deliver water from the containment recirculation sump to the SI pump and CCP1/CCP2 suctions, and to the CSS headers. The time required for changeover to recirculation was evaluated to
demonstrate that it can be compl eted during the time allowed.
6.3.3.27.1.2 System Test - Phase II The Phase II test was conducted during hot functional testing with the RCS at hot operating conditions.
High-pressure safety injection (CCP1 and C CP2) was tested by manually initiating safety injection and monitoring compon ents for correct ali gnment, autostarts of pumps, and delivery of water from the RWST to the reactor vessel through the high-pressure
safety injection branch lines.
Response time data were obtained for components being tested to demonstrate that they met or exceeded acceptance criteria.
6.3.3.27.1.3 System Test - Phase III
During Phase III tests, the SI pumps and accumulator check valves were tested. The test was conducted during the cooldown pha se of hot functional testing as the required RCS pressures were reached.
SI pumps were tested by manually in itiating safety injection with the RCS pressure at a value below the shutoff head of the pumps.
SI pump autostart and delivery of water from the RWST to the reactor vessel via SI pump injection flowpaths were checked.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-50 Revision 23 December 2016 Response time data were obtained for components being tested to demonstrate that they meet or exceed acceptance criteria.
Accumulator check valve operation was verified as RCS pressure decreased to a value below the accumulator pressure setpoint.
The accumulator discharge isolation valves were closed as soon as flow through the check valves had been verified to minimize the
6.3.3.27.2 Preoperational Testing Conformance with Regulatory Guide 1.79, June 1974 The ECCS tests described above meet the requirements of Regulatory Guide 1.79, June 1974, except in the following instances:
During the hot flow test (Section 6.3.3.27.1.3/Paragraph C.3.a(2) of Regulatory Guide 1.79, June 1974), feedwater flow from the AFW pumps is blocked in order to avoid a
temperature and pressure transient from this cause in the RCS. The pumps are started
on the "S" signal but will be run on recircul ation. Flow from the AFW pumps to the steam generators is verified as part of the feedwater system tests. The quantity of
water injected into the RCS by CCP1 and CCP2 during this test is limited by pressurizer
water level, rather than limiting the quantity to avoid reducing the number of design stress cycles. Calculations indicate that the injection nozzles are subjected to
essentially the full thermal shock by the time any meaningful data can be obtained from this test.
During the recirculation phase of the SI pumps low-pressure test at ambient conditions with the reactor vessel open (Section 6.3.3.27.1.1/Paragraph C.3.b(2) of Regulatory Guide 1.79, June 1974), temporary piping is installed from the refueling canal into the containment recirculation sump. This temporary piping bypasses the coarse but not the fine sump screen. The pressure drop across the coarse and fine screens is less than one thousandth of a foot of water head, whic h is impractical to measure and which will not compromise the pump NPSH.
During the testing of the accumulators under ambient conditions to verify flowrates (Section 6.3.4.2.1/Paragraph C.3.c(1) of Regulatory Guide 1.79, June 1974), the
accumulator discharge is initiated by openi ng the accumulator isolation valves, not by rapidly reducing RCS pressure. The discharge flowrate is calculated from the change of accumulator pressure with time, not from the change of accu mulator level with time.
Accumulator discharge tests are not repeated for normal and emergency power supplies; operation of the accumulator isolation valves with both normal and emergency power supplies is demonstrated as a part of other tests.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-51 Revision 23 December 2016 6.3.3.28 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 ECCS post-accident instrumentation for meeting Regulatory Guide 1.97, Revision 3, requirements consist of RHR system flow indication, SI pump discharge flow indication, CCP1 and CCP2 injection header flow indication, RHR heat exchanger outlet temperature indication, containment recirculation sump water level and temperature indication, accumulator tank level and pressure indication, accumulator isolation valve
position indication, RWST level indication, containment isolation valve position
indication, and subcooling margin indication (refer to Table 7.5-6).
DCPP uses the RVLIS processors to calculate RCS subcooling. Refer to Section 7.5.2.2.1 for details of the display, calculator, and inputs.
6.3.3.29 NUREG-0737 (Items I.C.1, I.D.2, II.B.2, II.F.1, I I.F.2, II.K.3.30, II.K.3.31, III.D.1.1), November 1980 -
Clarification of TMI Action Plan Requirements Item I.C.1 - Guidance for the Evaluation and Development of Procedures for Transients and Accidents: NUREG-0737, Supplement 1, January 1983 provides the ECCS requirements for I.C.1:
Section 7.1(b) - Upgraded emergency operating procedures have been implemented in
accordance with the Westinghouse Owners Group (WOG) developed generic
emergency response guidelines. The WOG is now known as the Pressurized Water Reactor Owners Group (PWROG).
Item I.D.2 - Plant Safety Parameter Display Console: NUREG-0737, Supplement 1, January 1983 provides the ECCS requirements for I.D.2:
Section 4.1(f)(v) - Containment Conditions: Containment recirculation sump water level
indication is provided on secondary displays of the safety parameter display system (SPDS). The SPDS dis plays are available in t he control room, TSC, and EOF (refer to Section 7.5.2.10).
Item II.B.2 - Design Review of Plant Shielding and Environmental Qualification of Equipment for Space/Systems Which May Be Used in Postaccident Operations: The
switchgear rooms are sufficiently shielded from external sources of radiation such that
personnel access and occupancy would not be unduly limited by the radiation
environment caused by a degraded core accident.
Item II.F.1 - Additional Accident Monitoring Instrumentation:
Position (5) - Continuous instrumentation to monitor containment recirculation sump
water level is provided in the control room (r efer to Sections 6.3.3.4.4.3, and 7.5.2.1.3; DCPP UNITS 1 &
2 FSAR UPDATE 6.3-52 Revision 23 December 2016 Figures 7.5-1, and 7.5-1B). Instrument ranges and accuracies are provided in Tables 7.5-2 and 7.5-4.
Item II.F.2 - Instrumentation for Detection of Inadequate Core Cooling: The instrumentation for detection of inadequate core cooling includes the subcooled margin
monitors, core exit thermocouple system, and RVLIS. Refer to Section 7.5.2.2 for
further discussion.
Item II.K.3.30 - Revised Small-Break Loss-Of-Coolant-Accident Methods to Show
Compliance with 10 CFR Part 50, Appendix K: The methodology for calculating
SBLOCAs (i.e., the NOTRUMP model) was submitted generically by the WOG (now
PWROG), of which PG&E is a participating member. The NOTRUMP model was
approved by the NRC Staff for use on DCPP Unit 1 and Unit 2. Refer to Section 15.3.1
for further discussion.
Item II.K.3.31 - Plant-Specific Calculations to Show Compliance with 10 CFR Part 50.46: Plant specific calculations were performed for DCPP using an NRC approved
model (NOTRUMP) for evaluating SBLOCAs.
These calculations determined the acceptance criteria of 10 CFR 50.46 are met for DCPP. Refer to Section 15.3.1 for
further discussion.
Item III.D.1.1 - Integrity of Systems Outside Containment Likely to Contain Radioactive
Material for Pressurized-Water Reactors and Boiling-Water Reactors: Pressure
containing portions of the ECCS are tested periodically to check for leakage. This testing includes the portions of the system that would circulate radioactive water from
the containment recirculation sump. The requirements for a leakage reduction program from reactor coolant sources outside containment are included in the Technical Specifications (Reference 10). Inservice valve leakage requirements are specified in
the IST Program. Refer to Section 6.2.4 for additional information on the CIS and
Section 6.3.3.14 for additional dis cussion on ECCS leakage testing.
6.3.3.30 Generic Letter 89-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance PG&E Design Class I MOVs are designed to function with a pressure differential across the valve disk determined in accordance with Generic Letter 89-10, June 1989. ECCS
MOVs, except the accumulator isolation valves, are subject to the requirements of Generic Letter 89-10, June 1989, and associated Generic Letter 96-05, September
1996, Periodic Verification of Design-Basis Capability of Safety-Related Motor-
Operated Valves, and meet the requirements of the DCPP MOV Program Plan. The
accumulator isolation valves have no active or credited safety function; therefore, they
are not included in DCPPs formal MOV Program Plan.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-53 Revision 23 December 2016 6.3.3.31 Generic Letter 95-07, August 1995 - Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves PG&E Design Class I power-operated gate valves in the ECCS that were determined to be susceptible to pressure lo cking were modifie d by installing bonn et cavity leakoffs with block valves to the high pre ssure inlet lines to prevent pressure locking. No power-operated gate valves in the ECCS were found su sceptible to the rmal binding.
6.3.3.32 Generic Letter 96-06, September 1996 - Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident
Conditions Generic Letter 96-06, September 1996 identified the potential of thermally induced overpressurization of isolated water-filled piping sections in containment that could
jeopardize the ability of accident mitigating systems to perform their safety functions and
could also lead to a breach of containment integrity via bypass leakage.
In the evaluation to determine potentially affected piping sections, DCPP identified two
ECCS piping sections. The evaluation concluded that no corrective actions were
necessary on ECCS piping as one section was deemed to meet ANSI B31.1-1967
allowable pressure values and the other sec tion was deemed to have an isolation valve (air-operated globe valve) whose design prevents overpressurization.
For the section of piping that was deemed to have an isolation valve whose design
prevents overpressurization, PG&E performed a bench test on a representative globe
valve from their warehouse. The prototype test demonstrated that all 3/8-, 3/4- and 1-inch installed globe valves could be credited to relieve pressure prior to their piping section pressure stresses exceeding ANSI B31.1-1973 Summer Addenda allowable
design values and, therefore, this section did not need physical modification.
6.3.3.33 Generic Letter 97-04, October 1997
- Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment
Heat Removal Pumps The general methodology for calculating NPSH is discussed in Section 6.3.3.26.
Evaluation of head loss associated with the ECCS suction strainers is discussed in Section 6.3.3.35. Adequate NPSH is shown to be available for all ECCS pumps as follows:
(1) RHR Pumps The NPSH of the RHR pumps was evaluated for normal plant shutdown
operation, and for both the injection and recirculation modes of operation
for the design basis accident. Recirculation operation gives the limiting
NPSH requirement. The NPSH evaluation was based on all pumps (i.e.,
both RHR, CCP1 and CCP2, both SI, and both CSS pumps) operating at DCPP UNITS 1 &
2 FSAR UPDATE 6.3-54 Revision 23 December 2016 the maximum design (runout) flowrates. The minimum available and required NPSH values for this pump are given in Table 6.3-11.
(2) Safety Injection and Centrifugal Charging Pumps 1 and 2 The NPSH for the SI pumps and CCP1/CCP2 was evaluated for both the
injection and recirculation modes of operation for the design basis
accident. The end of the injection mode of operation gives the limiting
NPSH available. The limiting NPSH was determined from the elevation
head and vapor pressure of the water in the RWST, which is at
atmospheric pressure, and the pressure drop in the suction piping from
the tank to the pumps. The NPSH evaluation is based on all pumps
operating at the maximum design flowrates. Following switchover to the
recirculation mode, adequate NPSH is supplied from the containment
recirculation sump by the booster action of the RHR pumps. The
minimum available and required NPSH for these pumps are given in
Table 6.3-11.
6.3.3.34 Generic Letter 98-04, July 1998
- Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System After a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment PG&E has implemented controls for procurement, application, and maintenance of Service Level 1 protective coatings used inside containment. The requirements of 10
CFR Part 50, Appendix B, are implemented through specification of appropriate technical and quality requirements for the Service Level 1 coatings program which includes ongoing maintenance activities.
PG&E has implemented a Coating Quality M onitoring Program that includes a thorough visual inspection of selected portions of the coatings inside the containment. As localized areas of degraded coatings are identified, those areas are evaluated and
scheduled for repair or replacement, as necessary. The periodic condition assessments, and the resulting repair/replacement activities, assure that the amount of Service Level
1 coatings which may be susceptible to detachment from the substrate during a LOCA
event is minimized. PG&E conducts condition assessments of Service Level 1 coatings
inside containment every refueling outage.
DCPP Technical Specifications (Reference
- 10) require periodic verification by visual inspection that the containment recirculation sump suction inlet trash racks and screens
are not restricted by debris and show no evidence of structural distress or abnormal
corrosion. DCPP plant procedures require internal inspection of sump plenums and RHR suction piping under specific circumstances. DCPP plant procedures ensure no
loose debris (rags, trash, clothing, insulation, plastics, etc.) exists inside containment
that could be transported to the containment recirculation sump in the event of a LOCA.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-55 Revision 23 December 2016 6.3.3.35 Generic Letter 2004-02, September 2004 - Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors The evaluation of insulation and other debris affecting containment recirculation sump availability following a LOCA was completed based on the requirements and guidance of Generic Letter 2004-02, September 2004, which was issued by the NRC to address
the potential impact of debris blockage on emergency recirculation during design basis
accidents at pressurized water reactors. The evaluation is based on the guidance
provided in NEI 04-07, Revision 0, "Pressurized Water Reactor Sump Performance
Evaluation Methodology," and the subsequently issued safety evaluation of that guidance by the NRC.
- NEI 04-07, along with the NRC safety evaluation, provides an acceptable methodology to evaluate and resolve the potential impact of debris
blockage on the emergency recirculation strainer. This methodology, along with subsequent industry guidance, provides a conservative
approach to evaluate the following main topics associated with post-
accident strainer performance:
- Upstream Effects
- Debris Generation
- Debris Transport
- Head Loss
- Downstream Effects Upstream Effects
The objective of the upstream effects assessment is to evaluate the flowpaths upstream
of the containment strainer for holdup of inventory which could reduce flow to and
possibly starve the strainer.
This evaluation was performed in accordance with the recommendations contained
within NEI 04-07 to identify those flowpaths that could result in the holdup of water not
previously considered. These flowpaths included those areas into which CSS and RCS
break flow would enter.
After holdup from the refueling cavity drain, ductwork, the portion of the reactor cavity
below the 91 foot elevation, and holdup curbs had been conservatively estimated and
the basis for no other significant sources of liquid holdup had been established, it was
determined that all other water return flowpaths have sufficiently large openings to
prevent the holdup of significant quantities of water that could challenge the DCPP UNITS 1 &
2 FSAR UPDATE 6.3-56 Revision 23 December 2016 containment minimum water level analysis. Therefore, the remaining water level is still sufficient to provide the containment minimum water level.
The required flowpaths for return of water to the containment recirculation sump pool include the refueling canal drains, the stairwells connecting the various elevations of containment, the reactor coolant drain tank hatch cover, and the openings (doorways)
within the crane wall.
The refueling canal is provided with a single 8-inch drain line (refer to Figure 3.2-19
Sheets 1 and 2, grid 37-C) with a sealed open valve (HCV-111) and blind flange that is
removed when the refueling canal is not in use. The drain is closed only when the
refueling canal is in use. The refueling cana l drain for both DCPP Unit 1 and Unit 2 is covered with a recessed deck grating that is flush with the refueling canal floor. The
grating is protected by a raised basket with openings which are sufficiently large to
prevent any credible debris from blocking this flowpath. Therefore, there is no expected
blockage of the refueling canal drain. The upper internals laydown area is within the
refueling canal and this area is slightly recessed below the nominal refueling canal floor.
This is an area of potential holdup of water and it has been estimated that this area
could holdup approximately 244 ft
- 3. This volume is not credited in the minimum containment water level.
The reactor coolant drain tank hatch cover is designed to provide a flow path for
injected ECCS water, and spilled RCS fluid, to the containment recirculation sump from a break inside the biological shield wall.
The break will flow through the reactor annulus space, fill the portion of the reactor cavity below the 91 foot elevation, and start to fill the
floor at the 91 foot elevation through this hatch cover. The fluid will then flow out of the openings in the crane wall and to the containment recirculation sump. For breaks other than inside the biological shie ld wall, the reactor coolant drain tank hatch cover is not required to function and is not credited.
As a result of the evaluations performed and physical changes completed it was
determined that the upstream effects analysis provides the necessary level of
assurance that the required volume of water will be available to the containment
recirculation sump for the function to meet the applicable requirements as set forth in NEI 04-07 and Generic Letter 2004-02, September 2004.
Debris Generation
Debris generation analysis has two primary inputs. The plant accident analysis
identifies the postulated accidents that require RHR strainer operation by the ECCS in
the recirculation mode from the containment recirculation sump. An accurate inventory
of debris source materials addressing type, quantity, location, and characteristics of the
materials is also required.
The purpose of the debris generation evaluation is to determine which breaks have the
potential to challenge the sump operation in a post-accident scenario. Break locations DCPP UNITS 1 &
2 FSAR UPDATE 6.3-57 Revision 23 December 2016 are postulated based upon which location gi ves both the most fibrous debris (e.g., insulation) and the worst combination of debris with regard to expected debris transport and head loss behavior given the respective zone of influence (ZOI) of the debris. The
ZOI represents the zone where a given high-energy line break will generate debris that
will be transported to the strainer. The locations will be used to determine total debris
generated. The debris generated is then assigned size distributions and defined by
material characteristics. The methodology as provided in NEI 04-07 was generally
followed along with the recommendations from the safety evaluation as applicable.
For break selection, the only exception taken to NEI 04-07 and the safety evaluation
was the use of the criterion specifying "every five feet" as described in the safety
evaluation. Due to the volume and configuration of DCPP's containment, the
overlapping ZOIs essentially covered the same locations. The approach used was to determine the limiting debris generation locations (based on ZOIs) and then determine
the quantity and types of debris within the ZOI. This simplification of the process did not
reduce the debris generation potential for the worst case conditions as described in NEI
04-07 and the safety evaluation.
Through a review of the breaks evaluated it was determined that all breaks generate
similar quantities of debris from erosion of unjacketed fibrous materials, latent dirt/dust, miscellaneous debris (stickers, tags, labels, tape), coatings in the ZOI (particulate), and
unqualified coating chips. Therefore, breaks that present the greatest challenge to post-accident sump performance are breaks that generate limiting amounts of cal-sil and
fibrous debris. All areas with a significant potential to generate fibrous debris (Loop 2
crossover leg, pressurizer surge line, and pressurizer loop seal lines) have been
analyzed. All areas with a significant potential to generate cal-sil debris (hot-leg, cold-leg, and crossover legs on all four loops) have been analyzed. Debris quantities have been calculated for any location which generates substantial quantities of fibrous
insulation or cal-sil insulation.
Debris Transport
The purpose of the debris transport evaluation is to estimate the fraction of debris that
would be transported from debris sources within containment to the sump strainers.
The methodology used in the transport analysis is based on the NEI 04-07, Volume 1, guidance for refined analyses as modified by the refined methodologies suggested in
Appendices III, IV, and VI of NEI 04-07, Volume 2. The specific effect of each of four modes of transport was analyzed for each type of debris generated. These modes of
transport are:
- Blowdown transport - the vertical and horizontal transport of debris to all areas of containment by the break jet;
- Washdown transport - the vertical (downward) transport of debris by the CSS and break flows;
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-58 Revision 23 December 2016
- Pool fill-up transport - the transport of debris by break and CSS flows from the RWST to regions that may be active or inactive during recirculation; and
- Recirculation transport - the horizontal transport of debris from the active portions of the recirculation pool to the sump screens by the flow through
the ECCS.
The logic tree approach was then applied for each type of debris determined from the
debris generation calculation. The logic tree used by DCPP is somewhat different than
the baseline logic tree provided in NEI 04-07, Volume 1. This departure was made to
account for certain nonconservative assumptions identified in NEI 04-07, Volume 2, including the transport of large pieces, the potential for washdown debris to enter the pool after inactive areas have been filled, and the direct transport of debris to the sump
screens during pool fill-up. Also, the generic logic tree was expanded to account for a
more refined debris size distribution.
As part of DCPPs debris reduction modifications, debris interceptors were installed in
all three crane wall doors. These debris interceptors are vertically mounted perforated
stainless steel plates (18-inches tall, 11 gauge, with 1/8-inch diameter holes) with a
horizontal lip (10 inches, also perforated stainless steel plate) that projects into the flow.
Prototypical debris interceptors were tested at critical parameters (expected fluid
velocity, flood height and turbulent kinetic energy [TKE]) to determine the performance of the debris interceptors. As was shown in testing, debris which transports to the debris interceptor by tumbling along the floor will be stopped by the interceptor. Debris
which is suspended near the debris interceptor is assumed to transport over the
interceptor, with the exception of paint chips.
This testing replicated an accurate reflective metal insulation (RMI) debris bed in front of
the interceptor, and suspended 9-mil unqualified coatings chips and 2-mil high heat aluminum chips (in separate tests) uniformly throughout the flow stream. The test
showed that the debris interceptor is effective in capturing a portion of the 9-mil coatings
chips and 2-mil coatings chips, even with sufficient TKE to suspend them at the
interceptor.
Head Loss
As there are no acceptable analytical methods availab le for the selection and sizing of a suitable strainer, the resolution of Generic Letter 2004-02, September 2004 for DCPP
was an evolution of iterations of head loss testing, fiber bypass testing, and debris
mitigation. A base debris loading was obtained from the existing debris within containment. As the replacement screen size was limited due to the space constraints
and fiber bypass limitations, various debris mitigation options were considered. The
resulting debris loads were determined and subsequent head loss and bypass tests
were performed to verify strainer performance.
This iteration process was repeated, as DCPP UNITS 1 &
2 FSAR UPDATE 6.3-59 Revision 23 December 2016 required to obtain successful results. The ultimate resolution was the screen head loss and fuel bottom nozzle head loss testing which confirmed the ability to maintain a
coolable core on recirculation with debris-laden fluid.
Sump strainer head losses were determined through a combination of testing and
analysis. The testing performed was designed to assure that a conservative design
basis head loss would be determined for DCPP. The head losses associated with the
portion of the strainers downstream of the perforated plate screens (the strainer
plenums and RHR piping entrance) were established using analytical methods.
The testing performed included use of debris loadings representative of the design basis debris loads and included fiber, particulate, coating chips, and chemical
precipitate debris. This testing program provides the basis for all strainer head losses
and covers both clean-screen and debris-laden conditions.
In addition to strainer head loss testing, DCPP has performed fiber bypass testing to
determine the quantity of fibrous debris which could potentially bypass the strainer and
be capable of forming a debris bed on the fuel bottom nozzle.
Downstream Effects
The purpose of the downstream effects evaluation is to evaluate the effects of debris
carried downstream of the containment recirculation sump strainers on the function of
the ECCS and CSS in terms of potential wear of components and blockage of flow
streams.
The following specific downstream effects evaluations have been completed:
- Debris ingestion evaluation
- Blockage of equipment in the ECCS/CSS flow paths
- Erosive wear of ECCS/CSS valves
- Wear and abrasion on auxiliary equipment
- Fuel and vessel evaluation
The debris ingestion evaluation determined the quantity and size of debris which may
bypass the containment sump strainer assemblies, and the concentration of this debris
in the sump pool following a high energy line break. The output of this evaluation is
used in the subsequent downstream evaluations.
The blockage evaluation determined that there are no blockage/pluggin g issues for existing piping, valves, instrumentation lines, orifices, eductors, heat exchanger tubes, and CSS nozzles. As part of the resolution for Generic Letter 2004-02, September DCPP UNITS 1 &
2 FSAR UPDATE 6.3-60 Revision 23 December 2016 2004, new ECCS screens were installed in DCPP Unit 1 and Unit 2. The new screens were specified to be fabricated from stainless steel plates with holes of 3/32-inch
perforations. Although the blockage evaluation was performed on the previous screen
configuration with 1/8-inch round openings, the blockage evaluation was reviewed and
determined to be conservative for the new replacement screen with nominal 3/32-inch
round openings. A post installation inspection was performed on the replacement
screens to verify that there were no gaps between the joints of any two adjacent
surfaces greater than the nominal hole or gap size. The potential for blockage of the
RVLIS is not included in this evaluation.
DCPP has a Westinghouse designed RVLIS for which Reference 17 states there is no blockage concern due to the debris ingested
through the sump strainer assembly during recirculation.
The erosive wear evaluation determines the downstrea m effects of sump debris with respect to erosive wear on the valves in the ECCS and CSS at DCPP Unit 1 and Unit 2 using the methodology of Reference 17.
As required by the WCAP methodology, a detailed erosive wear evaluation was
required for 12 ECCS throttle valves, Valves 8822A-D, 8904A-D, and 8810A-D. Erosion
to prevent erosive wear from significantly impacting the flow rate through the valves.
The results of this evaluation show that all valves pass the erosion evaluation using the
depleting debris concentration evaluation.
Wear and abrasion of auxiliary equipment evaluation addresses wear and abrasion from debris ingestion on the DCPP auxiliary equipment. This includes the effects of abrasive and erosive wear on applicable pumps, heat exchangers, orifices, and spray nozzles in
the ECCS and CSS, following the methodology in Reference 17.
Erosion is defined as the gradual wearing away of material on an object due to particles
impinging on the surface of the object. Abrasion is defined as the gradual wearing away
of material on an object due to friction of par ticles rubbing the surface of the object.
For heat exchangers, orifices, and spray nozzles, the two concerns raised by debris
ingestion are plugging (previously addressed) and failure due to erosive wear. Failure
of the heat exchangers, orifices, and spray nozzles to maintain system performance
could occur as a result of loss of wall material caused by erosive wear.
The DCPP heat exchangers, orifices, and spray nozzles were evaluated for the effects
of erosive wear for a constant debris concentration as determined in the debris
ingestion evaluation. The erosive wear on these components was determined to be insufficient to affect the system performance.
For pumps, the concern raised by debris ingestion through the sump strainer assembly
during recirculation is failure due to abrasive and erosive wear. Three aspects of pump
operability are potentially affected by debris ingestion including hydraulic performance, DCPP UNITS 1 &
2 FSAR UPDATE 6.3-61 Revision 23 December 2016 mechanical shaft seal assembly performance, and mechanical performance (vibration) of the pump.
For the DCPP ECCS pumps, the effect of debris ingestion through the sump strainer
assembly on three aspects of operability, including hydraulic performance, mechanical
shaft seal assembly performance, and mechanical performance (vibration) of the
pumps, were evaluated. The hydraulic and mechanical performances of the pumps
were determined to not be affected by the recirculating sump debris for the 30-day
mission time of the pumps.
There has been no demonstration that the ECCS pump primary seals would fail during a postulated LOCA. The 40-hour testing referenced in Section 8.1.3 of Reference 17
showed that the seals did not fail when tested. Mechanical pump seals at DCPP were
not considered to fail as a result of the downstream debris after a postulated LOCA.
Such seals would still be subject to a postulated single passive failure of the pressure boundary. Section 6.3.3.6.2.8 describes the detection and isolation capabilities to
minimize the effects of a post-LOCA recirculation loop leakage.
Fuel and Vessel
The objective of the fuel and vessel downstre am evaluation is to determine the effects that debris carried downstream of the containment sump strainer assembly and into the
reactor vessel has on core cooling.
The following specific fuel and vessel downstream effects evaluation have been
performed:
- Vessel blockage
- Fuel blockage, bottom nozzle tests
- Loss-of-Coolant Deposition Analysis Model (LOCADM) analysis The vessel blockage evaluation determined the potential for reactor vessel blockage
from debris carried downstream of the containment recirculation sump strainer
assembly. In addition to locations at the core inlet and exit, other possible locations for blockage within the reactor vessel internals which migh t affect core cooling were assessed. The smallest clearance in the reactor vessel exclusive of the core was found to be 0.52 inches and 0.46 inches for DCPP Unit 1 and Unit 2, respectively. These
dimensions are approximately five times greater than the dimension of the strainer
holes in the containment recirculation sump screen.
Therefore, any debris that could make it through the 3/32-inch holes in the strainer
would not challenge the limiting (smallest) clearances in the vessel.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-62 Revision 23 December 2016 DCPP performed fuel bottom nozzle head loss test to determine the effects of debris carried downstream of the containment sump strainer assembly onto the fuel assembly.
The test and evaluation is an alternate assessment of fuel blockage performed for
DCPP. DCPP is taking exception to the Reference 17 screening evaluation method. A series of fuel assembly bottom nozzle head loss tests were performed.
The test article for the fuel bottom nozzle head loss tests consisted of a simulated core
support plate, a bottom nozzle, a P-grid, an intermediate support grid, simulated fuel
rods and simulated control rods. Fuel bottom nozzle head loss tests were conducted
using the actual fibrous debris which bypassed the test sector during the fiber bypass
tests with maximum particulate debris (it was conservatively assumed that 100 percent
of the particulate debris which arrives at the strainer also arrives at the fuel bottom nozzle). Unqualified inorganic zinc and unqualified high heat aluminum coatings were
conservatively assumed to fail as particulate debris when conducting the fuel bottom
nozzle head loss tests, and were conservatively assumed to fail as chips when
conducting strainer head loss tests. Fuel bottom nozzle head loss tests conservatively
included all chemical precipitate debris.
The fuel bottom nozzle head loss effects were evaluated by Westinghouse through a
comparison between the measured head loss of the test data and available driving head for the various DCPP LOCA scenarios.
The Westinghouse comparisons showed that sufficient driving head is available to
match the head loss due to debris buildup, t herefore, adequate flow will enter the core to match boil-off, and the core will remain covered. Because the core remains covered, Westinghouse concluded that no late heatup occurs, and the maximum local oxidation, the corewide oxidation, and the peak claddi ng temperature calculations for the traditional LOCA analyses are still consi dered applicable.
The LOCADM evaluation used the LOCADM code from Reference 18 to predict the
growth of fuel cladding deposits and to determine the clad/oxide interface temperature
that results from coolant impurities entering the core following a LOCA.
The stated acceptance criterion is that the maximum cladding temperature maintained during periods when the core is covered will not exceed a core average clad
temperature of 800°F. This acceptance basis is applied after the initial quench of the
core and is consistent with the long-term core cooling requirements stated in
10 CFR 50.46 (b)(4) and 10 CFR 50.46 (b)(5).
An additional acceptance criterion is to demonstrate that the total debris deposition on
the fuel rods (oxide plus crud plus precipitate) is less than 50 mils. This is based on the
maximum acceptable deposition thickness before bridging of adjacent fuel rods by
debris is predicted to occur. Debris accumulation in the fuel was observed at the lower
grid locations during testing. The testing showed that the bridging that occurred at the
grids was acceptable, and that flow through the accumulated debris bed was sufficient
to ensure cooling of the fuel.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-63 Revision 23 December 2016 The evaluation was performed with the LOCADM code using DCPP specific data. The
results of this evaluation show that the calculated fuel cladding deposits and clad/oxide interface temperature do not challenge the acceptance criteria.
For the minimum sump water volume cases, LOCADM was also run with increased
quantities of debris - in accordance with the bump-up factor methodology. The bump-
up factor had a negligible effect on both the total thickness and fuel cladding
temperature.
The results of these evaluations show that DCPP can maintain adequate long-term core cooling post-LOCA.
6.3.3.36 Generic Letter 2008-01, January 2008 - Managing Gas Accumulation in Emergency Core Cooling, Decay H eat Removal, and Containment Spray Systems Comprehensive piping isometric drawing re views were performed to evaluate ECCS piping for the potential for gas accumulation and transport to the suction of the ECCS pumps. Plant modifications (e.g., the replacement of air operated valves with manual valves, and the installation of high point vents, void headers, etc.) were implemented to
prevent gas intrusion/accumulation in the ECCS.
The DCPP Gas Intrusion Program prevents and manages gas accumulation to within design and licensing requirements. System venting procedures ensure appropriate accumulated gas removal following system breaches and during normal operation.
DCPP plant procedures include instructions and controls for performing ultrasonic
testing (UT) of ECCS piping to detect gas accumulation. The ECCS is periodically
inspected per DCPP Technical Specifications (Reference 10) to verify that ECCS piping is full of water.
Design drawing details support fill and vent activities and periodic venting of gas accumulation during normal plant operations. Drawings and procedures provide
guidance for evaluating on-line maintenance includ ing details of flushing capability to preclude gas intrusion into system piping that cannot be vented during refill operations.
6.3.3.37 IE Bulletin 79-06A (Position 8), April 1979 - Review of Operational Errors and System Misalignments Id entified During the Three Mile Island Incident Position (8) (subsequently NUREG-0737, November 1980, Item II.K.1.5, Safety Related Valve Position):
Critical manual valves in the ECCS are seale d in position and a check list is maintained for inspection on a typical audit basis. When the ESF system operates, the misalignment of any remotely-operated critical valve in the ECCS will be shown by a DCPP UNITS 1 &
2 FSAR UPDATE 6.3-64 Revision 23 December 2016 monitor light on the main control board.
All ECCS PG&E Design Class I valves wh ich are operated remotely and whose purpose is to open or close (rather than throttle flow) have position indicating lights on the main
control board. Valves with power removed from their motor operators during normal
operation have continuously energized positi on indicating lights on the main control board which are redundant to the monitor lig hts that indicate misalignment of any remotely-operated critical valve discussed above.
6.3.3.38 IE Bulletin 80-18, July 1980 - Maintenance of Adequate Minimum Flow Thru Centrifugal Charging Pumps Following Secondary Side High
Energy Line Rupture In order to ensure adequate minimum flow is available to the ECCS CCPs, DCPP maintains the CCP recirculation line isolatio n MOVs in the open position under all pumping conditions, except for post-LOCA recirculation mode when they are closed.
The S signal no longer provides automatic closure of these valves, as that signal was
removed from the valves in response to IE Bulletin 80-18, July, 1980.
6.3.3.39 NRC Bulletin 88-04, May 1988
- Potential Safety-Related Pump Loss
To prevent pump to pump interaction as a result of differences between pump flow
characteristics, check valves were installed downstrea m of the RHR heat exchangers.
During minimum flow operation the check valve will prevent the stronger pump from
deadheading/reversing flow into the weaker pump, thereby maintaining minimum
required recirculation flow.
An evaluation concluded that the existing pump minimum flow rates are adequate for all the pumps evaluated and no changes to hardware or operating procedures would be
required, with the exception of the RHR pumps during an inadvertent safety injection.
When the RHR pumps are operating in this mode, they could potentially experience
unusual wear and aging due to the flow hydraulic effects. This wear and aging is long
term in nature and is expected to result in gradual wear to the pumps. Wear and aging
of the pumps can be detected through monitoring and trending pump performance
parameters, vibration levels, and bearing temperatures. Increased maintenance and
part replacement may occur due to the described wear and aging effects.
6.3.3.40 NRC Bulletin 88-08, June 1988 - Thermal Stresses in Piping Connected to Reactor Coolant Systems The charging injection header, for both DCPP Unit 1 and Unit 2, has been modified to include a recirculation line back to the charging pump suction. This passive recirculation
line continuously vents valve seepage and accompanying pressure build-up in the charging injection header during normal operations. During accident conditions, the
small size and large resistance of the recirculation line limits the recirculation flow to an
acceptable negligible value. The charging injection header is also periodically verified to DCPP UNITS 1 &
2 FSAR UPDATE 6.3-65 Revision 23 December 2016 be less than RCS pressure, and is procedurally directed to be depressurized via venting if RCS pressure is approached. The periodic verification of the charging injection header
pressure serves as a redundant means to prevent unacceptable thermal stresses due to
temperature stratifications induced by leaking valves in the event of blockage or
maintenance to the recirculation line.
In a similar manner, the charging injection bypass line is periodically verified to be less
than RCS pressure; thereby eliminating the possibi lity of undetected leakage from this line. If RCS pressure is approached, the charging injection bypass line is procedurally
directed to be vented.
6.3.3.41 NRC Bulletin 2003-01, June 2003 - Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors DCPP Unit 1 and Unit 2 had two configurations where the flow paths to the containment recirculation sump were possibly susceptible to the issue of blockage due to
the accumulation of debris, as discussed in NRC Bulletin 2003-01, June 2003. These
were the refueling cavity drain (at elevation 99 feet 6 inches) and the three doors
installed in the crane wall (at elevation 91 feet 0 inches).
To prevent potential blockage due to the accumulation of debris, the refueling cavity
drain was modified with a raised drain screen and the three crane wall doors were
modified with replacement bars that are less restrictive to flow. This configuration allows the passage of most of the floating debris without causing a blockage of the flow path.
The doorframes function as debris curbs.
A licensee controlled program provides administrative control for the crane wall doors to ensure they perform their intended safety function. DCPP plant procedures ensure no
loose debris (rags, trash, clothing, insulation, plastics, etc.) exists inside containment and provide instructions to personnel not to use fibrous insulation inside containment as
a replacement for ref lective metal or calci um silicate insulation witho ut prior authorization.
6.3.3.42 Branch Technical Position EICSB 18, November 1975 - Application of the Single Failure Criterion to Manually-Controlled Electrically-Operated Valves During the safety review of the operating license application for the DCPP, the AEC (now NRC) regulatory staff adopted the position that failures of the type discussed in
Section 6.3.3.5.4, that could lead to spurious movement of passive MOVs, must be considered in relation to satisfying the single failure criterion. The regulatory staff's
position, as stated in Branch Technical Position EICSB 18, November 1975, considers
removal of electric power an acceptable means, under certain conditions, of satisfying
the single failure criterion. As a consequence of the regulatory staff's requirements, electric power will be removed from certain ECC S and RHR valves during normal operation. These valves are listed in Table 6.3-12. When electric power is removed
from the valve operators, power is still supp lied to position indication circuitry so that DCPP UNITS 1 &
2 FSAR UPDATE 6.3-66 Revision 23 December 2016 there is continuous, redundant position indication on the control board. Redundant position indication is provided by two sets of lights: (a) red and green position lights that
indicate open or closed, and (b) white monitor lights that illuminate when the valves are
not in their proper locked-out position.
6.3.4 Tests and Inspections To demonstrate the readiness and operability of the ECCS, all of the components are subjected to periodic tests and inspections. Preoperational performance tests of ECCS
components were conducted in the manufacturer's shop. An initial system flow test was
performed to demonstrate proper components functioning.
Refer to Sections 6.3.3.7, 6.3.3.13 through 6.3.3.16, 6.3.3.18, 6.3.3.24 and 6.3.3.25 for details regarding tests and inspections of the ECCS.
6.3.4.1 Quality Control Tests and inspections were carried out during fabrication of each of the ECCS
components. These tests were conducted and documented in accordance with the
quality assurance program discussed in Chapter 17.
HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED.
6.3.4.2 Preoperational System Tests
These tests evaluated the hydraulic and mec hanical performance of the passive and active components involved in the injection mode by demonstrating that they have been installed and adjusted so they will operate in accordance with the intent of the design.
The tests were divided into two categories: component tests and integrated system
test. The components tests were divided into the following five sub-categories:
(a) valve and pump actuation, (b) accumulator injection, (c) RHR pump, (d) SI pump, and (e) CCP1/CCP2 performance tests. The integrated systems tests are described in
Section 6.3.3.27.1.
6.3.4.2.1 Component Tests
The actuation tests verified: the operability of all ECCS valves initiated by the "S" signal, the phase A contain ment isolation signal ("T"), and the Phase B containment isolation signal ("P"), the operability of all safety feature pump circuitry down through the pump breaker control circuits, and the proper operation of all valve interlocks.
Sequencing and timing tests were conducted to verify that the ECCS components will be aligned properly to perform their intended functions.
The objective of the accumulator injection test was to verify that the injection lines were free from obstruction and that the accumula tor check valves operate correctly. The test objectives were met by a low-pressure blowdown of each accumulator. The test was DCPP UNITS 1 &
2 FSAR UPDATE 6.3-67 Revision 23 December 2016 performed with the reactor head and internals removed. The acceptance criteria for the accumulator blowdown test were based on equaling or exceeding a calculated curve;
the curve simulated the system line resistances (L/D).
The primary intent of the accumulator blowdown tests was to verify these calculated discharge line resistances.
The purpose of the three pump performance s tests (RHR pumps, SI pumps, and CCP1/CCP2) was to evaluate the hydraulic a nd mechanical performance of the pumps delivering through the flowpath required for emergency core cooling. These tests were
divided into two parts: pu mp operation under miniflow conditions and pump operation at full flow conditions.
The predicted system resistances were verified by measuring the flow in each piping branch, as each pump delivered from the RWST to the open reactor vessel, and
adjustments were made where necessary to ensure that flow was distributed properly
among branches.
During flow tests, each system was checked to ensure that there is sufficient minimum
total line resistance to preclude runout from overloading the motor of any pump. At the
completion of the flow tests, the total pu mp flow and relative flow between the branch lines were compared with the system acceptable flows.
Each system was accepted only after demonstration of proper actuation of all
components and after de monstration of flow delivery of all components within design requirements.
6.3.4.3 Containment Recirculation Sump and Screen Inspection No periodic testing is performed for the containment recirculation sump. However, the
sump and screens are inspected during eac h regularly scheduled refueling outage, and after any maintenance that could result in sufficient debris to block the sump screens.
Work area inspections are performed at the conclusion of maintenance activities
anywhere in the containment to ensure that debris that could block the sump screens is removed.
6.3.5 Instrumentation Applications Refer to Section 6.3.3.4 for the instrument applications related to ECCS.
6.
3.6 REFERENCES
- 1. Deleted in Revision 22.
- 2. Deleted in Revision 22.
- 3. Deleted in Revision 22.
DCPP UNITS 1 &
2 FSAR UPDATE 6.3-68 Revision 23 December 2016
- 4. Deleted in Revision 22.
- 5. Deleted in Revision 22.
- 6. Environmental Testing of Engine ered Safety Features Related Equipment (NSSS-Standard Scope), WCAP-7744, Volume I, August 1971.
- 7. Deleted in Revision 22.
- 8. DELETED
- 9. Diablo Canyon Power Plant - Inservice Inspection Program Plan - The Third 10 Year Inspection Interval.
- 10. Technical Specifications, Diablo Canyon Power Plant Units 1 and 2, Appendix A to License Nos. DPR-80 and DPR-82, as amended.
- 11. Deleted in Revision 22.
- 12. IEEE-Std-279, Criteria for Protection Systems for Nuclear Power Generating Stations, 1971.
- 13. Deleted in Revision 22.
- 14. Deleted in Revision 22.
- 15. Deleted in Revision 22.
- 16. License Amendment Nos. 199 (DPR-80) and 200 (DPR-82), "Technical Specification 3.5.4, Refueling Water Storage Tank (RWST)," USNRC, March 26, 2008.
- 17. Rinkacs, W.J., et al., Evaluation of Downstream Sump Debris Effects in Support of GSI-191, WCAP-16406-P, Revision 1 (Proprietary), August 2007.
- 18. Rinkacs, W.J., et al., Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid, WCAP-16793-NP, Revision 0 (Non-Proprietary), May 2007.
DCPP UNITS 1 &
2 FSAR UPDATE 6.4-1 Revision 23 December 2016 6.4 HABITABILITY SYSTEMS The DCPP habitability systems are associated with the control room and the onsite technical support center (TSC).
Both facilities are designed to be habitable throughout the course of a design basis
accident and the resulting radiological condit ions, except that the TSC system is manually activated. In addition, the control room is designed to be habitable throughout
the course of a hazardous chemical release.
6.4.1 CONTROL ROOM The DCPP control room, located at elevation 140 feet of the auxiliary building, is common to Unit 1 and Unit 2. The associated habitability systems provide for access and occupancy of the control room during normal operating conditions, radiological
emergencies, hazardous chemical emergencies, and fire emergencies. Control room
habitability is supported by administrative pro cedures, shielding, the ventilation and air conditioning system, the fire protection system, kitchen facilities, and sanitary facilities
Normal operating and post-accident control room operating, emergency, and
administrative procedures are contained in the DCPP Plant Manual. The several
volumes of the Plant Manual are listed in Section 13.5.
6.4.1.1 Design Bases 6.4.1.1.1 General Design Criterion 2, 1967 - Performance Standards
The control room habitability systems are designed to withstand the effects of or are protected against natural phenomena, such as earthquakes, flooding, tornados, winds, and other local site effects.
6.4.1.1.2 General Design Criterion 3, 1971 - Fire Protection The control room habitability systems are designed and located to minimize, consistent
with other safety requirements, the probability and effect of fires and explosions.
6.4.1.1.3 General Design Criterion 4, 1967 - Sharing of Systems The control room habitability systems and components are not shared by the DCPP
units unless it is shown safety is not impaired by the sharing.
6.4.1.1.4 General Design Criterion 11, 1967 - Control Room A control room is provided from which actions to maintain safe operational status of the
plant can be controlled. The control room is designed to support safe shutdown and to
maintain safe shutdown from the control room or from an alternate location if control
room access is lost due to fire or other causes. The control room provides adequate DCPP UNITS 1 &
2 FSAR UPDATE 6.4-2 Revision 23 December 2016 radiation protection to permit access without radiation exposures to personnel in excess of 10 CFR Part 20 limits under normal conditions.
6.4.1.1.5 General Design Criterion 12, 1967 - Instrumentation and Control Systems Instrumentation and controls related to control room habitability are provided to monitor
and maintain applicable variables within prescribed operating ranges.
6.4.1.1.6 General Design Criterion 17, 1967 - Monitoring Radioactivity Releases Radiation monitoring instrumentation is provided to monitor radioactive releases
entering each control room normal air intake and pressurization system intake. Area
radiation monitoring is provided in the control room.
6.4.1.1.7 General Design Criterion 19, 1971 - Control Room The control room is designed to permit access and occupancy of the control room under
accident conditions without personnel receiving radiatio n exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of a design basis
accident.
6.4.1.1.8 General Design Criterion 37, 1967 - Engineered Safety Features Basis for Design The control room habitability systems are designed to provide backup to the safety
provided by the core design, the reactor coolant pressure boundary, and their protection systems. 6.4.1.1.9 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 control room habitability systems is provided by a combination of physical separation, fire-rated
barriers, and automatic suppression (except in the control room) and detection.
Section III.J - Emergency Lighting: Emergency lighting or Battery Operated Lights (BOLs) are provided in the control room and associated areas required to safely shut
down a unit in the event of a fire.
Section III.L - Alternative and Dedicated Shutdown Capability: Safe shutdown
capabilities are provided in the control room and at an alternate location via the hot
shutdown panel or locally at the 480-V switchgear, for equipment powered by the 480-V system required for the safe shutdown of the plant, in the event of a fire.
DCPP UNITS 1 &
2 FSAR UPDATE 6.4-3 Revision 23 December 2016 6.4.1.1.10 Regulatory Guide 1.52, Revision 0, June 1973 - Design, Testing, and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Adsorption Units of Ligh t-Water-Cooled Nuclear Power Plants The control room heating and ventilation system (CRVS) design, testing, and
maintenance comply with applicable Re gulatory Guide 1.52, June 1973, requirements with exceptions as noted in Table 9.4-2.
6.4.1.1.11 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 Instrumentation is provided to monitor control room emergency ventilation damper
position failure on engineered safety features (ESF) actuation.
6.4.1.1.12 Regulatory Guide 1.197, Revision 0, May 2003 - Demonstrating Control Room Envelope Integrity At Nuclear Power Reactors Control room envelope (CRE) integrity testing is conducted such that:
(1) An integrated in-leakage test (i.e., the American Society for Testing and Materials [ASTM] Standard E741-2000 method) is conducted in concert with the component test.
(2) The results of the two methods correlate; and
(3) The components tested account for no less than 95 percent of the CRE in- leakage as determined by the integrated in-leakage test.
6.4.1.1.13 NUREG-0737 (Items II.B.2 and II I.D.3.4), November 1980 - Clarification of TMI Action Plan Requirements Item II.B.2 - Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces/Systems Which May Be Used in Postaccident Operations:
Adequate access to the control room is provided by design changes, increased
permanent or temporary shielding, or post-accident procedural controls.
Item III.D.3.4 - Control-Room Habitability R equirements: The control room is designed
to ensure operators are adequately protected against the effects of accidental release of
toxic and radioactive gases such that the nuclear power plant can be safely operated or shut down under design basis accident conditions.
DCPP UNITS 1 &
2 FSAR UPDATE 6.4-4 Revision 23 December 2016 6.4.1.1.14 Generic Letter 2003-01, June 2003 - Control Room Habitability The control room meets the applicable hab itability regulatory requirements and the habitability systems are designed, constructed, configured, operated, and maintained in
accordance with the facilitys design and l icensing bases, with emphasis on ensuring:
(1) The most limiting unfiltered in-leakage (and the filtered in-leakage if applicable) into the CRE is no more than the value assumed in the design basis radiological analyses for control room habitability.
(2) The most limiting unfiltered in-leakage into the CRE is incorporated into the hazardous chemical assessments. This in-leakage may differ from the value assumed in the design basis radiological an alyses. The reactor control capability is maintained from either the control room or alternate shutdown location in the event of smoke.
(3) Technical specifications verify the integrity of the CRE, and the assumed in-leakage rates of potentially contaminated air.
6.4.1.2 System Description The design bases for the functional design of control room habitability systems for both normal and emergency radiological hazards were:
HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED.
(1) 10 CFR 20.1 through 20.601, Standards for Protection Against Radiation (pre-1994; compliance with current requirements of Part 20 is addressed in Chapter 12):
...no licensee shall possess, use, or transfer licensed material in such a
manner as to cause any individual in a restricted area to receive in any
period of one calendar quarter from radioactive material and other sources
of radiation in the licensee's possession a dose in excess of the standards specified in the following table:
Rems per calendar quarter
Whole body; head and trunk, active blood-forming organs; lens of eyes, or gonads 1-1/4 Hands and forearms; feet and ankles 18-3/4 Skin of whole body 7-1/2 (2) GDC 19, 1971 - Refer to Sections 6.4.1.1.7 and 6.4.1.3.7 for discussion
DCPP UNITS 1 &
2 FSAR UPDATE 6.4-5 Revision 23 December 2016 (3) The National Council on Radiation Protection and Measurements (NCRP)
Report No. 39 (1971), Basic Radiation Protection Criteria:
It is compatible with the risk concept to accept exposures leading to doses considerably in excess of those appropriate for lifetime use when
recovery from an accident or major operational difficulty is necessary.
Saving of life, measures to circumvent substantial exposures to population
groups or even preservation of valuable installations may all be sufficient
cause for accepting above-normal exposures. Dose limits cannot be
specified. They should be commensurate with the significance of the
objective, and held to the lowest practicable level that the emergency
permits.
As described in Section 12.1.2, control room shielding consists of concrete walls, floor, and roof. Control room shielding design radiation exposure limits are consistent with
GDC 19, 1971.
The CRVS is a redundant, PG&E Design Class I system. The PG&E Design Class I systems included in the CRVS are the control room heating, ventilation, and air
conditioning (CRHVAC) system and the control room pressurization system (CRPS).
The third system included in the CRVS, the plant process computer (PPC) room air
conditioning system, is PG&E Design Class II. Sections 9.4 and 12.2 describe the
CRVS.
Control room communications are described in Section 9.5.2.
Kitchen and sanitary facilities are shared by Unit 1 and Unit 2 and are designed to support operating personnel during normal operating conditions and for the duration of
an accident.
6.4.1.3 Safety Evaluation 6.4.1.3.1 General Design Criterion 2, 1967 - Performance Standards The structures (auxiliary and turbine building s) that form the CRE are PG&E Design Class I or QA Class S (refer to Section 3.8
). These buildings or applicable portions thereof are designed to withstand the effects of winds and tornados (refer to Section 3.3), floods and tsunami (refer to Section 3.4), external missiles (refer to Section 3.5), earthquakes (refer to Section 3.7), and other applicable natural phenomena, and to protect the CRE and its safety functions from damage due to these events.
Refer to Section 9.4.1 for evaluation of CRVS components.
DCPP UNITS 1 &
2 FSAR UPDATE 6.4-6 Revision 23 December 2016 6.4.1.3.2 General Design Criterion 3, 1971 - Fire Protection The control room habitability systems are designed to the fire protection guidelines of
Branch Technical Position AP CSB 9.5-1 (refer to Appendix 9.5B, Table B-1).
The adequacy of the control room fire protection system is evaluated in Sections 7.7.1 and 9.5.1.
6.4.1.3.3 General Design Criterion 4, 1967 - Sharing of Systems The control room is common to DCPP Unit 1 and Unit 2 and therefore requires sharing
of SSCs between units. The CRVS is shared between Unit 1 and Unit 2. In addition, CRPS pressurization is shared by the control room and the TSC. The sharing of these
systems is addressed in Section 9.4.1.3.3.
6.4.1.3.4 General Design Criterion 11, 1967 - Control Room Control room habitability is provided by shielding, the CRVS, and the fire protection system. The adequacy of control room shielding is evaluated for normal operating
conditions in Chapter 11 and Section 12.1.
The adequacy of the CRVS is evaluated for normal operating conditions in Chapter 11 and Sections 9.4.1 and 12.1; for hazardous
chemical emergencies in Section 9.4.1; and for fire emergencies in Sections 9.4.1 and 9.5.1. 6.4.1.3.5 General Design Criterion 12, 1967 - Instrumentation and Control
Systems Controls for control room habitability components are provided for system operation.
Instrumentation is provided for monitoring habitability system parameters during normal
operations and accident conditions (refer to Sections 6.4.1.3.6 and 6.4.1.5).
6.4.1.3.6 General Design Criterion 17, 1967 - Monitoring Radioactivity Releases
Main control room normal air intake radiation monitors and CRPS intake radiation monitors are provided to detect radioactivity in the air flow into the control room
ventilation and pressurization systems. A control room area radiation monitor is provided to monitor radiation in the control room environs.
6.4.1.3.7 General Design Criterion 19, 1971 - Control Room
The control room habitability systems permit access and occupancy for operating the plant without personnel receiving radiation exposures in excess of GDC 19, 1971, limits
for the duration of a design basis accident. The adequacy of control room shielding is
evaluated for post-accident conditions in Sections 12.1 and 15.5. The adequacy of the CRVS is evaluated for radiological emergencies in Section 15.5. Note that for the postulated fuel handling accident in the fuel handling building, an alternate source term
is assumed per 10 CFR 50.67 (refer to Section 15.5.22).
DCPP UNITS 1 &
2 FSAR UPDATE 6.4-7 Revision 23 December 2016 6.4.1.3.8 General Design Criterion 37, 1967 - Engineered Safety Features Basis for Design
By providing for operator access and occupancy so that the plant can be maintained in a safe condition under accident conditions, the control room habitability system ESF
function performed by the CRVS provides the capabi lity to control the airborne radioactive material that could enter the control room atmosphere in the event of a
postulated loss-of-coolant accident (LOCA) to acceptable levels (refer to Section 6.1.2).
6.4.1.3.9 10 CFR Part 50, Appendix R (Sections III.
G, III.J, and I II.L) - Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979 Section III.G - The control room is provided with fire protection features that limit fire
damage and support habitability and operator occupancy for safe reactor operation in
compliance with 10 CFR Part 50, Appendix R, Section III.G requirements. The control room is constructed of noncombustible, fire-resistant, and fire-retardant materials; and is
separated from the rest of the plant by minim um three-hour fire barriers. The existence of doors in the perimeter walls and the absence of an area-wide fixed fire protection
system have been evaluated and found acceptable.
Extinguishers are located within the control room, and hose stations are located in adjacent rooms. Smoke detectors are
provided in control room cabinets and consoles containing redundant safe shutdown
cabling. Self-Contained Breathing Apparatus (SCBA) units for operators are readily
available in the control room complex. The operator can isolate the control room
manually. Also, smoke removal is facilitated by the capability to operate the ventilation system on a once-through basis (refer to Appendix 9.5B, item F.2).
Section III.J - The installed emergency lighting system in the control room provides an acceptable margin of safety equivalent to that provided by 10 CFR Part 50, Appendix R, Section III.J requirements (refer to Appendices 9.5D and 9.5B, item D.5).
Section III.L - Post-fire safe shutdown capa bility is provided that meets the requirements of 10 CFR Part 50, Appendix R, Section III.
L (refer to Appendix 9.5E).
6.4.1.3.10 Regulatory Guide 1.52, Revision 0, June 1973 - Design, Testing, and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Adsorption Units of Ligh t-Water-Cooled Nuclear Power Plants
The CRVS is the ESF system that provides the capability to control the airborne radioactive material that could enter the control room atmosphere during design basis
events to acceptable levels. The extent of compliance and noted exceptions to
Regulatory Guide 1.52, June 1973, design, testing, and maintenance criteria are
discussed in Table 9.4-2. Requirements for the performance of ventilation filter testing
are stated in the DCPP Technical Specifications.
DCPP UNITS 1 &
2 FSAR UPDATE 6.4-8 Revision 23 December 2016 6.4.1.3.11 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 Damper failure indication in the control room provides status of the Unit 1 and Unit 2
CRPS fan suction and discharge motorized dampers in response to an ESF actuation (refer to Table 7.5-6, item 63).
6.4.1.3.12 Regulatory Guide 1.197, Revision 0, May 2003 - Demonstrating Control Room Envelope Integrity At Nuclear Power Reactors
Required testing for CRE integrity includes:
(1) An integrated test for total CRE in-leakage per ASTM E741-2000,
(2) The integrated test is conducted in concert with and correlated with component testing, and
(3) The components tested should account f or no less than 95 percent of the CRE in-leakage as determined by the integrated in-leakage test.
Testing includes peer reviews to identify in-leakage vulnerabilities, quantitative testing methods, and verification, prior to testing, of t he consistency of air sources and ventilation system flow rates with the licensing basis. The requirements for compliance
with Regulatory Guide 1.197, May 2003, are described in the DCPP Technical
Specifications.
6.4.1.3.13 NUREG-0737 (Items II.B.2 and II I.D.3.4), November 1980 - Clarification of TMI Action Plan Requirements Plant shielding and the CRVS are designed to maintain control room habitability under
accident conditions as discussed in the following sections:
Item II.B.2 - Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces/Systems Which May Be Used in Postaccident Operations: The
adequacy of control room shielding is evaluated for post-accident conditions in Sections 12.1 and 15.5.
Item III.D.3.4 - Control-Room Habitability Requirements: The adequacy of the CRVS is
evaluated for radiological emergencies in Section 15.5, for hazardous chemical
emergencies in Section 9.4.1, and for fire emergencies in Sections 9.4.1 and 9.5.1.
A minimum of ten SCBA units (five for each unit) are provided in the control room.
DCPP UNITS 1 &
2 FSAR UPDATE 6.4-9 Revision 23 December 2016 6.4.1.3.14 Generic Letter 2003-01, June 2003 - Control Room Habitability Testing is performed to demonstrate compliance with guidance provided by Regulatory Guide 1.197, May 2003, using the alignment that would result in the greatest
consequence to the control room operator. The purpose of operating the CRVS in
mode 4 (the pressurization mode) is to limit radiation exposure to control room personnel in the event of a radiation release. An outside air supply damper and an
exhaust damper powered by a common power supply are intentionally failed opened as
the single active failure condition for each test configuration. The test alignment
involves operating one train in mode 4. The opposite train is placed in mode 3, recirculation, but all fans are shut off.
The DCPP CRE is designed to minimize unfiltered in-leakage. Consistent with
Regulatory Guide 1.197, May 2003, Section 1.4, Test Results and Uncertainty, the test uncertainty value is not included in results showing in-leakage to be less than 100
standard cubic feet per minute. DCPP Technica l Specifications verify the integrity of the CRE with respect to the in-leakage rates of potentially contaminated air assumed in the
accident analysis.
There are no offsite or onsite hazardous chemicals that would pose a credible threat to
DCPP control room habitability. Therefore, e ngineered controls for the control room are not required to ensure habitability against a hazardous chemical threat and no amount
of assumed unfiltered in-leakage is incorporated into PG&Es hazardous chemical
assessment.
The DCPP assessment of alternate shutdown capability in the event of fire and related smoke effects is discussed in Section 9.5 and associated appendices. SCBA units are provided within the CRE for operator use and portable fans for use with temporary
power are appropriately staged to allow operators to provide ventilation in the event that
a loss-of-power event has occurred.
6.4.1.4 Tests and Inspections Testing of the control room habitability systems is discussed in the following sections:
(1) CRVS Section 9.4.1 (2) Fire protection system Section 9.5.1 (3) Communication system Section 9.5.2
Surveillance requirements for inspection and testing of plant equipment are contained in the Technical Specifications (Reference 3) and the Plant Manual. These requirements ensure that performance capability is maintained throughout the lifetime of the plant.
DCPP UNITS 1 &
2 FSAR UPDATE 6.4-10 Revision 23 December 2016 6.4.1.5 Instrumentation Applications Smoke detector and radiation detector instrumentation employed for monitoring and
actuation of the control room habitability systems are discussed in the following
sections:
(1) CRVS Section 9.4.1 (2) Fire protection system Section 9.5.1 (3) Communication systems Section 9.5.2
Design details and logic of the instrumentation are discussed in Chapter 7.
6.4.2 TECHNICAL SUP PORT CENTER The onsite TSC, located on the upper levels of the buttresses on the west side of the
Unit 2 turbine building, is common to Unit 1 and Unit 2. The associated habitability systems provide for access and occupancy of the TSC during normal plant operating
conditions, fire emergencies, and, with manual activation, throughout the course of a
design basis accident. To this end, administrative procedures and shielding, as well as
the ventilation and air conditioning, and the fire protection systems, are used.
The TSC is sized to accommodate a minimum of 20 PG&E and 5 Nuclear Regulatory
Commission (NRC) personnel as well as necessary data and information displays. It
serves as the onsite NRC emergency headquarters. Access to the control room is via the east door of the TSC, across the Unit 2 turbine building at elevation 104 feet, and then to the control room at elevation 140 feet via the elevator or stairway on the east
side of the turbine building.
Instrumentation in the TSC is capable of providing displays of vital plant parameters
throughout the course of a DBA and personnel have the capability for transmitting
technical information between the control room and the TSC by telephone and process
computer printout.
6.4.2.1 Design Bases
6.4.2.1.1 General Design Criterion 4, 1967 - Sharing of Systems The TSC habitability systems are not shared by the DCPP units unless safety is shown
not to be impaired by the sharing.
6.4.2.1.2 10 CFR Part 20 - Standards for Protection Against Radiation TSC personnel are protected from radiation sources such that doses are maintained
below limits prescribed in 10 CFR Part 20.
DCPP UNITS 1 &
2 FSAR UPDATE 6.4-11 Revision 23 December 2016 6.4.2.1.3 10 CFR 50.47 - Emergency Plans Adequate emergency facilities and equipme nt to support the emergency response are provided and maintained.
6.4.2.1.4 NUREG-0737 (Items II.B.2 and III.A.1.2), November 1980 - Clarification of TMI Action Plan Requirements Item II.B.2 - Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces/Systems Which May Be Used in Postaccident Operations:
Adequate access to the TSC is provided by increased permanent or temporary
shielding.
Item III.A.1.2 - Upgrade Emergency Support Facilities: NUREG-0737, Supplement 1, January 1983 provides the requirements for III.A.1.2 as follows:
Section 8.2.1(e) - The TSC is envi ronmentally controlled to provide room air temperature, humidity, and cleanliness appropriate for personnel and equipment.
Section 8.2.1(f) - The TSC is provided with radiologic al protection and monitoring equipment necessary to assure that radiation exposure to any person working in the
TSC would not exceed 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.
6.4.2.2 System Description The TSC is designated to be habitable throughout the course of a design basis
accident. The outside walls, with steel bulkhead doors, form an airtight perimeter
boundary. The TSC structure is designed to PG&E Design Class III. For seismic
qualification, refer to the DCP P Q-List (Reference 8 of Section 3.2).
The TSC has the manual capability to isolate the area from the outside and to
recirculate air by the air conditioning system (refer to Section 9.4.11). The hazardous chemical release warning will have to be received from the control room to enable those in the TSC to manually isolate the area from the outside.
The TSC is provided with its own PG&E Design Class II heating, ventilation and air conditioning (HVAC) system. It is not seismically qualified and is fed from a non-Class 1E power source, although the air cleanup portion of the system has the capability to
be supplied power from a Class 1E bus.
The PG&E Design Class I CRPS system provides a redundant supply of pressurization air to the TSC ventilation system. The
CRPS connecting ductwork is designed to P G&E Design Class 1 and the TSC ventilation fans, and filter units are designed to PG&E Design Class II. For seismic
qualification, refer to the DCPP Q-List (Reference 8 of Section 3.2). Sections 12.2 and
9.4.11 describe the TSC HVAC system.
DCPP UNITS 1 &
2 FSAR UPDATE 6.4-12 Revision 23 December 2016 TSC fire protection features are designed considering the standards of the National Fire
Protection Association, as described in Section 9.5.1. A minimum of ten self-contained
breathing apparatuses (five for each unit) are provided in the TSC.
The TSC includes provisions to monitor important plant parameters. The TSC
computers provide all necessary plant and h ealth physics data to offsite facilities.
The TSC is tied to the radiological monitoring network such that a laboratory, located
adjacent to the TSC, is set aside for analytical work. The principal purpose of this
facility is to provide minimum onsite analytical capability in the event that the normal
facilities are unavailable.
Normal operating and post-accident TSC administrative procedures are discussed and
evaluated in the DCPP Manual, in Chapters 12 and 13, and in the Emergency Plan.
The TSC communications are described in Section 9.5.2.
6.4.2.3 Safety Evaluation
6.4.2.3.1 General Design Criterion 4, 1967 - Sharing of Systems The TSC habitability systems are common to Unit 1 and Unit 2 and therefore require
sharing of SSCs between units. Because th e TSC habitability systems serve no safety functions, sharing between units does not impair safety functions. The CRPS is shared
by the control room and the TSC. Sharing of the CRPS by the control room and the TSC is addressed in Section 9.4.1.3.3.
6.4.2.3.2 10 CFR Part 20 - Standards for Protection Against Radiation TSC personnel are protected from external radiation dose to the extent that doses are
maintained within the limits specified in 10 CFR Part 20. Compliance with 10 CFR Part
20 for occupational dose to TSC personnel is discussed in Sections 12.1 (shielding) and 12.2 (ventilation).
6.4.2.3.3 10 CFR 50.47 - Emergency Plans A TSC that meets applicable requirements is provided and maintained in support of emergency response (refer to Sections 6.4.2, 6.4.2
.2 and 6.4.2.3.4). Accessibility to the records of the as-built plant conditions and la yout of structures, systems, and components is provided.
DCPP UNITS 1 &
2 FSAR UPDATE 6.4-13 Revision 23 December 2016 6.4.2.3.4 NUREG-0737 (Items II.B.2 and III.A.1.2), November 1980 - Clarification of TMI Action Plan Requirements Item II.B.2, Design Review Of Plant Shield ing And Environmental Qualification Of Equipment For Space/Systems Which May Be Used In Postaccident Operations (originally Recommendation 2.1.6.b of NUREG-0578 [Reference 1]) - The TSC is
designed to meet the criteria for shielding provided in N UREG-0737, Item II.B.2.
Adequate shielding is provided to permit access to vital areas, including the control
room and TSC. Utilizing the guidelines of GDC 19, 1971, and the occupancy factors contained in Standard Review Plan 6.4, the TSC shie lding design radiation dose rate limits were established at 10 mrem/hr for direct radiation and 5 mrem/hr for airborne particulate and gaseous releases (internal to TSC). The total dose rate to any individual
in the TSC is thus limited to 15 mrem/hr, from a time period beginning 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after start of the design basis accident to 30 days later. The adequacy of shielding for the TSC has
been evaluated for normal and post-accident conditions as described in Section 12.1.
Item III.A.1.2, Upgrade Emergency Support Facilities - NUREG-0737, Supplement 1, January 1983 provides the requir ements for III.A.1.2.
The TSC is designed to meet the criteria for habitability provided in NUREG-0737, Supplement 1, items 8.2.1(e) and 8.2.1(f) (Reference 4). The guidance of NUREG-0696, 1981 (Reference 2), cited by NUREG
-0737, Supplement 1, is followed regarding ventilation, filtration, radiation monitoring, and radiation protection. The TSC HVAC
system is designed to PG&E Desi gn Class II (Section 9.4.11.3).
Section 8.2.1(e) - The adequacy of the TSC ventilation system has been evaluated for normal and post-accident operating conditions as described in Sections 9.4.11 and 12.2
and for fire emergencies as described in Sections 9.4.1 and 9.5.1.
The adequacy of TSC fire protection features is evaluated in Section 9.5.1.
Section 8.2.1(f) - TSC shielding and the ventilation system prevent post-accident doses
inside the TSC from exceeding 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.
The TSC is provided the capability for monitoring direct radiation and airborne radioactive contaminants. The monitors will provide warning if the radiation levels in the TSC reach potentially dangerous levels.
6.4.2.4 Tests and Inspections Preoperational testing of TSC habitability systems is discussed in the following sections:
(1) Ventilation and air conditioning system Section 9.4.11 (2) Fire protection system Section 9.5.1 DCPP UNITS 1 &
2 FSAR UPDATE 6.4-14 Revision 23 December 2016 (3) Communication systems Section 9.5.2 6.4.2.5 Instrumentation Applications Instrumentation and habitability support equipment associated with the TSC are
addressed in the Emergency Plan.
6.
4.3 REFERENCES
- 1. NUREG-0578, TMI Short-term Lessons Learned Requirements, U. S. Nuclear Regulatory Commission, 1979.
- 2. NUREG-0696, Functional Criteria for Emergency Response Facilities, U. S.
Nuclear Regulatory Commission, 1981.
- 3. Technical Specifications, Diablo Canyon Power Plant Units 1 and 2, Appendix A to License Nos. DPR-80 and DPR-82, as amended.
- 4. NUREG-0737 Supplement 1, Clarification of TMI Action Plan Requirements -
Requirements for Emergency Response Capability, U. S. Nuclear Regulatory Commission, 1983.
DCPP UNITS 1 &
2 FSAR UPDATE 6.5-1 Revision 23 December 2016 6.5 AUXILIARY FEEDWATER SYSTEM The auxiliary feedwater (AFW) system serves as a backup supply of feedwater to the
secondary side of the steam generators (SGs) when the main feedwater system is not available, thereby maintaining the heat sink capabi lities of the steam generators. As an Engineered Safety Features (ESF) system, the AFW system is directly relied upon to
prevent core damage and reactor coolant system (RCS) overpressurization in the event
of transients such as a loss of normal feedwater or a secondary system pipe rupture, and to provide a means for plant cooldo wn following any plant transient.
6.5.1 DESIGN BASES 6.5.1.1 General Design Criterion 2, 1967 - Performance Standards The AFW system is designed to withstand the effects of, or is protected against, natural phenomena, such as earthquakes, flooding, tornados, winds, and other local site effects.
6.5.1.2 General Design Criterion 3, 1971 - Fire Protection The AFW system is designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions.
6.5.1.3 General Design Criterion 4, 1967 - Sharing of Systems The AFW system and components are not shared by the DCPP Units unless safety is
shown to not be impaired by the sharing.
6.5.1.4 General Design Criterion 11, 1967 - Control Room The AFW system is designed to support actions to maintain and control the safe operational status of the plant from the con trol room or from an alternate location if control room access is lost due to fire or other causes.
6.5.1.5 General Design Criterion 12, 1967 -
Instrumentation and Control Systems Instrumentation and controls are provided as required to monitor and maintain AFW system variables within prescribed operating ranges.
6.5.1.6 General Design Criterion 21, 1967 - Single Failure Definition The AFW system is designed to tolerate a single failure during the period of recovery following an accident without loss of its protective function, including multiple failures resulting from a single event, which is treated as a single failure.
DCPP UNITS 1 &
2 FSAR UPDATE 6.5-2 Revision 23 December 2016 6.5.1.7 General Design Criterion 37, 1967 - Engineered Safety Features Basis for Design The AFW system is designed to provide back-up to the safety provided by the core design, the reactor coolant pressure boundary, and their protection systems in the event
of a design basis accident.
The AFW system is designed to ensure sufficient supplies of condensate-grade
auxiliary feedwater are available to support natural circulation cooldown (Generic Letter 81-21, May 1981).
The AFW system is designed to prevent stea m binding of the AFW pumps (Generic Letter 88-03, February 1988).
6.5.1.8 General Design Criterion 38, 1967 - Reliability and Testability of Engineered Safety Features The AFW system is designed to provide high functional reliability and ready testability.
6.5.1.9 General Design Criterion 40, 1967 - M issile Protection (Dynamic Effects)
The AFW system is designed to be protected against dynamic effects and missiles that might result from plant equipment failures.
6.5.1.10 General Design Criterion 41, 1967 - Engineered Safety Features Performance Capability The AFW system is designed to provide sufficient performance capability to accommodate a partial loss of installed capacity, such as a failure of a single active
component, and still perform its required safety function.
6.5.1.11 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment The AFW system is provided with leakage detection, isolation, and containment capabilities having redundancy, relia bility, and performance capabilities which reflect the importance to safety of isolating this system. The AFW system is provided with a
capability to test periodically the operabi lity of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits.
6.5.1.12 General Design Criterion 57, 1971 - Closed System Isolation Valves The AFW system contains piping connected to containment penetrations that are neither part of the reactor coolant pressure boundary nor connected directly to the
containment atmosphere. These penetrations are Group D containment isolation and
are provided with one local or remote-manual valve outside the containment.
DCPP UNITS 1 &
2 FSAR UPDATE 6.5-3 Revision 23 December 2016 6.5.1.13 Auxiliary Feedwater System Safety Function Requirements (1) Decay Heat Removal The AFW system supplies sufficient water to the SGs to remove decay heat and maintain adequate SG inventory.
6.5.1.14 10 CFR 50.49 - Enviro nmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants AFW system components that require EQ are qualified to the requirements of 10 CFR 50.49.
6.5.1.15 10 CFR 50.55a(f) - Inservice Testing Requirements AFW system ASME Code components are tested to the requirements of 10 CFR 50.55a(f)(4) and 50.55a(f)(5) to the extent practical.
6.5.1.16 10 CFR 50.55a(g) - Inservice Inspection Requirements AFW system ASME Code components are inspected to the requirements of 10 CFR 50.55a(g)(4) and a(g)(5) to the extent practical.
6.5.1.17 10 CFR 50.62 - Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants
The AFW system is designed to initiate upon receipt of a signal from the Anticipated Transient Without Scram (ATWS) Mitigation System Actuation Circuitry (AMSAC).
6.5.1.18 10 CFR 50.63 - Loss of All Alternating Current Power The AFW system is required to perform its s afety function of decay heat removal in the event of a Station Blackout.
6.5.1.19 10 CFR Part 50 Appendix R (Sections III.G, J, and L) - Fire Protection Program for Nuclear Power Facilities Op erating Prior to January 1, 1979 Section III.G - Fire Protection of Safe Shutd own Capability: The AFW system is designed with fire protection features that are capable of limiting fire damage so that one train of AFW necessary to achieve and maintain hot shutdown conditions from
either the control room or hot shutdown panel (HSP) is free of fire damage. Fire protection of the AFW system is provided by a combination of physical separation, fire-
rated barriers, and/or automatic suppression and detection.
DCPP UNITS 1 &
2 FSAR UPDATE 6.5-4 Revision 23 December 2016 Section III.J - Emergency Lighting: Emergency lighting or Battery Operated Lights (BOLs) are provided in areas where operation of the AFW system may be required to
safety 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 an alternate location via the HSP.
6.5.1.20 Regulatory Guide 1.97, Revision 3 - Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident The AFW system provides instrumentation to monitor AFW flow and Condensate Storage Tank (CST) level indication during and following an accident.
6.5.1.21 NUREG-0737 (Items II.E.1.1 and I I.E.1.2), November 1980 - Clarification of TMI Action Plan Requirements Item II.E.1.1 - Auxiliary Feedwater System Evaluation: The AFW system is designed such that AFW suction flow will not be interrupted by the failure of a common valve.
Additionally, the AFW system is designed to provide a train of AFW independent of on-site and off-site ac power and with sufficient redundancy to ensure that only one (1)
train of AFW is required to achieve and maintain safe shutdown.
Item II.E.1.2 - Auxiliary Feedwater System Initiation and Flows: The AFW system is designed to automatically initiate and is designed to the requirements of IEEE 279-1971.
6.5.1.22 Generic Letter 89-10, June 1989 - Safety-Related Motor-Operated Valve Testing and Surveillance The AFW systems PG&E Design Class I and position changeable motor-operated valves (MOVs) meet the requirements of Generic Letter 89-10, June 1989, and
associated Generic Letter 96-05, September 1996.
6.5.2 SYSTEM DESCRIPTION Following a reactor trip, decay heat is dissipated by evaporating water in the steam generators and venting the generated steam either to the condensers through the 40 percent steam dump valves or to the atmosphere through the steam generator safety valves or the 10 percent atmospheric dump valves. Steam generator water inventory must be sufficient to ensure adequate heat transfer and decay heat removal. The AFW
system must be capable of functioning for extended periods, allowing time either to
restore main feedwater flow or to proceed with an orderly cooldown of the reactor
coolant to 350°F where the RHR system can assume the burden of decay heat removal (refer to Section 5.5.6).
DCPP UNITS 1 &
2 FSAR UPDATE 6.5-5 Revision 23 December 2016 AFW system flow and emergency water supply capacity must be sufficient to remove core decay heat, reactor coolant pump heat, and sensible heat during the plant
cooldown. The AFW system can also be used to maintain the steam generator water
level above the tubes following a LOCA. The water head in the steam generators
prevents leakage of fission products from the RCS into the secondary side once the
RCS is depressurized.
The AFW system is comprised of three indep endent pump trains. Two trains consist of motor-driven AFW pumps backed by Class 1E power supplies, and one train consists of
a turbine-driven AFW pump with a Class 1E 125-Vdc steam inlet admission valve.
The motor-driven AFW pumps are each aligned to two steam generators. Flow to the
steam generators is modulated by PG&E Design C lass I, electro-hydraulic level control valves powered from Class 1E power suppli es. These valves also provide for pump runout protection.
The turbine-driven AFW pump is aligned to all four steam generators and contains
motor-operated level control valves; however, these valves do not automatically modulate. AFW is provided to the pumps from the PG&E Design Class I CST, which is
backed by the PG&E Design Class I fire water storage tank (FWST), and by the PG&E
Design Class II raw water storage reservoirs. The branch connection on two main
steam lines for the auxiliary feed pump turbine is provided with isolation valves and
In the unlikely event of a complete loss of the preferred power supply and main
generator electrical power to the station, decay heat removal would continue to be ensured by the availability of one turbine-driven, and two motor-driven AFW pumps (powered by the standby power source), and steam discharged to atmosphere through
the steam generator power-operated relief valves and/or the spring-loaded safety
valves. The system is shown in simplified form in Figure 6.5-1. For the detailed piping
schematic, refer to Figure 3.2-3, Sheets 3 and 4.
6.5.2.1 Equipment and Component Descriptions 6.5.2.1.1 Water Sources The minimum CST volume alone is sufficient to perform the plant cooldown described in
Section 6.5.3.7 and to address NRC Generic Letter 81-21, May 1981, postulated worst-
case natural circulation cooldown.
In the event the CST becomes exhausted, additional cooling water supplies are
available to maintain hot standby conditions or to bring the plant to cold shutdown.
These additional long-term cooling water sources use both existing piping systems and
pumps, along with temporary portable pump driver units and hoses. Two million gallons
of water will be available from the raw water reservoir for both units following exhaustion of preferential water sources (Reference 4). The FWST is the PG&E Design Class I DCPP UNITS 1 &
2 FSAR UPDATE 6.5-6 Revision 23 December 2016 backup water source for AFW. The design basis of the AFW system is to cooldown the RCS utilizing the PG&E Design Class I AFW sources to the point where the RHR system may be relied upon to complete the cooldown of the Unit.
The additional sources, listed in order of preference according to water quality, are as
follows:
(1) Unit 1 and 2 CST (supply from un-affected unit if water inventory is not required for that unit)
(2) Main condenser hotwells (using condensate pumps)
(3) Fire water transfer tank (4) FWST (5) Main condenser hotwells (using portable fire pumps)
(6) Raw water storage reservoirs (5 million gallons)
(7) Pacific Ocean (via auxiliary saltwater system)
The above order of preference, although desirable relative to control of steam generator
secondary side water chemistry, is not necessarily the preferred order in response to a
plant transient requiring rapid operator response. The operating procedures identify an
order of preference that is based on ensuring rapid alignment of the long-term cooling water supply.
The various long-term cooling water sources and their connections to the AFW system are shown schematically in Figure 6.5-2. Water systems are discussed in Section 9.2.
Connections and valving arrangements are provided to interconnect permanent plant systems by means of special-use hoses as follows:
(1) ASW system at the inlet water box of the CCW heat exchanger to the turbine building fire water system and then into the FWST (2) Raw water storage reservoir to the plant raw water supply line (3) Condenser hotwells to the turbine building fire water system and then into the FWST (4) Fire water system crosstie (through PG&E Design Class I piping) to the AFW system.
DCPP UNITS 1 &
2 FSAR UPDATE 6.5-7 Revision 23 December 2016 The available hoses and portable pumps (not permanently connected to existing systems) are stored in structures that have been verified to survive the postulated
Hosgri seismic event.
6.5.2.1.2 Auxiliary Feedwater Pumps and Controls Each of the two steam supply lines to the turbine-driven AFW pump is provided with a separate, normally open, PG&E Design Class I motor-operated isolation valve with Instrument Class IA control circuitry and a non-return valve. The non-return valves provide protection against potential cross-connection between the steam lines. A
normally closed, motor-operated stop valve is located in the steam supply line to the turbine inlet. During normal operation, the steam supply line is pressurized up to this
stop valve, with steam available to operate the turbine-driven AFW pump when a control
signal is received to open the stop valve. The turbine-driven AFW pump can deliver a
net flow of 780 gpm to all four (4) steam generators.
The four motor-driven AFW pumps (two per Unit) are powered from the Class 1E 4.16-
kV buses. They are available for standby service when there is insufficient steam to
operate the turbine-driven AFW pump, or when the turbine-driven AFW pump is
unavailable. Each motor-driven AFW pump can deliver a net flow of 390 gpm to two
steam generators. Note that the flow rates of 780 and 390 gpm represent the minimum
required flow rates of the turbine- and motor-driven AFW pumps at a steam generator
back-pressure corresponding to the lowest steam generator safety valve set pressure, plus 3 percent for setpoint tolerance and 5 psi for accumulation.
Controls for the AFW system are described in Sections 7.1 through 7.7. In addition to the manual actuation of the AFW pumps, the following signals provide for automatic actuation of the motor-driven AFW pumps:
(1) Two-out-of-three low-low level signals in any one steam generator (2) Trip of both main feedwater pumps (3) Safety injection signal (4) Transfer to diesel without safety injection signal (5) AMSAC
The turbine-driven AFW pump automatic actuation signals are:
(1) Two-out-of-three low-low level signals in any two steam generators (2) Undervoltage on one-out-of-two relays on both RCP buses (loss of offsite power)
DCPP UNITS 1 &
2 FSAR UPDATE 6.5-8 Revision 23 December 2016 (3) AMSAC The steam generator blowdown isolation valv es and the blowdown sample isolation valves are tripped shut whenever an AFW pump is started automatically.
6.5.2.2 Design Conditions The reactor plant conditions that impose PG&E Design Class I performance
requirements on the AFW system are as follows:
(1) Loss of normal feedwater transient (a) Loss of normal feedwater with offsite power available (b) Loss of offsite power to the station auxiliaries (2) Major secondary system pipe ruptures (c) Feedline rupture (d) Steam line rupture (inside containment)
(3) Loss of all ac power (4) Small break Loss-of-coolant accident (SBLOCA)
(5) Cooldown
Each of these conditions is discussed in more detail in the following sections.
6.5.2.2.1 Loss of Normal Feedwater Transients Design basis loss of normal feedwater transients are caused by:
(1) Interruptions of the main feedwater system flow due to a malfunction in the feedwater or condensate system
(2) Loss of offsite power with the consequential shutdown of the system pumps, auxiliaries, and controls
Loss of Normal Feedwater (Wi th Offsite Power Available)
Loss of normal feedwater transients are characterized by a reduction in steam
generator water level that results in a reactor t rip, a turbine trip, and auxiliary feedwater actuation by the protection system logic. Following reactor trip from a high initial power DCPP UNITS 1 &
2 FSAR UPDATE 6.5-9 Revision 23 December 2016 level, the power quickly falls to decay heat levels. The steam generator water levels continue to decrease, progressively uncovering the steam generator tubes as decay
heat is transferred and discharged in the form of steam either through the 40 percent dump valves to the condenser or to the atmosphere through the steam generator safety valves or the 10 percent atmospheric dump valves. The reactor coolant temperature increases since the residual heat exceeds that dissipated through the steam generators.
With increased temperature, reactor coolant volume expands and begins to fill the
pressurizer. Without the addition of sufficient AFW, further expansion wi ll result in liquid being discharged through the pressurizer safety and/or relief valves.
If the temperature rise and the resulting volumetric expansion of the primary coolant are permitted to continue, then the following ma y occur: (a) pressurizer safety valve capacities may be exceeded causing overpressurization of the RCS, and/or (b) the
continuing loss of fluid from the primary coolant system may result in bulk boiling in the RCS system and eventually in core uncovering, loss of natural circulation, and core
damage. If such a situation were to occur, the ECCS would not be effective because
the primary coolant system pressure exceeds the shutoff head of the safety injection
pumps, the nitrogen overpressure in the accumulator tanks, and the design pressure of
the RHR loop.
Loss of Offsite Power to the Station Auxiliaries The loss of offsite power transient differs from a simple loss of normal feedwater in that emergency power sources must be relied upon to operate vital equipment. The loss of
power to the motor-driven circulating water pumps results in a loss of condenser
vacuum and, therefore, of use of the 40 percent condenser dump valves. Hence, steam generated by decay heat is relieved through the steam generator safety valves or the
power-operated relief valves. The loss of normal feedwater and loss of offsite power
transient analyses are similar with the exception that reactor coolant pump heat input is
not a consideration in the loss of offsite power transient following loss of power to the
reactor coolant pump bus.
The loss of normal feedwater transient was the original basis for the minimum flow
required for the smallest capacity single AFW system pump. Each pump was originally
sized so that any single pump will provide sufficient flow against a conservative steam
generator safety valve set pressure with 3 percent tolerance and 5 psi accumulation to
prevent liquid relief from the pressurizer. The Loss of Normal Feedwater (LONF)
analysis requires that at least two motor-driven AFW pumps provide at least 600 gpm of
AFW (assuming a single failure of the turbine-driven AFW pump) to four steam
generators to prevent pressurizer overfilling. Refer to Sections 6.5.3.7 and 15.2.8.2 for
further discussion on LONF.
DCPP UNITS 1 &
2 FSAR UPDATE 6.5-10 Revision 23 December 2016 6.5.2.2.2 Major Secondary System Pipe Ruptures Feedwater Line Rupture A feedwater line rupture results in the loss of feedwater flow to the steam generators
and the complete blowdown of one steam generator within a short time if the rupture
occurs downstream of the last non-return valve in the main or AFW system piping to an individual steam generator. A feedwater line rupture may also cause spilling of AFW through the break due to the fact that the AFW system branch line may be connected to
the main feedwater line in the region of the postulated break. Such situations can result
in the injection of a disproportionately large fraction of the total AFW system flow (the
system preferentially pumps water to the lowest pressure region) to the faulted loop
rather than to the effective steam generators, which are at relatively high pressure.
System design provides for terminating, limiting, or minimizing that fraction of AFW flow, which is delivered to a faulted loop or spilled through a break to ensure that sufficient
flow is delivered to the remaining effective steam generator(s).
Main Steam Line Rupture Inside Containment Main steam line rupture accident conditions are characterized initially by a plant
cooldown and, for breaks inside c ontainment, by increasing containment pressure and temperature. Auxiliary feedwater is not required during the early phase of the transient.
However, modeling AFW flow to the faulted loop contributes to an excessive release of mass and energy to containment, maximizing the peak containment pressure. In this
way, these steam line rupture conditions establish the upper limit on AFW flow delivered
to a faulted loop and the time required to isolate the faulted steam generator.
Eventually, however, the RCS heats up again and AFW flow is required for the non-
faulted loops, but at a lower rate than the loss of feedwater transients described
previously. Provisions in the design of the AFW system limit, control, or terminate AFW
flow to the faulted loop as necessary to prevent containment overpressurization
following a steam line break inside containment, or, for steam leads 3 and 4, to maintain
the temperature profile in the GE/GW area within analyzed limits, and to ensure
minimum flow to the remaining intact loops.
6.5.2.2.3 Loss of All AC Power The loss of all ac power is postulated as resulting from accident conditions wherein not
only onsite and offsite ac power is lost, but also emergency ac power is lost as an
assumed common mode failure. Battery power for operation of protection circuits is
assumed available. The impact on the AFW system is the necessity for providing both
AFW pump power and a control source that are not dependent on ac power (refer to
Section 6.5.3.21) and which are capable of maintaini ng the plant at hot shutdown until ac power is restored. In the event of a Loss of All AC Power, decay heat removal would
continue to be ensured through the availability of one turbine-driven AFW pump.
DCPP UNITS 1 &
2 FSAR UPDATE 6.5-11 Revision 23 December 2016 6.5.2.2.4 Small-Break Loss-of-Coolant Accident The LOCAs do not impose any flow requirements on the AFW system that are in excess
of those required by the other accidents addressed in this section.
Small-Break LOCAs cause relatively slow rates of decrease in RCS pressure and liquid
volume. The principal contribution from the AFW system following a small-break LOCA
is essentially the same as the system's function during hot shutdown or following a
spurious safety injection signal which trips the reactor. Maintaining a water level
inventory in the secondary side of the steam generators provides a heat sink for
removing decay heat and establishes the capability for providing a buoyancy head for
natural circulation. The primary contributor to heat removal during a small-break LOCA, however, is through the break. The AFW system may be used to assist in system
cooldown and depressurization fol lowing a small-break LOCA while bringing the reactor to a cold shutdown condition.
6.5.2.2.5 Cooldown The AFW system is required to cool down the RCS from normal zero load temperature
to a hot leg temperature of approximately 350°F. This is the maximum temperature
recommended for placing the RHR system into service. The RHR system completes
the cooldown to cold shutdown conditions.
Cooldown may be required following expected transients, following an accident such as a main feedline break, or prior to refueling or plant maintenance. If the reactor trips following extended operation at rated
power level, the AFW system delivers sufficient feedwater to remove decay heat and
reactor coolant pump heat following reactor trip while maintaining steam generator water level. Following transients or accidents, the recommended cooldown rate is consistent with expected needs and at the same time does not impose additional
requirements on the capacities of the AFW pumps, considering a single failure.
6.5.2.3 Applicable Codes and Classifications All AFW pumps and their appropriate piping and valves are PG&E Design Class I. The
AFW system fittings and piping are designe d to ANSI Code for Pressure Piping B31.1 and B31.7, as appropriate. The system valves were originally designed to the
requirements of the ASME Pump and Valve Code, 1968 Draft. Later additions were designed to ASME B&PV Code,Section III.
6.5.3 SAFETY EVALUATION 6.5.3.1 General Design Criterion 2, 1967 - Performance Standards The AFW system components are located at the 100 foot elevation of the auxiliary
building, a PG&E Design Class I structure (refer to Figure 1.2-6). The au xiliary building is designed to withstand the effects of winds and tornados (refer to Section 3.3), floods and tsunamis (refer to Section 3.4), external missiles (refer to Section 3.5), earthquakes DCPP UNITS 1 &
2 FSAR UPDATE 6.5-12 Revision 23 December 2016 (refer to Section 3.7). This desi gn protects the AFW SSCs, ensuring their design functions will be performed.
Portions of the AFW system are not contained within a building and are exposed directly
to potential wind and tornado loads and have been evaluated. Loss of this equipment
does not compromise the capability of shutting down the plant safely due to the
availability of a train of AFW completely lo cated within qualified structures (refer to
Section 3.3.2.3).
The AFW system SSCs are designed to perform their safety functions under the effects
of earthquakes. The PG&E Design Class I portion of the AFW system is seismically
qualified.
Flooding of PG&E Design Class I equipment due to an AFW line rupture would not occur because of the relatively low flowrates and the location of the system. The
consequences of postulated pipe rupture outside the containment, including the
postulated rupture of AFW system lines, are discussed in Section 3.6.
6.5.3.2 General Design Criterion 3, 1971 - Fire Protection The AFW system is designed to the fire protection guidelines of Branch Technical Position 9.5-1 (refer to Appendix 9.5B Table B-1).
6.5.3.3 General Design Criterion 4, 1967 - Sharing of Systems The Units 1 and 2 CSTs are cross-tied through a 4-inch line. The effective elevation of the nozzles have been raised above the Technical Specification volume requirement for AFW due to the addition of internal plenums, thereby ensuring that failure of the crosstie
line cannot reduce the condensate storage capacity below the minimum required
volume to ensure steam generator makeup from the AFW pumps. Refer to Section 9.2.6.3 for a discussion on the Condensate Storage Facilities.
There is no direct connection between the raw water supply header in the plant and the
CST such that any single failure of a component could cause the loss of both CST
inventory and reservoir water. Refer to Section 9.2.3.2.2 for a discussion on the raw water reservoir. The FWST provides a PG&E Design Class I source of backup water to Units 1 and 2 for the AFW system. This source of water (the FWST) is connected to the
common supply header from both the Raw Water Reservoir and the FWST to both
Units AFW pumps. A normally closed PG&
E Design Class I valve separates the PG&E Design Class II Raw Water Reservoir supply and the PG&E Design Class I FWST
supply. This ensures that a failure of the PG&E Design Class II Raw Water Reservoirs
piping will not result in drain down of the FWST or the prevention of the FWST from supplying backup PG&E Design Class I AFW. The piping from the FWST is provided
with isolation valves such that a failure in this line will not prevent sufficient backup supplies of AFW from being provided by other backup AFW sources. As discussed in DCPP UNITS 1 &
2 FSAR UPDATE 6.5-13 Revision 23 December 2016 Section 6.5.3.7, there is a sufficient volume of water available and maintained in the CSTs for the AFW system to perform the worst-case natural circulation cooldown.
The turbine-driven AFW pump leakoffs and bearing cooling water return lines are piped
to the PG&E Design Class II common auxil iary steam drain receiver tank. Isolation
valves and check valves in these drain lines ensure that a failure of the tank or PG&E
Design Class II piping will not impair the operation of either Units AFW system.
There is no sharing between units that would prevent either Unit 1 or Unit 2 AFW
system from performing its design function.
6.5.3.4 General Design Criterion 11, 1967 - Control Room Manual initiation for each train exists in the control room. The manual initiation system is installed in the same manner as the automatic initiation system.
One PG&E Design Class I AFW flow indicator is provided for each of four steam
generators. Indication is provided at the main control board and the HSP (refer to
Section 7.5.1.6). For Unit 2 only, Instru ment Class II AFW backup flow indicators are provided at the HSP for loops 3 and 4.
These flow indicators monitor the flow from the turbine-driven AFW pump and the
motor-driven AFW pumps.
Additional indication of AFW flow is provided by PG&E Design Class I steam generator wide-range level indication (refer to Section 7.5.2.6).
This provides recording on the main control board and indication on the HSP.
The AFW system primary water supply source, the CST, contains redundant, PG&E
Design Class I level recording with indicati on in the main control board (refer to Table 7.5-6). PG&E Design Class II CST level indication is provided at the HSP.
Additional controls for components of the AFW system are available in the control room
and at the HSP. Controls for the AFW level control valves, start/stop switches for the motor-driven AFW pump, control of the steam admission valve for the turbine-driven
AFW pump, and AFW pump discharge pressure indication are all provided in the control
room and at the HSP. In addition to cooling the RCS down to RHR entry conditions (Mode 4), the AFW system allows the plant to remain in hot standby (Mode 3) for an
extended period of time if desired by operation and/or required for safe shutdown (refer
to Section 7.4.2.1).
Annunciation is provided in the main control room for AFW pump discharge piping
temperature to prevent steam binding of the AFW pumps (refer to Section 6.5.3.7).
Controls for the AFW system are provided in the control room such that the AFW
system may be operated to perform its design function.
DCPP UNITS 1 &
2 FSAR UPDATE 6.5-14 Revision 23 December 2016 6.5.3.5 General Design Criterion 12, 1967 -
Instrumentation and Control Systems All automatic initiating signals and circuits are installed in accordance with IEEE 279-1971 (Reference 1) and are PG&E Design Class I and redundant (refer to Section 7.3).
As discussed in Section 7.3 and shown in Figures 7.3-8 and 7.3-17 the motor-driven
AFW pumps are started by closure of the solid-state protection system (SSPS) output
relay and one of the timers. The relay is actuated by safety injection initiation or low-low level in any steam generator. The timers provide automatic starting sequences after
bus transfer either with or without safety injection. Each pump is started by a separate
relay or timer from redundant SSPS trains A or B. The motor-driven pumps are also automatically started by trip of both main feedwater pumps, or an AMSAC signal.
The turbine-driven AFW system pump is started by opening steam supply valve FCV-
- 95. As shown in Figure 7.3-18, this valve is opened by one of the SSPS output relays.
One of these relays starts on loss of offsite power and the other on low-low level in any
two steam generators. The turbine-driven pump is also started by an AMSAC signal.
The initiating sensors are powered from separate and redundant nuclear
instrumentation and control panels, each of which is supplied by either the Class 1E 120-Vac power supply or Class 1E 125-Vdc batteries. Each of the two redundant SSPS
trains is supplied by a separate Class 1E power source.
Instrumentation is provided in the motor-driven AFW pump discharge line to sense low
pump discharge pressure indicative of a depressurized steam generator. In a low pump discharge pressure situation, control valves are automatically throttled to prevent pump runout. This automatic action limits flow to any depressurized steam generator.
No such instrumentation is provided for the turbine-driven AFW pump. Manual action by the plant operator is required to terminate flow to a depressurized steam generator.
6.5.3.6 General Design Criterion 21, 1967 - Single Failure Definition No single failure in the manual initiation portion of the circuit can result in the loss of the AFW system function (refer to Section 7.3.2.
1.1 and Figures 7.3-17 and 7.3-18 for the circuitry).
No single failure in the automatic portion of the circuit will result in loss of the capability to manually initiate the AFW system from the control room.
All automatic initiating signals and circuits are installed in accordance with IEEE 279-
1971 (Reference 1) and are PG&E Design Class I and redundant (refer to Section 7.3).
A single failure that results in the failure of multiple components, such as the failure of
an EDG, will not result in a failure of the AFW system to perform its design function.
DCPP UNITS 1 &
2 FSAR UPDATE 6.5-15 Revision 23 December 2016 The AFW system is designed with a train that is completely independent of the preferred power supply or standby power supply (refer to Section 6.5.3.21). Isolation valves are provided in the water supply lines to the AFW pumps such that a single failure of a valve in the suction line of the AFW pumps will not result in the failure of the remaining pumps.
6.5.3.7 General Design Criterion 37, 1967 - Engineered Safety Features Basis for Design Analyses have been performed for the limiting transients that define the AFW system
performance requirements. Specifically, they include:
(1) Loss of normal feedwater (2) Loss of Offsite Power to the Station Auxiliaries (3) Rupture of a Main Feedwater Pipe (4) Rupture of a Main Steam Pipe Inside Containment
(5) Small Break Loss-of-Coolant Accident
In addition, specific calculations for DCPP Units 1 and 2 were performed to determine
plant cooldown flow (storage capacity) requirements.
The loss of all ac power was evaluated by comparison with the transient results of a loss of offsite power, assuming an available AFW pump having a diverse (non-ac) power supply. The SBLOCA analysis incorporates system flow requirements defined by
other transients and is, therefore, not perform ed to determine AFW system flow requirements. Each of the above analyses is explained further below.
Loss of Normal Feedwater A loss of normal feedwater was analyzed in Section 15.2.8 to show that two motor-
driven AFW system pumps delivering at least 600 gpm of AFW flow to four steam generators does not result in pressurizer over-filling. Furthermore, the peak RCS
pressure remains below the criterion for Condition II transients and no fuel failures
occur.
Table 6.5-2 summarizes the assumptions used in the Chapter 15 analysis. All main
feedwater flow to the steam generators is terminated at event initiation. Reactor trip is
assumed to occur when the water level in any steam generator reaches the low-low
level trip setpoint. AFW flow from both motor-driven pumps initiates within 60 seconds
after receiving a low-low level signal in any steam generator. The analysis assumes that the plant is initially operating at 102 percent (calorimetric error) of the Nuclear DCPP UNITS 1 &
2 FSAR UPDATE 6.5-16 Revision 23 December 2016 Steam Supply System (NSSS) design rating shown in Table 6.5-2, and includes a conservative assumption in defining decay heat and stored energy in the RCS.
Both the loss of normal feedwater and loss of offsite power analyses demonstrate that
there is considerable margin with respect to pressurizer over-filling (refer to Sections
15.2.8 and 15.2.9).
A better-estimate analysis is performed to address the reliability of the AFW system.
This analysis is similar to that described above for the Chapter 15 analysis, but assuming that only a single motor-driven AFW system pump supplies a minimum of 390
gpm to two of the four steam generators.
The cases considered in this additional analysis assume better-estimate conditions for several key parameters, including initial
power level, decay heat, RCS temperature, pressurizer pressure, and the low-low steam generator water level reactor trip setpoint. The results of this better-estimate
analysis demonstrate that there is margin to pressurizer over-filling. While this analysis demonstrates that the AFW system remains highly reliable, the DCPP licensing basis requires that at least two AFW pumps delivering at least 600 gpm to four steam generators is required for this event.
Loss of Offsite Power to the Station Auxiliaries The AFW system is initiated for a loss of offsite power to the station auxiliaries transient
as discussed in Section 15.2.9. The same assumptions discussed above for the loss of normal feedwater transient apply to this analysis, except that power is assumed to be
lost to the reactor coolant pumps following reactor trip.
As with all ESF equipment, the ac motor-driven AFW pumps and all valves in the system are automatically and sequentially loaded on the emergency buses on loss of
offsite power.
Rupture of Main Feedwater Pipe The double-ended rupture of a main feedwater pipe downstream of the main feedwater line check valve was analyzed (refer to Sections 15.4.2.2 and 15.4.2.4). Table 6.5-2 summarizes the assumptions used in these analyses. A reactor trip is assumed to occur when the faulted steam generator reaches the low-low level trip setpoint (adjusted
for errors). The initial power rating assumed in the feedline break analysis is 102
percent of the NSSS design rating.
For the analysis in Section 15.4.2.2, although the AFW system at DCPP Units 1 and 2 would allow delivery of AFW to two intact loops automatically in 1 minute, no AFW flow
is assumed until 10 minutes after the break. At this time it is assumed that the operator has isolated the AFW system from the break and flow from one motor-driven AFW
pump of 390 gpm (total) to two steam generators commences. As discussed in Section
15.4.2.2, the analysis assumes a single failure of the most limiting component, the
turbine-driven AFW pump, and assumes that all flow from the motor-driven AFW pump DCPP UNITS 1 &
2 FSAR UPDATE 6.5-17 Revision 23 December 2016 aligned to the faulted steam generator is lost through the break. The AFW flow is asymmetrically split between two of the three unaffected steam generators. The
analysis demonstrates that the re actor coolant remains subcooled, assuring that the core remains covered with water and no bulk boiling occurs in the hot leg.
For the analysis in Section 15.4.2.4, the limiting single failure is the failure of a PG&E Design Class I pressurizer PORV. It is assumed the motor-driven AFW pump that is initially aligned to two intact SGs delivers 390 gpm total AFW flow to two of the three intact SGs at 1 minute after the trip. It is also assumed that operator actions are taken within 10 minutes to isolate the faulted steam generator. For this reason, the other motor-driven AFW pump, which is initially aligned to both the third intact SG and the faulted SG, is assumed to deliver an additional 195 gpm of AFW flow to the third intact SG at 10 minutes after the trip.
Similarly, the turbine-driven AFW pump, which is initially aligned to all four SGs, is assumed to deliver an additional total 585 gpm of AFW flow to all three intact SGs at 10 minutes after the trip. The analysis demonstrates that water relief through the pressurizer safety valves is precluded.
Rupture of a Main Steam Pipe Inside Containment Because the result of the steam line break transient is an initial RCS cooldown, the
AFW system does not have a requirement to remove he at in the short term. However, addition of AFW to the faulted steam generator will increase the secondary mass
available for release to the containment thus maximizing the peak containment pressure
following a steam line break inside containment. This transient is performed at four
power levels for several break sizes. AFW is assumed to be initiated at the time the SI
setpoint is reached. The AFW flowrate to the faulted SG is maximized based on flow from both motor-driven AFW pumps and the turbine-driven AFW pump where runout protection is not credited. Table 6.5-2 summarizes the assumptions used in this
analysis. At 10 minutes after the break, it is assumed that the operator has isolated the AFW system from the faulted steam generator, which subsequently blows down to
ambient pressure. This assumption for operator action is also used for temperature
profile development for main steam line breaks outside containment. Refer to Section
6.2D.3 for further discussion on a main steam line break inside containment.
Small Break Loss of Coolant Accident A SBLOCA is described in UFSAR Section 15.3.1. The AFW system plays a minor role in response to an SBLOCA. This is due to the primary means of RCS heat removal
occurring through the break. However, once the SG secondary sides are isolated, the
boiling process essentially ceases unless the safety valves are lifting. This means, excluding the safety valves, the only heat removal mechanism through the SGs will be
through sensible heat gain of the AFW mass addition. Therefore, the significance of
AFW flow with respect to the SBLOCA transient is considered small. For this transient, only 390 gpm of AFW flow is assumed to be provided by one motor-driven AFW pump
divided equally to four SGs (97.5 gpm per SG).
Normally AFW flow provided by one motor-driven AFW pump would be asymmetrically split between two SGs; however, since the NOTRUMP model used to analyze a SBLOCA event cannot explicitly model DCPP UNITS 1 &
2 FSAR UPDATE 6.5-18 Revision 23 December 2016 this condition, a lumped loop model was used.
Refer to Section 15.3.1 for a discussion on the SBLOCA event.
Natural Circulation Cooldown Maximum and minimum flow requirements from the previously discussed transients
meet the flow requirements of plant cooldown. This operation, however, defines the
basis for tank size, based on the required cooldown duration, maximum decay heat
input, and maximum stored heat in the system. The AFW system partially cools the
RCS to the point where the RHR system may complete the cooldown . Table 6.5-2
shows the assumptions used to determine th e cooldown heat capacity of the AFW system.
The minimum CST volume alone is sufficient to perform the plant cooldown described in
Section 6.5.2.2.5 and to address a NRC Generic Letter 81-21, May 1981, postulated
worst-case natural circulation cooldown.
Due to Unit 2 being converted to a T cold reactor vessel head design, the natural circulation cooldown rates, and subsequent water volume requirement, between the two
Units is different. With the Unit 2 T cold upper head design, a cooldown rate of up to 50°F per hour can be used. With Unit 1 being a T hot upper head design, a reduced cooldown rate of 25°F per hour is required to maintain sub-cooling in the reactor vessel upper
head region. The natural circulation cooldown analysis has shown that with the two
reactor vessel head designs, the worst case conditions for each Unit occurred when Unit 2 was held in hot standby for two hours followed by a four hour cooldown at a rate
of 50°F per hour. The worst-case natural circulation cooldown for Unit 1 was determined to be when Unit 1 was held in hot standby for one hour followed by an eight hour cooldown at 25 °F per hour.
For a worst-case natural circulation cooldown, 196,881 gallons for Unit 1 and 163,058
gallons for Unit 2 are required to cooldown the RCS to 350°F (Mode 4, RHR entry
temperature conditions). An additional volume of 3,119 gallons for Unit 1 and 2,942 gallons for Unit 2 are reserved for allow ed leakage through internal plenums at CST connections and for margin. The inventory of the CST at the minimum Technical
Specification usable volumes of 200,000 gallons for Unit 1 and 166,000 gallons for Unit
2 envelops this total required amount. The usable reserved inventory in both CSTs was
increased from 164,678 gallons to 224,860 gallons (Reference 3). In addition to cooling
the RCS down to RHR entry conditions, the AFW system allows the plant to remain in
hot standby (Mode 3) for an extended period of time if desired by operation and/or
required for safe shutdown. Holding Unit 1 in hot standby for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> requires 140,584
gallons. Holding Unit 2 in hot sta ndby for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> requires 140,703 gallons.
The Technical Specifications do not permit the RCS to be heated above 350°F without
at least 200,000 gallons for Unit 1 and 166,000 gal lons for Unit 2 of usable water in the CST.
DCPP UNITS 1 &
2 FSAR UPDATE 6.5-19 Revision 23 December 2016 Therefore, the volume of water maintained in the CST, as ensured by the Technical Specifications, is sufficient for all Design Bas is Accidents for which the AFW system is required to perform its design function.
In case of an incident, such as a small break in the reactor coolant loop concurrent with a loss of offsite and main generator power, the plant can remain at the hot standby
condition for a period of time that depends on the amount of water available in the CST.
Cooldown delay guidance is found in the plant operating procedures, which, under
certain conditions (e.g., control room inaccessibility), allow a delay in beginning
cooldown. Backup sources of AFW are available as described in Section 6.5.2.1.1.
Table 6.5-1 summarizes the criteria used for the AFW system general design bases for
various plant conditions.
Steam Binding of the AFW Pumps In accordance with GL 1988-03, February 1988, the AFW system is equipped with permanently installed temperature sensors on the AFW discharge lines to each steam
generator main feedwater line. The temperature sensors detect high temperatures
caused by main feedwater system back-leakage through the AFW discharge check
valves. This hot water back-leakage could result in steam binding of the AFW pumps.
Annunciation is provided in the m ain control room if the AFW pipe fluid temperatures reach 200 °F. Operations responds to this annunciation by venting the AFW pump casings until the temperatures are reduced. Also, the AFW pumps may be run to feed
forward to displace the hot water in the pipes with cold CST water and attempt to
reduce check valve leakage by cycling the valve.
6.5.3.8 General Design Criterion 38, 1967 - Reliability and Testability of Engineered Safety Features The AFW system initiation signals and circuitry are testable. Such testability is included in the surveillance test procedures for the plant as delineated in the Technical Specifications.
The AFW system piping also has a periodic inservice inspection (ISI) program in
accordance with the ASME B&PV Code,Section XI (refer to Section 5.2.3.15).
6.5.3.9 General Design Criterion 40, 1967 -
Missile Protection (Dynamic Effects)
The AFW system is protected from missiles, pipe whip, or jet impingement from the
rupture of any nearby high-energy line (refer to Sections 3.5 and 3.6).
The AFW system is protected by barriers and restraints from the dynamic effects of a ruptured pipe outside the containment.
DCPP UNITS 1 &
2 FSAR UPDATE 6.5-20 Revision 23 December 2016 An analysis, using the methodology presented in Reference 1, shows that cooldown using the AFW system will not be prevented by any postulated auxiliary steam line break within an AFW system compartment.
Rupture of a Steam Supply Line to the Turbine-Driven AFW Pump The double-ended rupture of a turbine-driven AFW pump steam supply line in the GE/GW area (downstream of the non-return valves associated with steam supply isolation valves, FCV-37/38, and upstream of steam supply stop valve, FCV-95) will not
result in loss of all AFW flow.
The postulated break would render the turbine driven AFW pump No. 1 inoperable due to loss of steam supply. FCV-37 and FCV-38 are capable of remote manual closure.
FCV-37 is located outside of the GE/GW area and would not be subjected to a harsh
environment. FCV-38 is located within the GE/GW area and is qualified for operation in a harsh environment. Therefore, the break can be isolated from the main steamline.
The resulting increased temperature in the GE/GW area would cause the E/H actuated
level control valves (LCVs) associated with motor driven AFW pump No. 3 to fail due to
a harsh environment. However, due to the small rupture, motor-driven AFW pump No.
3 will not trip due to pump runout and subsequent overcurrent.
The E/H actuated LCVs associated with motor-driven AFW pump No. 2 are located
outside of the GE/GW area and would not be subjected to a harsh environment.
If either of the steam supply valves (FCV-37/38) fails to close, isolation of feedwater to the respective SG will effectively isolate the AFW steam supply line break, thus ensuring at least one motor-driven AFW pump is available to provide sufficient AFW
flow to two intact loops.
Isolation of the faulted SG within 10 minutes ensures the GE/GW area does not exceed
its analyzed temperature profile.
6.5.3.10 General Design Criterion 41, 1967 - Engineered Safety Features Performance Capability The single failure of any active component in the AFW system will not prevent the system from performing its design function (refer to Sections 6.5.3.6 and 6.5.3.7).
6.5.3.11 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment The AFW system isolation valves required for containment closure are periodically tested as part of the MOV program for operability in accordance with GDC 54, 1971 (refer to Section 6.5.3.2.2); however, since t he AFW piping must remain in service following an accident, these valves are not tested for leakage (refer to Section 6.2.4 for DCPP UNITS 1 &
2 FSAR UPDATE 6.5-21 Revision 23 December 2016 leakage testing of the containment isolation system). Leakage detection is provided in the AFW discharge lines that sense back leakage of main feedwater as discussed in
Section 6.5.3.4 and 6.5.3.7.
6.5.3.12 General Design Criterion 57, 1971 - Closed System Isolation Valves The AFW system containment penetrations include the AFW pump discharge lines to the main feedwater lines (Penetrations 1, 2, 3, and 4) and the steam supply lines to the
turbine-driven AFW pump (Penetrations 6 and 7).
These penetrations are classified as Group D containment isolation because they are
lines that must remain in service following an accident and, therefore, do not isolate
automatically.
Refer to Section 6.2.4.1 and Table 6.2-39 for additional information on these
penetrations and their containme nt isolation capabilities.
6.5.3.13 Auxiliary Feedwater System Safety Function Requirements (1) Decay Heat Removal Auxiliary feed pumps are provided and designed to ensure complete reactor decay heat removal under plant transient and accident conditions, including loss of power and loss of the normal heat sink (the condenser circulating water), while maintaining minimum
water levels within the steam generator. The AFW system may be used for plant startup and for a controlled shutdown.
6.5.3.14 10 CFR 50.49 - Enviro nmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants AFW system 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. 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 valves, switches, and flow
transmitters and are listed on the EQ Master List.
6.5.3.15 10 CFR 50.55a(f) - Inservice Testing Requirements All AFW pumps and their appropriate piping and valves are PG&E Design Class I (refer to Section 3.2) and PG&E Quality Class II or III. The IST requirements for these
components are contained in the IST Program Plan.
DCPP UNITS 1 &
2 FSAR UPDATE 6.5-22 Revision 23 December 2016 6.5.3.16 10 CFR 50.55a(g) - Inservice Inspection Requirements The AFW system piping also has a periodic inservice inspection (ISI) program in
accordance with the ASME B&PV Code,Section XI.
6.5.3.17 10 CFR 50.62 - Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants The AFW system is automatically initiated as part of the ATWS Mitigation System Actuation Circuitry (refer to Section 7.6.2.3).
6.5.3.18 10 CFR 50.63 - Loss of All Alternating Current Power In the event of a loss of all alternating current power, decay heat is removed from the core by natural circulation of the reactor coolant. This heat is then transferred to the secondary side of the steam generators and discharged to the atmosphere through the
10% atmospheric dump valves.
Makeup feedwater to the steam generators is provided by the motor-driven AFW pump
on Bus F (refer to Section 6.5.3.21). If Bus F or Bus H is the bus being used, then, by procedure, the respective motor-driven AFW pump is preferred over the turbine-driven
AFW pump.
The AFW system may be used to cool the plant down to hot standby conditions (Mode 3). 6.5.3.19 10 CFR Part 50 Appendix R (Sections III.G, J, and L) - Fire Protection Program for Nuclear Power Facilities Op erating Prior to January 1, 1979 Section III.G - Fire Protection of Safe Shutd own Capability: A fire protection review of the AFW system electrical cable and control wiring has shown that no single postulated fire can prevent the AFW system from perform its design function to bring the plant to
cold shutdown. This is due to physical separation of redundant electrical buses, combined with the ability to control the AFW system from either the main control board, the HSP, 4.16-kV switchgear, or locally at the valves.
Tables 9.5G-1 and 9.5G-2 for DCPP Units 1 and 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.
Specifically, 1 of 3 AFW pumps, the turbine-driven AFW pump steam isolation valves (when the turbine-driven AFW pump is the
one pump relied upon), the steam generator AFW supply level control valves (only the
specific valves associated with each pump), and the water supply (CST or raw water
storage reservoir) with associated valving are the minimum required equipment to bring
the plant to a cold shutdown condition.
DCPP UNITS 1 &
2 FSAR UPDATE 6.5-23 Revision 23 December 2016 Section III.J - Emergency Lighting: Emerge ncy lighting or BOLs are provided in areas where operation of the AFW system may be required to safely shutdown the Unit
following a fire as defined by 10 CFR Part 50, Appendix R, Section III.J.
Section III.L - Alternative and Dedicated Shut down Capability: Safe shutdown capabilities are provided in the control room and at an alternate location via the HSP (refer to Section 7.4) as defined by 10 CFR Part 50, Appendix R, Section III.L. The
ability to safely shut down the plant following a fire in any fire area is summarized in
Section 4.0 of Appendix 9.5A.
6.5.3.20 Regulatory Guide 1.97, Revision 3 - Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident AFW system post-accident instrumentation for meeting Regulatory Guide 1.97, Revision
3, requirements consist of flow indication to al l four (4) steam generators and CST level indication.
One AFW flow indicator is provided for each of four steam generators. The indicators
are PG&E Design Class I (refer to Section 7.5.2.6 and Table 7.5-6). Indication is provided at the main control board and the HSP.
Two separate critical instrument power buses are used for the four flow indicators, with two flow indicators on each bus. The flow from the turbine-driven AFW pump is
monitored by the same indicators that monitor the motor-driven AFW pump flow.
6.5.3.21 NUREG-0737 (Items II.E.1.1 and I I.E.1.2), November 1980 - Clarification of TMI Action Plan Requirements Item II.E.1.1 - Auxiliary Feedwater System Evaluation: The AFW pumps takes water from the CST, which is the preferred source of AFW. The CST provides redundant level indication and low level alarms in the control room. The purpose of the low-low level alarm is to notify the operator that the AFW supply is running low and must be aligned to an alternate water supply.
The alarm provides the operator with at least a 20 minute supply of water for the auxiliary feed pumps at a net flowrate of 880 gpm. The turbine-driven AFW pump has a net flow of 780 gpm available to supply the steam generators. One normally open, manual valve in the common suction piping of the AFW pumps is secured in the open position to prevent interruption of AFW flow. The train of AFW provided by the turbine-driven AFW pump is designed to be completely independent from a standby and preferred ac power source. This train consists of a steam supply stop valve powered from a Class 1E 125-Vdc bus, automatic AFW system actuation instrumentation powered from a vital instrument ac bus (powered by station batteries through an inverter), and steam generator level and AFW flow indication instrumentation powered from station batteries. Driving steam for the turbine-driven AFW pump is taken from two of the four main steam lines upstream of the main steam isolation valves and is exhausted to the atmosphere. Only one steam supply is required for turbine operation. However, steam must always be available from both steam lines during plant operation to preclude a loss of all steam supplies due to any single failure incident. Therefore, in accordance with Item II.E.1.1, the AFW DCPP UNITS 1 &
2 FSAR UPDATE 6.5-24 Revision 23 December 2016 system is designed with one train of AFW that contains a pump power and control source not dependent on ac power and that can provide sufficient AFW flow to the steam generators to bring the plant to hot shutdown conditions.
The motor-driven AFW pumps are powered from the Class 1E buses. They are available for standby service when there is insufficient steam to operate the turbine-
driven AFW pump, or when the turbine-driven AFW pump is unavailable (refer to
Section 8.3.1.1.3). Each motor-driven AFW pump can deliver a net flow of 390 gpm to
two steam generators.
Note that the flow rates of 780 and 390 gpm represent the minimum required flow rates
of the turbine- and motor-driven AFW pumps at a steam generator back-pressure
corresponding to the lowest steam generator safety valve set pressure, plus 3 percent
for setpoint tolerance and 5 psi for accumulation.
Item II.E.1.2 - Auxiliary Feedwater System Initiation and Flows: The AFW system level control valves are normally open and require no actions for system operation. The AFW
initiation circuitry is part of the ESF, and as such, is installed in accordance with IEEE
Standard 279-1971 (refer to Section 7.3) and meets the requirements of Item II.E.1.2.
6.5.3.22 Generic Letter 89-10, June 1989 - Safety Related Motor-Operated Valve Testing and Surveillance The AFW system 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.
6.5.4 TESTS AND INSPECTIONS Refer to Sections 6.5.3.8 and 6.5.3.16 6.5.5 INSTRUMENTATION REQUIREMENTS Refer to Sections 6.5.2.1.2, 6.5.3.4, 6.5.3.5, and 6.5.3.20.
6.
5.6 REFERENCES
- 1. IEEE 279-1971, Criteria for Protection Systems for Nuclear Power Generating Stations.
- 2. Technical Specifications, Diablo Canyon Power Plant Units 1 and 2, Appendix A to License Nos. DPR-80 and DPR-82, as amended.
2 FSAR UPDATE 6.5-25 Revision 23 December 2016
- 4. PG&E Letter to the NRC, Review of Systems and Equipment Necessary to Accomplish a Safe Shutdown Following a Major Earthquake, dated January 26, 1978. 6.5.7 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 UNIT 1 & 2 FSAR UPDATE TABLE 6.1-1 Sheet 1 of 12 Revision 22 May 2015 APPLICABLE DESIGN BASIS CRITERIA CRITERIA TITLE APPLICABILITY Engineered Safety Features Containment Functional Design Containment Heat Removal Systems Containment Air Purification and Cleanup Systems Containment Isolation System Combustible Gas Control in Containment Emergency Core Cooling System Control Room Habitability System Technical Support Center Habitability System Auxiliary Feedwater System Section 6.2.1 6.2.2 6.2.3 6.2.4 6.2.5 6.3 6.4.1 6.4.2 6.5 1. General Design Criteria Criterion 2, 1967 Performance Standards X X X X X X X Criterion 3, 1971 Fire Protection X X X X X Criterion 4, 1967 Sharing of Systems X X X Criterion 4, 1987 Environmental and Dynamic Effects Design Bases X Criterion 10, 1967 Containment X X X Criterion 11, 1967 Control Room X X X X X X X Criterion 12, 1967 Instrumentation and Control System X X X X X X X Criterion 15, 1967 Engineered Safety Features Protection Systems X Criterion 17, 1967 Monitoring Radioactivity Releases X X Criterion 19, 1971 Control Room X X X Criterion 21, 1967 Single Failure Definition X X X X X Criterion 37, 1967 Engineered Safety Features Basis for Design X X X X X X Criterion 38, 1967 Reliability and Testability of Engineered Safety Features X X X X Criterion 40, 1967 Missile Protection X X X X X DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.1-1 Sheet 2 of 12 Revision 22 May 2015 CRITERIA TITLE APPLICABILITY Engineered Safety Features Containment Functional Design Containment Heat Removal Systems Containment Air Purification and Cleanup Systems Containment Isolation System Combustible Gas Control in Containment Emergency Core Cooling System Control Room Habitability System Technical Support Center Habitability System Auxiliary Feedwater System Section 6.2.1 6.2.2 6.2.3 6.2.4 6.2.5 6.3 6.4.1 6.4.2 6.5 1. General Design Criteria (contd.)
Criterion 41, 1967 Engineered Safety Features Performance Capability X X X X X Criterion 42, 1967 Engineered Safety Features Components Capability X X X X Criterion 43, 1967 Accident Aggravation Prevention X Criterion 44, 1967 Emergency Core Cooling Systems Capability X Criterion 45,1967 Inspection of Emergency Core Cooling Systems X Criterion 46, 1967 Testing of Emergency Core Cooling Systems Components X Criterion 47, 1967 Testing of Emergency Core Cooling Systems X Criterion 48, 1967 Testing of Operational Sequence of Emergency Core Cooling Systems X Criterion 49, 1967 Containment Design Basis X X X X Criterion 52, 1967 Containment Heat Removal Systems X
DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.1-1 Sheet 3 of 12 Revision 22 May 2015 CRITERIA TITLE APPLICABILITY Engineered Safety Features Containment Functional Design Containment Heat Removal Systems Containment Air Purification and Cleanup Systems Containment Isolation System Combustible Gas Control in Containment Emergency Core Cooling System Control Room Habitability System Technical Support Center Habitability System Auxiliary Feedwater System Section 6.2.1 6.2.2 6.2.3 6.2.4 6.2.5 6.3 6.4.1 6.4.2 6.5 1. General Design Criteria (contd.)
Criterion 53, 1967 Containment Isolation Valves X Criterion 54, 1967 Containment Leakage Rate Testing X Criterion 54, 1971 Piping Systems Penetrating Containment X X X X X Criterion 55, 1967 Containment Periodic Leakage Rate Testing X Criterion 55, 1971 Reactor Coolant Pressure Boundary Penetrating Containment X X Criterion 56, 1971 Primary Containment Isolation X X X X Criterion 57, 1971 Closed System Isolation Valves X X Criterion 58, 1967 Inspection of Containment Pressure-Reducing Systems X Criterion 59, 1967 Testing of Containment Pressure-Reducing Systems X DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.1-1 Sheet 4 of 12 Revision 22 May 2015 CRITERIA TITLE APPLICABILITY Engineered Safety Features Containment Functional Design Containment Heat Removal Systems Containment Air Purification and Cleanup Systems Containment Isolation System Combustible Gas Control in Containment Emergency Core Cooling System Control Room Habitability System Technical Support Center Habitability System Auxiliary Feedwater System Section 6.2.1 6.2.2 6.2.3 6.2.4 6.2.5 6.3 6.4.1 6.4.2 6.5 1. General Design Criteria (contd.)
Criterion 60, 1967 Testing of Containment Spray Systems X Criterion 61, 1967 Testing of Operational Sequence of Containment Pressure-Reducing Systems Components X Criterion 62, 1967 Inspection of Air Cleanup Systems X X Criterion 63, 1967 Testing of Air Cleanup Systems Components X X Criterion 64, 1967 Testing of Air Cleanup Systems X X Criterion 65, 1967 Testing of Operational Sequence of Air Cleanup Systems X X Criterion 70, 1967 Control of Releases of Radioactivity to the Environment X X 2. 10 CFR Part 20 Part 20 Standards for Protection Against Radiation X
DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.1-1 Sheet 5 of 12 Revision 22 May 2015 CRITERIA TITLE APPLICABILITY Engineered Safety Features Containment Functional Design Containment Heat Removal Systems Containment Air Purification and Cleanup Systems Containment Isolation System Combustible Gas Control in Containment Emergency Core Cooling System Control Room Habitability System Technical Support Center Habitability System Auxiliary Feedwater System Section 6.2.1 6.2.2 6.2.3 6.2.4 6.2.5 6.3 6.4.1 6.4.2 6.5 3. 10 CFR Part 50 50.44 Combustible Gas Control for Nuclear Power Reactors X 50.46 Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Plants X 50.47 Emergency Plans X 50.49 Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants X X X X 50.55a(f)
Inservice Testing Requirements X X X X X 50.55a(g)
Inservice Inspection Requirements X X X X X 50.62 Requirements of Reduction of Risk from Anticipated Transients Without Scram (ATWS)
Events for Light-Water-Cooled Nuclear Power Plants X 50.63 Loss of All Alternating Current Power X DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.1-1 Sheet 6 of 12 Revision 22 May 2015 CRITERIA TITLE APPLICABILITY Engineered Safety Features Containment Functional Design Containment Heat Removal Systems Containment Air Purification and Cleanup Systems Containment Isolation System Combustible Gas Control in Containment Emergency Core Cooling System Control Room Habitability System Technical Support Center Habitability System Auxiliary Feedwater System Section 6.2.1 6.2.2 6.2.3 6.2.4 6.2.5 6.3 6.4.1 6.4.2 6.5 3. 10 CFR Part 50 (contd.)
Appendix J, Option B Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors X X Appendix K, Part I.A ECCS Evaluation Models, Sources of Heat during the LOCA X Appendix R Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979 X X 4. Atomic Energy Commission (AEC) Safety Guides Safety Guide 1, November 1970 Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps X DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.1-1 Sheet 7 of 12 Revision 22 May 2015 CRITERIA TITLE APPLICABILITY Engineered Safety Features Containment Functional Design Containment Heat Removal Systems Containment Air Purification and Cleanup Systems Containment Isolation System Combustible Gas Control in Containment Emergency Core Cooling System Control Room Habitability System Technical Support Center Habitability System Auxiliary Feedwater System Section 6.2.1 6.2.2 6.2.3 6.2.4 6.2.5 6.3 6.4.1 6.4.2 6.5 5. Regulatory Guides Regulatory Guide 1.7, Revision 2, November 1978 Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident X Regulatory Guide 1.52, Revision 0, June 1973 Design, Testing, and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants X Regulatory Guide 1.79, June 1974 Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors 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 X X X DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.1-1 Sheet 8 of 12 Revision 22 May 2015 CRITERIA TITLE APPLICABILITY Engineered Safety Features Containment Functional Design Containment Heat Removal Systems Containment Air Purification and Cleanup Systems Containment Isolation System Combustible Gas Control in Containment Emergency Core Cooling System Control Room Habitability System Technical Support Center Habitability System Auxiliary Feedwater System Section 6.2.1 6.2.2 6.2.3 6.2.4 6.2.5 6.3 6.4.1 6.4.2 6.5 5. Regulatory Guides (contd.)
Regulatory Guide 1.163, September 1995 Performance-Based Containment Leak-Test Program X X Regulatory Guide 1.197, Revision 0, May 2003 Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors X 6. NRC NUREG NUREG-0737, November 1980 Clarification of TMI Action Plan Requirements X X X X X X X 7. NRC Generic Letters Generic Letter 89-10, June 1989 Safety-Related Motor-Operated Valve Testing and Surveillance X X X X X X Generic Letter 95-07, August 1995 Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves X
DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.1-1 Sheet 9 of 12 Revision 22 May 2015 CRITERIA TITLE APPLICABILITY Engineered Safety Features Containment Functional Design Containment Heat Removal Systems Containment Air Purification and Cleanup Systems Containment Isolation System Combustible Gas Control in Containment Emergency Core Cooling System Control Room Habitability System Technical Support Center Habitability System Auxiliary Feedwater System Section 6.2.1 6.2.2 6.2.3 6.2.4 6.2.5 6.3 6.4.1 6.4.2 6.5 7. NRC Generic Letters (contd.)
Generic Letter 96-06, September 1996 Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions X X Generic Letter 97-04, October 1997 Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps X X Generic Letter 98-04, July 1998 Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System after a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment X
DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.1-1 Sheet 10 of 12 Revision 22 May 2015 CRITERIA TITLE APPLICABILITY Engineered Safety Features Containment Functional Design Containment Heat Removal Systems Containment Air Purification and Cleanup Systems Containment Isolation System Combustible Gas Control in Containment Emergency Core Cooling System Control Room Habitability System Technical Support Center Habitability System Auxiliary Feedwater System Section 6.2.1 6.2.2 6.2.3 6.2.4 6.2.5 6.3 6.4.1 6.4.2 6.5 7. NRC Generic Letters (contd.)
Generic Letter 2003-01, June 2003 Control Room Habitability X Generic Letter 2004-02, September 2004 Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors X Generic Letter 2008-01, January 2008 Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems X DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.1-1 Sheet 11 of 12 Revision 22 May 2015 CRITERIA TITLE APPLICABILITY Engineered Safety Features Containment Functional Design Containment Heat Removal Systems Containment Air Purification and Cleanup Systems Containment Isolation System Combustible Gas Control in Containment Emergency Core Cooling System Control Room Habitability System Technical Support Center Habitability System Auxiliary Feedwater System Section 6.2.1 6.2.2 6.2.3 6.2.4 6.2.5 6.3 6.4.1 6.4.2 6.5 8. Bulletins IE Bulletin 79-06A, April 1979 Review of Operational Errors and System Misalignments Identified During the Three Mile Island Incident X IE Bulletin 80-18, July, 1980 Maintenance of Adequate Minimum Flow Thru Centrifugal Charging Pumps Following Secondary Side High Energy Line Rupture X NRC Bulletin 88-04, May 1988 Potential Safety-Related Pump Loss X NRC Bulletin 88-08, June 1988 Thermal Stresses in Piping Connected to Reactor Coolant Systems X NRC Bulletin 2003-01, June 2003 Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors X DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.1-1 Sheet 12 of 12 Revision 22 May 2015 CRITERIA TITLE APPLICABILITY Engineered Safety Features Containment Functional Design Containment Heat Removal Systems Containment Air Purification and Cleanup Systems Containment Isolation System Combustible Gas Control in Containment Emergency Core Cooling System Control Room Habitability System Technical Support Center Habitability System Auxiliary Feedwater System Section 6.2.1 6.2.2 6.2.3 6.2.4 6.2.5 6.3 6.4.1 6.4.2 6.5 9. Branch Technical Position Branch Technical Position EICSB 18, November 1975 Application of the Single Failure Criterion to Manually-Controlled Electrically-Operatred Valves X
DCPP UNIT 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 6.2-25 CONTAINMENT HEAT REMOVAL SYSTEMS DESIGN CODE REQUIREMENTS Component Code Valves ANSI B16.5-1968
Piping (including headers and spray nozzles)
- PG&E Design Class I portions ANSI B31.7-1969 with 1970 Addenda
- PG&E Design Class II portions ANSI B31.1-1967
Containment Spray Pump ASME P&V III-1968 (a)
Refueling Water Storage Tank AWWA D100-1967 (b) (a) Draft ASME Code for Pumps and Valves for Nuclear Power, November 1968.
(b) ASME BPVC Section VIII-1974, Allowable Stresses Used for Design
DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.2-26 Sheet 1 of 3 Revision 22 May 2015 CONTAINMENT HEAT REMOVAL SYSTEMS DESIGN PARAMETERS Containment Spray Pump Type Horizontal Centrifugal Number (per unit) 2 Design pressure, psig 275 (c) Design temperature, °F 275 (c) Design flowrate, gpm 2600 Design head, ft 450 Material Type 316 stainless steel
Containment Spray Nozzle
Number (per unit) 343 (Unit 1)/
342 (Unit 2)
Type Spraco 1713A Flow per nozzle at 40 psi p, gpm 15.2 Material Type 304 stainless steel
Refueling Water Storage Tank
Number (per unit) 1 Total available tank volume (includes only
usable volume)
(a), gal 450,000 Minimum Technical Specifications required volume (includes usable and unusable volume), gal 455,300 Accident analysis volume (assumed), gal 350,000 Boron concentration, ppm 2300-2500 Design temperature, °F 100 Design pressure, psig Atmospheric Operating pressure, psig Atmospheric Material Austenitic stainless steel with reinforced
concrete shroud
Containment Fan Coolers
Number (per unit) 5 Fan type Centrifugal Bearing monitors Vibration
DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.2-26 Sheet 2 of 3 Revision 22 May 2015 Containment Fan Coolers (Continued)
Normal Mode Operation (b) Accident Mode Operation (b) Speed, rpm 1,200 600 Capacity, cfm 110,000 47,000 Static pressure at 0.075 lb/ft 3, in. water 7.3 3.75 Containment atmosphere pressure, psig 0 47 Containment atmosphere temperature, °F 120 271 Containment atmosphere density, lb/ft 3 0.0685 0.175 Brake horsepower 275 103 Name plate horsepower 300 100
Component cooling water flow to motor heat exchanger, gpm
50 30 (min)
Motor Assembly Number (per unit) 5 Type 460 V, 3 phase, 60 Hz, two speed, single winding Bearing monitor Vibration and temperature Winding monitor RTDs Service factor 1 Heat exchange cooling media Component cooling water
Cooling Coil Assembly Number (per unit) 5
Type Plate-finned Tube material Copper Fin material Copper Fins per inch 8.5 Tube thickness, in. 0.035 Fin thickness, in. 0.008 Tube normal OD, in. 0.625 Tube length, in. 114 Vertical drain pan spacing, ft 3.25 Pan drain diameter, in. 2 Assembly drain diameter, in. 8 Drain pipe, Sch 10 Assembly frame material Steel Drain pan material Steel
DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.2-26 Sheet 3 of 3 Revision 22 May 2015 Containment Fan Coolers (Continued)(d) Normal Accident Mode
Operation (b) Operation (b) Heat removal minimum, Btu/hr 3.14 x 10 6 81 x 10 6 Steam-air flow, cfm 110,000 47,000 Steam-air inlet temperature, °F 120 271 Steam-air outlet temperature, °F 92.5 269 Total pressure, psig 0 47
Air density, lb/ft 3 0.0685 0.0677 Steam density, lb/ft 3 - 0.1073 Condensation rate, gpm 0 180.7 Air face velocity, fpm 645 275 Static pressure drop (0.075 lb/ft 3), in. water 2.0 0.3 clean, 0.74 dirty Cooling water flow, gpm 2000 2000 Cooling water inlet temperature, °F 90 125 Cooling water outlet temperature, °F 94 212 Pressure drop, ft water 8.8 8.8 Coil tube side foul factor 0.0005 0.0005 Water velocity, fps 3.43 3.43
The moisture separator and HEPA filter were deleted in Revision 9.
(a) Usable volume includes the water above the outlet pipe. Unusable water includes the water below the outlet.
(b) CFCU data are shown for an illustration of performance at typical operating points.
Design minimum CCW flow to the CFCU cooling coils is nominally 1600 gpm for both
normal and limiting accident modes. Design maximum CCW flow to the CFCU cooling
coils is 2500 gpm for accident modes. Temperatures, heat removal rates, and other
parameters will vary dynamically according to the accident conditions.
(c) Specified values for containment spray pump design pressure and temperature are maximum values and do not designate concurrent design conditions for the pump.
(d) Values for CFCUs with Westinghouse Sturtevant coils are historical and for normal mode operation are based off a maximum flow of 110,000 CFM. Values for accident mode operation are based off a flow rate of 47,000 CFM. The operating values depend on the operating point in the acceptable range of airflow rates for both modes. Under normal conditions the range is 68,000-110,000 CFM, under accident conditions the range of airflow rates is 34,000-57,000 CFM. At airflow rates less than 47,000 CFM, the heat removal performance is still acceptable (See Westinghouse Calc. Note CN-CRA-12-5).
DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.2-27 Revision 22 May 2015 SINGLE FAILURE ANALYSIS - CONTAINMENT HEAT REMOVAL SYSTEMS Component Malfunction Comments and Consequences A. Spray Nozzles Clogged Large number of nozzles precludes clogging of a significant number.
B. Pumps Containment spray pump Fails to start Two pumps provided.
Operation of one required. C. Automatically Operated Valves:
(9001A/B): (Open on coincidence of two out of four high-high
containment pressure signals)
Containment spray pump discharge isolation valve Fails to open Two complete systems provided. Operation of one
required.
D. Valves Operated From Control Room for Spray Recirculation, if
Used (9003A/B):
Containment spray header isolation valve from residual
heat exchanger discharge Fails to open Two complete systems provided. Operation of spray
recirculation not required.
E. Containment Fan Coolers Fails to start Five fan coolers provided.
Two required for minimum
safety feature operation.
DCPP UNIT 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 6.2-29 SPRAY ADDITIVE SYSTEM DESIGN PARAMETERS Eductors Quantity 2 Eductor Inlet (motive)
Operating fluid Borated Water Operating temperature Ambient Eductor Suction fluid NaOH concentration, wt% 30 Specific gravity ~ 1.3 Viscosity (design), cp ~ 10 Operating temperature Ambient
Spray Additive Tank
Number 1 Total Volume (nominal), gal.
4000 NaOH Concentration, wt% 30 Design Temperature, °F 300 Internal Design Pressure, psig 14 Operating Temperature, °F 100 Operating Pressure, psig 5 (a) Material Stainless Steel (a) During normal operating and test conditions, the tank is pressurized with nitrogen at 5 psig.
During preoperational testing or following a postulated accident, the nitrogen supply system will
supply sufficient nitrogen to maintain this pressure as the tank empties. In the postulated event
that the nitrogen supply system is inoperative, the tank pressure would fall below atmospheric
pressure as the tank empties. Vacuum breakers are provided for this occurrence.
DCPP UNIT 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 6.2-30 SPRAY ADDITIVE SYSTEM - CODES USED IN SYSTEM DESIGN Spray Additive Tank ASME B&PV,Section VIII Code Class 3 Valves ANSI B16.5 Piping (including headers and spray nozzles)
Design Class I portions ANSI B31.7 Design Class II portions ANSI B31.1
Eductors ASME B&PV,Section III
DCPP UNIT 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 6.2-36 PARAMETERS AND RESULTS FOR SPRAY IODINE REMOVAL ANALYSIS DURING INJECTION PHASE OPERATION (a) Best Minimum Design Parameter Estimate Expected Case Power, MWt 3568 3568 3568 Containment free vol, ft 3 2.6 x 10 6 2.6 x 10 6 2.6 x 10 6 Unsprayed volume, % 17 17 17 No. pumps operating 2 1 1 Spray pump flowrate, gpm 5300 2650 2600 Containment pressure, psig 25 25 47 Containment temperature, °F 233 233 271 Spray fall height, ft 132.75 128 128 Spray solution pH 10.2 (b) 10.2 (b) 9.5 Results Exponential removal constant for the
spray system (hr-1) 92 36 31 (c) (a) Data provided in this table are for spray iodine removal analysis only. A different set of data used in the containment integrity analysis are provided in Appendix 6.2D.
(b) Based on 2000 ppm boron concentration in the RWST. Only the Design Case has been updated for the change to a minimum 2300 ppm boron concentration in the RWST.
(c) Although a subsequent safety evaluation showed that the Design Case coefficient of 31 hr
-1 (for 2600 gpm spray header flow) should be reduced to approximately 29 hr
-1 (for 2466 gpm spray header flow), the potential offsite dose increase due to this change is extremely small and can be considered insignificant (Reference 50).
DCPP UNIT 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 6.2-37 SPRAY FALL HEIGHTS IN THE CONTAINMENT Area (a), ft 2 Drop Fall Height (b), ft 1,075 58 1,400 93 100 105 225 122 9,900 128 150 145 2,550 177
(a) Area represents portion of operating deck covered by spray nozzles located at a particular elevation above the operating deck.
(b) Measured from 268 ft minimum average nozzle elevation. Average spray fall height: 128 ft.
DCPP UNIT 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 6.2-38 SPRAY ADDITIVE SYSTEM SINGLE FAILURE ANALYSIS
Component
Malfunction Comments and Consequences Automatically Operated Valves:
(Open on coincidence of
two-out-four high-high signals)
Spray additive tank outlet isolation valve Fails to open Two parallel provided (8994A/B). Operation of one required.
DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 1 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes 1 Feedwater 57 Yes 1 (II) - - - - - - - - - - - - - - - (NE/SA) A FW-FCV-438 Gte Mtr O C Yes O As is F N C W Hot 1,26 B FW-140 Gte Man O D No O As is - Y O W Hot 7,26
- Closed System Cls - I D - - - - Y - W Hot 2 2 Feedwater 57 Yes 1 (II) - - - - - - - - - - - - - - - (NE/SA) A FW-FCV-439 Gte Mtr O C Yes O As is F N C W Hot 1,26 B FW-147 Gte Man O D No O As is - Y O W Hot 7,26
- Closed System Cls - I D - - - - Y - W Hot 2 3 Feedwater 57 Yes 1 (II) - - - - - - - - - - - - - - - (NE/SA) A FW-FCV-440 Gte Mtr O C Yes O As is F N C W Hot 1,26 B FW-153 Gte Man O D No O As is - Y O W Hot 7,26
- Closed System Cls - I D - - - - Y - W Hot 2 4 Feedwater 57 Yes 1 (II) - - - - - - - - - - - - - - - (NE/SA) A FW-FCV-441 Gte Mtr O C Yes O As is F N C W Hot 1,26 B FW-157 Gte Man O D No O As is - Y O W Hot 7,26
- Closed System Cls - I D - - - - Y - W Hot 2 5 Main steam 57 Yes 1 (I) - - - - - - - - - - - - - - - (NE/SA) E MS-FCV Air O C Yes O As is M N C G Hot 3,4,26 F MS-FCV-25 Glb Air O C Yes C Closed M N C G Hot 4,26 G MS-PCV-19 Glb Air O D Yes C Closed - N C G Hot 5,26 H MS-RV-3, 4, 5, 6, 222 Rlf Spr O D No C - - N C G Hot 6,26 - Closed System Cls - I D - - - - Y - G Hot 2 6 Main steam 57 Yes 2 (I) - - - - - - - - - - - - - - - (NE/SA) A MS-FCV Air O C Yes O As is M N C G Hot 3,4,26 B MS-FCV-24 Glb Air O C Yes C Closed M N C G Hot 4,26 C MS-PCV-20 Glb Air O D Yes C Closed - N C G Hot 5,26 D MS-RV-7, 8, 9, 10, 223 Rlf Spr O D No C - - N C G Hot 6,26 E MS-FCV-37 Gte Mtr O D Yes O As is R-M Y O G Hot 26
- Closed System Cls - I D - - - - Y - G Hot 2 7 Main steam 57 Yes 2 (I) - - - - - - - - - - - - - - - (NE/SA) A MS-FCV Air O C Yes O As is M N C G Hot 3,4,26 B MS-FCV-23 Glb Air O C Yes C Closed M N C G Hot 4,26 C MS-PCV-21 Glb Air O D Yes C Closed - N C G Hot 5,26 D MS-RV-11,12,13,14,224 Rlf Spr O D No C - - N C G Hot 6,26 DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 2 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes E MS-FCV-38 Gte Mtr O D Yes O As is R-M Y O G Hot 26 - Closed System Cls - I D - - - - Y - G Hot 2 8 Main steam 57 Yes 1 (I) - - - - - - - - - - - - - - - (NE/SA) E MS-FCV Air O C Yes O As is M N C G Hot 3,4,26 F MS-FCV-22 Glb Air O C Yes C Closed M N C G Hot 4,26 G MS-PCV-22 Glb Air O D Yes C Closed - N C G Hot 5,26 H MS-RV-58,59,60,61,225 Rlf Spr O D No C - - N C G Hot 6,26 - Closed System Cls - I D - - - - N - G Hot 2 9 Compone nt 57 Yes 3 (V) - - - - - - - - - - - - - - 7 Cooling Water to Fan Coolers (SA) H CCW-169 But Man O D No O As is - Y O W Cold 7,26 - Closed System Cls - I D - - - - Y - W Cold 47 10 Compone nt 57 Yes 3 (V) - - - - - - - - - - - - - - 7 Cooling Water to Fan Coolers (SA) H CCW-177 But Man O D No O As is - Y O W Cold 7,26 - Closed System Cls - I D - - - - Y - W Cold 47 11 Compone nt 57 Yes 3 (V) - - - - - - - - - - - - - - 7 Cooling Water to Fan Coolers (SA) H CCW-469 But Man O D No O As is - Y O W Cold 7,26 - Closed System Cls - I D - - - - Y - W Cold 47 12 Compone nt 57 Yes 3 (V) - - - - - - - - - - - - - - 7 Cooling Water to Fan Coolers (SA) H CCW-477 But Man O D No O As is - Y O W Cold 7,26 DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 3 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes - Closed System Cls - I D - - - - Y - W Cold 47 13 Compone nt 57 Yes 3 (V) - - - - - - - - - - - - - - 7 Cooling Water to Fan Coolers (SA) H CCW-185 But Man O D No O As is - Y O W Cold 7,26 - Closed System Cls - I D - - - - Y - W Cold 47 14 Compone nt 57 Yes 3 (VI) - - - - - - - - - - - - - - 7 Cooling Water from Fan Coolers (SA) I CCW-176 But Man O D No O As is - Y O W Cold 7,26 - Closed System Cls - I D - - - - Y - W Cold 47 15 Compone nt 57 Yes 3 (VI) - - - - - - - - - - - - - - 7 Cooling Water from Fan Coolers (SA) I CCW-184 But Man O D No O As is - Y O W Cold 7,26 - Closed System Cls - I D - - - - Y - W Cold 47 16 Compone nt 57 Yes 3 (VI) - - - - - - - - - - - - - - 7 Cooling Water from Fan Coolers (SA) I CCW-476 But Man O D No O As is - Y O W Cold 7,26 - Closed System Cls - I D - - - - Y - W Cold 47 17 Compone nt 57 Yes 3 (VI) - - - - - - - - - - - - - - 7 Cooling Water from Fan Cooler s I CCW-484 But Man O D No O As is - Y O W Cold 7,26 DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 4 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes (SA) - Closed System Cls - I D - - - - Y - W Cold 47 18 Compone nt 57 Yes 3 (VI) - - - - - - - - - - - - - - 7 Cooling Water from Fan Coolers (SA) I CCW-192 But Man O D No O As is - Y O W Cold 7,26 - Closed System Cls - I D - - - - Y - W Cold 47 19 Compone nt 55 Yes 4 (I) - - - - - - - - - - - - - - - Cooling Water to A CCW-FCV-356 But Mtr O A Yes O As is P N C W Cold -
Reactor Coolant B CCW-585 Chk - I A No O - - N C W Cold -
Pumps (ES) 20 Compone nt 55 Yes 4 (III) - - - - - - - - - - - - - - - Cooling Water F CCW-FCV-363 But Mtr O A Yes O As is P N C W Cold - from Reactor E CCW-FCV-749 But Mtr I A Yes O As is P N C W Cold - Coolant Pumps H CCW-581 Chk - I A No O - - N C W Cold - (ES) 21 Compone nt 55 Yes 4 (II) - - - - - - - - - - - - - - - Cooling Water C CCW-FCV-750 Glb Mtr I A Yes O As is P N C W Hot - from Reactor D CCW-FCV-357 Glb Mtr O A Yes O As is P N C W Hot - Coolant Pumps G CCW-670 Chk - I A No O - - N C W Hot - (ES) 22 CCW to Excess 57 No 4 (IV) - - - - - - - - - - - - - - 8 Letdown I CCW-695 Chk - O C No C - - N C W Cold 8 DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 5 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes Heat Exchanger (NE) - Closed System Cls - I C - - - - N - W Cold - 23 CCW from 57 Yes 4 (V) - - - - - - - - - - - - - - - Excess J CCW-FCV-361 But Air O C Yes C Closed T N C W Cold - Letdown Heat Exchanger (NE) - Closed System Cls - I C - - - - N - W Cold - 24 Residual Heat 55 Yes 5 (I) - - - - - - - - - - - - - - - Removal No. 1 A SI-8818A Chk - I B No C - - Y C W Hot 26 Cold Leg B SI-8818B Chk - I B No C - - Y C W Hot 26 Injection (SA) C SI-8809A Gte Mtr O B Yes O As is R-M Y C W Hot 7,9,26,40 M SI-8885A (U1) Glb Air I E - C Closed - N C W Hot 22,26, 41 Closed System Cls - O B - - - - Y - W Hot 10 25 Residual Heat 55 Yes 5 (III) - - - - - - - - - - - - - - - Removal No. 2 I SI-8818C Chk - I B No C - - Y C W Hot 26 Cold Leg J SI-8818D Chk - I B No C - - Y C W Hot 26 Injection (SA) K SI-8885B (U1) Glb Air I E - C Closed - N C W Hot 22,26, 41 L SI-8809B Gte Mtr O B Yes O As is R-M Y C W Hot 7,9,26,40 - Closed System Cls - O B - - - - Y - W Hot 10 26 Residual Heat 55 Yes 5 (II) - - - - - - - - - - - - - - - Removal Hot D RHR-8716A Gte Mtr O B Yes O As is R-M Y C W Hot 7,9,26,40 Leg Injection (SA) G RHR-8716B Gte Mtr O B Yes O As is R-M Y O W Hot 7,9,26,40 H RHR-8703 Gte Mtr I B Yes C As is R-M Y O W Hot 7,11, 26 - Closed System Cls - O B - - - - Y - W Hot 10 DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 6 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes 27 Reactor Coolant 55 Yes 6 (I) - - - - - - - - - - - - - - - System Loop 4 A RHR-8701 Gte Mtr I D Yes C As is R-M N C W Hot 7,11, 26 Recirculation (SA) H RHR-1012 Glb Man O D No C - - N C W Hot 11,26,4 5 I RHR-1014 Glb Man O D No C - - N C W Hot 11,26,4 5 B SI-8980 Gte Mtr O D Yes O As is R-M Y C W Hot 11,7, 26 C RHR-8700A Gte Mtr O D Yes O As is R-M Y C W Hot 11,7, 26 D RHR-8700B Gte Mtr O D Yes O As is R-M Y C W Hot 11,7, 26 28 Containm ent 56 Yes 6 (II) - - - - - - - - - - - - - - - Sump E SI-8982A Gte Mtr O D Yes C As is - Y O W Hot 12,26 Recirculation (SA) - Closed System Cls - O D - - - - Y - W Hot 10 29 Containm ent 56 Yes 6 (III) - - - - - - - - - - - - - - - Sump F SI-8982B Gte Mtr O D Yes C As is - Y O W Hot 12,26 Recirculation (SA) - Closed System Cls - O D - - - - Y - W Hot 10 30 Containm ent 56 Yes 7 (I) - - - - - - - - - - - - - - - Spray System A CS-9011B Chk - I D No C - - Y C W Cold - (SA) B CS-9001B Gte Mtr O D Yes C As is R-M Y C W Cold 11,13 C CS-32 Glb Man O E No C - - N C W Cold D CS-9003B Gte Mtr O D Yes C As is R-M Y C W Cold 7,11, 26,48 31 Containm ent 56 Yes 7 (II) - - - - - - - - - - - - - - - Spray System E CS-9011A Chk - I D No C - - Y C W Cold - (SA) F CS-9003A Gte Mtr O D Yes C As is R-M Y C W Cold 7,11, 26,48 G CS-9001A Gte Mtr O D Yes C As is R-M Y C W Cold 11,13 H CS-31 Glb Man O E No C - - N C W Cold
DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 7 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes 32 Spare - - - - - - - - - - - - - - - - - 26 33 Safety Injection 55 Yes 8 (I) - - - - - - - - - - - - - - - System (SA) B SI-8835 Gte Mtr O D Yes O As is R-M Y C W Cold 7,11, 26 G SI-8823 (U1) Glb Air I E - C Closed - N C W Cold 22,26,4 1 34 Safety Injection 55 Yes 8 (II) - - - - - - - - - - - - - - - System (SA) D SI-8801A, 8801B Gte Mtr O D Yes C As is S Y O W Cold 11,14,2 6 I SI-8969 Glb Man O E No C - - N C W Cold 26,48 35 Regenerat ive 55 Yes 9 (I) - - - - - - - - - - - - - - -
Heat Exchanger A CVCS-8149A Glb Air I A Yes O Closed T N C W Hot -
to Letdown Heat B CVCS-8149B Glb Air I A Yes O Closed T N C W Hot - Exchanger (NE) C CVCS-8149C Glb Air I A Yes O Closed T N C W Hot - D CVCS-8152 Glb Air O A Yes O Closed T N C W Hot - 36 Normal Charging 55 Yes 9 (II) - - - - - - - - - - - - - - 34 to Regenerat ive E CVCS-8378C Chk - I A No O - - N C W Cold 26 Heat Exchanger F CVCS-8107 Gte Mtr O A Yes O As is S N C W Cold 26 (SA) 37 Steam Generator 57 Yes 1 (III) - - - - - - - - - - - - - - - Blowdown (NE) J MS-FCV-151 Glb Air O C Yes O Closed T N C W Hot 26,55 I MS-FCV-760 Glb Air I C Yes O Closed M N C W Hot 2,4, 26,48 - Closed System Cls - I C - - - - - - W Hot 2 38 Steam 57 Yes 1 (III) - - - - - - - - - - - - - - -
DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 8 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes Generator Blowdown (NE) J MS-FCV-154 Glb Air O C Yes O Closed T N C W Hot 26,55 I MS-FCV-761 Glb Air I C Yes O Closed M N C W Hot 2,4, 26,48 - Closed System Cls - I C - - - - - - W Hot 2 39 Steam Generator 57 Yes 1 (III) - - - - - - - - - - - - - - - Blowdown (NE) J MS-FCV-157 Glb Air O C Yes O Closed T N C W Hot 26,55 I MS-FCV-762 Glb Air I C Yes O Closed M N C W Hot 2,4, 26,48 - Closed System Cls - I C - - - - - - W Hot 2 40 Steam Generator 57 Yes 1 (III) - - - - - - - - - - - - - - - Blowdown (NE) J MS-FCV-160 Glb Air O C Yes O Closed T N C W Hot 26,55 I MS-FCV-763 Glb Air I C Yes O Closed M N C W Hot 2,4, 26,48 - Closed System Cls - I C - - - - - - W Hot 2 41 Reactor Coolant 55 Yes 10 (I) - - - - - - - - - - - - - - -
Pump Seal Water A CVCS-8368A Chk - I B No O - - Y O W Cold - Supply (ES) - Closed System Cls - O B - - - - Y - W Cold 10 42 Reactor Coolant 55 Yes 10 (I) - - - - - - - - - - - - - - -
Pump Seal Water A CVCS-8368B Chk - I B No O - - Y O W Cold - Supply (ES) - Closed System Cls - O B - - - - Y - W Cold 10 43 Reactor Coolant 55 Yes 10 (I) - - - - - - - - - - - - - - -
Pump Seal Water A CVCS-8368C Chk - I B No O - - Y O W Cold - Supply - Closed System Cls - O B - - - - Y - W Cold 10 DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 9 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes (ES) 44 Reactor Coolant 55 Yes 10 (I) - - - - - - - - - - - - - - -
Pump Seal Water A CVCS-8368D Chk - I B No O - - Y O W Cold - Supply (ES) - Closed System Cls - O B - - - - Y - W Cold 10 45 Reactor Coolant 55 Yes 10 (II) - - - - - - - - - - - - - - -
Pump Seal Water B CVCS-8112 Gte Mtr I A Yes O As is T N C W Cold - Return (NE) C CVCS-8109 Chk - I A No O - - Y O W Cold - D CVCS-8100 Gte Mtr O A Yes O As is T N C W Cold - 46 Refueling Canal 56 Yes 3 (II) - - - - - - - - - - - - - - -
Recirculat ion C LWS-8796 Dia Man I E No C - - N C W Cold - (NE) D LWS-8787 Dia Man O E No C - - N C W Cold - 47 Refueling Canal 56 Yes 3 (IV) - - - - - - - - - - - - - - - Return (NE) F LWS-8795 Dia Man I E No C - - N C W Cold - G LWS-8767 Dia Man O E No C - - N C W Cold - 48 Spare - - - - - - - - - - - - - - - - - 26 49 Containm ent 56 Y es 3 (I) - - - - - - - - - - - - - - -
Sump Discharge A LWS-FCV-500 Bal Air I A Yes O Closed T N C W Cold 56 (NE) B LWS-FCV-501 Bal Air O A Yes O Closed T N C W Cold - 50 Reactor Coolant 56 Yes 12 (IV) - - - - - - - - - - - - - - - Drain Tank G LWS-FCV-254 Bal Air O A Yes O Closed T N C W Hot - Discharge (NE) H LWS-FCV-253 Bal Air I A Yes O Closed T N C W Hot 56 DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 10 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes 51A Nitrogen Supply 56 Yes 8 (III) - - - - - - - - - - - - - - - Header to E SI-8880 Glb Air O A Yes O Closed T N C G Cold -
Accumulators F SI-8916 Chk - I A No O - - N C G Cold - (NE) 51B Safety Injection 55 Yes 13 (I) - - - - - - - - - - - - - - - A SI-8871 Glb Air I A Yes C Closed T N C W Cold - System B SI-8883 Glb Air O A Yes C Closed T N C W Cold - Test Line (NE) C SI-8961 Glb Air O A Yes C Closed T N C W Cold - D SI-161 Glb Man O E No C - - N C W Cold 48 51C Reactor Coolant 56 Yes 12 (II) - - - - - - - - - - - - - - - Drain Tank C LWS-FCV-256 Bal Air O A Yes O Closed T N C G Cold - Vent (NE) D LWS-FCV-255 Bal Air I A Yes O Closed T N C G Cold - 51D Reactor Coolant 56 Yes 12 (III) - - - - - - - - - - - - - - - Drain Tank to Gas E LWS-FCV-257 Bal Air O A Yes C Closed T N C W Cold - Analyzer (NE) F LWS-FCV-258 Bal Air I A Yes O Closed T N C W Cold - 52A Pressurizer Relief 55 Yes 14 (III) - - - - - - - - - - - - - - - Tank Makeup E R CS-8029 Bal Air O A Yes O Closed T N C W Cold - (NE) F RCS-8046 Chk - I A No O - - N C W Cold - 52B Pressurizer Relief 55 Yes 14 (II) - - - - - - - - - - - - - - - Tank Nitrogen C RCS-8047 Chk - I A No O - - N C G Cold - Supply (NE) D RCS-8045 Dia Air O A Yes O Closed T N C G Cold - 52C Steam Generator 56 Yes 2 (II) - - - - - - - - - - - - - - -
DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 11 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes Nitrogen Supply F MS-902 Gte Man O E No C - - N C G Cold 26 (NE) G MS-5200 Chk - I E No C - - N C G Cold 26 52D Reactor Coolant 56 Yes 12 (I) - - - - - - - - - - - - - - - Drain Tank A LWS-FCV-260 Bal Air O A Yes O Closed T N C G Cold - Nitrogen Supply B LWS-60 Chk - I A No O - - N C G Cold - (NE) 52E Containm ent H 2 56 Yes 24 (I) - - - - - - - - - - - - - - - Monitor Supply A VAC-FCV-235 Glb Sol I E Yes C Closed - Y O G Cold 28,29 (NE) B VAC-FCV-236 Glb Sol O E Yes C Closed - Y O G Cold 28,29 52F Containm ent H 2 56 Yes 24 (II) - - - - - - - - - - - - - - - Monitor Return C VAC-FCV-237 Glb Sol O E Yes C Closed - Y O G Cold 28,29 (NE) D VAC-252 Chk - I E No C - - Y O G Cold 29 52G Containm ent 56 Yes 15 (I) - - - - - - - - - - - - - - - Pressure PT-937 A Sealed Bellows Sbl - I - - - - - - - G Cold 16 (SA) B Sealed Instrument Sin - O - - - - - - - G Cold 16 52H Containm ent 56 Yes 15 (I) - - - - - - - - - - - - - - - Pressure PT-932 A Sealed Bellows Sbl - I - - - - - - - W Cold 16 (SA)(abandon in place)
B Sealed Instrument Sin - O - - - - - - - W Cold 16 53A Steam Generator 1 57 Yes 1 (IV) - - - - - - - - - - - - - - - Blowdow n K MS-FCV-250 Glb Air O C Yes O Closed T N C W Hot 26,55 Sample (NE) I MS-FCV-760 Glb Air I C Yes O Closed M N C W Hot 2,4, 26,48 - Closed System Cls - I C - - - - - - W Hot 2
DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 12 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes 53B Steam Generator 2 57 Yes 1 (IV) - - - - - - - - - - - - - - - Blowdow n K MS-FCV-248 Glb Air O C Yes O Closed T N C W Hot 26,55 Sample (NE) I MS-FCV-761 Glb Air I C Yes O Closed M N C W Hot 2,4, 26,48 - Closed System Cls - I C - - - - - - W Hot 2 53C Steam Generator 3 57 Yes 1 (IV) - - - - - - - - - - - - - - - Blowdow n K MS-FCV-246 Glb Air O C Yes O Closed T N C W Hot 26,55 Sample (NE) I MS-FCV-762 Glb Air I C Yes O Closed M N C W Hot 2,4, 26,48 - Closed System Cls - I C - - - - - - W Hot 2 53D Steam Generator 4 57 Yes 1 (IV) - - - - - - - - - - - - - - - Blowdow n K MS-FCV-244 Glb Air O C Yes O Closed T N C W Hot 26,55 Sample (NE) I MS-FCV-763 Glb Air I C Yes O Closed M N C W Hot 2,4, 26,48 - Closed System Cls - I C - - - - - - W Hot 2 54 Instrument Air 56 Yes 16 (I) - - - - - - - - - - - - - - - Header (NE) A AIR-I-587 Chk - I A No O - - N C G Cold - B FCV-584 Bal Air O A Yes O Closed T N C G Cold - E AIR-I-585 Dia Man O E No C - - N C G Cold 48 55 Spare - - - - - - - - - - - - - - - - 26 56 Service Air 56 Yes 16 (II) - - - - - - - - - - - - - - - Header (NE) C AIR-S-114 Chk - I E No C - - N C G Cold - - D AIR-S-200 Bal Man O E No C - - N C G Cold - 57 Containm ent 56 Yes 24 (III) - - - - - - - - - - - - - - - External H 2 E VAC-FCV-669 Gte Mtr O E Yes C As is - Y C G Cold 28,38 Recombiners (SA) F VAC-FCV-659 Gte Mtr I E Yes C As is - Y C G Cold 28,38 DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 13 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes 58 Mini-Equipment - - - - Not a piping penetration - - - - - - - - - - - - -
Hatch (NE) 59A Pressurizer Liquid 55 Yes 17 (II) - - - - - - - - - - - - - - -
Sample (NE) C NSS-9355A Glb Air I A Yes O Closed T N C W Hot - D NSS-9355B Glb Air O A Yes O Closed T N C W Hot - 59B Hot Leg Sample 55 Yes 17 (III) - - - - - - - - - - - - - - - (NE) E NSS-9356A Glb N 2 I A Yes O Closed T Y C W Hot - - F NSS-9356B Glb N 2 O A Yes O Closed T Y C W Hot - 59C Accumulator 55 Yes 17 (IV) - - - - - - - - - - - - - - -
Sample (NE) G NSS-9357A Glb Air I A Yes C Closed T N C W Cold - H NSS-9357B Glb Air O A Yes C Closed T N C W Cold - 59D Containm ent 56 Yes 15 (I) - - - - - - - - - - - - - - - Pressure Transmitter A Sealed Instrument Sin - I - - - - - - - W Cold 16 PT-938 (SA) B Sealed Bellows Sbl - O - - - - - - - W Cold 16 59E RV Level 55 Yes 25 (I) - - - - - - - - - - - - - - - Instrumen tation A Sealed Bellows Sbl - I - - - - - - - W Cold 16 Transmitt er LIS-1310 (SA) B Hydraulic Isolators Hys - O - - - - - - - W Cold 16 59F RV Level 55 Yes 25 (I) - - - - - - - - - - - - - - - Instrumen tation A Sealed Bellows Sbl - I - - - - - - - W Cold 16 Transmitt er LIS-1311 (SA) B Hydraulic Isolators Hys - O - - - - - - - W Cold 16 DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 14 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes 59G RV Level 55 Yes 25 (I) - - - - - - - - - - - - - - - Instrumen tation A Sealed Bellows Sbl - I - - - - - - - W Cold 16 Transmitt er LIS-1312 (SA) B Hydraulic Isolators Hys - O - - - - - - - W Cold 16 59H Containm ent 56 Yes 15 (I) - - - - - - - - - - - - - - - Pressure Transmitt ers A Sealed Instrument Sin - I - - - - - - - W Cold 16 PT-933 & PT-935 (SA) B Sealed Bellows Sbl - O - - - - - - - W Cold 16 60 Mini-Equipment - - - - Not a piping penetration - - - - - - - - - - - - -
Hatch (NE) 61 Containm ent 56 Yes 18 (II) - - - - - - - - - - - - - - - Purge Supply D VAC-FCV-660 But Air I A Yes C Closed T N C G Cold 18 (NE) E VAC-FCV-661 But Air O A Yes C Closed T N C G Cold 18 62 Containm ent 56 Yes 18 (III) - - - - - - - - - - - - - - - Purge Exhaust F VAC-RCV-11 But Air I A Yes C Closed T N C G Cold 18 (NE) G VAC-RCV-12 But Air O A Yes C Closed T N C G Cold 18 63 Containm ent 56 Yes 18 (I) - - - - - - - - - - - - - - - Pressure and A VAC-FCV-662 But Air I A Yes C Closed T N C G Cold 18 Vacuum Relief B VAC-FCV-663 But Air O A Yes C Closed T N C G Cold 18 (NE) C VAC-FCV-664 But Air O A Yes C Closed T N C G Cold 18 J Spectacle Flange Spf - O A - N/A - - N/A N/A G Cold 26 64 Fuel Transfer - - 3 (III) - - - - - - - - - - - - - - 26 DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 15 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes Tube (NE) Not a piping penetration - - - - - - - - - - - - - 65 Personnel Hatch (NE) - - - Not a piping penetration - - - - - - - - - - - - - 66 Emergenc y - - - Not a piping penetration - - - - - - - - - - - - - Personnel Hatch (NE) 67 Equipment Hatch - - - Not a piping penetration - - - - - - - - - - - - - (NE) - - - - - - - - - - - - - - - - 68 Containment Air 56 Yes 19 (II) - - - - - - - - - - - - - - -
Sample (NE ) C VAC-FCV-679 Bal Air O A Yes O Closed T N C G Cold 18 D VAC-FCV-678 Bal Air I A Yes O Closed T N C G Cold 18 69 Containment Air 56 Yes 19 (I) - - - - - - - - - - - - - - -
Sample (NE) A VAC-FCV-681 Bal Air O A Yes O Closed T Y C G Cold 18 B VAC-21 Chk - I A No O - - Y C G Cold - 70 Auxiliary Steam 56 Yes 20 (I) - - - - - - - - - - - - - - - Supply (NE) A AXS-26 Gte Man O E No C - - N C G Hot - B AXS-208 Chk - I E No C - - N C G Hot - 71 Relief Valve 56 Yes 14 (IV) - - - - - - - - - - - - - - - (NE) Header G RCS-8028 Chk - I E No C - - N C G Hot - H CVCS-RV-8125, Rlf - O E No C - - N C G Hot 20,26 H SI-RV-8851,8853A,8853B, Rlf - O E No C - - N C G Hot 20,26 H 8856A,8856B,8858 Rlf - O E No C - - N C G Hot 20,26 H CS-RV-9007A,9007B Rlf - O E No C - - N C G Hot 20,26 DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 16 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes I RCS-512 Glb Man O E No C - - N C G Hot - 72 Spare - - - - - - - - - - - - - - - - - 26 73 Spare - - - - - - - - - - - - - - - - - 26 74 Spare - - - - - - - - - - - - - - - - - 26 75 Safety Injection 55 Yes 21 (I) - - - - - - - - - - - - - - - Pump 2 Discharge A SI-8802B Gte Mtr O D Yes C As is R-M Y O W Cold 7,11, 26 (SA) 76A Pressurize r 55 Yes 17 (I) - - - - - - - - - - - - - - - Steam A NSS-9354A Glb Air I A Yes C Closed T Y C G Hot - Sample B NSS-9354B Glb Air O A Yes C Closed T Y C G Hot - 76B Pressurizer Relief 55 Yes 14 (I) - - - - - - - - - - - - - - - Tank Gas Analyzer A RCS-8034B Glb Air O A Yes C Closed T N C G Cold - (NE) B RCS-8034A Glb Air I A Yes C Closed T N C G Cold - 76C Deadweight Tester 55 Yes 22 (I) - - - - - - - - - - - - - - - (Abandoned in A Sealed Bellows Sb l - I - - - - - - - W Cold 16 place. Tester B U1: 1-07P-8085B Glb Man O E No C - - N C W Cold 26 removed.) (NE) - U2: 1/4" Instrument cap - - O - - - - - - - - - 16 76D Containm ent 56 Yes 15 (I) - - - - - - - - - - - - - - - Pressure Transmitt er A Sealed Bellows Sbl - I - - - - - Y - W Cold 16 PT-934 (SA) B Sealed Instrument Sin - O - - - - - Y - W Cold 16 76E Spare 56 Yes 11 (III)
Connectio - U1: 3/8 Inst Cap - - I - - - - - - - - - 26 DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 17 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes n - U1: 3/8 Inst Cap - - O - - - - - - - - - 26 - - - U2: 1 Welded Pipe Cap - - - - - - - - - - - - 26 77 Safety Injection 55 Yes 21 (II) - - - - - - - - - - - - - - - Pump 1 Discharge (SA) E SI-8802A Gte Mtr O D Yes C As is R-M Y O W Cold 7,11, 26 78A Containm ent H 2 56 Yes 24 (I) - - - - - - - - - - - - - - - Monitor Supply (SA) A VAC-FCV-238 Glb Sol I E Yes C Closed - Y O G Cold 28,29 B VAC-FCV-239 Glb Sol O E Yes C Closed - Y O G Cold 28,29 78B Containm ent H 2 56 Yes 24 (II) - - - - - - - - - - - - - - - Monitor Return C VAC-FCV-240 Glb Sol O E Yes C Closed - Y O G Cold 28,29 (SA) D VAC-253 Chk - I E No C - - Y O G Cold 29 78C Containm ent 56 Yes 15 (I) - - - - - - - - - - - - - - - Pressure Transmitt er A Sealed Bellows Sbl - I - - - - - - - W Cold 16 PT-936 (SA) B Sealed Instrument Sin - O - - - - - - - W Cold 16 78D Spare 56 Yes 11 (III)
Connectio n - U1: 3/8 Inst Plug - - I - - - - - - - - - 26 - U1: 3/8 Inst Plug - - O - - - - - - - - - 26 - - - U2: 1 Welded Pipe Cap - - - - - - - - - - - - 26 78E Spare Connectio n - - - 1 Welded Pipe Cap - - - - - - - - - - - - 26 78F Spare Connectio n - - - 1 Welded Pipe Cap - - - - - - - - - - - - 26 DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 18 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes 78G Spare - Connectio n - - - 1 Welded Pipe Cap - - - - - - - - - - - - 26 78H Spare Connectio n - - - 1 Welded Pipe Cap - - - - - - - - - - - - 26 79 Fire Water (NE) 56 Yes 23 (I) - - - - - - - - - - - - - - - A FP-FCV-633 Glb Air O A Yes O Closed T N C W Cold - B FP-180(U1),FP-867(U2) Chk - I A No O - - N C W Cold - 80A Spare Instrumen t 56 Yes 11 (II) - - - - - - - - - - - - - - - (U1) Test Line (Unit-1) - Instrument Plug - - I E No - - - N - G Cold 26 - VAC-301 Glb Man O E No C - - N C G Cold 26 80A Spare Instrumen t 56 Yes 11 (IV) - - - - - - - - - - - - - - - (U2) Test Line (Unit-2) - VAC-303 Glb Man I E No C - - N C G Cold 26 - VAC-301 Glb Man O E No C - - N C G Cold 26 80B Spare Instrumen t 56 Yes 11 (II) - - - - - - - - - - - - - - - (U1) Test Line (Unit-1) - Instrument Plug - - I E No - - - N - G Cold 26 - VAC-302 G l b Man O E No C - - N C G Cold 26 80B Spare Instrumen t 56 Yes 11 (IV) - - - - - - - - - - - - - - - (U2) Test Line (Unit-2) - VAC-304 Glb Man I E No C - - N - G Cold 26 - VAC-302 Glb Man O E No C - - N C G Cold 26 80C Spare Instrumen t 56 Yes 11 (III) - - - - - - - - - - - - - - -
DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 19 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes Test Line - Instrument Plug - - I E No - - - N C G - 26 - Instrument Plug - - O E No - - - N C G - 26 80D Containm ent 56 Yes 15 (I) - - - - - - - - - - - - - - - Pressure Transmitt er A Sealed Bellows Sbl - I - - - - - - - W Cold 16 PT-939 (SA) B Sealed Instrument Sin - O - - - - - - - W Cold 16 80E RV Level 55 Yes 25 (I) - - - - - - - - - - - - - - - Instrumen tation A Sealed Bellows Sbl - I - - - - - - - W Cold 16 Transmitt er LIS-1320 (SA) B Hydraulic Isolators Hys - O - - - - - - - W Cold 16 80F RV Level 55 Yes 25 (I) - - - - - - - - - - - - - - - Instrumen tation A Sealed Bellows Sbl - I - - - - - - - W Cold 16 Transmitt er LIS-1321 (SA) B Hydraulic Isolators Hys - O - - - - - - - W Cold 16 80G RV Level 55 Yes 25 (I) - - - - - - - - - - - - - - - Instrumen tation A Sealed Bellows Sbl - I - - - - - - - W Cold 16 Transmitt er LIS-1322 (SA) B Hydraulic Isolators Hys - O - - - - - - - W Cold 16 81 Containm ent 56 Yes 24 (III) - - - - - - - - - - - - - - - External H 2 E VAC-FCV-668 Gte Mtr O E Yes C As is - Y C G Cold 28,38 Recombiners (SA) F VAC-FCV-658 Gte Mtr I E Yes C As is - Y C G Cold 28,38 82A Post-Accident 56 Yes 25 (IV) - - - - - - - - - - - - - - -
DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 20 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes Sampling System G LWS-FCV-697 Glb Sol O A Yes C Closed - Y C W Cold 28,29 Reactor Cavity H LWS-FCV-696 Glb Sol I A Yes C Closed - Y C W Cold 28 Sump (NE) 82B Post-Accident 56 Yes 25 (II) - - - - - - - - - - - - - - -
Sampling System Containm ent Air C VAC-FCV-698 Glb Sol I A Yes C Closed - Y C G Cold 28,29 Supply (NE) D VAC-FCV-699 Glb Sol O A Yes C Closed - Y C G Cold 28,29 82C Post-Accident 56 Yes 25 (III) - - - - - - - - - - - - - - -
Sampling System E VAC-FCV-700 Glb Sol O A Yes C Closed - Y C G Cold 28,29 Containment Air Return (NE) F VAC-116 Chk - I A No C - - Y C G Cold 29 82D Spare Piping 56 Yes 11 (II) - - - - - - - - - - - - - - - (U1) Connectio n - Blind Flange - - I E No C - - N C G Cold 26
- VAC-1-680 Gte Man O E No C - - N C G Cold 26 82D Chilled Water - - 7 (III) - - - - - - - - - - - - - - 52 (U2) Supply (NE) (abandon in place) 82E Spare Instrumen t 56 Yes 11 (III) - - - - - - - - - - - - - - - Test Line - Instrument Plug - - I E No - - - - - G Cold 26 - Instrument Plug - - O E No - - - - - G Cold 26 DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 21 of 24 Revision 23 December 2016 Pentr. Nos.
System (Safety Priority) (Note 33)
Appli- cable GDC GDC Confor- mance Figure 6.2-19 Sheet No.
Vlv Ltr Valve ID Number (Note 51)
Valve Type (Note 30)
Operator Type (Note 31)
Cntmt Locat. (Note 32)
PG&E Piping Group Control Room Indication Normal Position (Note 35)
Power Fail. Position Trip On (Note 25) Used After LOCA (Note 36) Long Term Post- LOCA Position (Note 37)
Fluid (Note 23)
Temp (Note 24)
Notes 83A Spare Piping 56 Yes 11 (II) - - - - - - - - - - - - - - - (U1) Connectio n - Blind Flange - - I E No C - - N C G Cold 26 - VAC-1-681 Gte Man O E No C - - N C G Cold 26 83A Chilled Water - - 7 (IV) - - - - - - - - - - - - - - 52 (U2) Return (NE) (abandon in place) 83B Hydrogen Purge 56 Yes 11 (I) - - - - - - - - - - - - - - - Supply (NE) A VAC-200,201 Chk - I E No C - - Y O G Cold 38 B VAC-1,2 Gte Man O E No C - - Y C G Cold 38 84 Spare - - - - - - - - - - - - - - - - - 26 Notes: 1. Trip on feedwater isolation. (refer to Table 6.2-40, Item 4.)
- 2. Steam generator secondary side is missile-protected closed system. The closed loop system includes all pressure boundary test, vent, drain, and instrument valves; instruments; blanked flange connections; and steam generator tubes.
- 3. Reverse check (main steam isolation).
- 4. Trip on steam line isolation. (refer to Table 6.2-40, Item 3.)
- 5. Safety-related function. Valve trip is related to systems safety function.
- 6. PG&E considers the five main steam relief valves to be containment isolation valves.
- 7. Safety-related function demands that this valve not automatically isolate.
- 8. Penetration does not meet 1971 GDC but does meet the GDC 53, 1967 which was applicable at time of construction commitment.
- 9. This valve is not considered as the automatic isolation barrier. The barrier is provided by the closed system. The valve does have provision for remote manual isolation should the situation require it.
- 10. PG&E considers the closed system outside containment as an automatic isolation barrier.
- 11. Provision for remote manual isolation exists should the situation require it. Safety-related function demands that this valve not isolate. It conforms with the intent of the GDC as it affects maintenance of the containment boundary.
DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 22 of 24 Revision 23 December 2016 12. The protective chamber is considered outside containment. The motor-operated valves are the isolation valves outside containment. 13. Valve opens on containment spray signal.
- 14. Valve opens on SIS signal.
- 15. Deleted in Revision 17
- 16. Containment isolation effected by completely sealed instrument system. This penetration is not leak tested.
- 17. Deleted in Revision 22
- 18. Containment vent isolation trip. (refer to Table 6.2-40, Item 5)
- 19. The fuel transfer tube is not considered to be a piping penetration, but rather a Type B test penetration. The quick-opening hatch is double gasketed with a test connection allowing pressurization between the gaskets for Type B testing. The portion of the transfer tube inside the containment is considered to be part of the containment liner; the portion of the transfer tube outside the containment is not considered to be part of the containment boundary.
- 20. The relief valves are considered as normally closed containment isolation valves.
- 21. Deleted.
- 22. Administratively controlled valve which is treated in the same manner as a sealed closed valve. Control room indication is only active when the valve is not administratively cleared.
- 23. W = Water; G = Gas
- 24. Hot - over 200
ºF; Cold - 200
ºF or less
- 25. R-M = Remote Manual; S = Safety Injection; T = Containment Isolation Signal, Phase A; P = Containment Isolation Signal, Phase B; F = Feedwater Isolation Signal; M = Main Steam Isolation Signal.
- 26. Testability not required under Option B of Appendix J to 10 CFR Part 50.
- 27. Deleted in Revision 17.
- 28. This device is used for post-accident monitoring or control and must not be isolated by a containment isolation signal.
- 29. Multiple penetration number usage results from multiple tubes running though the guard pipe of a single penetration.
- 30. The following abbreviations are used: Gte = Gate Cls = Closed system Sbl = Sealed bellows Spf = Spectacle flange Glb = Globe Chk = Check Blf = Blind flange Rlf = Relief Dia = Diaphragm Sin = Sealed instruments But = Butterfly Bal = Ball Hys = Hydraulic isolators
- 31. The following abbreviations are used:
Man = Manual Mtr = Motor Air = Air E/H = Electrohydraulic Spr = Spring N 2 = Nitrogen Sol = Solenoid
- 32. "I" is used for inside and "O" for outside.
- 33. Safety-related priority designation for each penetration per Section 6.2.4.4.15 is as follows: ES = Essential SA = Safety
DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 23 of 24 Revision 23 December 2016 NE = Nonessential
- 34. Penetration 36 is not used by a safety system for accident mitigation. However, flow through this Penetration may be required to achieve safe shutdown following a Hosgri earthquake or an Appendix R fire.
- 35. Normal Position column: (A) C = Closed, O = Open (B) For check valves, position is stated open if flow passes through the check valve during normal operation, otherwise the valve is stated closed. (C) Normal configuration applies to the following plant conditions:
(a) Modes 1-4, applicable T.S. 3.6.1 and 3.6.3. (b) Mode 6, applicable ECG 42.1. (D) If valve is normally open it is designated Open or if only periodically opened for fulfillment of its function during the Normal plant configurations, then a Closed designator is used.
(E) If valve is normally closed and opened only in support of testing, under administrative control per T.S. 3.6.3, or for stroke testing of the valve itself, then a Closed designator is used. (F) Relief valves are assigned a Closed designator. 36. Used After LOCA column: (A) N = No, Y = Yes (B) For check valves, if valve passes flow at any point following the accident, then a Yes designator is used. (C) For valves that change position on a safeguards signal, this change is not considered a use, that is, the time the safeguards signal is received is not considered after the accident. (D) Use is principally an indicator of a valve passing flow at any point after the accident. 37. Long Term Post-LOCA Position column:
(A) C = Closed, O = Open (B) The column pertains to a post accident condition, long term core cooling. (C) The assumed accident is a primary system LOCA with Containment isolation Phase A and B signals generated, system depressurization below 150 PSIG, corresponding to a RHR pump injection flow of greater than 200 GPM. The accident is assumed to progress through injection, cold leg recirculation, and to hot leg recirculation for the long term. (D) The hot leg injection flowpath is injection to RHR hot legs 1 & 2, SI pump hot legs 1,2,3, & 4, and Charging cold legs 1,2,3, & 4; the condition established by EOP E-1.4, (no RNOs entered).
(E) Containment temperature has been reduced to near ambient conditions. (F) Containment pressure has been reduced to near atmospheric conditions. (G) Primary system/containment recirculation sump temperature has been reduced to below 200
°F. (H) Although used periodically, PASS valves are considered to be normally closed.
- 38. Valve may be used following a LOCA only in the event of a failure of both internal hydrogen recombiners.
- 39. Deleted in Revision 22
- 40. Valve is not subject to Type C leakage tests because it is required to be in service post accident and the line pressure upstream of the CIV is greater than the post-accident pressure inside containment.
- 41. Deleted in Revision 22
- 42. Deleted in Revision 22
- 43. Deleted in Revision 22
- 44. Deleted in Revision 22.
- 45. This penetration does not require testing per Appendix J section II.H because the penetration does not provide a direct connection to the atmosphere during normal operation, is not required to close automatically, and is not required to operate intermittently. This penetration is filled with water from the submerged sump preventing gas escape to the outside and meets the test exception per Appendix J section II.c.3.
- 46. Deleted in Revision 22
- 47. The Containment Fan Coolers are a closed system inside containment. CCW does not communicate directly with the containment atmosphere. The closed loop system includes all pressure boundary test, vent, drain and instrument valves; instruments; blanked flange connections; and CFCU tubing.
DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 24 of 24 Revision 23 December 2016
- 48. Valve is included as a credited CIV because it was included on the original Tech Specs for the plant license.
- 49. Deleted in Revision 22
- 50. Deleted in Revision 22
- 51. This column lists the credited containment isolation valve or barrier that is used to demonstrate conformance to the PG&E Piping Group and the applicable GDC. Associated isolation devices required to maintain containment leakage integrity of penetrations and closed systems such as vent, drain test, instrumentation and branch line valves, blind flanges, caps or other passive devices are not shown They are administratively controlled to be in their proper isolation configuration by plant procedures and per TS 3.6.3.
- 52. This penetration is abandoned in place through the installation of welded plugs in the penetration.
- 53. Deleted in Revision 22
- 54. Deleted in Revision 22
- 55. The Steam Generator Blowdown System does not communicate directly with the containment atmosphere or reactor coolant pressure boundary. This valve is required to close when the auxiliary feedwater pumps are started in order to maintain steam generator inventory.
- 56. Valve is installed with a hole drilled through the ball allowing unidirectional pressure relief to meet the requirements of Generic Letter 96-06, September 1996 (refer to Section 6.2.4.4.17).
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2-40 Sheet 1 of 2 Revision 22 May 2015 OPERATING CONDITIONS FOR CONTAINMENT ISOLATION Item No. Functional Unit Operating Conditions 1 Containment isolation Phase A a. High containment pressure
- b. Pressurizer low pressure
- c. Low steamline pressure
- d. Manual 2 Containment isolation Phase B a. High-high containment pressure
- b. Manual 3 Steam line isolation a. Low steamline pressure
- b. High steamline pressure rate
- c. High-high containment pressure
- d. Manual 4 Feedwater line isolation a. High containment pressure
- b. Low steamline pressure
- c. Pressurizer low pressure
- d. Steam generator high-high level
- e. Manual SIS
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2-40 Sheet 2 of 2 Revision 22 May 2015 Item No. Functional Unit Operating Conditions 5 Containment ventilation isolation a. Containment exhaust detectors
- b. Safety injection activation
- c. Manual Phase A or Manual Phase B or Manual Spray Actuation
DCPP UNITS 1 & 2 FSAR UPDATE Revision 14 November 2001 TABLE 6.2-41 POST-LOCA TEMPERATURE TRANSIENT USED FOR ALUMINUM AND ZINC CORROSION Time Interval, sec Temperature, °F 0 - 10 240 10 - 30 265 30 - 750 258 750 - 2000 253 2000 - 10,000 250 10,000 - 20,000 245 20,000 - 50,000 240 50,000 - 75,000 230 75,000 - 100,000 220 100,000 - 200,000 210 200,000 - 400,000 198 400,000 - 600,000 185 600,000 - 800,000 175 800,000 - 1,200,000 160 1,200,000 - 2,000,000 153 2,000,000 - 3,000,000 145 3,000,000 - 5,000,000 138 5,000,000 - 8,640,000 129
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2-42 Sheet 1 of 2 Revision 14 November 2001 PARAMETERS USED TO DETERMINE HYDROGEN GENERATION Plant Thermal Power Rating 3,425 MWt Containment Temperature Prior to Accident 120°F Containment Free Volume 2,550,000 ft 3 Weight Zirconium Cladding 43,300 lb
Hydrogen Recombiner Flowrate 100 scfm Corrodible Metals Aluminum and zinc
Core Cooling Solution Radiolysis Sources Percent of total halogens retained in the core 50.00 Percent of total noble gases retained in the core 0.00 Percent of other fission products retained in the core 99.00
Energy Distribution Percent of total decay energy - gamma 50.00 Percent of total decay energy - beta 50.00
Energy Absorption by Core Cooling Solution Percent of gamma energy absorbed by solution 10.00 Percent of beta energy absorbed by solution 0.00
Hydrogen Production Molecules H 2 produced per 100 eV energy absorbed by solution 0.50 Sump Solution Radiolysis Sources
Percent of total halogens released to sump solution 50.00 Percent of noble gases released to sump solution 0.00 Percent of other fission products released to sump solution 1.00
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2-42 Sheet 2 of 2 Revision 14 November 2001 Energy Absorption by Sump Solution Percent of total energy (beta and gamma) which is absorbed by the sump solution 100.00 Hydrogen Production Molecules of hydrogen produced per 1000 eV of energy absorbed by the sump solution 0.50 Long-term Aluminum Corrosion Rate 200 mils/year
Aluminum Inventory in Containment 3585 lb (amount used in analyses) 15,988 ft 2
Zinc Inventory in Containment (amount used in analysis) 58,449 lb 397,000 ft 2
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 6.2-43 CORE FISSION PRODUCT ENERGY AFTER OPERATION WITH EXTENDED FUEL CYCLES Core Fission Product Energy (a) Time After Reactor Trip, days Energy Release Rate, watts/MWt x 10 3 Integrated Energy Release, watts days/MWt x 10 4 1 5.11 0.696 5 3.41 2.28 10 2.72 3.80 20 2.00 6.11 30 1.66 7.92 40 1.47 9.48 50 1.33 10.9 60 1.21 12.2 70 1.12 13.3 80 1.02 14.4 90 0.943 15.4 100 0.868 16.2
(a) Assumes 50% core halogens +99% other fission products and no noble gases.
Values are for total ( and ) energy.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 6.2-44 FISSION PRODUCT DECAY DEPOSITION IN SUMP SOLUTION Sump Fission Product Energy (a) Energy Integrated Time After Release Energy Reactor Trip, Rate, Release, days watts/MWt x 10 watt-days/MWt x 10 3 1 25.6 0.535 5 8.17 1.02 10 5.35 1.35 15 3.80 1.57 20 2.91 1.75 30 2.06 1.99 40 1.69 2.18 60 1.30 2.47 80 1.04 2.70 100 0.837 2.88 (a) Considers release of 50 percent of core halogens, no noble gases, and 1 percent of other fission products to the sump solution.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 6.2-45
SUMMARY
OF HYDROGEN ACCUMULATION DATA (with no recombination)
Time of Total Production Volume Percent Occurrence, Rate, Hydrogen days scfm 2.09 1 10.9 2.65 2 8.18 3.09 3 6.45 3.48 4 6.23 3.83 5 5.03 4.13 6 4.90 5.10 10 3.19 5.96 15 2.62
6.68 20 2.42 9.77 50 1.66 11.54 75 1.22 12.95 100 1.11
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2-47 Page 1 of 2 Revision 11 November 1996 CONTAINMENT REFLECTIVE INSULATION (a) Line No. Size, in. Length, ft. Line Designation 1 29 19 Reactor Coolant Out Loop 1 2 29 19 Reactor Coolant Out Loop 2 3 29 19 Reactor Coolant Out Loop 3 4 29 19 Reactor Coolant Out Loop 4 5 31 22 Reactor Coolant PP Suction Loop 1 6 31 22 Reactor Coolant PP Suction Loop 2 7 31 13 Reactor Coolant PP Suction Loop 3 8 1 3 Reactor Coolant PP Suction Loop 4 9 27-1/2 21 Reactor Coolant PP Discharge Loop 1 10 27-1/2 4 Reactor Coolant PP Discharge Loop 2 11 27-1/2 23 Reactor Coolant PP Discharge Loop 3 12 27-1/2 23 Reactor Coolant PP Discharge Loop 4 13 4 111 Loop 1 spray line 14 4 84 Loop 2 spray line 15 4 96 Pressurizer spray line 16 14 65 Pressurizer surge line 24 3 38 Letdown Line Loop 2 50 3 9 Charging Line Loop 3 109 14 45 Hot Leg Recirc. Before V-8702 235 6 13 Safety Injection Loop 1 Hot Leg 236 6 10 Safety Injection Loop 2 Hot Leg 237 6 8 Safety Injection Loop 3 Hot Leg 238 6 11 Safety Injection Loop 4 Hot Leg 246 3 6 Charging Line Loop 4 253 10 41 Accumulator Injection Loop 1 254 10 35 Accumulator Injection Loop 2 255 10 14 Accumulator Injection Loop 3 256 10 30 Accumulator Injection Loop 4 958 2 4 Loop 1 cold leg drain RCDT 959 2 6 Loop 2 cold leg drain RCDT 960 2 2 Loop 3 cold leg drain RCDT 961 2 3 Loop 4 cold leg drain RCDT 1665 14 21 Loop 4 hot leg before V 8701 1992 1-1/2 3 Boron Inj Tk. Out Loop 2 cold 1993 1-1/2 3 Boron Inj Tk. Out Loop 3 cold
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2-47 Page 2 of 2 Revision 11 November 1996 Line No. Size, in. Length, ft. Line Designation 3844 6 35 RHR PP 1-1 Inj Cold Leg 1 3845 6 48 RHR PP 1-1 Inj Cold Leg 2 3846 6 48 RHR PP 1-1 Inj Cold Leg 3 3847 6 48 RHR PP 1-1 Inj Cold Leg 4
3855 2 4 SI PPS Cold Leg Loop 1 recirc.
(a) The information contained in this table is "representative" of Units 1 and 2. Data reflecting actual conditions for any individual line may be somewhat different than that presented, or may change with plant modifications. This table will not be revised to reflect these individual conditions or changes.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2-48 Sheet 1 of 2 Revision 19 May 2010 CONTAINMENT CONVENTIONAL INSULATION Line No. Insul. Size, in. Length, ft. Line Designation 23 IV 12 45 Pressurizer relief header 50 IV 3 64 Charging Line Loop 3 51 IV 2 77 Charging line auxiliary spray 63 V 1 218 Excess Letdown Loop 2 214 V 3/4 3 Loop 1 hot leg sample 215 V 3/4 1 Loop 4 hot leg sample 225 V 28 94 Steam Gen. 1-4 steam outlet 226 V 28 96 Steam Gen. 1-3 steam outlet 227 V 28 96 Steam Gen. 1-2 steam outlet 228 V 28 94 Steam Gen. 1-1 steam outlet 246 IV 3 115 Charging Line Loop 4 508 III 8 226 RHR PP 1-1 Inj Cold Leg 1 + 2 509 III 8 77 RHR PP 1-1 Inj Cold Leg 3 + 4 528 III 2-1/2 123 Reactor coolant drain tk. PP disch.
554 IV 16 54 Steam Gen. 1 feedwater supply 555 IV 16 57 Steam Gen. 2 feedwater supply 556 IV 16 53 Steam Gen. 4 feedwater supply 557 IV 16 54 Steam Gen. 3 feedwater supply 692 III 3/4 4 Cold Leg Loop 3 + 4 test line 927 III 14 33 Loop 4 hot leg to RHR PPS 1012 V 2 3 Steam Gen. 1 blowdown out N/S 1017 2 3 Steam Gen. 2 blowdown out N/S 1020 V 2 14 Steam Gen. 3 blowdown out N/S 1038 V 2 8 Steam Gen. 4 blowdown out N/S 1040 V 2-1/2 235 Steam G en. 1-1 blowdown tank hdr. 1041 V 2-1/2 189 Steam G en. 1-2 blowdown tank hdr. 1042 V 2-1/2 85 Steam G en. 1-3 blowdown tank hdr. 1043 V 2-1/2 116 Steam G en. 1-4 blowdown tank hdr. 1059 V 2 32 Steam Gen. 1 blowdown out S/N 1060 V 2 17 Steam Gen. 2 blowdown out S/N 1061 IV 2 3 Steam Gen. 3 blowdown out S/N 1062 V 2 15 Steam Gen. 4 blowdown out S/N 1167 III 4 51 RHR hot leg RV outlet 1169 V 3/4 7 Loop 1 spray line bypass 1170 V 3/4 7 Loop 2 spray line bypass 1675 V 3/8 51 Loop 1 hot leg sample 1676 V 3/8 6 Loop 4 hot leg sample
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2-48 Sheet 2 of 2 Revision 19 May 2010 Line No. Insul. Size, in. Length, ft.
Line Designation 1901 III 1 8 Loop 2 V8076 + V8074B Lkoff. H 1999 III 3/4 10 SIS Accum. 1 test 2000 III 3/4 13 SIS Accum. 2 test 2001 III 3/4 60 SIS Accum. 3 test 2002 III 3/4 28 SIS Accum. 4 test 2158 IV 3/4 12 Regen. hx. channel temp. relief 2176 III 1 35 Leakoff header line to PRT 2385 III 4 202 SIS RV outlet header to PRT 2523 III 1/2 4 RHR Suction Vlv. 2 Loop 4 leakoff 2524 III 1/2 18 RHR Suction Vlv. 1 Loop 4 leakoff 2638 III 1 57 Cont. aux. steam supply to el. 91 2766 III 1/2 2 PCV-455 A leakoff line 2773 III 1/2 4 PCV-455 B leakoff line 2998 III 4 4 SIS RV outlet header to PRT 2999 III 4 5 SIS RV outlet header to PRT 3094 V 3/4 5 RHR Loop 4 V-8702 8702 therm 3095 III 3/4 5 RHR Loop 4 V-8701 8702 therm 3214 I 2 73 Incore chiller chill wtr. sup.
3215 I 2 73 Incore chiller chill wtr. ret.
3407 III 1/2 6 Loop 2 Letdown V-8076 leakoff 3729 III 2-1/2 14 REAC Clnt. Dr. PPS Disch. Header 3844 II 6 70 RHR PP 1-1 inj cold leg 1 3845 II 6 59 RHR PP 1-1 inj cold leg 2 3900 III 1 174 Reactor head aux. steam 3936 I 2 2 Incore chiller chill wtr. ret.
4399 III 1/2 1 4400 III 1/2 7 4402 III 1/2 5 4406 III 1 4 4407 III 1 2
(a) This line is contained completely within the hot leg insulation.
(b) The information contained in this table is "representative" of Units 1 and 2. Data reflecting actual conditions for any individual line may be somewhat different than that presented, or
may change with plant modifications. This table will not be revised to reflect these individual conditions or changes.
DCPP UNITS 1 & 2 FSAR UPDATE Sheet 1 of 2 Revision 23 December 2016 TABLE 6.2-57 CONTAINMENT SUBCOMPARTMENT ANALYSES DOUBLE-ENDED PRESSURIZER SURGE LINE BREAK AT THE HOT LEG CONNECTION MASS AND ENERGY RELEASES Time Seconds Mass Flow Lbm/Sec Energy Release Btu/Sec .0000E+00
.0000E+00
.0000E+00
.1010E-02
.4634E+04
.2822E+07
.2020E-02
.8154E+04
.4968E+07
.4020E-02
.1135E+05
.6943E+07
.6030E-02
.1267E+05
.7788E+07
.1104E-01
.1857E+05
.1148E+08
.1201E-01
.1926E+05
.1191E+08
.1404E-01
.1970E+05
.1221E+08
.1804E-01
.1816E+05
.1134E+08
.2305E-01
.2121E+05
.1327E+08
.3503E-01
.2317E+05
.1459E+08
.4007E-01
.2250E+05
.1421E+08
.5809E-01
.2386E+05
.1515E+08
.6803E-01
.2351E+05
.1496E+08
.7805E-01
.2276E+05
.1453E+08
.8503E-01
.2313E+05
.1478E+08
.1010E+00
.2489E+05
.1587E+08
.1051E+00
.2497E+05
.1593E+08
.1720E+00
.2140E+05
.1376E+08
.1870E+00
.2019E+05
.1304E+08
.2370E+00
.2020E+05
.1305E+08
.2731E+00
.1905E+05
.1235E+08
.3070E+00
.1707E+05
.1115E+08
.3530E+00
.1593E+05
.1047E+08
.7550E+00
.1565E+05
.1027E+08
.8100E+00
.1583E+05
.1037E+08
.8300E+00
.1568E+05
.1027E+08
.8400E+00
.1586E+05
.1038E+08
.8550E+00
.1555E+05
.1019E+08
.9000E+00
.1564E+05
.1024E+08
.9350E+00
.1614E+05
.1053E+08
.9500E+00
.1590E+05
.1039E+08
.9751E+00
.1619E+05
.1056E+08
.9850E+00
.1657E+05
.1079E+08
.1000E+01
.1785E+05
.1155E+08
.1020E+01
.1711E+05
.1110E+08
.1060E+01
.1679E+05
.1091E+08 DCPP UNITS 1 & 2 FSAR UPDATE Sheet 2 of 2 Revision 23 December 2016 TABLE 6.2-57 CONTAINMENT SUBCOMPARTMENT ANALYSES DOUBLE-ENDED PRESSURIZER SURGE LINE BREAK AT THE HOT LEG CONNECTION MASS AND ENERGY RELEASES Time Seconds Mass Flow Lbm/Sec Energy Release Btu/Sec .1080E+01
.1631E+05
.1062E+08
.1150E+01
.1629E+05
.1060E+08
.1180E+01
.1702E+05
.1102E+08
.1200E+01
.1691E+05
.1096E+08
.1220E+01
.1734E+05
.1121E+08
.1260E+01
.1635E+05
.1062E+08
.1300E+01
.1659E+05
.1077E+08
.1340E+01
.1712E+05
.1108E+08
.1420E+01
.1670E+05
.1083E+08
.1470E+01
.1698E+05
.1100E+08
.1900E+01
.1652E+05
.1075E+08
.2400E+01
.1528E+05
.1002E+08
.3000E+01
.1483E+05
.9772E+07
- - The LOCA M&E in this table was increased by 10% for use in the subcompartment pressurization calculations. WCAP-8264-P-A contains the basis for the 10% increase.
DCPP UNIT 1 & 2 FSAR UPDATE Sheet 1 of 2 Revision 23 December 2016 TABLE 6.2-58 CONTAINMENT SUBCOMPARTMENT ANALYSES DOUBLE ENDED PRESSURIZER SPRAY LINE BREAK AT THE TOP OF THE PRESSURIZER MASS AND ENERGY RELEASES Time Seconds Mass Flow Lbm/Sec Energy Release Btu/Sec .0000E+00
.0000E+00
.0000E+00
.1040E-02
.3410E+02
.1772E+05
.2010E-02
.7114E+02
.3695E+05
.3010E-02
.1087E+03
.5644E+05
.4020E-02
.1449E+03
.7520E+05
.5080E-02
.1811E+03
.9395E+05
.6030E-02
.2113E+03
.1096E+06
.7060E-02
.2415E+03
.1251E+06
.8070E-02
.2681E+03
.1387E+06
.9000E-02
.2895E+03
.1497E+06
.1003E-01
.3110E+03
.1607E+06
.1104E-01
.3285E+03
.1696E+06
.1209E-01
.3433E+03
.1771E+06
.1304E-01
.3537E+03
.1824E+06
.1502E-01
.3676E+03
.1894E+06
.1809E-01
.3957E+03
.2041E+06
.2305E-01
.4347E+03
.2248E+06
.2713E-01
.4730E+03
.2456E+06
.3212E-01
.5367E+03
.2803E+06
.4204E-01
.6965E+03
.3686E+06
.4702E-01
.7684E+03
.4090E+06
.5203E-01
.8232E+03
.4408E+06
.6102E-01
.8809E+03
.4769E+06
.7504E-01
.9400E+03
.5199E+06
.7805E-01
.9637E+03
.5358E+06
.8005E-01
.9869E+03
.5502E+06
.8901E-01
.1125E+04
.6326E+06
.9410E-01
.1190E+04
.6707E+06
.9908E-01
.1223E+04
.6911E+06
.1230E+00
.1308E+04
.7701E+06
.1271E+00
.1340E+04
.7929E+06
.1420E+00
.1489E+04
.8905E+06
.1460E+00
.1513E+04
.9067E+06
.1510E+00
.1524E+04
.9183E+06
.1580E+00
.1509E+04
.9195E+06
.1670E+00
.1529E+04
.9408E+06
.1840E+00
.1599E+04
.9943E+06 DCPP UNIT 1 & 2 FSAR UPDATE Sheet 2 of 2 Revision 23 December 2016 TABLE 6.2-58 CONTAINMENT SUBCOMPARTMENT ANALYSES DOUBLE ENDED PRESSURIZER SPRAY LINE BREAK AT THE TOP OF THE PRESSURIZER MASS AND ENERGY RELEASES Time Seconds Mass Flow Lbm/Sec Energy Release Btu/Sec .6102E-01
.8809E+03
.4769E+06
.7504E-01
.9400E+03
.5199E+06
.7805E-01
.9637E+03
.5358E+06
.8005E-01
.9869E+03
.5502E+06
.8901E-01
.1125E+04
.6326E+06
.9410E-01
.1190E+04
.6707E+06
.9908E-01
.1223E+04
.6911E+06
.1230E+00
.1308E+04
.7701E+06
.1271E+00
.1340E+04
.7929E+06
.1420E+00
.1489E+04
.8905E+06
.1460E+00
.1513E+04
.9067E+06
.1510E+00
.1524E+04
.9183E+06
.1580E+00
.1509E+04
.9195E+06
.1880E+00
.1603E+04
.9991E+06
.2050E+00
.1576E+04
.9980E+06
.2281E+00
.1563E+04
.1003E+07
.2390E+00
.1529E+04
.9878E+06
.2800E+00
.1365E+04
.8975E+06
.3011E+00
.1272E+04
.8374E+06
.3220E+00
.1213E+04
.7950E+06
.3431E+00
.1176E+04
.7690E+06
.3841E+00
.1157E+04
.7567E+06
.4960E+00
.1152E+04
.7525E+06
.1470E+01
.1147E+04
.7487E+06
.3000E+01
.1128E+04
.7373E+06
- - The LOCA M&E in this table was increased by 10% for use in the subcompartment pressurization calculations. WCAP-8264-P-A contains the basis for the 10% increase.
DCPP UNIT 1 & 2 FSAR UPDATE Sheet 1 of 2 Revision 23 December 2016 Table 6.2-59 CONTAINMENT SUBCOMPARTMENT ANALYSES DOUBLE-ENDED RESIDUAL HEAT REMOVAL SUCTION LINE BREAK AT THE RCS HOT LEG CONNECTION MASS AND ENERGY RELEASES Time Seconds Mass Flow Lbm/Sec Energy Release Btu/Sec .0000E+00
.0000E+00
.0000E+00
.1000E-05
.1548E+05
.9356E+07
.1010E-02
.1588E+05
.9604E+07
.2040E-02
.1635E+05
.9882E+07
.4010E-02
.1720E+05
.1040E+08
.6050E-02
.1804E+05
.1090E+08
.1404E-01
.2111E+05
.1275E+08
.1603E-01
.2181E+05
.1318E+08
.2003E-01
.2300E+05
.1389E+08
.2805E-01
.2518E+05
.1521E+08
.3029E-01
.2570E+05
.1552E+08
.3030E-01
.1022E+05
.6172E+07
.3301E-01
.1076E+05
.6500E+07
.3702E-01
.1150E+05
.6943E+07
.4109E-01
.1214E+05
.7332E+07
.4901E-01
.1295E+05
.7816E+07
.5703E-01
.1361E+05
.8215E+07
.6509E-01
.1412E+05
.8529E+07
.7305E-01
.1444E+05
.8721E+07
.1140E+00
.1529E+05
.9230E+07
.1300E+00
.1548E+05
.9342E+07
.1430E+00
.1532E+05
.9243E+07
.1930E+00
.1323E+05
.7967E+07
.2060E+00
.1277E+05
.7695E+07
.2191E+00
.1243E+05
.7484E+07
.2320E+00
.1229E+05
.7405E+07
.2471E+00
.1233E+05
.7433E+07
.2631E+00
.1227E+05
.7387E+07
.2791E+00
.1201E+05
.7233E+07
.3270E+00
.1074E+05
.6463E+07
.3430E+00
.1042E+05
.6270E+07
.3590E+00
.1032E+05
.6210E+07
.3890E+00
.1042E+05
.6273E+07
.4250E+00
.1029E+05
.6190E+07
.4650E+00
.1005E+05
.6045E+07
.4780E+00
.1011E+05
.6086E+07 DCPP UNIT 1 & 2 FSAR UPDATE Sheet 2 of 2 Revision 23 December 2016 Table 6.2-59 CONTAINMENT SUBCOMPARTMENT ANALYSES DOUBLE-ENDED RESIDUAL HEAT REMOVAL SUCTION LINE BREAK AT THE RCS HOT LEG CONNECTION MASS AND ENERGY RELEASES Time Seconds Mass Flow Lbm/Sec Energy Release Btu/Sec .5401E+00
.1099E+05
.6613E+07
.5601E+00
.1110E+05
.6680E+07
.6100E+00
.1111E+05
.6683E+07
.6750E+00
.1162E+05
.6983E+07
.7500E+00
.1159E+05
.6960E+07
.8450E+00
.1207E+05
.7235E+07
.9851E+00
.1225E+05
.7334E+07
.1150E+01
.1234E+05
.7378E+07
.1320E+01
.1218E+05
.7289E+07
.1600E+01
.1147E+05
.6892E+07
.1680E+01
.1120E+05
.6742E+07
.2300E+01
.1037E+05
.6354E+07
.2500E+01
.1003E+05
.6184E+07
.3000E+01
.9022E+04
.5649E+07
- - The LOCA M&E in this table was increased by 10% for use in the subcompartment pressurization calculations. WCAP-8264-P-A contains the basis for the 10% increase.
DCPP UNIT 1 & 2 FSAR UPDATE Sheet 1 of 2 Revision 23 December 2016 Table 6.2-60 CONTAINMENT SUBCOMPARTMENT ANALYSES DOUBLE-ENDED ACCUMULATOR LINE BREAK AT THE RCS COLD LEG CONNECTION MASS AND ENERGY RELEASES Time Seconds Mass Flow Lbm/Sec Energy Release Btu/Sec .0000E+00
.0000E+00
.0000E+00
.1020E-02
.3693E+04
.1900E+07
.2020E-02
.6600E+04
.3386E+07
.3020E-02
.8247E+04
.4219E+07
.4030E-02
.9129E+04
.4656E+07
.6030E-02
.9893E+04
.5020E+07
.1003E-01
.1047E+05
.5276E+07
.1400E-01
.1228E+05
.6194E+07
.1601E-01
.1274E+05
.6419E+07
.1900E-01
.1289E+05
.6479E+07
.2302E-01
.1277E+05
.6379E+07
.2504E-01
.1300E+05
.6481E+07
.3109E-01
.1411E+05
.6981E+07
.4210E-01
.1449E+05
.7014E+07
.5501E-01
.1383E+05
.6560E+07
.7304E-01
.1482E+05
.6848E+07
.9313E-01
.1451E+05
.6516E+07
.1080E+00
.1579E+05
.7031E+07
.1121E+00
.1591E+05
.7066E+07
.1180E+00
.1576E+05
.6959E+07
.1301E+00
.1527E+05
.6587E+07
.1591E+00
.1563E+05
.6591E+07
.1800E+00
.1565E+05
.6483E+07
.2011E+00
.1546E+05
.6269E+07
.2240E+00
.1588E+05
.6413E+07
.2690E+00
.1584E+05
.6263E+07
.2930E+00
.1594E+05
.6270E+07
.3170E+00
.1560E+05
.6059E+07
.4111E+00
.1582E+05
.6105E+07
.4310E+00
.1571E+05
.6041E+07
.4900E+00
.1608E+05
.6227E+07
.7400E+00
.1634E+05
.6395E+07
.1210E+01
.1570E+05
.6145E+07
.1700E+01
.1609E+05
.6446E+07
.1720E+01
.1649E+05
.6654E+07
.1730E+01
.1609E+05
.6450E+07
DCPP UNIT 1 & 2 FSAR UPDATE Sheet 2 of 2 Revision 23 December 2016 Table 6.2-60 CONTAINMENT SUBCOMPARTMENT ANALYSES DOUBLE-ENDED ACCUMULATOR LINE BREAK AT THE RCS COLD LEG CONNECTION MASS AND ENERGY RELEASES Time Seconds Mass Flow Lbm/Sec Energy Release Btu/Sec .1740E+01
.1590E+05
.6347E+07
.1810E+01
.1630E+05
.6571E+07
.1860E+01
.1607E+05
.6456E+07
.1870E+01
.1620E+05
.6527E+07
.1880E+01
.1654E+05
.6714E+07
.1890E+01
.1650E+05
.6692E+07
.1900E+01
.1625E+05
.6559E+07
.1930E+01
.1614E+05
.6508E+07
.1950E+01
.1629E+05
.6591E+07
.1980E+01
.1607E+05
.6481E+07
.2000E+01
.1632E+05
.6618E+07
.2100E+01
.1618E+05
.6558E+07
.2500E+01
.1619E+05
.6625E+07
.3000E+01
.1579E+05
.6481E+07
- - The LOCA M&E in this table was increased by 10% for use in the subcompartment pressurization calculations. WCAP-8264-P-A contains the basis for the 10% increase.
DCPP UNIT 1 & 2 FSAR UPDATE Sheet 1 of 2 Revision 23 December 2016 TABLE 6.2-61 CONTAINMENT SUBCOMPARTMENT ANALYSES NODE DESCRIPTIONS FOR SUBCOMPARTMENT MODEL
%&'()*+,%&'()*
+,%&'()*+%,
DCPP UNIT 1 & 2 FSAR UPDATE Sheet 2 of 2 Revision 23 December 2016 TABLE 6.2-61 CONTAINMENT SUBCOMPARTMENT ANALYSES NODE DESCRIPTIONS FOR SUBCOMPARTMENT MODEL
%%%%%&%'%(%)%*
DCPP UNIT 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 6.2-62 CONTAINMENT SUBCOMPARTMENT ANALYSES VOLUMES FOR SUBCOMPARTMENT MODE L
DCPP UNIT 1 & 2 FSAR UPDATE Sheet 1 of 3 Revision 23 December 2016 TABLE 6.2-63 CONTAINMENT SUBCOMPARTMENT ANALYSES SUBCOMPARTMENT MODEL FLOW PATH CHARACTERISTICS
DCPP UNIT 1 & 2 FSAR UPDATE Sheet 2 of 3 Revision 23 December 2016 TABLE 6.2-63 CONTAINMENT SUBCOMPARTMENT ANALYSES SUBCOMPARTMENT MODEL FLOW PATH CHARACTERISTICS
DCPP UNIT 1 & 2 FSAR UPDATE Sheet 3 of 3 Revision 23 December 2016 TABLE 6.2-63 CONTAINMENT SUBCOMPARTMENT ANALYSES SUBCOMPARTMENT MODEL FLOW PATH CHARACTERISTICS
"~ #~ ~!~#~$!"! "#"#~$ $~ ~~~~~~~~~~!~"~!~#~#~$
DCPP UNIT 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 6.2-64 CONTAINMENT SUBCOMPARTMENT ANALYSES LOCA PEAK DIFFERENTIAL PRESSURES Break Differential Pressure Across (Node Numbers)
Design Differential Pressure, psi Calculated Peak Differential Pressure, psi Calculated Absolute Pressure At Time of Peak Differential, psia Figure Pressurizer Surge Line at the RCS Hot Leg Connection Loop Compartment Crane Wall (Node 3 to 12) 15.0 5.90 21.87 6.2-66 Pressurizer Surge Line at the RCS Hot Leg Connection Loop Compartment Operating Deck Elev. 140, inside Crane Wall (Node 3 to 7) 15.0 5.86 21.87 6.2-72 Pressurizer Surge Line at the RCS Hot Leg Connection Loop Compartment Floor / Lower reactor cavity (Node 3 to 33) 6.0 5.80 21.87 6.2-74 Pressurizer Surge Line at the RCS Hot Leg Connection Reactor Cavity Primary Shield Wall (Node 3 to 31) 6.0 4.93 21.87 6.2-73 Pressurizer Spray Line at the Top of the Pressurizer Pressurizer Enclosure Pressurizer Enclosure Wall (Node 38 to 7) 4.0 0.96 16.95 6.2-68 DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.2-65 CONTAINMENT SUBCOMPARTMENT ANALYSES MAIN STEAM LINE BREAK OUTSIDE THE CRANE WALL AT THE CONTAINMENT PENETRATION MASS AND ENERGY RELEASES Revision 23 December 2016 Time (sec) Forward Flow (lbm/sec)
Enthalpy (Btu/lbm) Energy Rate (Btu/sec)
Time (sec) Reverse Flow (lbm/sec) Enthalpy (Btu/lbm) Energy Rate (Btu/sec)
Time (sec) Total Flow (lbm/sec) Total Energy Rate (Btu/sec) 0.0 6203.4 1192.2 7.396x10 6 0.0 7740.4 1192.2 9.228x10 6 0.0 13943.8 16.624x10 6 0.096 6203.4 1192.2 7.396x10 6 7740.4 1192.2 9.228x10 6 0.096 13943.8 16.624x10 6 0.097 2914.8 1192.2 3.475x10 6 7740.4 1192.2 9.228x10 6 0.097 10655.2 12.703x10 6 0.964 2914.8 1192.2 3.475x10 6 7740.4 1192.2 9.228x10 6 0.964 10655.2 12.703x10 6 0.965 10271.2 571.5 5.870x10 6 7740.4 1192.2 9.228x10 6 0.965 18011.6 15.098x10 6 10271.2 571.5 5.870x10 61.391 7740.4 1192.2 9.228x10 6 1.391 18011.6 15.098x10 6 10271.2 571.5 5.870x10 61.392 27275.7 571.5 15.588x10 6 1.392 37546.9 21.458x10 6 DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.2-66 CONTAINMENT SUBCOMPARTMENT ANALYSES MAIN FEEDWATER LINE BREAK INSIDE THE STEAM GENERATOR ENCLOSURE AT THE INLET TO THE STEAM GENERATOR MASS AND ENERGY RELEASED Revision 23 December 2016 Time (sec) Reverse Flow (lbm/sec)
Enthalpy (Btu/lbm) Energy Rate (Btu/sec)
Time (sec) Forward Flow (lbm/sec) Enthalpy (Btu/lbm) Energy Rate (Btu/sec)
Time (sec) Total Flow (lbm/sec) Total Energy Rate (Btu/sec) 0.0 8920.1 545.6 4.867x10 6 0.0 5315.5 413.9 2.200x10 6 0.0 14235.6 7.067x10 6 8920.1 545.6 4.867x10 6 0.268 5315.5 413.9 2.200x10 6 0.268 14235.6 7.067x10 6 8920.1 545.6 4.867x10 6 0.269 5315.5 413.9 2.200x10 6 0.269 14235.6 7.067x10 6 18.60 8920.1 545.6 4.867x10 6 5315.5 413.9 2.200x10 6 18.60 14235.6 7.067x10 6 18.61 0.0 NA 0.0 5315.5 413.9 2.200x10 6 18.61 5315.5 2.200x10 6 DCPP UNIT 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 6.2-67 CONTAINMENT SUBCOMPARTMENT ANALYSES Non-LOCA PEAK DIFFERENTIAL PRESSURES Break Differential Pressure Across (Node Numbers)
Design Differential Pressure, psi Calculated Peak Differential Pressure, psi Calculated Absolute Pressure At Time of Peak Differential, psia Figure Main Feedwater Line Break Inside the Steam Generator Enclosure Steam Generator Enclosure Steam Generator Shield Wall (Elev.
152 to 157) (Node 36 to 7) 20.0 19.54 33.24 6.2-70 Main Steam Line Break Outside the Crane Wall Crane Wall (Node 10 to 1) 15.0 8.43 24.43 6.2-71 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.3-1 Sheet 1 of 3 Revision 22 May 2015 EMERGENCY CORE COOLING SYSTEM COMPONENT PARAMETERS Accumulators Number (per unit) 4 Design Pressure, psig 700 Design Temperature, °F 300 Operating Temperature, °F 50-150 Normal Operating Pressure, psig 621.5 Minimum Operating Pressure, psig (c) 579 Total Volume, ft 3 1,350 each Nominal Water Volume, ft 850 Boric Acid Concentration Nominal, ppm 2,350 Minimum, ppm 2,200 Relief Valve Setpoint, psig 700 Centrifugal Charging Pumps (CCP1 and 2) (Design parameters for these pumps are given in Table 9.3-5.)
Safety Injection Pumps Number (per unit) 2 Design Pressure, psig 1,700 Design Temperature, °F 300 Design Flowrate, gpm 425 Design Head, ft 2,500 Max Flowrate, gpm 675 Head at Max Flowrate, ft 1,500 Discharge Pressure at Shutoff Head, psig 1,520 Motor Rating, hp (a) 400 Residual Heat Removal Pumps (Design parameters for these pumps are given in
Table 5.5-10)
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.3-1 Sheet 2 of 3 Revision 22 May 2015 Residual Heat Exchangers (Design parameters for these heat exchangers are given in Table 5.5-10)
Refueling Water Storage Tank Number (per unit)
1 Total available tank volume (includes only usable
volume)(b), gal 450,000 Minimum Technical Specifications
required volume (includes usable
and unusable volume), gal 455,300 Accident analysis volume (assumed) 350,000 Boron Concentration, ppm 2300-2500 Design Pressure, psig Atmospheric Operating Pressure, psig Atmospheric Design Temperature, ºF 100 Material Austenitic
stainless
steel with
reinforced
concrete shroud Valves (1) All Motor-Operated Valves That Must
Function on Safety
Injection ("S") Signal (a) Up to and including 8
inches (excluding SI-
8805 A&B, CVCS-8107 and
CVCS-8108) Maximum opening or closing time, sec 10 (b) CVCS-8107 and CVCS 8108 Maximum opening and closing time, sec 14 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.3-1 Sheet 3 of 3 Revision 22 May 2015 (c) SI-8805 A&B Maximum opening or closing time, sec 11 (d) Over 8 inches Minimum opening or closing rate, in./min 60 (2) All Other Motor-Operated Gate Valves Up to and
Including 8 Inches Minimum opening or closing rate, in./min 12 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED (3) Original purchase specification leakage criteria. Inservice leakage requirements are specified in the valve Inservice Testing Program.
(a) Conventional globe valves Disk leakage, cc/hr/in. of nominal pipe size 3 Backseat leakage (when open), cc/hr/in. of stem diameter 1 (b) Gate valves Disk leakage, cc/hr/in. of nominal pipe size 3 Backseat leakage (when open), cc/hr/in. of
stem diameter 1 (c) Check valves Disk leakage, cc/hr/in. of nominal pipe size 3 (d) Diaphragm valves Disk leakage None (e) Pressure relief Disk leakage, cc/hr/in. of nominal pipe size 3 (f) Accumulator check valves Disk leakage, cc/hr/in. of nominal pipe size 3 (a) 1.15 service factor not included.
(b) Usable volume includes the water above the outlet pipe. Unusable water includes the water below the outlet.
(c) This minimum SI accumulator pressure is the value that is used in the accident analysis in Chapter
- 15. (Note that more conservative values may appear in other documents such as Technical Specifications, operating procedures, etc.)
DCPP UNITS 1 & 2 FSAR UPDATE Revision 12 September 1998 TABLE 6.3-2 EMERGENCY CORE COOLING SYSTEM DESIGN CODE REQUIREMENTS Component Code Accumulators ASME B&PV,Section III (a) Class C
Refueling Water Storage Tank Valves AWWA D100(c) USAS B16.5, MSS-SP-66 and
ASME B&PV,Section III (a)
Piping (b) - Design Class I portions (excluding Code Class A and @)
ANSI B31.7 - Design Class I, Code Classes A and @
portions and Design Class II portions
Pumps Charging ASME B&PV,Section III (a) Residual heat removal ASME B&PV,Section III (a) Safety injection ASME B&PV,Section III (a)
(a) Draft Code November 1968 Edition.
(b) See Q-List (Reference 8 of Section 3.2) for piping classification.
(c) ASME B&PV Code,Section VIII, allowable stresses used for design.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.3-3 Sheet 1 of 2 Revision 12 September 1998 MATERIALS OF CONSTRUCTION EMERGENCY CORE COOLING SYSTEM COMPONENTS COMPONENT MATERIAL
Accumulators Carbon steel, clad with austenitic stainless steel
Refueling Water Storage Tank Austenitic stainless steel with reinforced concrete shroud
Pumps (parts in contact with coolants) Centrifugal charging Austenitic stainless steel Safety injection Martensitic stainless steel Residual heat removal Austenitic stainless steel or equivalent corrosion-resistant material
Residual Heat Exchangers Shell Carbon steel Shell end cap Carbon steel Tubes Austenitic stainless steel Channel Austenitic stainless steel Channel cover Austenitic stainless steel Tube sheet Forged carbon steel with austenitic stainless steel weld overlay
Valves Motor-operated valves containing radioactive fluids and pressure-containing parts
Austenitic stainless steel or equivalent
Body-to-bonnet bolting and nuts Low-alloy steel, austenitic stainless steel, or 17-4PH stainless
Seating surfaces Stellite No. 6 or equivalent
Stems Austenitic stainless steel or 17-4PH stainless
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.3-3 Sheet 2 of 2 Revision 12 September 1998 COMPONENT MATERIAL
Motor-operated Valves Containing Non-radioactive, Boron-free Fluids
Body, bonnet, and flange Carbon steel
Stems Corrosion resistant steel
Diaphragm Austenitic stainless steel
Parts contacting borated water Austenitic stainless steel
Clapper arm shaft Corrosion resistant steel
Relief Valves
Stainless steel bodies Stainless steel
Carbon steel bodies Carbon steel
All nozzles, disks, spindles, and guides Austenitic stainless steel
Bonnets for stainless steel valves without a balancing
bellows Stainless steel
All other bonnets Carbon steel
Piping All piping in contact with borated water Austenitic stainless steel
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.3-5 Sheet 1 of 6 SAFETY INJECTION TO RECIRCULATION MODE; SEQUENCE AND TIMING OF MANUAL CHANGEOVER Revision 20 November 2011 Actuation Time for Total Time Operation Elapsed Time Action Status Item sec sec min LOCA 0:00
Safety injection signal-all RHR, SI and charging pumps (CCP1 and CCP2) in operation 0:20 Spray initiation 0:40
RWST low level alarm and RHR pumps trip; initiate
recirculation changeover 17:56 IMPLEMENT - Appendix EE 0 10 18:06 CUT IN - Series Contactors **
8974A, 8809A, 8982A, 8982B, 8974B, 8809B 0 10 18:16 VERIFY - Safety injection signal reset **
5 10 18:26 VERIFY - Containment isolation Phase A and Phase B reset ** 5 10 18:36 CHECK - Both ASW pumps running 0 5 18:41 VERIFY - CCW heat exchanger saltwater inlet valves open **
FCV-602(603) (Note 3) 5 10 18:51 VERIFY - CCW heat exchanger CCW outlet valves open **
FCV-430(431) (Note 3) 5 10 19:01 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.3-5 Sheet 2 of 6 Revision 20 November 2011 OPEN - Component cooling water to RHR heat exchanger 2 ** FCV-364 10 15 19:16 OPEN - Component cooling water to RHR heat exchanger 1 ** FCV-365 10 15 19:31 STOP - Centrifugal charging pump CCP3 **
5 10 19:41 PA Announcement **
0 10 19:51 Dispatch Operators to locally close breakers for 8976 and 8980 **
52-1H/2H-20 52-1F/2F-31 0 20 20:11 VERIFY - RHR Pump 2 stopped 0 5 18:11 CLOSE - RHR pump suction valve from RWST 8700B 120 125 20:16 VERIFY - RHR Pump 1 stopped 0 5 20:21 CLOSE - RHR pump suction valve from RWST 8700A 120 125 22:26 CLOSE - RHR crosstie isolation valves 8716A 8716B 20 20 30 22:56 CHECK - RCS Pressure less than 1500 PSIG 0 5 23:01 CLOSE - SI pump miniflow block valves 8974A 8974B 10 10 20 23:21 CLOSE - CCP recirculation valves 8105 8106 10 10 20 23:41 CHECK - recirculation sump level
> 92.0' LI-940 LI-941 5 5 10 23:51 CHECK - RHR pump 2 stopped 0 5 23:56 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.3-5 Sheet 3 of 6 Revision 20 November 2011 CHECK - RHR P2 suction from RWST closed 8700B 0 5 24:01 OPEN - RHR P2 suction from sump 8982B 30 35 24:36 VERIFY - RHR HX 1-2 in Service per App. EE 5 5 24:41 START - RHR-P2 RHR-P2 1 15 24:56 OPEN - SI pumps suction from RHR HX2 8804B 20 25 25:21 CHECK - RCS Pressure less than 1500 PSIG 0 5 25:26 VERIFY - SI pumps running 0 5 25:31 CHECK - RHR Pump 2 motor current less than 57 amps AND stable 0 5 25:36 OPEN - Cross-connect line from SI pump 1 to charging pumps (CCP1 and CCP2) 8807A 8807B 20 20 30 26:06 VERIFY - CCP1 and CCP2 running Changeover of a single train complete at 26:11 0 5 26:11 CHECK - RHR pump 1 stopped 0 5 26:16 CHECK - RHR P1 suction from RWST closed 8700A 0 5 26:21 OPEN - RHR P1 suction from sump 8982A 30 35 26:56 VERIFY - RHR HX 1 in Service per App. EE 5 5 27:01 START - RHR P1 RHR-P1 1 15 27:16 OPEN - SI pump suction from RHR HX1 Changeover of both trains complete at 27:41
8804A 20 25 27:41 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.3-5 Sheet 4 of 6 Revision 20 November 2011 CHECK- RCS Pressure less than 1500 PSIG 0 5 27:46 VERIFY - SI pumps running 0 5 27:51 CHECK - RHR Pump 1 motor current less than 57 amps AND stable 0 5 27:56 CHECK - at least one RHR pump running RHR Pump 0 5 28:01 CLOSE - Charging pump suction RWST isolation 8805A 8805B 11 11 21 28:22 CLOSE - SI pump suction RWST isolation 8976 20 25 28:47 CLOSE - RHR pump suction from RWST 8980 25 30 29:17 Receive RWST low-low level
alarm 33:25 CHECK - both RHR pumps running 0 5 33:30 CHECK - PK01-18, Containment Spray Actuated On OR Cont. Pressure greater than 22 PSIG 0 10 33:40 CHECK - RWST level less than 4%
LI-920 LI-921 LI-922 5 5
5 15 33:55 RESET - Containment spray 5 10 34:05 STOP - Containment Spray (CS) pumps1 & 2 CS-P1 CS-P2 1 1 5 34:10 CLOSE - CS pump discharge to spray header valves System in recirculation mode 9001A 9001B 10 10 20 34:30 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.3-5 Sheet 5 of 6 Revision 20 November 2011 CHECK - Cont. Pressure greater than 22 PSIG 0 5 34:35 VERIFY - Both RHR trains in service 0 5 34:40 CLOSE - RCS from RHR-P1 Cold leg injection terminated 8809A 20 25 35:05 OPEN - RHR-P1 to spray header Recirculation sump to spray header 9003A 15 20 35:25 Notes to Table 6.3-5
- 1. Actuation Time: The estimated actuation time for a component to complete its function. For va lves, this is the maximum expected stroke time. In some cases, where the component is already expected to be in the desired position, no actuation time is added.
- 2. Time for Operation: The actuation time pl us the estimated operator action, if applicable.
- 3. If the valves are not already in the desired position the analys is assumes they are stroked concurrently while the crew cont inues in the procedure, since subsequent checks of ASW/CCW alignment are m ade at decision points further on in the procedure.
- All these steps from Appendix EE of Procedur e EOP E-1.3 can be performed in parallel wi th the steps that follow since they a dd up to 125 seconds, which is less than the 295 seconds (from 18:06 to 23:01) needed bef ore we close 8974A/B (this is the first step that depends on a step from Appendix EE).
The following assumptions are used for Table 6.3-5:
- 1. Double-ended reactor coolant pump suction LOCA.
- 2. Maximum safety features implemented:
- a. Pumps at flow limited by pipe friction 2 RHR pumps: 7300 gpm total during injection 1 RHR pump: 4600 gpm for 5 minutes during changeover, when assumi ng a single failure of one RHR pump to trip on low RWST leve l 2 SI pumps: 900 gpm total 2 Charging pumps (CCP1 and CCP2): 900 gpm total
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.3-5 Sheet 6 of 6 Revision 20 November 2011 b. 2 Containment spray pumps: 6550 gpm total duri ng injection, assuming 20 psig in containment 6800 gpm total during changeover, assuming 0 psig in containment
- c. Allowable 100 gpm leakage penalty from RWST
- 3. Refueling water storage tank.
- a. Maximum outflow: 15,750 gpm based on 2a and 2b during injection 8,700 gpm based on 2a and 2b during changeover, no single failure
- b. Volume available: 404,511 gal.
- c. Low-level alarm volume: 116,812 gal.
- d. Low-low level alarm volume: 0 gal.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 12 September 1998 TABLE 6.3-6 NORMAL OPERATING STATUS OF EMERGENCY CORE COOLING SYSTEM COMPONENTS FOR CORE COOLING Number of Safety Injection Pumps Operable 2 Number of Charging Pumps Operable 2 Number of Residual Heat Removal Pumps Operable 2 Number of Residual Heat Exchangers Operable 2 Refueling Water Storage Tank Available Volume, gal 350,000 Boron Concentration in Refueling Water Storage Tank, Maximum, ppm 2,500 Minimum, ppm 2,300 Boron Concentration in Accumulators, Maximum, ppm 2,500 Minimum, ppm 2,200 Number of Accumulators 4 Minimum Accumulator Pressure, psig (a) 579 Nominal Accumulator Water Volume, ft 3 850 System Valves, Interlocks, and Piping Required for the Above Components which are Operable All
(a) This minimum SI accumula tor pressure is the value that is used in the accident analysis in Chapter 15. (Note that more conservative values may appear in other documents such as Technical Specif ications, operating procedures, etc.)
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.3-7 Sheet 1 of 4 Revision 23 December 2016 SEQUENCE AND DELAY TIMES FOR STARTUP OF ECCS CHANGES ON THIS TABLE APPLY TO BOTH UNITS 1 & 2 Delay, sec References
Accident Actuation Signal(s)
Action Sequence (Subsystem or Component)
(h)
(i)
(j)
Design Performance Minimum ECCS Performance Assumed in Analysis
Section FSAR
Figures
Tables 1. Major Reactor 15.4.1 Coolant System Rupture (LOCA) a. Injection phase (g) Accumulator tank (g) (g) (g) 4 tanks, each with 850 ft 3 of borated water @ 600 psig Three tanks injecting into RCS;
one injecting into broken loop 6.3 (a) Containment isolation valves 1 1 10 Double barrier; fast automatic valve closure upon receipt of CIS A single active failure is
allowable 6.2.4 6.2-17, 6.2-18 & 6.2-19 8.3-4 (b) (d) ECCS required valves (k) (k) See Table 6.3-1 Rapid reliable system alignment or
isolation A single active failure is
allowable 6.3.2 7.3-22, 7.3-33 8.3-4 (b) (d) Centrifugal charging pumps -5 15 4-1/2 Two centrifugal charging pumps supply borated water into a single
injection flowpath splitting into 4
cold leg injection lines One pump required at
assumed flow 6.3.2, 9.3.4 7.3-4 8.3-4 (b) (d) Safety injection pumps Two pumps inject via a single path splitting into 4 cold leg injection
lines One pump delivering at
assumed flow 6.3.2 3.2-9 (b) (d) Residual heat removal pumps Two pumps inject into 4 cold legs, via 2 lines that each split into 2
cold leg injection lines One pump delivering at
assumed flow 6.3.2, 5.5.6 3.2-9 (b) (d) Component cooling water pumps 10/10/14 20/20/24 4-1/2 Two flowpaths; each 11,500 gpm
@ 130 ft One flowpath required at
assumed flow 9.2.2 7.3-7 8.3-4 (e) Auxiliary feed-water pumps 30/35 40/45 5 Two flowpaths; each 800 gpm
@ 2350 ft Not modeled for LBLOCA 6.5.2 7.3-8 8.3-4 (b) (d) Auxiliary salt-water pumps 14/14 24/24 5 Two flowpaths; each 11,000 gpm
@ 115 ft One flowpath required at
assumed flow 9.2.7 7.3-5 8.3-4 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.3-7 Sheet 2 of 4 CHANGES ON THIS TABLE APPLY TO BOTH UNTIS 1 & 2 Revision 23 December 2016 Delay, sec References
Accident Actuation Signal(s)
Action Sequence (Subsystem or Component)
(h)
(i)
(j)
Design Performance Minimum ECCS Performance Assumed in Analysis
Section FSAR
Figures
Tables (c) (l) Containment
spray pumps
Pump 1
Pump 2
26 26
36 36
1.7 1.7 Two flowpaths; each 2600 gpm
@ 450 ft One flowpath required at
assumed flow 6.2.2 7.3-11 8.3-4 b. Recirculation
phase (f) Operating personnel shift
system alignment
from injection
phase (Total switchover time is
approximately 10 min.
See 6.3.2) (Design performance for ECCS and related equipment as described in
1a above) A single failure is allowable 6.3.3 6.3-5 2. Major
Seccondary System Rupture (b) Action sequence similar to 1a
above. Operation
of ESF required.
Valves isolate
feedwater & steam Same as 1a above
Same as 1a above Same as 1a above with these further notes: Accumulator and
low head injection required only
in the severe cases. Since no
RCS rupture has occurred, all
four accumulators are functional 15.4.2 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.3-7 Sheet 3 of 4 CHANGES ON THIS TABLE APPLY TO BOTH UNTIS 1 & 2 Revision 23 December 2016
Accident Actuation Signal(s)
Action Sequence (Subsystem or Component)
(h)
(i)
(j)
Design Performance Minimum ECCS Performance Assumed in Analysis
Section FSAR
Figures
Tables 3. Steam Generator Tube Rupture Low pressurizer
pressure Same as 1a
above although no
containment
spray.
Additionally, automatic isolation
of individual steam
generator
blowdown valve
occurs due to
SGBD liquid
radiation monitor.
Injection and
charging flow
regulated to
maintain visible
pressurizer water
level. Auxiliary
feedwater to
affected SG
manually isolated.
Pressurizer reliefs
operated to
reduce RCS
pressure to less than the ruptured SG pressure.
Same as 1a above with
additional isolation done
within 30 minutes Same as 1a above Same as 1a above (but all four accumulators assumed
functional, AFW available). 15.4.3
pressure or high containmen
t pressure 15.3.1 a. Injection phase Same as 1a above Same as 1a
above Same as 1a above Same as 1a above Same as 1a above
- b. Recirculation Same as 1a above Same as 1a
above Same as 1a above Same as 1a above Same as 1a above
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.3-7 Sheet 4 of 4 CHANGES ON THIS TABLE APPLY TO BOTH UNTIS 1 & 2 Revision 23 December 2016 (a) Initiated by means of containment isolation signal, which occurs on containment high pressure (2 of 3, Phase A), containment high-high pressure (2 of 4, Phase B), low pressurizer pressure (Phase A) or on manual actuation (Phase A and/or B).
(b) Safety injection signal (SIS) actuates on any of the following: Low pressurizer pressure, high containment pressure, low s teamline pressure, or manual actuation.
(c) Containment spray signal, which occurs on containment high-high pressure (2 of 4), or manual actuation.
(d) Emergency diesel loading sequencer loads the diesel in accordance with the sequence shown in Tables 8.3-2 and 8.3-4. Also see Figures 8.3-9, 8.3-10, 8.3-11, and 8.3-16.
(e) Auxiliary feedwater autostart signal, which occurs with an SIS, SG low-low level, or tripping of both main feedwater pumps.
(f) Water level indication and alarms on the refueling water storage tank and in the containment sump provide ample warning to terminate the injection mode and begin the recirculation mode while the operating pumps still have adequate net positive suction head. Manual switchover by operating personnel changes the ECCS from injection to recirculation mode.
(g) All valves between the accumulators and the RCS are required to be open in Modes 1, 2, and 3; consequently, the accumulators inject as soon as the RCS pressure drops below the pressure (600 psia) of the accumulators.
(h) Electrical and instrumentation delay time after "S" signal with main generator power or offsite power available.
(i) Electrical and instrumentation delay time after "S" signal using diesel generator.
(j) Equipment startup time after receipt of signal.
(k) These delay times vary.
(l) DIT 68029121-1-0 and Westinghouse containment integrity analysis CN-CRA-14-6 and CN-CRA-14-7 established the limiting containment spray delay time to be based on the S signal rather than the P signal.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 12 September 1998 TABLE 6.3-8 EMERGENCY CORE COOLING SYSTEM SHARED FUNCTIONS EVALUATION Component Normal Operating Arrangement Accident Arrangement Refueling water
storage tank Lined up to suction of safety injection and residual heat removal pumps Lined up to suction of centrifugal charging, safety injection, and residual heat removal pumps.
Valves for realignment meet single failure criteria.
Centrifugal charging
pumps Lined up for charging service Lined up to charging injection header. Valves for realignment meet single failure criteria.
Residual heat
removal pumps Lined up to cold legs of reactor coolant piping Lined up to cold legs of reactor coolant piping.
Residual heat
exchangers Lined up for residual heat removal pump operation Lined up for residual heat removal pump operation.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 6.3-9 MAXIMUM POTENTIAL RECIRCULATION LOOP LEAKAGE EXTERNAL TO CONTAINMENT
Items Type of Leakage Control and Unit Leakage Rate Used in the Analysis Leakage to Atmosphere, cc/hr Leakage to Drain Tank, cc/hr
- 1. Residual Heat Removal Pumps Mechanical seal with leakoff - 10 cc/hr seal (a) 20 0
- 2. Safety Injection Pumps Same as residual heat removal pump 40 0
- 3. Charging Pumps Same as residual heat removal pump 40 0
- 4. Flanges:
- a. Pumps
Gasket - adjusted to zero leakage following any test; 10
drops/min/flange used(30 cc/hr). Due to leak tight flanges on
pumps, no leakage to atmosphere is assumed 0
0
- b. Valves bonnet to body (larger than 2 in.) 1200 0 c. Control valves 180 0
- d. Heat exchangers 240 0
- 5. Valves - Stem Leakoffs Backseated double-packing with leakoff - 1 cc/hr in. stem diameter used (see Table 6.3-1) 0 40
- 6. Misc. Small Valves Flanged body-packed stems -1 drop/min used (3 cc/hr). 150 0
- 7. Misc. Large Valves (larger than 2 in.) Double-packing 1 cc/hr/in. stem diameter used 40 0 TOTALS 1910 40
(a) Seals are acceptance tested to essentially zero leakage.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 6.3-10 ECCS RELIEF VALVE DATA
Description
Fluid Discharged Fluid Inlet Temp Normal, °F Set Pressure, psig Back Pressure Constant Psig Build-up Capacity N 2 supply to accumulators N 2 120 700 0 0 1500 scfm SIS pump discharge Water 100 1750 3 50 20 gpm RHR pumps SI line Water 120 600 3 50 400 gpm SI pumps suction header Water 120 220 3 50 20 gpm Accumulator to containment Water or N 2 gas 120 700 0 0 1500 scfm
DCPP UNITS 1 & 2 FSAR UPDATE Revision 18 October 2008 TABLE 6.3-11 NET POSITIVE SUCTION HEADS FOR POST-DBA OPERATIONAL PUMP (a)
Pump Flow and Condition Suction Source Minimum Available NPSH, ft Required NPSH, ft
Water Temp, °F Safety injection 675 gpm runout flow Refueling water
storage tank 31 29 100 max Centrifugal
charging 560 gpm runout flow Refueling water
storage tank 44 24 100 max Residual heat
removal 4500 gpm Refueling water storage tank 27 20 100 max Residual heat
removal 4900 gpm runout flow Containment sump 25 24 Saturated liquid Containment spray 3500 gpm runout flow Refueling water
storage tank 40 19 100 max (a) NPSH conservatively calculated without considering additional static suction head of water in the RWST.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.3-12 Sheet 1 of 2 Revision 16 June 2005 ECCS MOTOR-OPERATED VALVES WITH ELECTRIC POWER REMOVED DURING NORMAL POWER PLANT OPERATION Position Valve Identification
Service Description Normal Operation Injection Phase Cold Leg Recirc Phase Hot Leg Recirc Phase Power Restorable From Control Room 8703 RHR pump discharge to RCS hot leg loops Closed Closed Closed Open No 8802 A, B SIS pump discharge to RCS hot leg loops Closed Closed Closed Open No 8808 A, B, C, D Accumulator isolation Open Open Open Open No 8809 A 8809 B RHR pump discharge to RCS
cold leg loops Open Open Closed (b) Closed Yes Open Open Open Closed Yes
8835 SIS pump discharge to RCS cold leg loops Open Open Open Closed No 8974 A, B SIS pump miniflow Open Open Closed Closed Yes
8976 RWST supply to SIS pumps Open Open Closed Closed No
8980 RWST supply to RHR pumps Open Open Closed Closed No
8982 A, B Containment recirc. sump supply to RHR pumps Closed Closed Open Open Yes
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.3-12 Sheet 2 of 2 Revision 16 June 2005 Position Valve Identification
Service Description Normal Operation Injection Phase Cold Leg Recirc Phase Hot Leg Recirc Phase Power Restorable From Control Room 8992 NaOH spray additive supply Open Open Open Open No 8701 (a) Loop 4 hot leg RHR suction
valve 2 Closed Closed Closed Closed No 8702 (a) Loop 4 hot leg RHR suction
valve 1 Closed Closed Closed Closed No
(a) Valve required to function for a normal RHR cooldown not for ECCS.
(b) Closed if containment spray is required.
DCPP UNITS 1 & 2 FSAR UPDATE Table 6.3-13 Sheet 1 of 3 Revision 22 May 2015 SINGLE ACTIVE FAILURE ANALYSIS FOR EMERGENCY CORE COOLING SYSTEM COMPONENTS SHORT-TERM PHASE Component Malfunction Comments A. Accumulator Deliver to broken Totally passive system with one accumulator per loop. loop Evaluation based on one spilling accumulator.
B. Pump 1. Centrifugal charging Fails to start Two provided; evaluation based on operation of one.
- 2. Safety injection Fails to start Two provided; evaluation based on operation of one.
- 3. Residual heat removal Fails to start Two provided; evaluation based on operation of one.
- 4. Residual heat removal Fails to trip on Operator trips pump locally at the breaker. RWST Low Level C. Automatically Operated Valves
- 1. Charging injection isolation
- a. Inlet Fails to open Two parallel lines; one valve in either line required to open
- b. Outlet Fails to open Two parallel lines; one valve in either line required to open.
- 2. Centrifugal Charging Pumps
- a. Suction line from refueling Fails to open Two parallel lines; only one valve in either line is water storage tank required to open.
DCPP UNITS 1 & 2 FSAR UPDATE Table 6.3-13 Page 2 of 3 Revision 22 May 2015 SHORT-TERM PHASE (Continued)
Component Malfunction Comments b. Discharge line to the Fails to close Two valves in series; only one valve required to close. normal charging path
- c. Minimum flow line Fails to close Two valves in series; only one valve required to close.
- d. Suction line from Fails to close Two valves in series; only one valve required to close. volume control tank LONG-TERM PHASE A. Valves Operated from Control Room for Recirculation
- 1. Containment sump recirculation Fails to open Two parallel lines; only one valve in either line is isolation required to open.
- 2. Residual heat removal pump Fails to close Check valve in series with gate valve, operation of suction line from refueling only one valve required. water storage tank
- 3. Safety injection pump suction Fails to close Check valve in series with gate valve, operation of line from refueling water only one valve required. storage tank
- 4. Centrifugal charging pump Fails to close Check valve in series with two parallel gate valves. (CCP1 and 2) suction line from Operation of either the check valve or the gate valve refueling water storage tank required.
DCPP UNITS 1 & 2 FSAR UPDATE Table 6.3-13 Page 3 of 3 Revision 22 May 2015 LONG-TERM PHASE (Continued)
Component Malfunction Comments 5. Safety injection pump suction Fails to open Separate and independent high head injection path taking line at discharge of residual suction from discharge of residual heat exchanger No. 1. A heat exchanger No. 2 crossover line allows flow from one heat exchanger to reach both safety injection and charging pumps, if necessary.
- 6. Centrifugal charging pump Fails to open Separate and independent high head injection path taking (CCP1 and CCP2) suction line at suction from discharge of residual heat exchanger No. 2. A discharge of residual heat exchanger No. 1 crossover line allows flow from one heat exchanger to reach both safety injection and charging pumps, if necessary.
- 7. Centrifugal charging pump (CCP1 Fails to open Two parallel lines; only one valve in either line And CCP2) crossover line to safety is required to open. injection pump suction B. Pumps 1. Residual heat removal pump Fails to start Two provided. Evaluation based on operation of one.
- 2. Charging pump Fails to operate Same as short-term phase
- 3. Safety injection pumps Fails to operate Same as short-term phase
DCPP UNITS 1 & 2 FSAR UPDATE Table 6.3-14 Sheet 1 of 1 Revision 22 May 2015 EMERGENCY CORE COOLING SYSTEM RECIRCULATION PIPING PASSIVE FAILURE ANALYSIS LONG-TERM PHASE Flow Path Indication of Loss of Flow Path Alternate Flow Path Low Head Recirculation From containment sump to low head injection header via the residual heat removal pumps and the residual heat exchangers Reduced flow in the discharge line from one of the residual heat exchangers (one flow monitor in each discharge line).
Accumulation of water in a residual heat removal pump compartment or auxiliary building sump.
Via the independent identical low head flow path utilizing the second residual heat exchanger.
High Head Recirculation From containment sump to the high head injection header via residual heat removal pump residual heat exchanger, and the high head injection pumps
- 1) Increasing activity of the air exhausted from the RHR heat exchanger rooms or in the plant vent.
- 2) Accumulation of water in a residual heat removal pump compartment or the auxiliary building sump
- 3) Increasing ESF pump room temperature
- 4) Reduced ECCS flow rates From containment sump to the high head injection headers via alternate residual heat removal pump, residual heat exchanger, and the alternate high head charging pump
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 6.5-1 CRITERIA FOR AUXILIARY FEEDWATER SYSTEM DESIGN BASIS CONDITIONS Page 1 of 2 Condition or Transient Classification (a) Criteria (a) Additional Design Criteria Loss of Normal Feedwater (Refer to Section 15.2.8)
Condition II AFW capable of removing stored and residual heat to prevent pressurizer liquid
relief.(b) AFW automatically initiated on low-low SG
level.
Loss of Offsite Power to the
Station Auxiliaries (Refer to Section 15.2.9) Condition II AFW capable of removing stored and residual heat to prevent pressurizer liquid
relief. AFW automatically initiated on low-low SG
level.
Steamline Rupture (Mass &
Energy Release - Refer to
Section 6.2D.3) Condition IV N/A - not an AFW system design requirement.
AFW flow maximized for mass and energy
release.
Major Rupture of a Main
Feedwater Pipe (Refer to
Section 15.4.2.2) Condition IV AFW to provide assured source of feedwater to SGs for decay heat removal.
AFW flow to SGs assumed in 10 minutes
after reactor trip.
Major Rupture of a Main Feedwater Pipe for Pressurizer Filling (Refer to Section 15.4.2.4)
Condition IV AFW to provide assured source of feedwater to SGs for decay heat removal.
AFW flow to SGs assumed within 1 minute after reactor trip Loss of all ac power
N/A AFW to provide assured source of
feedwater to SGs for decay heat removal
independent of ac power.
AFW system turbine-driven pump train
independent of ac power.
Small-Break Loss of Coolant (Refer to Section 15.3.1) Condition III AFW provide 390 gpm.(d) N/A
Natural Circulation Cooldown Same as LONF AFW system provides an assured source of feedwater to SGs to prevent reactor
vessel head voiding Unit 1 - Hot Standby 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> cooldown
@ 25 °F per hour.
Unit 2 - Hot Standby 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
cooldown @ 50 °F per hour.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 6.5-1 CRITERIA FOR AUXILIARY FEEDWATER SYSTEM DESIGN BASIS CONDITIONS Page 2 of 2 (a) Ref: ANS N18.2 (This information provided for those transients analyzed in Chapter 15.)
(b) A better-estimate analysis has also been performed to demonstrate that the pressurizer does not fill with a single motor-driven auxiliary feedwater pump feeding 2 SGs a total of 390 gpm.
(c) Refer to Section 15.5 for 10 CFR 100 acceptance criteria for accident analysis dose consequences.
(d) An AFW flowrate of 97.5 gpm per SG is assumed in NOTRUMP based on 390 gpm divided evenly among 4 SGs since NOTRUMP cannot explicitly model asymmetric flow.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.5-2 Sheet 1 of 2 Revision 23 December 2016
SUMMARY
OF ASSUMPTIONS AFW SYSTEM DESIGN VERIFICATION
Transient Loss of Normal Feedwater (Loss of Offsite Power)
Natural Circulation Cooldown Major Rupture of a Main Feedwater Pipe Major Steam Line Break (b) (Containment) Small Break Loss of Coolant Accident a. Max NSSS power 102% of 3425 MW t 102% of 3411MW t 102% of 3425 MW t 102% of 3425 MW t 102% of 3411 MW t b. Time delay from event to Rx trip (Refer to Table 15.2-1) 2 sec (Refer to Table 15.4-8) Variable 4.7 sec c. AFW system actuation signal/time delay for AFW system flow Low-low SG level
1 minute Low-low SG Level
1 minute Low-low SG level (Refer to Table 15.4-8)
Assumed immediately
0 sec (no delay) Low pressurizer pressure SI
signal / 60 sec
- d. SG water level at time of reactor
trip. Low-low SG level
8 % narrow range span (NRS) Same as LOOP Low-low SG level 0% NRS N/A N/A
- e. Decay heat Figure 15.1-7 Figure 15.1-7 Figure 15.1-7 Figure 15.1-7 Figure 15.1-7
- f. AFW pump design pressure 1102 psig 1112 psia 1102 psig N/A 1130 psig
- g. Min. No. of SGs that must receive AFW flow 4 of 4 Same as LONF/LOOP 2 of 4 (Section 15.4.2.2) 3 of 4 (Section 15.4.2.4)
N/A 4 of 4 (c) h. Maximum AFW
temperature 100 °F 100 °F 100 °F 100 °F 100 °F DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.5-2 Sheet 2 of 2 Revision 23 December 2016
Transient Loss of Normal Feedwater (Loss of Offsite Power)
Natural Circulation Cooldown Major Rupture of a Main Feedwater Pipe Major Steam Line Break (b) (Containment) Small Break Loss of Coolant Accident i. Operator action None N/A 10 minutes to isolate the faulted SG 10 minutes to isolate
the faulted SG None j. AFW purge volume/
temperature 113 ft 3per loop/435° F 113 ft 3 per loop/435
°F 113 ft 3 per loop/435° F (Section 15.4.2.2),
425°F (Section 15.4.2.4) 0.0ft 3/ based on power N/A
- k. Normal blowdown None assumed None assumed None assumed None assumed None assumed
- l. Sensible heat Table 6.5-3 Table 6.5-3 Refer to cooldown N/A Refer to cooldown
- m. Time at standby/time to cooldown to RHR 2 hr/4 hr with offsite power
available (without offsite
power available refer to
Natural Circulation
Cooldown)
Unit 1 - 1 hr/8 hr @
25 °F
Unit 2 - 2 hr/4 hr @
50 °F N/A N/A N/A
- n. AFW flowrate 600 gpm (total) constant (minimum requirement Variable based on
maintaining SG
level at lower NR
level tap at SG
backpressure For Section 15.4.2.2, 390 gpm (total)
constant (after 10 minutes)(a) (minimum requirement)
For Section 15.4.2.4, refer to Table 15.4-8 569 gpm to 1588 gpm
varying due to faulted
SG pressure changes 390 gpm to 4 SGs (c)
(a) Minimum flow of 175.5 / 214.5 gpm to each of the two steam generators receiving AFW flow.
(b) A rupture of a main steam pipe inside containment does not impose any performance related requirements on the AFW system.
For the accident analysis, AFW flowrates were maximized to increase the mass and energy contributions from the AFW system (Refer to Section 6.2D
.3). (c) 390 gpm to four SGs was assumed to be provided by one motor-driven AFW pump. The approved NOTRUMP model cannot model asym metric flow, therefore the 390 gpm is assumed to be distributed equally among the four SGs
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 6.5-3
SUMMARY
OF SENSIBLE HEAT SOURCES (For Plant Cooldown by AFW system)
Primary Water Sources (initially at emergency safeguards design (ESD) power temperature and inventory)
- RCS fluid
- Pressurizer fluid (liquid and vapor)
Primary Metal Sources (initially at ESD power temperature)
- Reactor coolant piping, pumps and reactor vessel
- Pressurizer
- Steam generator tube metal and tubesheet
- Reactor vessel internals
Secondary Water Sources (initially at ESD power temperature and inventory)
- Steam generator fluid (liquid and vapor)
- Main feedwater purge fluid between steam generator and AFW system piping
Secondary Metal Sources (initially at ESD power temperature)
- All steam generator metal above tubesheet, excluding tubes
Revision 11 November 1996 FIGURE 6.2-10 CONTAINMENT SPRAY PUMP PERFORMANCE CURVE UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 6.2-12 CONTAINMENT SPRAY NOZZLE CUTAWAY UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE (Unit1)39SPRAYNOZZLES(Unit2)Revision 22 May 2015 FIGURE 6.2-14 GENERIC METHODOLOGY COMPARISON OF SPRAY REMOVAL MODEL AND CSE RESULTS (RUN A6)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 12 September 1998 Revision 11 November 1996 FIGURE 6.2-15 CONTAINMENT RECIRCULATION SUMP pH VERSUS TIME AFTER LOCA BEGINS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE FIGURE 6.2-16 CONTAINMENT EQUILI BRIUM ELEMENTAL IODINE PARTITION COEFFICIENT VERSUS TIME FOR MINIMUM SUMP pH CASE (2 ECCS TRAINS AND 1 SPRAY TRAIN)
FOR THREE TEMPERATURES UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 12 September 1998 Revision 11 November 1996 FIGURE 6.2-17 CONTAINMENT ISOLATION SYSTEM UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 6.2-18 (Sheet 1 of 2)
PENETRATION DIAGRAM LEGEND UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 6.2-18 (Sheet 2 of 2)
PENETRATION DIAGRAM LEGEND UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 1 OF 25)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 2 OF 25)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 3 OF 25)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 22 May 2015 Revision 22 May 2015 Revision 22 May 2015 TC D*V*D*V* - UNIT-2 ONLY
- H**I R H R S Y S T E M (C L O S E D S Y S T E M)(I ) RCS LOOP 4 RECIRC. (27)(II ) CONT. SUMP RECIRC. (28)(III) CONT. SUMP RECIRC. (29)# - UNIT-1 ONLY INSIDECONT.OUTSIDE CONT.M M TV TC M III F M TV TC E IICONTAINMENTRECIRC. SUMP PROTECTIVE CHAMBERPROTECTIVECHAMBER M C M M BRWST#CONCENTRICGUARD PIPE I D#TV#M TV A M TV 1MB0MBRCDT L O O P 4 R C STO PRT (RCS)G Revision 20 November 2011 FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 6 OF 25)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE IMB OMB D INSIDECONT.OUTSIDECONT.MISC. EQUIP.DRAIN TK.TV TV,TCRHR HT.EX.PRHR HT.EX.CONTAINMENT SPRAY PUMPS, VALVES AND RELATED EQUIPMENT FROMRWST THIS SYSTEM IS DESIGN CLASS IRCS AND PRTCLOSED SYSTEM TC A I TV C TC B M M D M TV,TC F H G M TV P TC MISC. EQUIP.
DRAIN TK.TC E II IV IIICOOLING COILWATER CHILLER TC TC TC TV S-1 T J S-1 T L S-1 S-1 T I T K0MB1MB TV(I ) CONTAINMENT SPRAY (30)(II ) CONTAINMENT SPRAY (31)(III) CHILLED WATER SUPPLY (82D)(IV ) CHILLED WATER RETURN (83A)
TC **-UNIT-2 ONLY
- Revision 21 September 2013 FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 7 OF 25) UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE OMB IMB 2 2TC C O L D L E G S (R C S)GSIS TEST I MTV BTCOMBIMB INSIDECONT.OUTSIDECONT.SIS INJECTIONPUMPS ATYP. OF 4 IITV ISCCHARGINGPUMPS M DTCTYP. OF 2 MTCTVCHARGINGPUMPS STV C C O L D L E G S (R C S)IIITCTC V V FTOACCUM.T ES-1TV N (I ) SIS COLD INJECTION (33)(II ) SIS COLD INJECTION (34)(III) N SUPPLY TO ACCUM. (51A)
S Revision 20 November 2011FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 8 OF 25) UNIT 1 DIABLO CANYON SITE FSAR UPDATE 2 2TC C O L D L E G S (R C S)GTO SIS TESTSTATION #2 I MTV BTCOMBIMB INSIDECONT.OUTSIDECONT.SIS INJECTIONPUMPS ATYP. OF 4 IITV ISCCHARGINGPUMPS M DTCTYP. OF 2 MTCTVCHARGINGPUMPS STV C C O L D L E G S (R C S)IIITCTC V V FTOACCUM.T ES-1TV N (I ) SIS COLD INJECTION (33)(II ) SIS COLD INJECTION (34)(III) N SUPPLY TO ACCUM. (51A)
S Revision 20 November 2011FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 8A OF 25) UNIT 2 DIABLO CANYON SITE FSAR UPDATE S.O.T A T B T C D PRT G TCFROMRCSREGENERATIVEHEATEXCHANGER I T D VD TVLETDOWNHEATEXCHANGER FROM RESID.HT REMOVALHT. EXCH.OUTLETCVCS (CLOSED SYSTEM)
REGEN.HEATEXCH.0MBTO RCS COLD LEG 3 (ALTERNATE) 0MB 0MBTO RCS COLD LEG 4 (NORMAL)TO RCSPRESSAUX. SPRAYTCTC E II M S M S F TVFROMCHARGING PUMPS (II ) REGEN. HEAT EXCHR.
CHARGING/AUX. SPRAY (36)(I ) LETDOWN LINE REGEN. HEAT EXCHR. TO LETDOWN HEAT EXCHR. (35)
V0MB1MB INSIDECONT.OUTSIDECONT. Revision 21 September 2013 FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 9 OF 25) UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE OMB OMB OMB OMB IMB FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 10 OF 25)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 IMB OMB Revision 11 November 1996 FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 11 OF 25)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 22 May 2015 FROM ACCUMULATORTANK (TYP. 4 LOOPS)FROM SISTEST CONN.FROM RCS COLD LOOP (TYP. 4 LOOPS)
SIS TEST LINETC A T I E PRESSUREINDICATORTVTEST CONN. FROM SI PUMP DISCH. HDR.
T B SC DTVTESTLINE T CTVLIQUID HOLD UP TANKINSIDE CONT.OUTSIDE CONT.( I ) TEST LINE FROM SAFETY INJECTION SYSTEM (51B)
Revision 20 November 2011FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 13 OF 25) UNIT 1DIABLO CANYON SITE FSAR UPDATE F
TO SISTEST STATION
SIS TEST LINETC A T I E PRESSUREINDICATORTVTEST CONN. FROM SI PUMP DISCH. HDR.
T B SC DTVTESTLINE T CTVLIQUID HOLD UP TANKINSIDE CONT.OUTSIDE CONT.(I ) TEST LINE FROM SAFETY INJECTION SYSTEM (51B)
Revision 20 November 2011FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 13A OF 25) UNIT 2 DIABLO CANYON SITE FSAR UPDATE F
Revision 22 May 2015 Revision 11 November 1996 FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 15 OF 25)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE 0MB1MB TC A I TV S-1 T B E TCINSTRUMENT AIR SUPPLY
(>100 PSIG)TO VALVES ANDINSTRUMENTSTO VALVES ANDINSTRUMENTSINSIDE CONT.OUTSIDE CONT.0MB1MB TC C II S-1 D TCINSIDE CONT.OUTSIDE CONT.A I R O U T L E T S SC VTC & TV COMPRESSED AIR SUPPLY (I ) INSTRUMENT AIR HEADER (54)(II ) SERVICE AIR HEADER (56)
FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 16 OF 25)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 IMB OMB OMB IMB Revision 22 May 2015 Revision 11 November 1996 FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 18 OF 25)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 19 OF 25)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 20 OF 25)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE C D HOT LEG 3 HOT LEG 4 M TV A TC B I SISTEST LINE TC SIS PUMP 2INSIDE CONT.OUTSIDE CONT.F G HOT LEG 1 HOT LEG 2 M TV E TC II TC SIS PUMP 1 (I ) SIS PUMP 2 DISCHARGE (75)(II ) SIS PUMP 1 DISCHARGE (77)
Revision 19 Ma y 2010 FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 21 OF 25)
UNIT 1 DIABLO CANYON SITE FSAR UPDATE C D HOT LEG 3 HOT LEG 4 M TV A TC B I TO SISTEST STATION
- 2 TC SIS PUMP 2INSIDE CONT.OUTSIDE CONT.F G HOT LEG 1 HOT LEG 2 M TV E TC II TC SIS PUMP 1 (I ) SIS PUMP 2 DISCHARGE (75)(II ) SIS PUMP 1 DISCHARGE (77)
Revision 19 Ma y 2010 FSAR UPDATE UNIT 2 DIABLO CANYON SITE FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 21A OF 25)
FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 22 OF 25)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008 TCTC TC I B A T TVINSIDECONT.OUTSIDECONT.1MB0MBS-1S-1FIRE WATERSUPPLY HDR.(I ) FIRE WATER SUPPLY (79)
FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 23 OF 25) UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 IMB OMB H 2 2 2DESIGN CLASS IIDESIGN CLASS II M F SV A TC D I SV B TV II SV C TV 2MONITOR III M ERECOMBINERPROVISIONCHPEXHAUSTSYSTEM INSIDECONT.OUTSIDECONT.(I ) CONTAINMENT H MONITOR SUPPLY (52E, 78A)(II ) CONTAINMENT H MO NITOR RETURN (52C, 78B)(III) CONTAINMENT H EXTERNAL RECOMBINER (57, 81)
ALL SYSTEMS ARE DESIGN CLASS I, EXCEPT AS NOTED.
FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 24 OF 25)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 16 June 2005 A I INSIDECONT.OUTSIDECONT.RCSHIGH VOLUMEBELLOWS SENSOR BHYDROLIC ISOLATOR INSTRUMENTATION1MB0MB S C II S-1 S D S-1 S E III F S-1 S G IV S-1 S H POST LOCASAMPLINGSYSTEMREACTORCAVITYSUMP D (I ) REACTOR VESSEL LEVEL INSTRUMENTATION (59E,59F,59G,80E,80F,80G)(II ) POST-LOCA SAMPLING SYSTEM CONT. AIR SUPPLY (82B)(III) POST-LOCA SAMPLING SYSTEM CONT. AIR RETURN (82C)(IV) POST-LOCA SAMPLING SYSTEM REACTOR CAVITY SUMP (82A)
FIGURE 6.2-19 PENETRATION DIAGRAM (SHEET 25 OF 25)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 IMB OMB Revision 11 November 1996 FIGURE 6.2-20 CONTAINMENT HYDROGEN PURGE SYSTEM PURGE STREAM UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 6.2-21 CONTAINMENT HYDROGEN PURGE SYSTEM SUPPLY STREAM UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE CONTAINMENTAIR SCEL-82 SCONTAINMENTAIRRETURN S TV TV TCCONTAINMENTAIR SCEL-83 SCONTAINMENTAIRRETURN S TV TV52E52F78A78BOUTSIDEINSIDECONTAINMENTNOTES:1. VALVES SHOWN FOR NORMAL PLANT OPERATION.2. ALL PIPING AND VALVES ARE DESIGN CLASS I, EXCEPT AS NOTED.3. HYDROGEN MONITORS ARE CLASS II.
TCDESIGN CLASS IIDESIGN CLASS II FIGURE 6.2-22 CONTAINMENT HYDROGEN PURGE SYSTEM HYDROGEN ANALYZER STREAM UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 16 June 2005 Revision 11 November 1996 FIGURE 6.2-23 MODEL B ELECTRIC HYDROGEN RECOMBINER CUTAWAY UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 6.2-24 ALUMINUM AND ZINC CORROSION RATE DESIGN CURVE UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE 0.0010.010.1 10.1110100GAS/LIQUID VOLUME RATIOAPPARENT G(H2), MOLECULES/100 eV2.5E+6 Rads/hr, pH = 8.62.5E+5 Rads/hr, pH = 9.46.1E+5 Rads/hr, pH = 9.4 Alkaline Sodium Borate Solution3000 ppm Boron - 72 deg F Revision 14 November 2001 FIGURE 6.2-25 RESULTS OF WESTINGHOUSE CAPSULE IRRADIATION TESTS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE 0 1 2 3 4
5 6 7 8
9 10 0 5 10 15 202530Time After LOCA (days)Hydrogen Concentration (volume %)No RecombinerRecombiner Started at 3.5 v/oRecombiner Started at 24 Hours4.0 v/o3.5 v/o FIGURE 6.2-26 POST-LOCA CONTAINMENT HYDROGEN CONCENTRATION UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 14 November 2001 0500001000001500002000002500003000003500000102030405060708090100Time after LOCA (days)Hydrogen Accumulation (scf)No RecombinerRecombination Started at 3.5 v/o Recombination Started at 24 Hours FIGURE 6.2-27 POST-LOCA HYDROGEN ACCUMULATION UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 14 November 2001 0 1 2 3 4 5 6 7 8 9 10 11 120102030405060708090100Time After LOCA (days)Hydrogen Production Rate (SCFM)Total CorrosionCore RadiolysisSump Radiolysis FIGURE 6.2-28 POST-LOCA HYDROGEN PRODUCTION UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 14 November 2001 0 25 50 75 100 125 150 175 2000102030405060708090100Time After LOCA (days)Hydrogen Accumulation (thousands of SCF)
Aluminum CorrosionZinc CorrosionTotal Corrosion FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2-29 POST-LOCA HYDROGEN ACCUMULATION FROM CORROSION OF MATERIAL INSIDE CONTAINMENT WITH NO RECOMBINER Revision 14 November 2001
Figures 6.2-53, 6.2-54, 6.2-55 and 6.2-56 Withheld From Public Disclosure in Accordance With 10 CFR 2.390
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2
-57 CONTAINMENT SUBCOMPARTMENT ANALYSES PRESSURIZER ENCLOSURE MODEL FROM ELEVATION 140 FEET TO 176 FEET
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2
-58 CONTAINMENT SUBCOMPARTMENT ANALYSES PRESSURIZER ENCLOSURE MODEL FROM ELEVATION 147 FEET 9 INCHES TO 166 FEET
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2
-59 CONTAINMENT SUBCOMPARTMENT ANALYSES PRESSURIZER ENCLOSURE MODEL FROM ELEVATION 140 FEET TO 147 FEET 9 INCHES FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2
-60 CONTAINMENT SUBCOMPARTMENT ANALYSES STEAM GENERATOR ENCLOSURE MODEL FROM ELEVATION 140 FEET TO 184 FEET
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2
-61 CONTAINMENT SUBCOMPARTMENT ANALYSES STEAM GENERATOR ENCLOSURE MODEL FROM ELEVATION 140 FEET TO 151 FEET 11-1/2 INCHES
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2
-62 CONTAINMENT SUBCOMPARTMENT ANALYSES STEAM GENERATOR ENCLOSURE MODEL FROM ELEVATION 151 FEET 11-1/2 INCHES TO 171 FEET
- For a detailed schematic of Nodes 19 through 30 & 38, see Figures 6.2-64 and 6.2-57 through 6.2-59 **For a detailed schematic of Nodes 34 through 37, see Figures 6.2-65 and 6.2-60 through 6.2-62 FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2-63 CONTAINMENT SUBCOMPARTMENT ANALYSES NODES AND FLOW PATHS FOR SUBCOMPARTMENT MODEL
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2
-64 CONTAINMENT SUBCOMPARTMENT ANALYSES NODES AND FLOW PATHS FOR SUBCOMPARTMENT MODEL
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2
-65 CONTAINMENT SUBCOMPARTMENT ANALYSES NODES AND FLOW PATHS FOR SUBCOMPARTMENT MODEL
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2
-66 CONTAINMENT SUBCOMPARTMENT ANALYSES MAXIMUM DIFFERENTIAL PRESSURE FOR PRESSURIZER SURGE LINE BREAK AT HOT LEG CONNECTION IN NODE 3
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2
-67 CONTAINMENT SUBCOMPARTMENT ANALYSES MAXIMUM DIFFERENTIAL PRESSURE FOR RHR SUCTION LINE BREAK AT HOT LEG CONNECTION IN NODE 6
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2
-68 CONTAINMENT SUBCOMPARTMENT ANALYSES MAXIMUM DIFFERENTIAL PRESSURE FOR PRESSURIZER SPRAY LINE BREAK AT THE TOP OF THE PRESSURIZER IN NODE 38
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2
-69 CONTAINMENT SUBCOMPARTMENT ANALYSES MAXIMUM DIFFERENTIAL PRESSURE FOR AN ACCUMULATOR LINE BREAK AT THE TANK NOZZLE IN NODE 17
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2
-70 CONTAINMENT SUBCOMPARTMENT ANALYSES MAXIMUM DIFFERENTIAL PRESSURE FOR MAIN FEEDWATER LINE BREAK AT THE STEAM GENERATOR INLET NOZZLE NODE 36
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2
-71 CONTAINMENT SUBCOMPARTMENT ANALYSES MAXIMUM DIFFERENTIAL PRESSURE MAIN STEAM LINE BREAK AT THE CONTAINMENT PENETRATION NODE 10
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2
-72 CONTAINMENT SUBCOMPARTMENT ANALYSES MAXIMUM DIFFERENTIAL PRESSURE FOR PRESSURIZER SURGE LINE BREAK AT HOT LEG CONNECTION IN NODE 3
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2
-73 CONTAINMENT SUBCOMPARTMENT ANALYSES MAXIMUM DIFFERENTIAL PRESSURE FOR PRESSURIZER SURGE LINE BREAK AT HOT LEG CONNECTION IN NODE 3
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 6.2
-74 CONTAINMENT SUBCOMPARTMENT ANALYSES MAXIMUM DIFFERENTIAL PRESSURE FOR PRESSURIZER SURGE LINE BREAK AT HOT LEG CONNECTION IN NODE 3
Revision 11 November 1996 FIGURE 6.3-1 RESIDUAL HEAT REMOVAL PUMP PERFORMANCE CURVES (TYPICAL)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008 FIGURE 6.3-2 CENTRIFUGAL CHARGING PUMPS 1 & 2 PERFORMANCE CURVES (TYPICAL)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 6.3-3 SAFETY INJECTION PUMP PERFORMANCE CURVES (TYPICAL)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE FIGURE 6.3-4 ALIGNMENT OF ECCS-RELATED COMPONENTS DURING INJECTION PHASE OF EMERGENCY CORE COOLING UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 20 November 2011[W-6.3 (1), SAPN 50585077] Replace existing Figure 6.3-4, Revision 20 with this revised Figure 6.3-4 in Revision 22.
Revision 22 May 2015 FIGURE 6.3-5 ALIGNMENT OF ECCS-RELATED COMPONENTS DURING RECIRCULATIONPHASE OF EMERGENCY CORE COOLING UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 20 November 2011 AUXILIARY FEEDWATER FLOW FOR PLAN T SHUTDOWN FROM 3568 MWT WITH ALL THREE AFW PUMPS IN OPERATION (FOR INFORMATION ONLY Revision 12 September 1998 FIGURE 6.5-3 (Sheet 1)
AUXILIARY FEEDWATER FLOW FOR PLANT SHUTDOWN UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE AUXILIARY FEEDWATER FLOW FOR PLANT SHUTDOWN FROM 3568 MWT WITH ONLY ONE 440 GPM MOTOR-DRIVEN AFW PUMP IN OPERATION (FOR INFORMATION ONLY)
FIGURE 6.5-3 (Sheet 2) AUXILIARY FEEDWATER FLOW FOR PLANT SHUTDOWN UNITS 1 AND 2 DIABL O C ANY O N S ITE F S AR UPDATERevision 20 November 2011 DCPP UNITS 1 &
2 FSAR UPDATE
APPENDIX 6.2D ANALYSIS OF LONG-TERM LOSS-OF-COOLANT ACCIDENTS AND MAIN STEAMLINE BREAK EVENTS
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-1 Revision 23 December 2016 6.2D.1 INTRODUCTION This appendix details the methodology for calculating the long-term mass and energy (M&E) releases and the resulting containment response subsequent to a hypothetical
loss-of-coolant accident (LOCA) or a main ste amline break (MSLB) in DCPP Unit 1 and Unit 2. Short-term LOCA-related M&E releases are used as input to the
subcompartment analyses and are discussed in Section 6.2.1.
The containment system is designed such that for all break sizes, up to and including
the double-ended severance of a reactor coolant pipe or secondary system pipe, the
containment peak calculated pressure is less than the containment design pressure.
The M&E releases from the LOCA and MSLB analyses are used as input to the
containment response analyses. The containment integrity analysis conclusions are
discussed in Sections 6.2D.3.2.7 (LOCA) and 6.2D.4.2.6 (MSLB). Section 6.2.1
includes an evaluation demonstrating that the containment satisfies the applicable
design requirements.
6.2D.2 COMPUTER CODES WHICH SUP PORT CURRENT ANALYSES Computer Codes which Support Current Analyses LOCA MSLB SATAN-VI X N/A WREFLOOD X N/A FROTH X N/A EPITOME X N/A RETRAN-02W (used with NRC limitations, refer to WCAP-14882-P-A) N/A X GOTHIC Version 7.2a X X The WCAP-10325-P-A (Reference 1) M&E release evaluation model is comprised of
M&E release versions of the following codes: SATAN-VI, WREFLOOD, FROTH, and
EPITOME. These codes were used to calculate the long-term LOCA M&E releases for
DCPP Unit 1 and Unit 2.
SATAN-VI calculates blowdown, the first portion of the thermal-hydraulic transient following break initiation, including pressure, enthalpy, density, M&E flow rates, and energy transfer between primary and secondary systems as a function of time.
The WREFLOOD code addresses the portion of the LOCA transient where the core
reflooding phase occurs after the primary coolant system has depressurized (blowdown) due to the loss of water through the break and when water supplied by the ECCS refills
the reactor vessel and provides cooling to the core. The most important feature of WREFLOOD is the steam/water mixing mod el (refer to Section 6.2D.3.1.4).
FROTH models the post-reflood portion of the transient. The FROTH code is used for
the steam generator heat addition calculation from the broken and intact loop steam
generators.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-2 Revision 23 December 2016 EPITOME continues the FROTH post-reflood portion of the transient from the time at
which the secondary equilibrates to containment design pressure to the end of the
transient. It also compiles a summary of data on the entire transient, including formal
instantaneous M&E release tables and M&E balance tables with data at critical times.
The Westinghouse steamline break M&E release methodology was approved by the
NRC (Reference 8) and is documented in WCAP-8822, Mass and Energy Releases
Following a Steam Line Rupture (Reference 9) and WCAP-8822-S2-P-A, Mass and
Energy Releases Following a Steam Line Rupture, Supplement 2 - Impact of Steam
Superheat in Mass/Energy Releases Following a Steamline Rupture of Dry and
Subatmospheric Containment Designs (Reference 11). WCAP-8822 forms the basis
for the assumptions used in the calculation of the M&E releases resulting from a
steamline rupture. The analysis documented herein uses the RETRAN-02W code, which is documented in WCAP-14882-P-A, RETR AN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses (Reference 12).
GOTHIC Version 7.2a is used to perform the containment response analyses for the LOCA and MSLB events.
6.2D.3 LONG-TERM LOSS-OF-COOLANT ACCIDENTS 6.2D.3.1 Long-Term LOCA Mass and Energy Release Analysis 6.2D.3.1.1 Acceptance Criteria There are no direct acceptance criteria for LOCA M&E releases. The analysis methods follow the guidelines provided by the USNRC with respect to the sources of M&E during the various phases of a large break LOCA tr ansient (refer to Section 6.2D.3.1.4).
The specific acceptance criteria for the containment response to a LOCA are discussed
in Section 6.2D.3.2.1.
6.2D.3.1.2 Introduction and Background Discussion of the short-term LOCA-related M&E releases which are used as input to the subcompartment analyses can be found in Section 6.2.1.2.
The uncontrolled release of pressurized high-temperature reactor coolant, termed a LOCA, would result in release of steam and water into the containment. This, in turn, would result in increases in the local subcompartment pressures, and an increase in the
global containment pressure and temperature. Therefore, there are both long- and short-term issues relative to a postulated LOCA that must be considered at the
conditions for DCPP Unit 1 and Unit 2 at the licensed core power of 3411 MWt.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-3 Revision 23 December 2016 The long-term LOCA M&E releases are analyzed to approximately 10 7 seconds and are utilized as input to the containment integrity analysis. The containment integrity analysis demonstrates the acceptability of the containment safeguards systems to mitigate the consequences of a hypothetical large-break LOCA. The containment safeguards systems must be capable of limiting the peak containment pressure to less
than the design pressure and to limit the temperature excursion to less than the
acceptance limits. Westinghouse generated the M&E releases using the March 1979
model, described in WCAP-10325-P-A (Reference 1). The Nuclear Regulatory
Commission (NRC) review and approval letter is included with WCAP-10325-P-A (Reference 1). Section 6.2D.3.1 discusses the long-term LOCA M&E releases
generated. The results of this analysis were provided for use in the containment
integrity analysis (refer to Section 6.2D.3.2).
The M&E release rates described in this section form the basis of further computations
to evaluate the containment following the postulated accident. Discussed in this section
are the long-term LOCA M&E releases for the hypothetical double-ended pump suction (DEPS) rupture with minimum safeguards and maximum safeguards and double-ended
hot-leg (DEHL) rupture break cases. These LOCA cases are used for the long-term
containment integrity analyses in Section 6.2D.3.2.
6.2D.3.1.3 Input Parameters and Assumptions The M&E release analysis is sensitive to the assumed characteristics of various plant
systems, in addition to other key modeling assumptions. Where appropriate, bounding
inputs are utilized and instrumentation uncer tainties are included. For example, the RCS operating temperatures are chosen to bound the highest average coolant temperature range of all operating cases and a temperature uncertainty allowance of
+5.0°F is then added. Nominal parameters are used in certain instances. For example, the RCS pressure in this analysis is based on a nominal value of 2,250 psia plus an
uncertainty allowance +42.0 psi. All input parameters are chosen consistent with
accepted analysis methodology.
Some of the most critical items are the RCS initial conditions, core decay heat, safety
injection flow, and primary and secondary metal mass and steam generator heat
release modeling. Specific assumptions concerning each of these items are discussed
in the following paragraphs. Tables 6.2D-1 and 6.2D-2 present key data assumed in
the analysis.
The core thermal power of 3479 MWt adjusted for calorimetric error (that is, 102 percent of 3411 MWt) was used in the analysis. As previously noted, RCS operating
temperatures to bound the highest average coolant temperature range were used as
bounding analysis conditions. The use of higher temperatures is conservative because
the initial fluid energy is based on coolant temperatures that are at the maximum levels
attained in steady-state operation. Additionally, an allowance to account for instrument
error and dead-band is reflected in the initial RCS temperatures. The selection of 2,250
psia plus uncertainty as the limiting pressure is considered to affect the blowdown DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-4 Revision 23 December 2016 phase results only, since this represents the initial pressure of the RCS. The RCS rapidly depressurizes from this value until the point at which it equilibrates with
containment pressure.
The rate at which the RCS blows down is initially more severe at the higher RCS
pressure. Additionally, the RCS has a higher fluid density at the higher pressure (assuming a constant temperature) and subsequently has a higher RCS mass available
for releases. Thus, 2,250 psia plus uncertainty was selected for the initial pressure as
the limiting case for the long-term M&E release calculations.
The selection of the fuel design features for the long-term M&E release calculation is
based on the need to conservatively maximize the energy stored in the fuel at the
beginning of the postulated accident (that is, to maximize the core stored energy). The
core stored energy, selected to bound the 17x17 fuel product used at DCPP Unit 1 and
Unit 2, was 3.50 full-power seconds (FPS). The margins in the core stored energy
include a statistical uncertainty in order to address the thermal fuel model and
associated manufacturing uncertainties and the time in the fuel cycle for maximum fuel
densification. Thus, the analysis very conservatively accounts for the stored energy in
the core.
A margin in the RCS volume of 3 percent (which is composed of a 1.6-percent
allowance for thermal exp ansion and a 1.4-percent allowance for uncertainty) was modeled.
A uniform steam generator tube plugging level of 0 percent was modeled. This
assumption maximizes the reactor coolant volume and fluid release by virtue of consideration of the RCS fluid in all steam generator tubes. During the post-blowdown period, the steam generators are active heat sources since significant energy remains in
the secondary metal and secondary mass that has the potential to be transferred to the
primary side. The 0 percent tube plugging assumption maximizes the heat transfer area
and, therefore, the transfer of secondary he at across the steam generator tubes.
Additionally, this assumption reduces the reactor coolant loop resistance, which reduces
Thus, the analysis conservatively accounts for the level of steam generator tube
plugging.
The secondary-to-primary heat transfer is maximized by assuming conservative heat
transfer coefficients. This conservative energy transfer is ensured by maximizing the
initial internal energy of the inventory in the steam generator secondary side. This
internal energy is based on full-power operation plus uncertainties.
Regarding safety injection flow, the M&E release calculation considered
configurations/failures to conservatively bound respective alignments. The limiting case is the failure of a train of the solid state protection system (SSPS). This
configuration/failure would credit minimum flow from one centrifugal charging pump (CCP1 or CCP2), one safety injection (SI) pump, and one residual heat removal (RHR)
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-5 Revision 23 December 2016 pump (refer to Table 6.2D-2). In addition, the containment backpressure is assumed to be equal to the containment design pressure. This assumption was shown in WCAP-
10325-P-A (Reference 1) to be conservative for the generation of M&E releases.
In summary, the following assumptions were employed to ensure that the M&E releases are conservatively calculated, thereby maximizing energy release to containment:
- Maximum expected operating temperature of the RCS (100-percent full-power conditions)
- Allowance for RCS temperature uncertainty (+5.0°F )
- Margin in RCS volume of 3 percent (which is composed of 1.6-percent allowance for thermal expansi on and 1.4-percent allowance for uncertainty)
- Core rated power of 3411 MWt
- Allowance for calorimetric error (+2.0 percent of power)
- Conservative heat transfer coefficients (that is, steam generator primary/secondary heat transfer, and RCS metal heat transfer)
- Allowance in core stored energy for effect of fuel densification
- A margin in core stored energy (statistical uncertainty to account for manufacturing tolerances)
- An allowance for RCS initial pressure uncertainty (+42.0 psi)
- A maximum containment backpressure equal to design pressure (47.0 psig)
- Steam generator tube plugging level (0-percent uniform)
- Maximizes reactor coolant volume and fluid release
- Maximizes heat transfer area across the steam generator tubes
- Reduces coolant loop resistance, which reduces upstream of the break for the pump suction breaks and increases break flow
Thus, based on the previously discussed conditions and assumptions, an analysis of
DCPP Unit 1 and Unit 2 was made for the release of M&E from the RCS in the event of
a LOCA at 3479 MWt.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-6 Revision 23 December 2016 6.2D.3.1.4 Description of Analyses and Evaluations The evaluation model used for the long-term LOCA M&E release calculations is
described in WCAP-10325-P-A (Reference 1).
This section presents the long-term LOCA M&E releases generated in support of DCPP
Unit 1 and Unit 2.
The guidance of NRC Standard Review P lan Section 6.2.1.3 was used to determine what should be addressed in the M&E release analysis for postulated LOCAs as
documented in WCAP-16638-P. Criteria applicable to DCPP are contained in 10 CFR Part 50, Appendix K, Part I.A.
To meet those requirements, the following were addressed by the LOCA M&E release analysis:
- Calculation of each phase of the accident
- Break size and location
- M&E release data
- Sources of energy These M&E releases are then subsequently used in the containment integrity analysis.
LOCA M&E Release Phases The containment system receives M&E releases following a postulated rupture in the
RCS. These releases continue over a time period, which, for the LOCA M&E analysis, is typically divided into four phases.
Blowdown - the period of time from accident initiation (when the reactor is at steady-
state operation) to the time that the RCS and containment reach an equilibrium state.
Refill - the period of time when the lower plenum is being filled by accumulator and
emergency core cooling system (ECCS) water.
At the end of blowdown, a large amount of water remains in the cold legs, downcomer, and lower plenum. To conservatively
consider the refill period for the purpose of containment M&E releases, it is assumed
that this water is instantaneously transferred to the lower plenum along with sufficient
accumulator water to completely fill the lower plenum. This allows an uninterrupted
release of M&E to containment. Thus, the refill period is conservatively neglected in the
M&E release calculation.
Reflood - begins when the water from the lower plenum enters the core and ends
when the core is completely quenched.
Post-reflood
- describes the period following the reflood phase. For the pump suction
break, a two-phase mixture exits the core, passes through the hot legs, and is DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-7 Revision 23 December 2016 superheated in the steam generators prior to ex iting the break as steam. After the broken loop steam generator cools, the break flow becomes two phase.
Break Size and Location Generic studies have been performed with respect to the effect of postulated break size
on the LOCA M&E releases. The double-ended guillotine break has been found to be limiting due to larger mass flow rates during the blowdown phase of the transient.
During the reflood and post-reflood phases, the break size has little effect on the
releases.
Three distinct locations in the RCS loop can be postulated for a pipe rupture for M&E
release purposes:
- Hot leg (between vessel and steam generator)
- Cold leg (between pump and vessel)
- Pump suction (between steam generator and pump)
The break locations analyzed for this program are the double-ended pump suction (DEPS) rupture with a total break area of (10.46 ft 2) and the double-ended hot leg (DEHL) rupture with a total break area of (9.17 ft 2). Break M&E releases have been calculated for the blowdown, reflood, and post-reflood phases of the LOCA for the
DEPS cases. For the DEHL case, the releases were calculated only for the blowdown.
The following information provides a discussion on each break location.
The DEHL rupture has been shown in previous studies to result in the highest
blowdown M&E release rates. Although the core flooding rate would be the highest for
this break location, the amount of energy released from the steam generator secondary
is minimal because the majority of the fluid that exits the core vents directly to containment bypassing the steam generators. As a result, the reflood M&E releases
are reduced significantly as compared to either the pump suction or cold-leg break
locations where the core exit mixture must pass through the steam generators before
venting through the break. For the hot-leg break, generic studies have confirmed that
there is no reflood peak (that is, from the end of the blowdown period the containment
pressure would continually decrease). Therefore, only the M&E releases for the hot-leg
break blowdown phase are calculated and presented.
The cold-leg break location has also been found in previous studies to be much less
limiting in terms of the overall containment energy releases. The cold-leg blowdown is
faster than that of the pump suction break, and more mass is released into the
containment. However, the core heat transfer is greatly reduced, and this results in a
considerably lower energy release into containment. Studies have determined that the
blowdown transient for the cold leg is, in general, less limiting than that for the pump suction break. During reflood, the flooding rate is greatly reduced and the energy
release rate into the containment is reduced. Therefore, the cold-leg break is bounded
by other breaks and no further evaluation is necessary.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-8 Revision 23 December 2016 The pump suction break combines the effects of the relatively high-core flooding rate, as in the hot-leg break, and the addition of the stored energy in the steam generators.
As a result, the pump suction break yields the highest energy flow rates during the post-
blowdown period by including all of the avail able energy of the RCS in calculating the releases to containment. Thus, only the DEHL and DEPS cases are used to analyze long-term LOCA containment integrity.
Application of Single-Failure Criterion An analysis of the effects of the single-failure criterion has been performed on the M&E release rates for each break analyzed. An inherent assumption in the generation of the
M&E release is that offsite power is lost. This results in the actuation of the emergency
diesel generators, required to power the safety injection system. This is not an issue for
the blowdown period, which is limited by the DEHL break.
Two cases have been analyzed to assess the effects of a single failure. The first case
assumes minimum safeguards safety injection (SI) flow based on the postulated single
failure of a train of the solid state protection system. This results in the loss of one train
of safeguards equipment. The other case assumes maximum safeguards SI flow based
on no postulated failures that would impact the amount of ECCS flow. The analysis of
the cases described provides confidence that the effect of credible single failures is
bounded.
The M&E releases for the DCPP Unit 2 cases are shown in Tables 6.2D-3 through
6.2D-9. The Unit 2 results bound the Unit 1 results.
Blowdown M&E Release Data The SATAN-VI code is used for computing the blowdown transient. The SATAN-VI
code utilizes the control volume (element) approach with the capability for modeling a
large variety of thermal fluid system configurations. The fluid properties are considered
uniform and thermo-dynamic equilibrium is assumed in each element. A point kinetics
model is used with weighted feedback effects. The major feedback effects include
moderator density, moderator temperature, and Doppler broadening. A critical flow
calculation for subcooled (modified Zaloudek
), two-phase (Moody), or superheated break flow is incorporated into the analysis. The methodology for the use of this model
is described in WCAP-10325-P-A (Reference 1).
Table 6.2D-3 presents the calculated M&E release for the blowdown phase of the DEHL
break without any pumped safety injection for DCPP Unit 2. For the hot leg break M&E release tables, break path 1 refers to the M&E exiting from the reactor-vessel side of the
break; break path 2 refers to the M&E ex iting from the steam-generator side of the break. Tables 6.2D-4 and 6.2D-5 present the M&E balance data for the Unit 2 DEHL
case.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-9 Revision 23 December 2016 Table 6.2D-6 presents the calculated M&E releases for the blowdown phase of the DEPS break. For the pump suction breaks, break path 1 in the M&E release tables
refers to the M&E exiting from the steam-generator side of the break. Break path 2 refers to the M&E exiting from the pump-side of the break.
Refill M&E Release Data
As noted earlier in the discussion of LOCA M&E release phases, the refill period is
conservatively neglected in the M&E release calculation.
Reflood M&E Release Data The WREFLOOD code is used for computing the reflood transient. The WREFLOOD
code consists of two basic hydraulic models: one for the contents of the reactor vessel
and one for the coolant loops. The two models are coupled through the interchange of
the boundary conditions applied at the vessel outlet nozzles and at the top of the
downcomer. Additional transient phenomena such as pumped SI and accumulators, reactor coolant pump performance, and steam generator release are included as
auxiliary equations that interact with the basic models as required. The WREFLOOD
code permits the capability to calculate variations during the core reflooding transient of
basic parameters such as core flooding rate, core and downcomer water levels, fluid
thermo-dynamic conditions (pressure, enthalpy, density) throughout the primary system, and mass flow rates through the primary system. The code permits hydraulic modeling
of the two flow paths available for discharging steam and entrained water from the core to the break, that is, the path through the broken loop and the path through the unbroken loops.
A complete thermal equilibrium mixing conditi on for the steam and ECCS injection water during the reflood phase has been assumed for each loop receiving ECCS water. This
is consistent with the usage and appl ication of the WCAP-10325-P-A (Reference 1)
M&E release evaluation model in contemporary analyses, for example, D. C. Cook
Docket 50-315 (Reference 3). Even though the WCAP-10325-P-A (Reference 1) model
credits steam/water mixing only in the intact loop and not in the broken loop, the
justification, applicability, and NRC approval for using the mixing model in the broken
loop has been documented (Reference 3).
Moreover, this assumption is supported by test data and is further discussed below.
The model assumes a complete mixing condition (that is, thermal equilibrium) for the steam/water interaction. The complete mixing process, however, is made up of two
distinct physical processes. The first is a two-phase interaction with condensation of steam by cold ECCS water. The second is a single-phase mixing of condensate and
ECCS water. Since the steam release is the most important influence to the
containment pressure transient, the steam condensation part of the mixing process is the only part that need be considered. (Any spillage directly heats only the sump.)
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-10 Revision 23 December 2016 The most applicable steam/water mixing test data have been reviewed for validation of the containment integrity reflood steam/water mixing model. This data was generated in 1/3-scale tests (Reference 4), which were the largest scale data available in 1975 and
thus most clearly simulated the flow regimes and gravitational effects that would occur
in a pressurized water reactor (PWR). These tests were designed specifically to study
the steam/water interaction for PWR reflood conditions.
A group of 1/3-scale tests corresponds directly to containment integrity reflood
conditions. The injection flow rates for this group cover all phases and mixing
conditions calculated during the reflood transient. The data from these tests were
reviewed and discussed in detail in WCAP-10325-P-A (Reference 1). For all of these
tests, the data clearly indicate the occurrence of very effective mixing with rapid steam
condensation. The mixing model used in the containment integrity reflood calculation is, therefore, wholly supported by the 1/3-scale steam/water mixing data.
Additionally, the following justification is also noted. The post-blowdown limiting break for the containment integrity peak pressure analysis is the pump suction double-ended
rupture break. For this break, there are two flow paths available in the RCS by which
M&E may be released to containment. One is through the outlet of the steam
generator, the other via reverse flow through the reactor coolant pump. Steam that is not condensed by ECCS injection in the intact RCS loops passes around the
downcomer and through the broken loop cold leg and pump in venting to containment.
This steam also encounters ECCS injection water as it passes through the broken loop
cold leg, complete mixing occurs and a portion of it is condensed. It is this portion of
steam that is condensed that is taken credit for in this analysis. This assumption is justified based upon the postulated break location, and the actual physical presence of the ECCS injection nozzle. A description of the test and test results are contained in WCAP-10325-P-A (Reference 1) and operating licens e Amendment No. 126 for D.C.
Cook (Reference 3).
Table 6.2D-7 presents the calculated M&E releases for the reflood phase of the pump
suction double-ended rupture with a single failure of a train of the solid state protection
system (SSPS) for DCPP Unit 2. The principal parameters during reflood are given in
Table 6.2D-8 for the bounding DEPS case.
Post-Reflood M&E Release Data The FROTH code (Reference 5) is used for computing the post-reflood transient. The
FROTH code calculates the heat release rat es resulting from a two-phase mixture present in the steam generator tubes. The M&E releases that occur during this phase are typically superheated due to the depress urization and equilibration of the broken loop and intact loop steam generators. During this phase of the transient, the RCS has
equilibrated with the containment pressure. However, the steam generators contain a secondary inventory at an enthalpy that is much higher than the primary side.
Therefore, there is a significant amount of reverse heat transfer that occurs. Steam is produced in the core due to core decay heat. For a pump suction break, a two-phase DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-11 Revision 23 December 2016 fluid exits the core, flows through the hot legs, and becomes superheated as it passes through the steam generator (Reference 6). Once the broken loop cools, the break flow
becomes two phase. During the FROTH calculation, ECCS injection is addressed for
both the injection phase and the recirculation phase. The FROTH code calculation
stops when the secondary side equilibrates to the saturation temperature (Tsat) at the containment design pressure; after this point the EPITOME code completes the steam
generator depressurization (refer to Sections 6.2D.3.1.4, Reflood Mass and Energy
Release Data and 6.2D.3.1.4, Decay Heat Model for additional information).
The methodology for the use of this model is described in WCAP-10325-P-A (Reference 1). The M&E release rates are calculated by FROTH and EPITOME until
the time of containment depressurization. After containment depressurization
(14.7 psia), the M&E release available to containment is generated directly from core
boil-off/decay heat.
Table 6.2D-9 presents the two-phase post-reflood M&E release data for the pump
suction double-ended break case with a single failure of a train of the SSPS.
Decay Heat Model ANS Standard 5.1 (Reference 7) was used in the LOCA M&E release model for DCPP
Unit 1 and Unit 2 for the determination of decay heat energy. This standard was
balloted by the Nuclear Power Plant Standards Committee in October 1978 and
subsequently approved. The official standard (Reference 7) was issued in August
1979. Table 6.2D-10 lists the decay heat generation rate used in the DCPP M&E
release analysis.
Based upon NRC staff review, (Safety Evaluation Report [SER] of the March 1979
evaluation model [Reference 1]), use of the ANS Standard-5.1, November 1979 decay
heat model was approved for the calculation of M&E releases to the containment
following a LOCA.
Significant assumptions in the decay heat generation rate for use in the LOCA M&E
releases analysis include the following:
- The decay heat sources considered are fission product decay and heavy element decay of U-239 and Np-239.
- The decay heat power from fissioning isotopes other than U-235 is assumed to be identical to that of U-235.
- The fission rate is constant over the operating history of maximum power level.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-12 Revision 23 December 2016
- The factor accounting for neutron capture in fission products has been taken from Equation 11 of Reference 7 up to 10,000 seconds and from Table 10 of Reference 7 beyond 10,000 seconds.
- The fuel has been assumed to be at full power for 10 8 seconds.
- The number of atoms of U-239 produced per second has been assumed to be equal to 70 percent of the fission rate.
- The total recoverable energy associated with one fission has been assumed to be 200 MeV/fission.
- An uncertainty of two sigma (two times the standard deviation) has been applied to the fission product decay.
Steam Generator Equilibration and Depressurization Steam generator equilibration and depressurization is the process by which secondary-
side energy is removed from the steam generators in stages. The FROTH computer
code calculates the heat removal from the secondary mass until the secondary
temperature is the T sat at the containment design pressure. After the FROTH calculations, the EPITOME code continues the FROTH calculation for steam generator
cooldown removing steam generator secondary energy at different rates (that is, first-
and second-stage rates). The first-stage rate is applied until the steam generator
reaches T sat at the user-specified intermediate equilibration pressure, when the secondary pressure is assumed to reach the actual containment pressure. Then the
second-stage rate is used until the final depressurization, when the secondary reaches
the reference temperature of T sat at 14.7 psia, or 212°F. The heat removal of the broken loop and intact loop steam generators are calculated separately.
During the FROTH calculations, steam generator heat removal rates are calculated
using the secondary side temperature, primary side temperature, and a secondary side
heat transfer coefficient determined using a modified McAdams correlation. Steam
generator energy is removed during the FROTH transient until the secondary side
temperature reaches T sat at the containment design pressure. The constant heat removal rate used during the first heat removal stage is based on the final heat removal
rate calculated by FROTH. The steam generator energy available to be released during
the first stage interval is determined by calculating the difference in secondary energy
available at the containment design pressure and that at the (lower) user-specified
intermediate equilibration pressure, assuming saturated conditions. This energy is then
divided by the first-stage energy removal rate, resulting in an intermediate equilibration
time. At this time, the rate of energy release drops substantially to the second-stage rate. The second-stage rate is determined as the fraction of the difference in secondary
energy available between the intermediate equilibration and final depressurization at
212°F, and the time difference from the time of the intermediate equilibration to the user-specified time of the final depressurization at 212°F. With current methodology, all DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-13 Revision 23 December 2016 of the secondary energy remaining after the intermediate equilibration is conservatively assumed to be released by imposing a mandatory cooldown and subsequent
depressurization down to atmospheric pressure at 3,600 seconds, that is, 14.7 psia and
212°F (the M&E balance tables have this point labeled as Available Energy).
Sources of M&E The sources of mass considered in the LOCA M&E release analysis are given in
Table 6.2D-4 for the DEHL breaks for DCPP Unit 2. The sources of mass for the DEPS
break case with the SSPS failure for DCPP Unit 2 are given in Table 6.2D-11. These
sources are the RCS, accumulators, and pumped SI.
The energy inventory considered in the DEHL breaks for DCPP Unit 2 is given in
Table 6.2D-5. The energy inventory for the DEPS break M&E release analysis for
DCPP Unit 2 is given in Table 6.2D-12. The energy sources are as follows:
- RCS water
- Accumulator water (all inject)
- Pumped SI water
- Decay heat
- Core-stored energy
- RCS metal (includes steam generator tubes)
- Steam generator metal (includes transition cone, shell, wrapper, and other internals)
- Steam generator secondary energy (includes fluid mass and steam mass)
- Secondary transfer of energy (feedwater into and steam out of the steam generator secondary)
The analysis used the following energy reference points:
Available energy:
212°F; 14.7 psia (energy available that could be released) Total energy content: 32°F; 14.7 psia (total internal energy of the RCS)
The M&E inventories are presented at the following times, as appropriate:
- Time zero (initial conditions)
- End of blowdown time
- End of refill time
- End of reflood time
- Time of broken loop steam generator equilibration to pressure setpoint
- Time of intact loop steam generator equilibration to pressure setpoint
- Time of full depressurization (3,600 seconds)
The energy release from the metal-water reaction rate is considered as part of the
WCAP-10325-P-A (Reference 1) methodology. Based on the way that the energy in the
fuel is conservatively released to the vessel fluid, the fuel cladding temperature does not
increase to the point where the metal-water reaction is significant. This is in contrast to DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-14 Revision 23 December 2016 the 10 CFR 50.46 analyses, which are biase d to calculate high fuel rod cladding temperatures and, therefore, a significant metal-water reaction. For the LOCA M&E release calculation, the energy created by the metal-water reaction value is small and is
not explicitly provided in the energy balance tables. The energy that is determined is
part of the M&E releases and is therefore already included in the overall M&E releases for DCPP.
The sequence of events for the LOCA transients is shown in Tables 6.2D-13 through
6.2D-15.
6.2D.3.1.5 Results The consideration of the various energy sources in the long-term M&E release analysis
provides assurance that all available sources of energy have been included in this analysis. The results of this analysis were provided for use in the LOCA containment
integrity analysis in Section 6.2D.3.2.
6.2D.3.2 Long-Term LOCA Containment Integrity Analysis
6.2D.3.2.1 Acceptance Criteria The containment response for design basis LOCA containment integrity is an ANS
Condition IV event, an infrequent fault. The relevant requirements to satisfy NRC
acceptance criteria are as follows:
- The peak calculated containment pressure should be less than the containment design pressure of 47 psig.
- The peak calculated containment average air temperature should be less than the containment design temperature of 271 °F
- The calculated pressure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> should be less than 50 percent of the peak calculated value. (This is related to the criteria for containment
leakage assumptions as affecting doses at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.)
Section 6.2.1 includes an evaluation demonstrating that the containment satisfies the
applicable design requirements.
6.2D.3.2.2 Introduction and Background
The purpose of the LOCA containment integrity analysis is to evaluate the bounding peak pressure and temperature of a design basis LOCA event inside containment and
to demonstrate the ability of the containment heat removal systems to mitigate the
accident. The impact of LOCA M&E releases on the containment pressure and
temperature are assessed to ensure that the containment pressure and temperature
remain below their respective design limits. The containment heat removal systems DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-15 Revision 23 December 2016 must also be capable of maintaining the environmental qualification (EQ) parameters to within acceptable limits.
The DCPP LOCA containment response analysis considers a spectrum of cases that
address differences between the individual DCPP Units, LOCA break locations, and postulated single failures (minimum and maximum safeguards). The limiting cases that
address the containment peak pressure case and limiting long-term EQ temperature are
presented in this section.
Calculation of the containment response followin g a postulated LOCA was analyzed by use of the digital computer code GOTHIC version 7.2a. The GOTHIC Technical Manual (Reference 13) provides a description of the governing equations, constitutive models, and solution methods in the solver. The GOTHIC Qualifications Report (Reference 14) provides a comparison of the solver results with both analytical solutions and
experimental data.
The GOTHIC containment modeling for DCPP is consistent with the NRC approved
Kewaunee evaluation model (Reference 15). Kewaune e and DCPP both have large dry containment designs with similar active heat removal capabilities. The latest code
version is used to take advantage of the diffusion layer model heat transfer option. This
heat transfer option was approved by the NRC (Reference 15) for use in Kewaunee
containment analyses with the condition that the ef fect of mist be excluded from what was earlier termed as the mist diffusion layer model. The GOTHIC containment
modeling for DCPP has followed the conditions of acceptance placed on Kewaunee.
The differences in GOTHIC code versions are documented in Appendix A of the
GOTHIC User Manual Release Notes (Reference 16). Version 7.2a is used consistently with the restrictions identified in Reference 15; none of the user-controlled
enhancements added to version 7.2a were implemented in the DCPP containment model. A description of the DCPP GOTHIC model is provided later in this section.
6.2D.3.2.3 Input Parameters and Assumptions The major modeling input parameters and assumptions used in the DCPP LOCA
containment evaluation model are identified in this section. The assumed initial
conditions and input assumptions associated with the fan coolers and containment
sprays are listed in Table 6.2D-17. The containment spray flow data used in the
analysis are presented in Table 6.2D-18. The function of the residual heat removal
system (RHR) during a LOCA is to remove heat from the core by way of the ECCS.
The ECCS recirculation and CCW system parameters are outlined in Table 6.2D-17.
The containment structural heat sink input is provided in Table 6.2D-19, and the
corresponding material properties are listed in Table 6.2D-20.
The LOCA containment analysis described here uses revised input and assumptions in
support of the current design, while addressing analytical conservatisms. The assumptions used in the M&E release input model and the containment pressure input
model are discussed in WCAP-10325-P-A (Reference 1) and WCAP-8264-P-A Revision DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-16 Revision 23 December 2016 1 (Reference 5). Significant assumptions contained in the LOCA containment integrity analysis include:
(1) For all the long term cases and the base ho t leg cases, there is a loss of offsite power coincident with the LOCA. For the hot leg break case with safety injection, off site power is available to allow the safety injection to begin during the
blowdown.
(2) In all cases, two containment fan cooler units (CFCUs), each from separate trains, are assumed to be unavailable due to maintenance (3) The long term decay heat steaming M&E release calculation assumes that:
(a) All the decay heat is released to the containment as steam to maximize the pressure and temperature of the containment vapor region, (b) 102% reactor power and ANS 1979 +2 sigma decay heat, (4) The bounding auxiliary saltwater (ASW) temperature is 64 °F
(a) It should be noted that a separate set of analyses assuming a 70 °F ocean water temperature addresses operation with an elevated ultimate heat sink
temperature (refer to Section 9.2.2.3.13).
The following are notable features of the current containment integrity analysis.
Decay heat steaming M&E release rates, after the end of the sensible heat release from the RCS and steam generators, are calculated each time step by GOTHIC using the
transient containment pressure and recirculation safety injection water temperature.
Non-condensable accumulator gas release is modeled in the GOTHIC model (refer to
Section 6.2D.3.2.5); no accumulator nitrogen gas additio n due to refill is considered in the analysis.
A recirculation system model that couples the RHR, CCW, CFCUs and auxiliary saltwater systems was developed. Detailed accounting of CCW flow rates through the
containment heat removal systems was used for the CFCUs, RHR heat exchangers, and miscellaneous CCW heat loads.
The DCPP LOCA containment response analysis considered a spectrum of cases. The
cases address break locations, and postulated single failures (minimum and maximum
safeguards) for each DCPP unit. Only the limiting cases, which address the
containment peak pressure and limiting long-term EQ temperature, are presented. The LOCA pressure and temperature response analyses were performed assuming a loss of
offsite power and a worst single failure (loss of one solid state protection system [SSPS]
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-17 Revision 23 December 2016 train, i.e., loss of one containment cooling train). The active heat removal available in the long term cooling case is:
- One containment spray pump during injection-phase only
- Two containment fan cooler units
- One ASW pump.
The Unit 2 DEHL break produces the overall bounding peak pressure at the end of the
blowdown. The calculation for the DEPS case was performed for a 116 day (1x10 7 second) transient in support of long term EQ temperatures. The sequence of events for the DEHL containment peak pressure case is shown in Table 6.2D-13 and the DEPS
long term EQ temperature case for Unit 1 and Unit 2 are shown in Tables 6.2D-14 and
6.2D-15, respectively.
6.2D.3.2.4 Description of Analyses and Evaluations Plant input assumptions (identified in Section 6.2D.3.2.3) are the same as, or slightly
more restrictive, than in the original analyses performed with the COCO code (Reference 18). Benchmarking between the DCPP COCO and GOTHIC models was
performed to confirm consistency in the implementation of the plant input values.
Noding Structure
The DCPP GOTHIC containment model is comprised of one control volume with
separate vapor and liquid regions. M&E releases, containment spray injection, and
sump water recirculation are modeled using boundary conditions. A cooler component
is used to model CFCUs heat removal. Injection of accumulator nitrogen during the
event is modeled with a boundary condition.
The component cooling water system model is comprised of three control volumes (CFCU cooling water, the hot side of the CCW system, and the cold side of the CCW system) and uses GOTHIC component models for the RHR and CCW heat exchangers.
A heater component models the CFCU heat transfer to the CCW water. Boundary
conditions model the CCW flow through the CFCUs, RHR heat exchangers, and
miscellaneous CCW heat loads.
Volume Input
Values for the volume, height, hydraulic diameter, and elevation are input for each
node. The containment is modeled as a single control volume. The lower bound free DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-18 Revision 23 December 2016 volume is 2,550,000 ft
- 3. The hydraulic diameter, height, and floor elevation input values are 24.1 ft, 166 ft, and 91 ft, respectively.
A conservatively calculated pool surface area is used to model interfacial heat and
mass transfer to liquid pools on the various floor surfaces in the containment volume.
The conductor representing the floor is essentially insulated from the vapor region after
the sump pool develops; however, there can still be condensation or evaporation from
the surface of the liquid pools. Using this method to model the interfacial heat and
mass transfer between the pools and the atmosphere was previously approved by the
NRC for the Kewaunee containment DBA an d equipment qualification analyses (Reference 15).
Initial Conditions
The containment initial conditions for containment integrity cases are:
- Pressure: 16.0 psia
- Relative Humidity: 18 percent
system. The system volumes are water solid and assumed to be initially at 50 psia and
90°F.
Flow Paths
Flow boundary conditions linked to functions that define the M&E release model the
LOCA break flow to the containment. The boundary conditions are connected to the
containment control volume via flow paths. The containment spray is modeled as a
boundary condition connected to the containment control volume via a flow path.
The flow rates through the flow paths are specified by the boundary conditions, so the
purpose of the flow path is to direct the flow to the proper control volume. The flow path
input is mostly arbitrary. Standard values are used for the area, hydraulic diameter, friction length, and inertia length of the flow path. Since this is a single volume lumped
parameter model, the elevation of the break flow paths is arbitrarily set to 100 ft and the elevation of the spray flow paths is arbitrarily set to 70 ft above the containment floor.
Heat Sinks
The structural heat sinks in the containment are modeled as GOTHIC thermal
conductors. The heat sink geometry data is based on conservatively low surface areas
and is summarized in Table 6.2D-19. A thin air gap is assumed to exist between the
steel and concrete for steel-jacketed heat sinks. A gap conductance of 10 Btu/hr-ft 2-°F is conservatively assumed between steel and concrete. The volumetric heat capacity
and thermal conductivity for the heat sink materials are summarized in Table 6.2D-20.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-19 Revision 23 December 2016 Heat and Mass Transfer Correlation
GOTHIC has several heat transfer coefficient options that can be used for containment
analyses. For the DCPP GOTHIC model, the direct heat transfer coefficient set is used
with the diffusion layer model mass transfer correlation for the heat sinks inside
containment. This heat transfer methodology was reviewed by the NRC and approved for use in containment DBA analyses in the Kewaunee analysis (Reference 15). The diffusion layer model correlation does not require the user to specify a revaporization
input value, as was done in previous analyses using the Uchida correlation.
Split heat transfer coefficients are used for the heat sinks representing walls and floors.
The split coefficient allows one thermal conductor to model heat transfer to both the
water and vapor regions. The submerged portions of conductors are essentially
insulated from the vapor after the pool develo ps. The fraction of the wall that is not submerged uses the vapor heat transfer coefficient as described above. GOTHIC
calculates the fraction of the walls that are submerged in the sump water. The floors
are submerged quickly.
Sump Recirculation
The calculated containment peak pressure and temperature occur before the transfer to
cold leg recirculation. However, a sump recirculation model comprised of simplified
RHR and CCW system models was added to the DCPP containment model for the
long-term LOCA containment pressure and temperature response calculation.
ECCS recirculation is actuated after a low RWST level signal and the ECCS takes suction from the containment sump. The RHR heat exchanger cools the water before it
is injected back into the reactor vessel. The RHR heat exchanger is cooled by CCW
and ASW provides the ultimate heat sink, cooling the CCW heat exchangers.
Switchover to hot leg recirculation is assumed to occur at 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.
6.2D.3.2.5 Boundary Conditions M&E Release
Section 6.2D.3.1 describes the long-term LOCA M&E release. The LOCA M&E release
rates are generated using the Westinghouse methodology (Reference 1). M&E
releases are calculated for both sides of the double-ended break in the coolant loop: the vessel side of the break and the steam generator side of the break. The M&E releases
are input to the GOTHIC containment model as mass flow rates and enthalpies via
boundary conditions connected to the containment volume with flow paths.
During blowdown, the liquid portion of the break flow is released as drops with an
assumed diameter of 100 microns (0.00394 inches). This is consistent with the DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-20 Revision 23 December 2016 methodology approved for Kewaunee (Reference 15) and is based on data presented in Reference 17. After blowdown, the liqui d release is assumed to be a continuous pour into the sump.
GOTHIC uses the M&E release tables from the time of accident initiation to 3,600 seconds, the time at which all energy in the primary heat structures and steam generator secondary system is assumed to be released/depressurized to atmospheric
pressure, (i.e., 14.7 psia and 212°F).
After primary system and secondary system energy have been released, the M&E releases to the containment are due to long-term steaming of decay heat. GOTHIC calculates the decay heat steaming M&E releases
within user defined control variables. The steaming calculations incorporate the
transient containment pressure and RHR recirculated ECCS enthalpy to calculate the
M&E release. The calculations are essentially the same as the Westinghouse
methodology previously approved by the NRC, except the calculations are performed within the GOTHIC code.
The ANS Standard 5.1 decay heat model (+2 sigma uncertainty) (Reference 7) is used
to calculate the long-term boil-off from the core. All the decay heat is assumed to
produce steam from the recirculated ECCS water. The remainder of the ECCS water is
returned to the sump region of the containment control volume. These assumptions are
consistent with the long-term M&E release methodology documented in Reference 1.
Containment Fan Coolers
The CFCUs are modeled with a GOTHIC cooler component. There are a total of five
CFCUs in three trains. In all cases, two CFCUs are assumed to be out of service for maintenance. An inherent assumption in the LOCA containment analysis is that offsite power is lost with the pipe rupture. This results in the actuation of three emergency
diesel generators (EDGs), powering the two trains of safeguards equipment. Startup of the EDGs delays the operation of the safeguards equipment that is required to mitigate
the transient. There are two trains of the SSPS that actuate the two trains of
emergency safeguards. The failure of one train of SSPS will fail one train of safeguards.
A minimum of two CFCUs are available and a maximum of three CFCUs are assumed to be available based on the single failure assumptions. In order to incorporate a fan cooler flow of 34,000 cfm (37,000 cfm for fan coolers with Super Radiator Coils), a conservative multiplier of 0.9600 was used on the fan cooler heat rate input calculated at 47,000 cfm.
Three long term cases are analyzed to assess the effects of single failures. The first
case assumes minimum safeguards based on the postulated single failure of an SSPS
train. This assumption results in the loss-of-one train of safeguards equipment. The
operating equipment is conservatively modeled as: two CFCUs, one containment spray pump, one train of RHR, and one CCW heat exchanger. The other two cases assume
maximum safeguards, in which both trains of SSP S are available. With the maximum safeguards cases, the single failure assumptions are the failure of one containment DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-21 Revision 23 December 2016 spray pump or the failure of one CFCU. The analysis of these three cases provides confidence that the effect of credible single failures is bounded.
For LOCA cases, the fan coolers in the containment evaluation model are modeled to actuate on the SI actuation, and begin removing heat from containment after a 64-second delay.
The CFCUs are cooled by CCW. The heat removal rate per containment fan cooler is
calculated as a function of containment steam saturation temperature, the CCW inlet
temperature and flow rate, and input to the GOTHIC cooler model. The heat removal
rate is multiplied by the number of CFCUs av ailable. The heat removed from the containment control volume is transferred to the CCW control volume receiving the flow
through the CFCUs using a coupled heater model.
Containment Spray System
The containment spray is modeled with a boundary condition. DCPP has two trains of
containment safeguards available, with one spray pump per train. An inherent assumption in the LOCA containment analysis is that offsite power is lost with the pipe
rupture. This results in the actuation of the three EDGs powering the two trains of
safeguards equipment. Startup of the EDGs delays the operation of the safeguards
equipment that is required to mitigate the transient.
Relative to the single failure criterion with respect to a LOCA event, one spray pump is
considered inoperable due to the SSPS failure (minimum safeguards case) or as a
single failure in a maximum safeguards case. In the maximum safeguards case, in which the single failure is assumed to be one CFCU, two spray pumps are available.
The containment spray actuation is modeled on the SI actuation, conservatively assumed to be 7.0 seconds. The sprays begin injecting 90°F water after a specified 100 second delay. The spray flow rate is a function of containment pressure and is presented in Table 6.2D-18. The containme nt spray is credited only during the injection phase of the transient and is terminated on a refueling water storage tank empty alarm
after switchover to cold leg recirculation at a time based on the number of SI and spray
pumps operating. The timing of recirculation and spray termination assumed in the
LOCA containment analysis are presented in Table 6.2D-17.
Accumulator Nitrogen Gas Modeling
The accumulator nitrogen gas release is modeled with a flow boundary condition in the
LOCA containment model. The nitrogen release rate was conservatively calculated by
maximizing the mass available to be injected. The nitrogen gas release rate was used as input for the GOTHIC function, as a specified rate over a fixed time period. Nitrogen
gas was released to the containment at a rate of 327.4 lbm/s. The release begins at
51.9 seconds, the minimum accumulator tank water depletion time.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-22 Revision 23 December 2016
6.2D.3.2.6 LOCA Containment Integrity Analysis Results
The containment pressure, steam temperature, and water (sump) temperature profiles
of the DEHL peak pressure case are shown in Figures 6.2D-3 through 6.2D-5.
Table 6.2D-13 provides the transient sequence of events for the DEHL transient.
The containment pressure, steam temperature, and water (sump) temperature profiles
of the DEPS long-term EQ temperature transient are shown in Figures 6.2D-6 through
6.2D-8(The peak DEPS values are from Unit 2).
The sequence of events for the Unit 1
and Unit 2 DEPS transients are presented in Tables 6.2D-14 and 6.2D-15, respectively.
The peak pressure (Figure 6.2D-6) for the D EPS case occurs at 23.5 seconds after the end of the blowdown. The fans begin to cool the containment at 71.0 seconds.
Containment sprays begin injecting at 107.0 seconds. The pressure comes down as the steam generators reach equilibrium with the containment environment, but spikes
up again at recirculation when the CCW temperature increases and the CCW flow rate
to the CFCUs decreases. The sensible heat release from the steam generator
secondary system and RCS metal is completed at 3600 seconds, but at 3798 seconds, the RWST reaches a low level alarm and spray flow is terminated. The containment
pressure increases for a time and then begins to decrease over the long term as the RHR heat exchangers and CFCUs remove the heat from the containment.
Table 6.2D-21 summarizes the containment peak pressure and temperature results and
pressure and temperature at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for EQ support and the acceptance limits for
these parameters.
A review of the results presented in Table 6.2D-21 shows that the analysis margin (analysis margin is the difference between the calculated peak pressure and
temperature and the acceptance limits) is maintained for DCPP. From the GOTHIC
analysis the containment peak pressure is 41.4 psig. At 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the maximum
containment pressure is 8.5 psig and the maximum temperature is 164.97°F.
6.2D.3.2.7 Conclusion The DCPP containment can adequately account for the M&E releases that would result
from a LOCA. The DCPP containment systems will provide sufficient pressure and
temperature mitigation capability to ensure that containment integrity is maintained.
- The peak calculated pressure is less than the containment design pressure of 47 psig
- The peak calculated containment average air temperature is less than the containment design temperature of 271 °F
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-23 Revision 23 December 2016
- The calculated pressure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is less than 50 percent of the peak calculated value 6.2D.4 LONG-TERM MAIN STEAMLINE BREAK INSIDE CONTAINMENT
6.2D.4.1 MSLB Mass and Energy Release Analysis
6.2D.4.1.1 Acceptance Criteria
There are no direct acceptance criteria for MSLB M&E releases. The analysis methods
follow the guidelines provided by the USNRC with respect to the sources of M&E during the various phases of a MSLB transient (refer to Section 6.2D.4.1.4).
The specific acceptance criteria for the containment response to a MSLB are discussed
in Section 6.2D.4.2.1.
6.2D.4.1.2 Introduction and Background The MSLB is classified as an American Nuclear Society (ANS) Condition IV event, an
infrequent fault. A MSLB occurring inside a reactor containment structure may result in
significant releases of high-energy fluid to the containment environment that could
produce high pressure conditions for extended period s of time. The magnitude of the releases following a MSLB is dependent upon the plant initial operating conditions and
the size of the rupture as well as the configuration of the plant steam system and the
containment design. There are competing effects and credible single failures in the postulated accident scenario used to determine the worst cases for containment
pressure and the associated containment temperature following a MSLB.
The DCPP MSLB and containment response analysis considers a spectrum of cases
that vary the initial power condition, break size, and the postulated single failure.
6.2D.4.1.3 Input Parameters and Assumptions Major assumptions affecting the M&E releases to containment are summarized below:
Initial Steam Generator Inventory
A high initial steam generator mass is assumed. The initial level corresponds to 75 percent narrow range span (NRS) at all power levels. This consists of a nominal level
of 65 percent NRS plus a steam generator water level control uncertainty of 10 percent
NRS.
Main Feedwater System
The rapid depressurization that occurs following a MSLB typically results in large
amounts of water being added to the steam generators through the main feedwater DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-24 Revision 23 December 2016 system. A rapid-closing main feedwater regulating valve (MFRV) near each steam generator limits this effect. The feedwater addition to the faulted steam generator is
maximized to be conservative since it increases the water mass inventory that will be
converted to steam and released from the break.
Following the initiation of the MSLB, main feedwater flow is conservatively modeled by assuming that sufficient feedwater flow is provided to match or exceed the steam flow
prior to reactor trip. The initial increase in feedwater flow is in response to the MFRV
opening in response to the steam flow/feedwater flow mismatch and the lower
backpressure on the feedwater pump as a result of the depressurizing steam generator.
This maximizes the total mass addition prior to feedwater isolation. The feedwater
isolation response time, following the SI signal, is assumed to be a total of 9 seconds, accounting for delays associated with signal processing plus the valve stroke time.
The feedwater in the unisolable feedl ine between the MFRV and faulted steam generator is also considered in the analysis. The hot main feedwater reaches saturated
conditions as the steam generator and feedline depressurize. The decrease in density
as flashing occurs causes most of the unisolable feedwater to enter the faulted steam
generator. This unisolable feedwater line volume of 208 ft 3 is an additional source of fluid that can increase the mass discharged out of the break.
Some cases postulate the MFRV on the faulted loop failing open. Refer to Section
6.2D.3.1.4 for information on the effects of this single failure.
Auxiliary Feedwater Within the first minute following a MSLB, an SI signal is generated. Immediately upon receipt of the SI signal, auxiliary f eedwater (AFW) is initiated. Addition of AFW to the faulted steam generator will increase the secondary mass available for release to the
containment. The AFW flowrate to the faulted steam generator is maximized based on flow from both motor-driven AFW pumps and the turbine-driven AFW pump. The AFW
flowrate is modeled as a function of steam g enerator pressure, varying from 569 gpm to 1588 gpm to the faulted steam generator.
Operator action is credited to terminate the AFW flow to the faulted steam generator
after 10 minutes.
Unisolable Steamline
The initial steam in the main steamline between the break and the main steamline
isolation valve (MSIV) and check valve (CV) is included in the M&E released from the
break. The MSIV/CV is considered a single plant component and is credited to prevent
reverse flow from the main steamline header and intact steam generators for most cases. Cases that postulate the failure of the MSIV/CV are discussed in
Section 6.2D.3.1.4.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-25 Revision 23 December 2016
Quality of the Break Effluent
The quality of the break effluent is assumed to be 1.0, corresponding to saturated steam
that is all vapor with no liquid. Although it is expected that there would be a significant
quantity of liquid in the break effluent for a full double-ended rupture, the all-vapor
assumption conservatively maximizes the energy addition to the containment
atmosphere.
Reactor Coolant System Assumptions
While the M&E released from the break is determined from assumptions that have been
discussed above, the rate at which the release occurs is largely controlled by the
conditions in the RCS. The major features of the primary side analysis model are
summarized below:
- Continued operation of the reactor coolant pumps maintains a high heat transfer rate to the steam generators.
- The model includes consideration of the heat that is stored in the RCS metal.
- Reverse heat transfer from the intact steam generators to the RCS is modeled as the temperature in the RCS falls below the intact steam
generator fluid temperature.
- Minimum flowrates are modeled from ECCS injection to conservatively minimize the amount of boron that provides negative reactivity feedback.
Both the high-head and intermediate-head SI systems are considered
available, with unborated purge volumes of 75.9 ft3 and 19.4 ft3, respectively. A failure of one train of ECCS is included for cases that also
model a failure on a containment safeguards train (refer to Section
6.2D.3.1.4).
- The initial NSSS thermal power output assumed is 3,425 MWt, which includes the thermal power generated by the reactor coolant pumps minus
heat losses to the containment and the letdown system.
- RCS average temperature is the full-power nominal value of 577.3°F (DCPP Unit 1) or 577.6°F (DCPP Unit 2) plus an uncertainty of +5.0°F.
- Core residual heat generation is assumed based on the 1979 ANS decay heat plus 2 sigma model (Reference 7).
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-26 Revision 23 December 2016
- Conservative core reactivity coefficients (e.g. moderator temperature) corresponding to end-of-cycle conditions with the most reactive rod stuck out of the core are assumed. This maximizes the reactivity feedback
effects as the RCS cools down as a result of the MSLB.
- All cases have credited a minimum shutdown margin of 1.6 percent k.
6.2D.4.1.4 Description of Analyses and Evaluations The following limitations in the NRC Safety Evaluation Report for WCAP-14882-P-A have been adhered to in the use of RETRAN-02W to analyze this event.
- The break flow model is the Moody model.
- Only steam (dry vapor) will exit the break, since perfect steam separation in the steam generators is assumed.
- The superheat in the steam released to the containment is not evaluated.
Any superheated conditions will be reset to be equal to the saturation
temperature.
Case Definitions and Single Failures There are many factors that influence the quantity and rate of the M&E release from the main steamline. To encompass these factors, a spectrum of cases varies the initial power level, break size and the single failure. This section summarizes the basis of the
cases that have been defined for DCPP.
The power level at which the plant is operating when the MSLB is postulated can cause
different competing effects that make it difficult to pre-determine a single limiting case.
For example, at higher power levels there is less initial water/steam in the steam generator, which is a benefit. However, at a higher power level there is a higher initial
feedwater flowrate, higher feedwater temperature, higher decay heat, and there is a
higher rate of heat transfer from the primary side, which are all penalties. Therefore, cases consider initial power levels varying from full power to zero power. The specific
initial power levels that are analyzed are 100, 70, 30 and 0 percent, as presented in
WCAP-8822. A calorimetric uncertainty of 2 percent is applied to the initial condition for
the full power case.
Most cases consider the largest possible break, a double-ended rupture immediately
downstream of the flow restrictor at the outlet of the steam generator. This break
conservatively bounds the plant response to a smaller break size. The effective forward
break area is limited by the 1.4 ft 2 cross-sectional area of the flow restrictor that is integral to the replacement steam generator.
The actual break area is the cross-
sectional area of the pipe, which is 3.67 ft
- 2.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-27 Revision 23 December 2016 A few cases also consider the effects of a smaller, split break which allows contributing steam from the main steamline header and the intact steam generators until the intact loop MSIVs close. This break size is defined to be the largest break that does not
generate a low main steamline pressure signal. Instead, split breaks have to rely on
high containment pressure signals to actuate SI (and reactor trip, feedline isolation, etc.)
and main steamline isolation. The split break is only considered when the faulted loop
MSIV/CV is postulated as the single failure. The split breaks have the penalty of a
higher integrated M&E released to the containment, but the smaller break size provides
a beneficial reduction in the rate that the M&
E is released.
Several single failures can be postulated that would impair the performance of various
MSLB protection systems. The single failures either reduce the heat removal capacity
of the containment safeguards systems or increase the energy release from the MSLB.
The single failures that have been postulated for DCPP are summarized below. The
analysis cases separately consider each single failure at each initial power level.
(1) Containment Safeguards Failure
This is a failure of one safeguards train.
The main impact is on the containment response analysis, where the active heat removal is reduced by the loss of one
train of fan coolers and one containment spray pump. This failure also causes the loss of one train of ECCS in the MSLB M&E release analysis.
(2) MFRV Failure
The MFRV is a fast-closing (7 second stroke time) valve in the feedwater system that is the preferred (fastest) method for terminating feedwater addition to the faulted steam generator during a MSLB. If the MFRV on the faulted loop fails
open, the back-up main feedwater isolation valve (MFIV) is credited to close 64
seconds after an SI setpoint is reached. The slower closure time creates the
possibility of additional pumped feedwater entering the faulted steam generator.
Although the main feedwater pumps trip on an SI signal, the condensate pumps
do not trip and can continue to provide pumped flow when the faulted steam
generator depressurizes below approx imately 625 psia.
(3) MSIV/CV Failure The MSIV/CV is considered a single plant component that is credited to function
when another single failure is postulated. When the MSIV/CV is postulated as
the single failure, this means the failure of both the forward flow isolating function (MSIV) and the reverse flow isolation function (CV) on the faulted loop.
Therefore the main steamline blowdown initially includes steam mass from the main steamline header and intact steam generators. Isolation of the break
occurs due to the closure of the MSIVs on the intact steam generators.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-28 Revision 23 December 2016
Protection Logic and Setpoints
The pertinent signals and setpoints that are actuated in these analyses are summarized
below.
The first SI signal is generated by a low steamline pressure signal in all double-ended rupture cases. The assumed setpoint is 458.7 psia (DCPP low steamline pressure
setpoint is 600 psig), with a lead/lag of 50/5 seconds. For split rupture cases, the first SI
signal that is credited in the analysis comes from the high containment pressure
setpoint (5.0 psig) (DCPP high containment pressure setpoint is 3 psig). The SI signal
is credited to cause:
- Start of auxiliary feedwater pumps
- Closure of MFRVs and MFIVs
Most cases isolate the MSLB to blowdown of a single steam generator by fast closure of the passive CV, which prevents reverse flow from the intact main steamline. For cases that model the MSIV/CV failure on the faulted loop, the closure of the intact MSIVs are
credited due to the low steamline pressure signal (for double-ended ruptures) or the high-high containment pressure signal (for split breaks)(DCPP high-high containment pressure setpoint is 22 psig which initiates Phase B containment isolation).
6.2D.4.1.5 MSLB Inside Containment Mass and Energy Release Results Sixteen MSLB cases were analyzed varying the initial power level, break size and the assumed single failure.
The limiting containment pressure case is double-ended rupture MSLB initiated from 70
percent power with the faulted loop MFRV failed open. The break flowrate is shown in
Figure 6.2D-1 and the break enthalpy is shown in Figure 6.2D-2. (refer to Section
6.2D.4.2 for the basis of this case being the limiting transient.) Section 6.2D.4.2 also
contains the sequence of events for this case, including primary, secondary, and
containment system actuations.
A sensitivity analysis was performed to consider the effects of the plant response for Unit 1 versus Unit 2. It was determined that the plant response to this event is DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-29 Revision 23 December 2016 essentially the same for either unit. The MSLB analysis results bound the plant response for either Unit 1 or Unit 2.
The MSLB inside containment event has been analyzed with conservative assumptions to maximize the M&E release from the break. The M&E releases from this analysis are used as input to the containment integrity analysis documented in Section 6.2D.4.2.
6.2D.4.2 Long-Term MSLB Containment Integrity Analysis 6.2D.4.2.1 Acceptance Criteria The containment response to a MSLB is analyzed to ensure that the containment pressure remains below the containment design pressure of 47.0 psig. There is no
explicit design temperature limit to be met for the MSLB containment response.
6.2D.4.2.2 Introduction and Background Containment integrity analyses are performed to ensure that pressure inside containment will remain below the containment building design pressure for a postulated secondary system pipe rupture. The M&E release analysis discussed in Section
6.2D.4.1 is input to this analysis.
6.2D.4.2.3 Input Parameters and Assumptions This section identifies the major input values that are used in the MSLB containment
response analysis. The assumed initial conditions and the input assumptions associated with the fan coolers and containment sprays are listed in Table 6.2D-22.
The containment thermal conductor input is provided in Table 6.2D-19, and the
corresponding material properties are listed in Table 6.2D-20.
6.2D.4.2.4 Description of Analyses and Evaluations The containment response analysis uses the GOTHIC computer code. The GOTHIC
Technical Manual (Reference 13) provides a description of the governing equations, constitutive models, and solution methods in the solver. The GOTHIC Qualifications
Report (Reference 14) provides a comparison of the solver results with both analytical
solutions and experimental data.
The DCPP GOTHIC containment evaluation model consists of a single lumped-
parameter node; the diffusion layer model is used for heat transfer to all structures in
the containment. Plant input assumptions (identified in Section 6.2D.3.2.3) are the
same as, or slightly more restrictive, than in the original analyses performed with the
COCO code (Reference 18). Benchmarking between the DCPP COCO and GOTHIC
models was performed to confirm consistency in the implementation of the plant input
values. The benchmarking results show that the GOTHIC model predicted similar
transient results.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-30 Revision 23 December 2016 This MSLB containment response analysis uses GOTHIC version 7.2a. This code version is used to take advantage of the diffusion layer model heat transfer option. This
heat transfer option was approved by the NRC (Reference 19) for use in Kewaunee
containment analyses with the condition that mist be excluded from what was earlier
termed as the mist diffusion layer model.
The GOTHIC containment modeling for DCPP has followed the conditions of acceptance placed on Kewaunee. Kewaun ee and DCPP both have large, dry containments. Changes in the GOTHIC code versions are detailed
in Appendix A of the GOTHIC User Manual Release Notes (Reference 16). Version 7.2a is used consistent with the restrictions identified in Reference 19; none of the user-controlled enhancements added to version 7.2a were implemented in the DCPP containment model.
6.2D.4.2.5 Results Sixteen MSLB cases were analyzed varying the initial power level and the assumed single failure. The M&E release from the break was calculated using the RETRAN-02W code (refer to Section 6.2D.4.1.4), while the containment pressure response was
determined with the GOTHIC code. The MSLB results are based on Unit 2
configuration, which was determined to be bounding for Unit 1.
The peak pressures and peak temperatures from the spectrum of cases are listed in Table 6.2D-23. The limiting peak pressure case is a MFRV failure at 70 percent power
assuming a full double-ended rupture. The sequence of events for this limiting case is
listed in Table 6.2D-24. The containment pressure and temperature transient for the
limiting case is shown in Figures 6.2D-9 and 6.2D-10, respectively.
6.2D.4.2.6 Conclusions The peak containment pressure is 42.4 psig, which is below the containment design pressure of 47.0 psig. The analysis demonstrates that the containment pressure remains below the containment design pressure throughout the transient for a
postulated secondary system pipe rupture. Thus the containment integrity has been
demonstrated and the applicable acceptance criteria are therefore met.
6.2D.5 REFERENCES
- 1. Westinghouse LOCA Mass and Energy Release Model for Containment Design
- March 1979 Version, WCAP-10325-P-A (Proprietary), WCAP-10326-A (Non-Proprietary), May 1983.
- 2. Deleted in Revision 22.
- 3. License Amendment No.126 for D. C.
Cook Nuclear Plant Unit 1, Operating License No. DPR-58 (TAC No. 71062), June 9, 1989.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-31 Revision 23 December 2016
- 4. Mixing of Emergency Core Cooling Water with Steam; 1/3-Scale Test and Summary, WCAP-8423, EPRI 294-2, Final Report, June 1975.
- 5. Westinghouse Mass and Energy Release Data for Containment Design, WCAP-8264-P-A, Rev. 1 (Proprietary), WCAP-8312-A, Rev. 2 (Non-Proprietary), August 1975.
- 6. Letter from Herbert N. Berkow, Director (NRC) to James A. Gresham (Westinghouse), Acceptance of Clarifications of Topical Report WCAP-10325-P-A, Westinghouse LOCA Mass and Energy Release Model for Containment Design - March 1979 Version, (TAC No. MC7980).
- 7. ANSI/ANS-5.1 1979, American National Standard for Decay Heat Power in Light Water Reactors, August 29, 1979.
- 8. Letter from Cecil O. Thomas (NRC), Acceptance for Referencing of Licensing Topical Report WCAP-8821(P)/8859(NP), TRANFLO Steam Generator Code
Description, and WCAP-8822(P)/8860(NP), Mass and Energy Release Following a Steam Line Rupture, August 1983.
- 9. R. E. Land, Mass and Energy Releases Following a Steam Line Rupture, WCAP-8822 (Proprietary), WCAP-8860 (Non-Proprietary), September 1976.
- 10. Deleted in Revision 22
- 11. J. C. Butler and P. A. Lin, Mass and Energy Releases Following a Steam Line Rupture, Supplement 2 - Impact of Steam Superheat in Mass/Energy Releases Following a Steamline Rupture for Dry and Subatmospheric Containment Designs, WCAP-8822-S2-P-A (Proprietary), September 1986
- 12. D. S. Huegel, et al. RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses, [RETRAN-02W] WCAP-14882-P-A (Proprietary), April 1999, WCAP-15234-A (Non-Proprietary),
May 1999.
- 13. GOTHIC Containment Analysis Package Technical Manual, Version 7.2a, NAI-8907-06, Rev. 16, January 2006.
- 14. GOTHIC Containment Analysis Package Qualification Report, Version 7.2a, NAI-8907-09, Rev. 9, January 2006.
- 15. License Amendment No. 169 for Kewaunee Nuclear Power Plant, Operating License No. DPR-43 (TAC No. MB6408), September 29, 2003.
- 16. GOTHIC Containment Analysis Package User Manual, Version 7.2a, NAI-8907-02, Rev. 17, January 2006.
DCPP UNITS 1 &
2 FSAR UPDATE 6.2D-32 Revision 23 December 2016
- 17. Brown and York, Sprays formed by Flashing Liquid Jets, AICHE Journal, Volume 8, #2, May 1962.
- 18. F. M. Bordelon and E. T. Murphy, Containment Pressure Analysis Code (COCO), WCAP-8327 (Proprietary), WCAP-8326 (Non-Proprietary), July 1974.
- 19. Letter from Anthony C. McMurtray (NRC) to Thomas Coutu (NMC), Enclosure 2, Safety Evaluation, September 29, 2003.
- 20. CN-CRA-12-5, Rev 0, Diablo Canyon Containment Integrity Reanalysis to Address CFCU Fan Flow and NSAL-11-5 Issues
- 21. CN-CRA-14-6, Rev 0, Diablo Canyon Units 1 & 2, Delay in CFCU Start and a Decrease in Containment Heat Removal Capability
- 22. CN-CRA-14-7, Rev 0, Diablo Canyon Units 1 & 2, Steamline Break Containment Response for Revised CFCU and CS
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 6.2D-1 SYSTEM PARAMETERS INITIAL CONDITIONS Parameters Value Core Thermal Power (MWt)
(1) 3479.0 RCS Total Flow Rate (lbm/sec) 37222.22 Vessel Outlet Temperature (°F) 615.1 Core Inlet Temperature (°F) 550.1 Vessel Barrel-Baffle Configuration (2) Upflow Initial Steam Generator Steam Pressure (psia) 903.0 Steam Generator Design Steam Generator Tube Plugging (%)
0 Initial Steam Generator Secondary Side Mass (lbm) 132,953.7 Assumed Maximum Containment Backpressure (psia) 61.7 Accumulator Water volume (ft
- 3) per accumulator 850.0 N 2 cover gas pressure (psia) 695.2 Temperature (°F) 120.0 SI Start Time, (sec) [total time from beginning of event which includes the maximum delay from reaching the setpoint]
42.0 Notes: 1. Core thermal power, RCS total flow rate, RCS coolant temperatures, and SG secondary-side mass include appropriate uncertainty and/or allowance. 2. The upflow configuration of Unit 2 bounds the downflow configuration of Unit 1.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 18 October 2008 TABLE 6.2D-2 SI FLOW MINIMUM SAFEGUARDS - BOTH UNITS RCS Pressure (psig) Total Flow (gpm)
Injection Mode (reflood phase) 0 4805.0 20 4558.8 40 4298.6 60 4017.4 80 3710.2 100 3363.0 120 2950.4 140 2403.8 160 1386.2 180 780.6 200 775.0 Cold-Leg Recirculation Flow 3252.3 Hot-Leg Recirculation Flow 3071.7 DCPP UNITS 1 & 2 FSAR TABLE 6.2D-3 Sheet 1 of 4 DEHL BREAK NO SAFETY INJECTION MASS AND ENERGY RELEASES DURING BLOWDOWN Revision 23 December 2016 Time(s) Break Path No. 1 Flow(1) Break Path No. 2 Flow(2) (lbm/s) Thousand (Btu/s) (lbm/s) Thousand (Btu/s) 0.000000E+00 0.000000E+00 0.000000E+00 0.000000E+00 0.000000E+00 1.062358E-03 4.421330E+04 2.795865E+04 4.421073E+04 2.795568E+04 2.116870E-03 4.513561E+04 2.854166E+04 4.484486E+04 2.835063E+04 1.011610E-01 4.558368E+04 2.916997E+04 2.631157E+04 1.660517E+04 2.012629E-01 3.325809E+04 2.154267E+04 2.336337E+04 1.466358E+04 3.015284E-01 3.328901E+04 2.151881E+04 2.074400E+04 1.287011E+04 4.018289E-01 3.233691E+04 2.087650E+04 1.940834E+04 1.185700E+04 5.016486E-01 3.179156E+04 2.051087E+04 1.860073E+04 1.118752E+04 6.010187E-01 3.175035E+04 2.047324E+04 1.806742E+04 1.071491E+04 7.010036E-01 3.170974E+04 2.045320E+04 1.764590E+04 1.033815E+04 8.020547E-01 3.143076E+04 2.030445E+04 1.731057E+04 1.003537E+04 9.019648E-01 3.098787E+04 2.006708E+04 1.705860E+04 9.802414E+03 1.001210E+00 3.054152E+04 1.983958E+04 1.686041E+04 9.616324E+03 1.101514E+00 3.017290E+04 1.967163E+04 1.671930E+04 9.473734E+03 1.201277E+00 2.999474E+04 1.963587E+04 1.659974E+04 9.353070E+03 1.301412E+00 2.976682E+04 1.957221E+04 1.656717E+04 9.287328E+03 1.401996E+00 2.943739E+04 1.943873E+04 1.658615E+04 9.255053E+03 1.501864E+00 2.901809E+04 1.924138E+04 1.662590E+04 9.239039E+03 1.602090E+00 2.857801E+04 1.902255E+04 1.668783E+04 9.239026E+03 1.701623E+00 2.821512E+04 1.884598E+04 1.676359E+04 9.250566E+03 1.802174E+00 2.792087E+04 1.871164E+04 1.684624E+04 9.269260E+03 1.901835E+00 2.758637E+04 1.854544E+04 1.692639E+04 9.290490E+03 2.001085E+00 2.715653E+04 1.830796E+04 1.699545E+04 9.309264E+03 2.101509E+00 2.666613E+04 1.802233E+04 1.705422E+04 9.325545E+03 2.201561E+00 2.621646E+04 1.776084E+04 1.710275E+04 9.339314E+03 2.301787E+00 2.583163E+04 1.754326E+04 1.714244E+04 9.350919E+03 2.401518E+00 2.545424E+04 1.732786E+04 1.717032E+04 9.358416E+03 2.501914E+00 2.503389E+04 1.707628E+04 1.718287E+04 9.359553E+03 2.602138E+00 2.459066E+04 1.680199E+04 1.718151E+04 9.354817E+03 2.702086E+00 2.415870E+04 1.653089E+04 1.716793E+04 9.344832E+03 2.801309E+00 2.376343E+04 1.628281E+04 1.714481E+04 9.330784E+03 2.901624E+00 2.339761E+04 1.605365E+04 1.711252E+04 9.312652E+03 3.001978E+00 2.304258E+04 1.582842E+04 1.707259E+04 9.291107E+03 3.101309E+00 2.267665E+04 1.558865E+04 1.702505E+04 9.266022E+03 3.201600E+00 2.232172E+04 1.535143E+04 1.696861E+04 9.236593E+03 3.301245E+00 2.200559E+04 1.513983E+04 1.690832E+04 9.205429E+03 DCPP UNITS 1 & 2 FSAR TABLE 6.2D-3 Sheet 2 of 4 DEHL BREAK NO SAFETY INJECTION MASS AND ENERGY RELEASES DURING BLOWDOWN Revision 23 December 2016 Time(s) Break Path No. 1 Flow(1) Break Path No. 2 Flow(2) (lbm/s) Thousand (Btu/s) (lbm/s) Thousand (Btu/s) 3.401754E+00 2.170027E+04 1.493186E+04 1.684303E+04 9.171848E+03 3.501091E+00 2.141111E+04 1.473030E+04 1.677380E+04 9.136342E+03 3.601199E+00 2.115167E+04 1.454707E+04 1.670002E+04 9.098566E+03 3.701428E+00 2.090290E+04 1.436717E+04 1.662281E+04 9.059060E+03 3.801487E+00 2.066214E+04 1.418769E+04 1.654094E+04 9.017175E+03 3.901424E+00 2.045532E+04 1.402983E+04 1.645629E+04 8.973907E+03 4.001087E+00 2.026789E+04 1.388271E+04 1.636891E+04 8.929290E+03 4.200070E+00 1.993782E+04 1.360944E+04 1.618343E+04 8.834705E+03 4.401104E+00 1.967616E+04 1.337388E+04 1.597713E+04 8.729512E+03 4.600449E+00 1.948249E+04 1.317866E+04 1.575242E+04 8.614931E+03 4.800778E+00 1.936158E+04 1.302737E+04 1.550211E+04 8.487160E+03 5.000381E+00 1.933507E+04 1.293779E+04 1.518745E+04 8.324583E+03 5.200546E+00 1.933505E+04 1.286676E+04 1.473428E+04 8.086715E+03 5.401286E+00 1.938190E+04 1.283109E+04 1.438802E+04 7.910627E+03 5.600623E+00 1.945952E+04 1.281432E+04 1.417807E+04 7.809886E+03 5.800968E+00 1.960274E+04 1.281936E+04 1.375662E+04 7.589370E+03 6.000423E+00 1.978556E+04 1.284097E+04 1.333628E+04 7.369933E+03 6.200714E+00 2.005063E+04 1.290744E+04 1.299385E+04 7.194837E+03 6.401191E+00 2.046116E+04 1.303434E+04 1.262515E+04 7.004489E+03 6.600091E+00 1.547757E+04 1.086055E+04 1.229050E+04 6.832616E+03 6.800170E+00 1.530222E+04 1.067566E+04 1.195706E+04 6.660028E+03 7.000022E+00 1.549690E+04 1.069490E+04 1.160642E+04 6.476845E+03 7.200609E+00 1.563658E+04 1.071619E+04 1.130097E+04 6.318775E+03 7.400177E+00 1.578609E+04 1.072046E+04 1.100946E+04 6.167301E+03 7.600533E+00 1.593356E+04 1.070815E+04 1.072214E+04 6.017282E+03 7.801034E+00 1.615061E+04 1.077797E+04 1.043611E+04 5.867256E+03 8.000298E+00 1.632238E+04 1.079936E+04 1.016568E+04 5.725268E+03 8.200516E+00 1.639457E+04 1.077841E+04 9.913078E+03 5.592690E+03 8.400596E+00 1.650113E+04 1.079726E+04 9.671393E+03 5.465471E+03 8.601646E+00 1.636407E+04 1.063906E+04 9.437511E+03 5.342136E+03 8.801463E+00 1.656999E+04 1.068611E+04 9.214691E+03 5.224589E+03 9.000441E+00 1.676613E+04 1.073297E+04 9.001371E+03 5.112124E+03 9.201846E+00 1.696588E+04 1.078521E+04 8.793862E+03 5.002893E+03 9.401161E+00 1.717369E+04 1.084538E+04 8.592599E+03 4.896921E+03 9.601131E+00 1.741410E+04 1.092530E+04 8.393214E+03 4.792029E+03 9.801950E+00 1.776281E+04 1.106534E+04 8.196650E+03 4.688971E+03 1.000068E+01 1.839142E+04 1.135395E+04 8.003922E+03 4.588161E+03 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2D-3 Sheet 3 of 4 DEHL BREAK NO SAFETY INJECTION MASS AND ENERGY RELEASES DURING BLOWDOWN Revision 23 December 2016 Time(s) Break Path No. 1 Flow(1) Break Path No. 2 Flow(2) (lbm/s) Thousand (Btu/s) (lbm/s) Thousand (Btu/s) 1.020140E+01 1.851964E+04 1.140232E+04 7.810060E+03 4.487047E+03 1.020362E+01 1.851891E+04 1.140156E+04 7.808551E+03 4.486299E+03 1.040086E+01 1.833975E+04 1.126025E+04 7.618512E+03 4.387523E+03 1.060050E+01 1.804366E+04 1.104730E+04 7.425827E+03 4.287746E+03 1.080127E+01 1.635122E+04 1.012610E+04 7.230032E+03 4.186827E+03 1.100122E+01 1.495018E+04 9.366321E+03 7.043013E+03 4.091326E+03 1.120105E+01 1.495589E+04 9.329105E+03 6.857841E+03 3.997307E+03 1.140234E+01 1.505645E+04 9.349147E+03 6.683749E+03 3.910070E+03 1.160104E+01 1.518993E+04 9.395547E+03 6.524124E+03 3.830459E+03 1.180020E+01 1.534510E+04 9.458602E+03 6.372764E+03 3.754714E+03 1.200164E+01 1.543409E+04 9.486852E+03 6.226401E+03 3.681190E+03 1.220032E+01 1.547409E+04 9.492495E+03 6.086830E+03 3.610949E+03 1.240054E+01 1.542898E+04 9.455210E+03 5.949447E+03 3.541731E+03 1.260148E+01 1.504712E+04 9.243750E+03 5.813945E+03 3.473712E+03 1.280136E+01 1.404974E+04 8.712991E+03 5.678712E+03 3.406016E+03 1.300163E+01 1.341850E+04 8.370817E+03 5.547233E+03 3.340880E+03 1.320039E+01 1.321660E+04 8.247448E+03 5.418112E+03 3.277430E+03 1.340088E+01 1.309170E+04 8.167251E+03 5.292336E+03 3.216163E+03 1.360132E+01 1.292232E+04 8.068071E+03 5.169267E+03 3.156194E+03 1.380186E+01 1.266645E+04 7.927508E+03 5.041325E+03 3.093249E+03 1.400069E+01 1.232011E+04 7.743278E+03 4.905249E+03 3.026344E+03 1.420193E+01 1.189811E+04 7.521688E+03 4.753275E+03 2.952577E+03 1.440144E+01 1.146758E+04 7.298080E+03 4.588016E+03 2.874162E+03 1.460128E+01 1.106268E+04 7.091303E+03 4.409904E+03 2.791234E+03 1.480068E+01 1.068885E+04 6.905159E+03 4.224909E+03 2.705894E+03 1.500023E+01 1.033246E+04 6.733536E+03 4.040939E+03 2.620500E+03 1.520003E+01 9.984955E+03 6.576100E+03 3.864887E+03 2.536414E+03 1.540118E+01 9.317167E+03 6.407015E+03 3.709484E+03 2.459092E+03 1.560109E+01 8.714224E+03 6.305064E+03 3.572296E+03 2.385381E+03 1.580010E+01 8.266204E+03 6.250359E+03 3.461553E+03 2.320330E+03 1.600106E+01 7.696312E+03 6.075593E+03 3.370571E+03 2.260741E+03 1.620100E+01 7.124064E+03 5.829973E+03 3.295617E+03 2.207467E+03 1.640032E+01 6.541715E+03 5.540556E+03 3.231165E+03 2.160269E+03 1.660015E+01 5.772028E+03 5.130904E+03 3.171035E+03 2.117409E+03 1.680034E+01 5.318274E+03 4.796410E+03 3.109013E+03 2.077173E+03 1.700009E+01 5.008771E+03 4.554054E+03 3.047012E+03 2.040953E+03 DCPP UNITS 1 & 2 FSAR TABLE 6.2D-3 Sheet 4 of 4 DEHL BREAK NO SAFETY INJECTION MASS AND ENERGY RELEASES DURING BLOWDOWN Revision 23 December 2016 Time(s) Break Path No. 1 Flow(1) Break Path No. 2 Flow(2) (lbm/s) Thousand (Btu/s) (lbm/s) Thousand (Btu/s) 1.720029E+01 4.694776E+03 4.337421E+03 2.981612E+03 2.004773E+03 1.740054E+01 4.381731E+03 4.131616E+03 2.903048E+03 1.963422E+03 1.760086E+01 4.072314E+03 3.927940E+03 2.813318E+03 1.922735E+03 1.780037E+01 3.798619E+03 3.740319E+03 2.712476E+03 1.883886E+03 1.800024E+01 3.577700E+03 3.585851E+03 2.599361E+03 1.844563E+03 1.820044E+01 3.434225E+03 3.452000E+03 2.475408E+03 1.802445E+03 1.840049E+01 3.302377E+03 3.322555E+03 2.348470E+03 1.759088E+03 1.860098E+01 3.169430E+03 3.190353E+03 2.224475E+03 1.715486E+03 1.880010E+01 3.027789E+03 3.048826E+03 2.107528E+03 1.672662E+03 1.900057E+01 2.868351E+03 2.906054E+03 1.998177E+03 1.631855E+03 1.920019E+01 2.709716E+03 2.766480E+03 1.896698E+03 1.594088E+03 1.940071E+01 2.575033E+03 2.627882E+03 1.800793E+03 1.559035E+03 1.960032E+01 2.452574E+03 2.501322E+03 1.706999E+03 1.524038E+03 1.980017E+01 2.292345E+03 2.388876E+03 1.616612E+03 1.492169E+03 2.000005E+01 2.129252E+03 2.281103E+03 1.521493E+03 1.465201E+03 2.020031E+01 1.963596E+03 2.170189E+03 1.420540E+03 1.438942E+03 2.040059E+01 1.821726E+03 2.070901E+03 1.307619E+03 1.410822E+03 2.060048E+01 1.686111E+03 1.969454E+03 1.207293E+03 1.373882E+03 2.080012E+01 1.567739E+03 1.881512E+03 1.144899E+03 1.329894E+03 2.100034E+01 1.455380E+03 1.774918E+03 1.122059E+03 1.301265E+03 2.120069E+01 1.367142E+03 1.681319E+03 1.122412E+03 1.285166E+03 2.140071E+01 1.268643E+03 1.567466E+03 1.120282E+03 1.271598E+03 2.160041E+01 1.184332E+03 1.468961E+03 1.086405E+03 1.251029E+03 2.180002E+01 1.099172E+03 1.358462E+03 1.007622E+03 1.208777E+03 2.200024E+01 8.315474E+02 1.041161E+03 8.745183E+02 1.068759E+03 2.220025E+01 5.277383E+02 6.618167E+02 7.617935E+02 9.355157E+02 2.240053E+01 3.415200E+02 4.291319E+02 5.945007E+02 7.313068E+02 2.260050E+01 2.190858E+02 2.748483E+02 4.904691E+02 6.069666E+02 2.280063E+01 0.000000E+00 0.000000E+00 3.511805E+02 4.352937E+02 2.300058E+01 0.000000E+00 0.000000E+00 1.792344E+02 2.229219E+02 2.320074E+01 0.000000E+00 0.000000E+00 0.000000E+00 0.000000E+00 Notes: 1. Mass and energy exiting from the reactor-vessel side of the break. 2. Mass and energy exiting from the steam-generator side of the break.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 6.2D-4 DEHL BREAK NO SAFETY INJECTION MASS BALANCE Time(s) 0.00 23.20 23.20 Mass (thousand lbm)
Initial In RCS and Accumulator 746.11 746.11 746.11 Added Mass Pumped Injection Total Added 0.0 0.0 0.0 0.0 0.0 0.0 ***Total Available***
746.11 746.11 746.11 Distribution Reactor Coolant Accumulator Total Contents 527.79 218.32 746.11 81.96 158.86 240.82 92.03 148.79 240.82 Effluent Break Flow ECCS Spill Total Effluent 0.00 0.00 0.00 505.27 0.00 505.27 505.27 0.00 505.27 ***Total Accountable***
746.11 746.09 746.09 *See Section 6.2D.3.1.4 for a description of the mass and energy balance tables. Some round-off and truncation exists.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 6.2D-5 DEHL BREAK NO SAFETY INJECTION ENERGY BALANCE Time(s) .00 23.20 23.20 Energy (Million Btu)
Initial Energy In RCS, Accumulator, SG 909.96 909.96 909.96 Added Energy Pumped Injection Decay Heat Heat from Secondary Total Added 0.00 0.00 0.00 0.00 0.00 8.22 -5.00 3.22 0.00 8.22 -5.00 3.22 *** Total Available ***
909.96 913.18 913.18 Distribution Reactor Coolant Accumulator Core Stored Primary Metal Secondary Metal SG Total Contents 307.96 19.59 21.81 155.40 111.31 293.89 909.96 21.10 14.26 8.19 146.48 109.70 288.66 588.38 22.00 13.35 8.19 146.48 109.70 288.66 588.38 Effluent Break Flow ECCS Spill Total Effluent 0.00 0.00 0.00 324.20 0.00 324.20 324.20 0.00 324.20 *** Total Accountable ***
909.96 912.58 912.58 *See Section 6.2D.3.1.4 for a description of the mass and energy balance tables. Some round-off and truncation exists.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2D-6 DEPS BREAK MASS AND ENERGY RELEASES DURING BLOWDOWN Sheet 1 of 5 Revision 23 December 2016 Time(s) Break Path No. 1 Flow(1) Break Path No. 2 Flow(2) (lbm/s) Thousand (Btu/s) (lbm/s) Thousand (Btu/s) 0.000000E+00 0.000000E+00 0.000000E+00 0.000000E+00 0.000000E+00 1.012155E-03 9.252190E+04 5.030486E+04 3.880431E+04 2.103613E+04 2.079687E-03 4.081721E+04 2.212773E+04 4.048589E+04 2.194649E+04 1.010149E-01 4.039251E+04 2.195494E+04 2.110978E+04 1.143701E+04 2.012392E-01 4.096699E+04 2.239887E+04 2.305680E+04 1.249944E+04 3.016804E-01 4.168919E+04 2.297532E+04 2.322227E+04 1.259865E+04 4.012768E-01 4.245918E+04 2.362371E+04 2.279445E+04 1.237826E+04 5.014424E-01 4.319495E+04 2.429389E+04 2.194319E+04 1.192387E+04 6.013018E-01 4.359878E+04 2.478882E+04 2.112506E+04 1.148468E+04 7.022743E-01 4.349955E+04 2.498848E+04 2.047199E+04 1.113286E+04 8.018839E-01 4.271585E+04 2.476116E+04 1.991531E+04 1.083234E+04 9.014520E-01 4.162552E+04 2.433228E+04 1.947506E+04 1.059479E+04 1.001292E+00 4.055578E+04 2.390234E+04 1.917617E+04 1.043403E+04 1.101543E+00 3.951127E+04 2.348347E+04 1.901385E+04 1.034785E+04 1.201758E+00 3.841998E+04 2.303971E+04 1.894217E+04 1.031017E+04 1.301335E+00 3.728278E+04 2.256854E+04 1.892713E+04 1.030310E+04 1.401155E+00 3.614663E+04 2.208747E+04 1.894157E+04 1.031162E+04 1.501798E+00 3.509478E+04 2.163480E+04 1.897107E+04 1.032800E+04 1.601713E+00 3.421789E+04 2.126281E+04 1.899927E+04 1.034346E+04 1.701023E+00 3.346560E+04 2.095155E+04 1.903190E+04 1.036135E+04 1.801546E+00 3.276563E+04 2.066905E+04 1.906613E+04 1.038022E+04 1.901489E+00 3.207894E+04 2.039198E+04 1.909244E+04 1.039482E+04 2.001166E+00 3.137079E+04 2.009829E+04 1.908695E+04 1.039204E+04 2.101973E+00 3.056791E+04 1.974044E+04 1.904905E+04 1.037167E+04 2.201918E+00 2.983342E+04 1.941747E+04 1.898670E+04 1.033809E+04 2.301623E+00 2.906627E+04 1.906466E+04 1.889729E+04 1.028983E+04 2.401150E+00 2.827822E+04 1.868987E+04 1.858292E+04 1.011806E+04 2.501416E+00 2.744303E+04 1.827727E+04 1.835090E+04 9.992681E+03 2.601523E+00 2.637734E+04 1.769804E+04 1.819343E+04 9.908046E+03 2.701054E+00 2.503526E+04 1.691292E+04 1.803376E+04 9.822080E+03 2.801538E+00 2.301249E+04 1.563913E+04 1.784718E+04 9.721403E+03 2.901077E+00 2.071230E+04 1.415471E+04 1.763414E+04 9.606299E+03 3.001147E+00 2.006955E+04 1.381607E+04 1.742081E+04 9.491216E+03 3.101327E+00 2.019607E+04 1.396411E+04 1.721577E+04 9.380798E+03 3.201589E+00 1.937606E+04 1.341843E+04 1.701253E+04 9.271432E+03 3.301334E+00 1.874670E+04 1.301718E+04 1.679817E+04 9.155889E+03 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2D-6 DEPS BREAK MASS AND ENERGY RELEASES DURING BLOWDOWN Sheet 2 of 5 Revision 23 December 2016 Time(s) Break Path No. 1 Flow(1) Break Path No. 2 Flow(2) (lbm/s) Thousand (Btu/s) (lbm/s) Thousand (Btu/s) 3.401203E+00 1.828567E+04 1.272827E+04 1.658303E+04 9.039954E+03 3.501188E+00 1.763436E+04 1.229725E+04 1.638264E+04 8.932213E+03 3.601297E+00 1.699057E+04 1.187174E+04 1.619298E+04 8.830374E+03 3.701527E+00 1.634982E+04 1.144691E+04 1.601211E+04 8.733377E+03 3.801280E+00 1.570858E+04 1.101978E+04 1.584464E+04 8.643747E+03 3.901184E+00 1.510531E+04 1.061852E+04 1.568558E+04 8.558731E+03 4.001285E+00 1.454678E+04 1.024840E+04 1.553119E+04 8.476258E+03 4.200344E+00 1.364992E+04 9.658193E+03 1.526236E+04 8.333218E+03 4.400461E+00 1.298935E+04 9.220547E+03 1.500957E+04 8.198719E+03 4.600081E+00 1.246184E+04 8.865649E+03 1.478485E+04 8.079567E+03 4.800520E+00 1.205610E+04 8.585036E+03 1.457353E+04 7.967578E+03 5.000148E+00 1.172034E+04 8.344687E+03 1.437991E+04 7.865314E+03 5.200246E+00 1.146733E+04 8.152774E+03 1.418011E+04 7.759525E+03 5.400347E+00 1.126195E+04 7.986499E+03 1.398521E+04 7.656539E+03 5.600211E+00 1.110929E+04 7.850742E+03 1.381060E+04 7.564359E+03 5.800925E+00 1.100068E+04 7.739160E+03 1.411973E+04 7.744306E+03 6.000424E+00 1.096464E+04 7.671271E+03 1.530308E+04 8.392624E+03 6.200824E+00 1.099064E+04 7.641513E+03 1.504573E+04 8.253777E+03 6.400746E+00 1.107517E+04 7.647963E+03 1.476541E+04 8.103967E+03 6.600282E+00 1.118138E+04 7.667223E+03 1.466591E+04 8.052643E+03 6.800970E+00 1.133501E+04 7.717190E+03 1.447111E+04 7.949621E+03 7.000510E+00 1.193072E+04 8.058007E+03 1.432975E+04 7.876542E+03 7.200167E+00 1.169015E+04 7.886052E+03 1.434763E+04 7.891389E+03 7.400278E+00 1.077083E+04 7.618760E+03 1.416369E+04 7.792666E+03 7.600314E+00 9.425007E+03 7.126252E+03 1.394847E+04 7.675799E+03 7.800761E+00 8.905678E+03 6.881320E+03 1.375589E+04 7.571474E+03 8.000278E+00 8.884489E+03 6.834976E+03 1.363074E+04 7.505109E+03 8.200306E+00 8.933121E+03 6.811539E+03 1.355253E+04 7.463369E+03 8.400263E+00 8.969356E+03 6.782147E+03 1.341046E+04 7.383984E+03 8.600921E+00 9.026595E+03 6.765638E+03 1.323944E+04 7.288037E+03 8.800923E+00 9.113672E+03 6.758857E+03 1.310265E+04 7.211740E+03 9.000819E+00 9.221286E+03 6.757776E+03 1.296445E+04 7.134939E+03 9.200841E+00 9.351532E+03 6.768901E+03 1.281140E+04 7.049831E+03 9.400370E+00 9.500649E+03 6.789450E+03 1.265876E+04 6.964813E+03 9.600385E+00 9.634203E+03 6.798282E+03 1.250945E+04 6.881762E+03 9.800733E+00 9.739468E+03 6.793960E+03 1.236770E+04 6.802965E+03 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2D-6 DEPS BREAK MASS AND ENERGY RELEASES DURING BLOWDOWN Sheet 3 of 5 Revision 23 December 2016 Time(s) Break Path No. 1 Flow(1) Break Path No. 2 Flow(2) (lbm/s) Thousand (Btu/s) (lbm/s) Thousand (Btu/s) 1.000034E+01 9.783865E+03 6.757637E+03 1.222972E+04 6.726037E+03 1.020022E+01 9.729238E+03 6.667920E+03 1.209631E+04 6.651525E+03 1.040008E+01 9.615334E+03 6.555336E+03 1.197774E+04 6.585208E+03 1.060056E+01 9.482281E+03 6.442565E+03 1.186170E+04 6.520128E+03 1.080006E+01 9.329076E+03 6.326956E+03 1.174497E+04 6.454605E+03 1.100096E+01 9.172094E+03 6.217775E+03 1.163586E+04 6.393530E+03 1.100218E+01 9.171177E+03 6.217152E+03 1.163521E+04 6.393166E+03 1.100341E+01 9.170260E+03 6.216529E+03 1.163456E+04 6.392801E+03 1.100463E+01 9.169345E+03 6.215914E+03 1.163390E+04 6.392436E+03 1.100585E+01 9.168431E+03 6.215294E+03 1.163325E+04 6.392070E+03 1.100707E+01 9.167519E+03 6.214681E+03 1.163259E+04 6.391700E+03 1.120110E+01 9.023630E+03 6.118759E+03 1.152602E+04 6.332097E+03 1.140019E+01 8.865687E+03 6.015470E+03 1.141487E+04 6.269979E+03 1.160050E+01 8.702383E+03 5.911213E+03 1.131075E+04 6.211891E+03 1.180084E+01 8.541833E+03 5.811101E+03 1.120434E+04 6.152488E+03 1.200019E+01 8.375662E+03 5.709694E+03 1.109930E+04 6.093814E+03 1.220043E+01 8.209419E+03 5.611355E+03 1.099869E+04 6.037670E+03 1.240006E+01 8.046150E+03 5.517244E+03 1.089649E+04 5.980579E+03 1.260108E+01 7.881466E+03 5.424035E+03 1.079598E+04 5.924426E+03 1.280039E+01 7.718821E+03 5.333581E+03 1.069874E+04 5.870198E+03 1.300054E+01 7.562658E+03 5.249283E+03 1.060003E+04 5.815084E+03 1.320034E+01 7.416179E+03 5.171624E+03 1.049863E+04 5.758525E+03 1.340101E+01 7.271644E+03 5.093622E+03 1.040210E+04 5.704791E+03 1.360006E+01 7.134836E+03 5.018972E+03 1.030282E+04 5.649594E+03 1.380005E+01 7.002797E+03 4.946589E+03 1.020627E+04 5.595929E+03 1.400101E+01 6.874522E+03 4.876558E+03 1.010893E+04 5.541868E+03 1.420035E+01 6.750774E+03 4.809434E+03 1.001363E+04 5.489012E+03 1.440070E+01 6.629884E+03 4.744007E+03 9.917720E+03 5.435941E+03 1.460076E+01 6.512911E+03 4.680452E+03 9.823523E+03 5.383951E+03 1.480022E+01 6.400260E+03 4.618759E+03 9.729319E+03 5.332120E+03 1.500103E+01 6.291234E+03 4.558519E+03 9.636747E+03 5.281414E+03 1.520031E+01 6.184864E+03 4.498873E+03 9.539283E+03 5.228168E+03 1.540081E+01 6.077813E+03 4.438157E+03 9.432446E+03 5.170221E+03 1.560049E+01 5.960578E+03 4.368594E+03 9.311637E+03 5.104554E+03 1.580117E+01 5.836925E+03 4.293759E+03 9.218488E+03 5.044265E+03 1.600086E+01 5.709824E+03 4.214848E+03 9.122408E+03 4.965471E+03 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2D-6 DEPS BREAK MASS AND ENERGY RELEASES DURING BLOWDOWN Sheet 4 of 5 Revision 23 December 2016 Time(s) Break Path No. 1 Flow(1) Break Path No. 2 Flow(2) (lbm/s) Thousand (Btu/s) (lbm/s) Thousand (Btu/s) 1.620058E+01 5.587617E+03 4.135735E+03 9.106010E+03 4.913600E+03 1.640125E+01 5.477640E+03 4.059340E+03 9.079253E+03 4.843633E+03 1.660063E+01 5.381762E+03 3.989510E+03 9.116459E+03 4.803105E+03 1.680111E+01 5.293340E+03 3.924600E+03 9.141485E+03 4.758903E+03 1.700146E+01 5.207144E+03 3.863129E+03 9.158340E+03 4.718763E+03 1.720114E+01 5.123663E+03 3.806274E+03 9.161252E+03 4.682141E+03 1.740038E+01 5.039230E+03 3.752181E+03 9.055504E+03 4.598112E+03 1.760033E+01 4.953065E+03 3.700900E+03 8.982615E+03 4.533026E+03 1.780120E+01 4.872939E+03 3.657670E+03 8.943335E+03 4.490926E+03 1.800104E+01 4.797113E+03 3.620798E+03 8.652300E+03 4.325661E+03 1.820150E+01 4.744428E+03 3.599257E+03 8.610273E+03 4.285440E+03 1.840098E+01 4.713919E+03 3.606751E+03 8.374818E+03 4.156671E+03 1.860029E+01 4.666316E+03 3.645949E+03 8.229559E+03 4.070613E+03 1.880018E+01 4.548645E+03 3.688666E+03 8.093555E+03 3.991217E+03 1.900086E+01 4.366806E+03 3.713797E+03 8.179343E+03 4.024514E+03 1.920041E+01 4.157564E+03 3.733778E+03 7.416306E+03 3.620298E+03 1.940109E+01 3.853495E+03 3.678385E+03 8.063020E+03 3.883554E+03 1.960014E+01 3.533753E+03 3.593126E+03 6.570183E+03 3.134617E+03 1.980101E+01 3.228186E+03 3.477395E+03 7.339075E+03 3.403570E+03 2.000052E+01 2.887077E+03 3.288436E+03 1.209313E+04 5.647652E+03 2.020091E+01 2.609423E+03 3.121453E+03 6.727976E+03 3.180710E+03 2.040094E+01 2.439007E+03 2.977250E+03 3.790347E+03 1.784955E+03 2.060005E+01 2.221885E+03 2.733147E+03 8.895489E+03 3.725262E+03 2.080075E+01 1.978307E+03 2.447160E+03 7.998499E+03 3.369649E+03 2.100027E+01 1.818310E+03 2.257756E+03 5.286446E+03 2.246938E+03 2.120034E+01 1.709551E+03 2.127575E+03 3.776545E+03 1.613664E+03 2.140036E+01 1.586928E+03 1.978608E+03 2.943544E+03 1.207452E+03 2.160000E+01 1.456885E+03 1.819756E+03 5.023900E+03 1.858291E+03 2.180062E+01 1.330542E+03 1.665168E+03 6.286374E+03 2.276864E+03 2.200064E+01 1.219142E+03 1.528474E+03 5.131978E+03 1.845515E+03 2.220020E+01 1.115549E+03 1.400790E+03 4.406142E+03 1.571743E+03 2.240054E+01 1.028179E+03 1.292863E+03 4.212132E+03 1.480872E+03 2.260055E+01 9.252404E+02 1.165147E+03 4.033567E+03 1.385548E+03 2.280061E+01 8.358047E+02 1.054029E+03 3.696047E+03 1.237628E+03 2.300001E+01 7.627867E+02 9.629636E+02 3.266675E+03 1.063116E+03 2.320024E+01 7.007720E+02 8.852832E+02 2.819002E+03 8.892898E+02 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2D-6 DEPS BREAK MASS AND ENERGY RELEASES DURING BLOWDOWN Sheet 5 of 5 Revision 23 December 2016 Time(s) Break Path No. 1 Flow(1) Break Path No. 2 Flow(2) (lbm/s) Thousand (Btu/s) (lbm/s) Thousand (Btu/s) 2.340068E+01 6.692970E+02 8.463311E+02 2.448506E+03 7.485734E+02 2.360008E+01 6.403119E+02 8.101765E+02 2.067099E+03 6.134657E+02 2.380041E+01 5.949544E+02 7.531772E+02 1.637313E+03 4.733364E+02 2.400079E+01 5.463386E+02 6.920027E+02 1.149306E+03 3.254831E+02 2.420033E+01 4.963731E+02 6.290783E+02 6.261196E+02 1.749727E+02 2.440036E+01 4.442113E+02 5.632914E+02 1.863815E+02 5.183787E+01 2.460038E+01 3.890625E+02 4.936335E+02 0.000000E+00 0.000000E+00 2.480069E+01 3.305772E+02 4.196723E+02 0.000000E+00 0.000000E+00 2.500051E+01 2.760499E+02 3.506891E+02 1.366954E+02 3.836163E+01 2.520014E+01 2.116667E+02 2.691038E+02 1.698429E+02 4.740880E+01 2.540063E+01 1.775000E+02 2.258605E+02 1.075425E+02 2.978558E+01 2.560069E+01 1.241001E+02 1.580873E+02 4.893733E+01 1.346882E+01 2.580059E+01 9.152902E+01 1.167691E+02 2.915437E+01 7.982361E+00 2.600015E+01 4.796144E+01 6.133642E+01 0.000000E+00 0.000000E+00 2.620025E+01 0.000000E+00 0.000000E+00 0.000000E+00 0.000000E+00 Notes: 1. Mass and energy exiting from the steam-generator side of the break (path 1). 2. Mass and energy exiting from the pump-side of the break (path 2).
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2D-7 Sheet 1 of 5 Revision 23 December 2016 DEPS BREAK MASS AND ENERGY RELEASES DURING REFLOOD MINIMUM SAFEGUARDS (SSPS FAILURE)
Time(s) Break Path No. 1 Flow(1) Break Path No. 2 Flow(2) (lbm/s) Thousand (Btu/s) (lbm/s) Thousand (Btu/s) 2.620025E+01 0.000000E+00 0.000000E+00 0.000000E+00 0.000000E+00 2.670025E+01 0.000000E+00 0.000000E+00 0.000000E+00 0.000000E+00 2.690025E+01 0.000000E+00 0.000000E+00 0.000000E+00 0.000000E+00 2.700025E+01 0.000000E+00 0.000000E+00 0.000000E+00 0.000000E+00 2.710025E+01 0.000000E+00 0.000000E+00 0.000000E+00 0.000000E+00 2.715025E+01 0.000000E+00 0.000000E+00 0.000000E+00 0.000000E+00 2.722525E+01 2.165524E+01 2.550970E+01 0.000000E+00 0.000000E+00 2.733775E+01 5.814556E+01 6.851338E+01 0.000000E+00 0.000000E+00 2.743775E+01 3.050782E+01 3.594081E+01 0.000000E+00 0.000000E+00 2.756275E+01 3.119871E+01 3.675424E+01 0.000000E+00 0.000000E+00 2.766275E+01 3.767301E+01 4.438228E+01 0.000000E+00 0.000000E+00 2.776275E+01 4.459477E+01 5.253904E+01 0.000000E+00 0.000000E+00 2.786275E+01 4.965626E+01 5.850383E+01 0.000000E+00 0.000000E+00 2.796275E+01 5.440227E+01 6.409731E+01 0.000000E+00 0.000000E+00 2.806275E+01 5.886939E+01 6.936254E+01 0.000000E+00 0.000000E+00 2.816275E+01 6.310501E+01 7.435530E+01 0.000000E+00 0.000000E+00 2.826275E+01 6.714333E+01 7.911587E+01 0.000000E+00 0.000000E+00 2.836275E+01 7.101016E+01 8.367465E+01 0.000000E+00 0.000000E+00 2.838775E+01 7.195254E+01 8.478573E+01 0.000000E+00 0.000000E+00 2.846275E+01 7.472578E+01 8.805553E+01 0.000000E+00 0.000000E+00 2.856275E+01 7.830655E+01 9.227776E+01 0.000000E+00 0.000000E+00 2.866275E+01 8.176593E+01 9.635718E+01 0.000000E+00 0.000000E+00 2.876275E+01 8.511521E+01 1.003071E+02 0.000000E+00 0.000000E+00 2.886275E+01 8.836394E+01 1.041387E+02 0.000000E+00 0.000000E+00 2.896275E+01 9.152037E+01 1.078618E+02 0.000000E+00 0.000000E+00 2.906275E+01 9.459160E+01 1.114846E+02 0.000000E+00 0.000000E+00 2.916275E+01 9.758387E+01 1.150146E+02 0.000000E+00 0.000000E+00 2.926275E+01 1.005027E+02 1.184582E+02 0.000000E+00 0.000000E+00 3.026275E+01 1.265132E+02 1.491583E+02 0.000000E+00 0.000000E+00 3.126275E+01 1.484303E+02 1.750465E+02 0.000000E+00 0.000000E+00 3.226275E+01 4.795427E+02 5.687037E+02 4.801765E+03 6.722576E+02 3.256275E+01 5.011402E+02 5.947591E+02 4.970418E+03 7.211898E+02 3.326275E+01 5.041333E+02 5.983989E+02 4.989726E+03 7.327193E+02 3.426275E+01 4.966638E+02 5.894510E+02 4.923798E+03 7.262894E+02 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2D-7 Sheet 2 of 5 Revision 23 December 2016 Time(s) Break Path No. 1 Flow(1) Break Path No. 2 Flow(2) (lbm/s) Thousand (Btu/s) (lbm/s) Thousand (Btu/s) 3.526275E+01 4.880143E+02 5.790889E+02 4.845609E+03 7.175836E+02 3.626275E+01 4.791370E+02 5.684565E+02 4.764119E+03 7.081898E+02 3.666275E+01 4.755969E+02 5.642175E+02 4.731347E+03 7.043605E+02 3.726275E+01 4.703337E+02 5.579160E+02 4.682363E+03 6.985993E+02 3.826275E+01 4.617388E+02 5.476282E+02 4.601740E+03 6.890457E+02 3.926275E+01 4.534123E+02 5.376647E+02 4.522925E+03 6.796472E+02 4.026275E+01 4.453781E+02 5.280540E+02 4.446239E+03 6.704639E+02 4.126275E+01 4.376392E+02 5.187993E+02 4.371815E+03 6.615220E+02 4.206275E+01 4.316596E+02 5.116503E+02 4.313924E+03 6.545524E+02 4.226275E+01 4.301935E+02 5.098977E+02 4.299680E+03 6.528358E+02 4.326275E+01 4.230315E+02 5.013378E+02 4.229806E+03 6.444074E+02 4.426275E+01 4.161413E+02 4.931050E+02 4.162131E+03 6.362330E+02 4.526275E+01 4.095097E+02 4.851833E+02 4.096579E+03 6.283057E+02 4.626275E+01 4.031233E+02 4.775566E+02 4.033063E+03 6.206165E+02 4.726275E+01 4.308891E+02 5.107277E+02 4.331879E+03 6.374699E+02 4.821275E+01 3.503389E+02 4.146454E+02 3.458540E+03 5.452824E+02 4.831275E+01 3.482873E+02 4.121656E+02 3.466671E+03 5.396508E+02 4.931275E+01 4.173104E+02 4.943417E+02 3.175222E+02 2.256082E+02 5.031275E+01 4.451279E+02 5.277542E+02 3.295084E+02 2.428155E+02 5.131275E+01 4.331521E+02 5.134402E+02 3.241187E+02 2.356100E+02 5.231275E+01 4.202809E+02 4.980568E+02 3.183594E+02 2.278838E+02 5.331275E+01 4.076300E+02 4.829435E+02 3.127126E+02 2.203218E+02 5.431275E+01 3.964120E+02 4.695485E+02 3.077136E+02 2.136435E+02 5.481275E+01 3.910326E+02 4.631274E+02 3.053198E+02 2.104518E+02 5.531275E+01 3.857790E+02 4.568579E+02 3.029845E+02 2.073416E+02 5.631275E+01 3.756297E+02 4.447497E+02 2.984807E+02 2.013526E+02 5.731275E+01 3.659328E+02 4.331859E+02 2.941875E+02 1.956553E+02 5.831275E+01 3.566617E+02 4.221344E+02 2.900927E+02 1.902321E+02 5.931275E+01 3.477929E+02 4.115664E+02 2.861852E+02 1.850667E+02 6.031275E+01 3.393043E+02 4.014551E+02 2.824547E+02 1.801444E+02 6.131275E+01 3.311674E+02 3.917663E+02 2.788886E+02 1.754475E+02 6.231275E+01 3.233722E+02 3.824873E+02 2.754814E+02 1.709675E+02 6.331275E+01 3.159035E+02 3.735999E+02 2.722258E+02 1.666939E+02 6.431275E+01 3.087464E+02 3.650860E+02 2.691146E+02 1.626164E+02 6.531275E+01 3.018869E+02 3.569286E+02 2.661411E+02 1.587256E+02 6.631275E+01 2.953122E+02 3.491120E+02 2.632979E+02 1.550124E+02 6.731275E+01 2.890055E+02 3.416163E+02 2.605531E+02 1.514666E+02 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2D-7 Sheet 3 of 5 Revision 23 December 2016 Time(s) Break Path No. 1 Flow(1) Break Path No. 2 Flow(2) (lbm/s) Thousand (Btu/s) (lbm/s) Thousand (Btu/s) 6.831275E+01 2.829973E+02 3.344771E+02 2.579486E+02 1.481060E+02 6.931275E+01 2.772468E+02 3.276458E+02 2.554630E+02 1.449030E+02 7.031275E+01 2.717317E+02 3.210958E+02 2.530862E+02 1.418439E+02 7.111275E+01 2.674830E+02 3.160510E+02 2.512601E+02 1.394959E+02 7.131275E+01 2.664430E+02 3.148161E+02 2.508138E+02 1.389223E+02 7.231275E+01 2.613721E+02 3.087964E+02 2.486414E+02 1.361323E+02 7.331275E+01 2.565109E+02 3.030269E+02 2.465651E+02 1.334685E+02 7.431275E+01 2.518518E+02 2.974984E+02 2.445810E+02 1.309256E+02 7.531275E+01 2.473875E+02 2.922019E+02 2.426855E+02 1.284985E+02 7.631275E+01 2.431109E+02 2.871292E+02 2.408751E+02 1.261825E+02 7.731275E+01 2.390144E+02 2.822709E+02 2.391461E+02 1.239726E+02 7.831275E+01 2.350922E+02 2.776201E+02 2.374954E+02 1.218646E+02 7.931275E+01 2.313383E+02 2.731696E+02 2.359201E+02 1.198544E+02 8.031275E+01 2.277460E+02 2.689114E+02 2.344171E+02 1.179380E+02 8.131275E+01 2.243029E+02 2.648306E+02 2.329809E+02 1.161082E+02 8.231275E+01 2.210094E+02 2.609276E+02 2.316109E+02 1.143641E+02 8.331275E+01 2.178609E+02 2.571970E+02 2.303049E+02 1.127024E+02 8.431275E+01 2.148522E+02 2.536325E+02 2.290602E+02 1.111198E+02 8.531275E+01 2.119783E+02 2.502280E+02 2.278744E+02 1.096130E+02 8.631275E+01 2.092343E+02 2.469779E+02 2.267450E+02 1.081788E+02 8.731275E+01 2.066155E+02 2.438763E+02 2.256699E+02 1.068142E+02 8.931275E+01 2.017352E+02 2.380974E+02 2.236733E+02 1.042822E+02 9.131275E+01 1.973026E+02 2.328495E+02 2.218678E+02 1.019948E+02 9.271275E+01 1.944485E+02 2.294709E+02 2.207093E+02 1.005281E+02 9.331275E+01 1.932848E+02 2.280935E+02 2.202378E+02 9.993153E+01 9.531275E+01 1.896509E+02 2.237926E+02 2.187688E+02 9.807358E+01 9.731275E+01 1.863718E+02 2.199121E+02 2.174474E+02 9.640343E+01 9.931275E+01 1.834203E+02 2.164198E+02 2.162609E+02 9.490492E+01 1.013128E+02 1.807619E+02 2.132745E+02 2.151950E+02 9.355949E+01 1.033128E+02 1.783711E+02 2.104462E+02 2.142383E+02 9.235254E+01 1.053128E+02 1.762355E+02 2.079200E+02 2.133846E+02 9.127593E+01 1.073128E+02 1.743340E+02 2.056709E+02 2.126246E+02 9.031799E+01 1.093128E+02 1.726478E+02 2.036766E+02 2.119502E+02 8.946830E+01 1.113128E+02 1.711585E+02 2.019152E+02 2.113538E+02 8.871700E+01 1.133128E+02 1.698493E+02 2.003670E+02 2.108281E+02 8.805514E+01 1.153128E+02 1.687048E+02 1.990136E+02 2.103669E+02 8.747454E+01 1.173128E+02 1.677105E+02 1.978379E+02 2.099643E+02 8.696770E+01 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2D-7 Sheet 4 of 5 Revision 23 December 2016 Time(s) Break Path No. 1 Flow(1) Break Path No. 2 Flow(2) (lbm/s) Thousand (Btu/s) (lbm/s) Thousand (Btu/s) 1.188128E+02 1.670554E+02 1.970633E+02 2.096974E+02 8.663185E+01 1.193128E+02 1.668533E+02 1.968242E+02 2.096147E+02 8.652780E+01 1.213128E+02 1.661208E+02 1.959581E+02 2.093133E+02 8.614860E+01 1.233128E+02 1.655019E+02 1.952263E+02 2.090556E+02 8.582439E+01 1.253128E+02 1.649861E+02 1.946164E+02 2.088375E+02 8.555000E+01 1.273128E+02 1.645632E+02 1.941164E+02 2.086546E+02 8.532002E+01 1.293128E+02 1.641954E+02 1.936815E+02 2.084907E+02 8.511387E+01 1.313128E+02 1.639049E+02 1.933381E+02 2.083562E+02 8.494477E+01 1.333128E+02 1.636866E+02 1.930800E+02 2.082492E+02 8.481017E+01 1.353128E+02 1.635336E+02 1.928991E+02 2.081670E+02 8.470675E+01 1.373128E+02 1.634397E+02 1.927880E+02 2.081072E+02 8.463147E+01 1.393128E+02 1.633990E+02 1.927400E+02 2.080675E+02 8.458158E+01 1.413128E+02 1.634064E+02 1.927487E+02 2.080461E+02 8.455458E+01 1.433128E+02 1.634571E+02 1.928086E+02 2.080410E+02 8.454818E+01 1.453128E+02 1.635466E+02 1.929144E+02 2.080507E+02 8.456030E+01 1.473128E+02 1.636711E+02 1.930616E+02 2.080737E+02 8.458908E+01 1.474128E+02 1.636782E+02 1.930700E+02 2.080751E+02 8.459092E+01 1.493128E+02 1.638268E+02 1.932457E+02 2.081085E+02 8.463279E+01 1.513128E+02 1.640012E+02 1.934519E+02 2.081509E+02 8.468611E+01 1.533128E+02 1.641896E+02 1.936746E+02 2.081992E+02 8.474668E+01 1.553128E+02 1.644007E+02 1.939242E+02 2.082561E+02 8.481826E+01 1.573128E+02 1.646323E+02 1.941980E+02 2.083210E+02 8.489975E+01 1.593128E+02 1.648821E+02 1.944933E+02 2.083929E+02 8.499012E+01 1.613128E+02 1.651481E+02 1.948079E+02 2.084712E+02 8.508846E+01 1.633128E+02 1.654286E+02 1.951395E+02 2.085551E+02 8.519393E+01 1.653128E+02 1.657219E+02 1.954863E+02 2.086441E+02 8.530578E+01 1.673128E+02 1.660266E+02 1.958466E+02 2.087376E+02 8.542333E+01 1.693128E+02 1.663414E+02 1.962188E+02 2.088352E+02 8.554599E+01 1.713128E+02 1.666651E+02 1.966016E+02 2.089364E+02 8.567321E+01 1.733128E+02 1.669967E+02 1.969937E+02 2.090408E+02 8.580449E+01 1.753128E+02 1.673353E+02 1.973939E+02 2.091482E+02 8.593943E+01 1.773128E+02 1.676799E+02 1.978015E+02 2.092581E+02 8.607762E+01 1.793128E+02 1.680299E+02 1.982154E+02 2.093703E+02 8.621875E+01 1.813128E+02 1.683847E+02 1.986349E+02 2.094846E+02 8.636252E+01 1.833128E+02 1.687437E+02 1.990594E+02 2.096009E+02 8.650867E+01 1.853128E+02 1.691065E+02 1.994884E+02 2.097188E+02 8.665699E+01 1.873128E+02 1.694727E+02 1.999214E+02 2.098383E+02 8.680729E+01 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2D-7 Sheet 5 of 5 Revision 23 December 2016 Time(s) Break Path No. 1 Flow(1) Break Path No. 2 Flow(2) (lbm/s) Thousand (Btu/s) (lbm/s) Thousand (Btu/s) 1.893128E+02 1.698418E+02 2.003579E+02 2.099593E+02 8.695944E+01 1.913128E+02 1.702138E+02 2.007978E+02 2.100816E+02 8.711329E+01 1.933128E+02 1.705884E+02 2.012408E+02 2.102051E+02 8.726877E+01 1.953128E+02 1.710176E+02 2.017482E+02 2.103509E+02 8.745078E+01 1.973128E+02 1.722048E+02 2.031523E+02 2.111980E+02 8.805146E+01 1.993128E+02 1.732382E+02 2.043744E+02 2.126468E+02 8.865913E+01 2.013128E+02 1.743215E+02 2.056557E+02 2.146654E+02 8.935882E+01 2.033128E+02 1.754164E+02 2.069508E+02 2.171236E+02 9.011824E+01 2.053128E+02 1.764819E+02 2.082110E+02 2.198982E+02 9.090494E+01 2.073128E+02 1.774732E+02 2.093836E+02 2.228862E+02 9.168800E+01 2.085128E+02 1.780188E+02 2.100290E+02 2.247492E+02 9.214683E+01 Notes: 1. Mass and energy exiting from the steam-generator side of the break (path 1). 2. Mass and energy exiting from the pump-side of the break (path 2).
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2D-8 Sheet 1 of 2 Revision 23 December 2016 DEPS SSPS CASE PRINCIPAL PARAMETERS DURING REFLOOD Time(s) Flooding Carryover Fraction Core Height (ft) Downcomer Height (ft)
Flow Fraction Total Injection Accumulator Spill Enthalpy (Btu/lbm) Temp (°F) Rate (in/s) (lbm/s) 26.2 184.8 0 0 0 0 0.25 0 0 0 0 26.9 182.6 21.428 0 0.54 1.64 0 8131.3 8131.3 0 89.74 27.2 180.8 25.482 0 1.03 1.58 0 8052.1 8052.1 0 89.74 28.4 180 2.629 0.302 1.5 5.16 0.334 7723.7 7723.7 0 89.74 29.3 180.1 2.543 0.414 1.62 8.05 0.352 7527.7 7527.7 0 89.74 32.6 180.5 4.899 0.627 2.01 16.12 0.602 6240.1 6240.1 0 89.74 34.3 180.8 4.513 0.675 2.24 16.12 0.6 5961.3 5961.3 0 89.74 36.7 181.4 4.18 0.704 2.5 16.12 0.596 5663 5663 0 89.74 42.1 183.7 3.75 0.727 3 16.12 0.584 5122.1 5122.1 0 89.74 46.3 185.9 3.525 0.733 3.35 16.12 0.575 4780.8 4780.8 0 89.74 47.3 186.5 3.681 0.735 3.43 16.12 0.588 5121.9 4595.2 0 86.48 48.2 187 3.188 0.734 3.5 16.12 0.544 4122.2 3575.3 0 85.54 48.3 187.1 3.183 0.734 3.51 16.12 0.55 4124.3 3576.9 0 85.53 49.3 187.7 3.606 0.737 3.58 16.02 0.593 532.7 0 0 58.05 50.3 188.3 3.725 0.738 3.66 15.74 0.597 522.1 0 0 58.05 54.8 191.8 3.313 0.738 4.01 14.63 0.589 534.9 0 0 58.05 62.3 198.8 2.81 0.738 4.5 13.32 0.575 549.4 0 0 58.05 71.1 207.9 2.401 0.736 5 12.39 0.558 559.5 0 0 58.05 81.3 218.5 2.086 0.735 5.51 11.86 0.54 565.8 0 0 58.05 92.7 228.3 1.87 0.735 6 11.7 0.523 569.7 0 0 58.05 105.3 237.1 1.735 0.736 6.5 11.85 0.51 571.9 0 0 58.05 118.8 245.1 1.663 0.739 7 12.22 0.502 572.9 0 0 58.05 133.3 252.4 1.629 0.743 7.52 12.72 0.5 573.3 0 0 58.05 147.4 258.5 1.618 0.747 8 13.26 0.5 573.4 0 0 58.05 157.3 262.4 1.617 0.751 8.34 13.65 0.501 573.3 0 0 58.05 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2D-8 Sheet 2 of 2 Revision 23 December 2016 Time(s) Flooding Carryover Fraction Core Height (ft) Downcomer Height (ft)
Flow Fraction Total Injection Accumulator Spill Enthalpy (Btu/lbm) Temp (°F) Rate (in/s) (lbm/s) 163.3 264.6 1.618 0.753 8.54 13.88 0.502 573.2 0 0 58.05 177.3 269.2 1.621 0.758 9 14.43 0.504 573 0 0 58.05 193.3 273.9 1.628 0.764 9.52 15.05 0.507 572.8 0 0 58.05 208.5 277.7 1.656 0.77 10 15.57 0.516 571.9 0 0 58.05 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2D-9 Sheet 1 of 3 Revision 23 December 2016 DEPS BREAK MASS AND ENERGY RELE ASES DURING POST-REFLOOD SSPS FAILURE Time(s) Break Path No. 1 Flow(1) Break Path No. 2 Flow(2) (lbm/s) Thousand (Btu/s) (lbm/s) Thousand (Btu/s) 2.086000E+02 2.099452E+02 2.622469E+02 3.708333E+02 1.285948E+02 2.136000E+02 2.091126E+02 2.612068E+02 3.716659E+02 1.286068E+02 2.186000E+02 2.091996E+02 2.613155E+02 3.715789E+02 1.283755E+02 2.236000E+02 2.083520E+02 2.602567E+02 3.724265E+02 1.283907E+02 2.286000E+02 2.084145E+02 2.603348E+02 3.723640E+02 1.281651E+02 2.336000E+02 2.075510E+02 2.592562E+02 3.732275E+02 1.281837E+02 2.386000E+02 2.075883E+02 2.593028E+02 3.731902E+02 1.279640E+02 2.436000E+02 2.067086E+02 2.582039E+02 3.740699E+02 1.279861E+02 2.486000E+02 2.067197E+02 2.582178E+02 3.740588E+02 1.277725E+02 2.536000E+02 2.058232E+02 2.570980E+02 3.749553E+02 1.277982E+02 2.586000E+02 2.058074E+02 2.570782E+02 3.749711E+02 1.275909E+02 2.636000E+02 2.048932E+02 2.559363E+02 3.758853E+02 1.276204E+02 2.686000E+02 2.048499E+02 2.558822E+02 3.759286E+02 1.274196E+02 2.736000E+02 2.039173E+02 2.547173E+02 3.768611E+02 1.274532E+02 2.786000E+02 2.038453E+02 2.546274E+02 3.769331E+02 1.272592E+02 2.836000E+02 2.028939E+02 2.534389E+02 3.778846E+02 1.272969E+02 2.886000E+02 2.027921E+02 2.533118E+02 3.779864E+02 1.271099E+02 2.936000E+02 2.026699E+02 2.531591E+02 3.781086E+02 1.269278E+02 2.986000E+02 2.016834E+02 2.519269E+02 3.790951E+02 1.269737E+02 3.036000E+02 2.015299E+02 2.517351E+02 3.792486E+02 1.267991E+02 3.086000E+02 2.013545E+02 2.515160E+02 3.794240E+02 1.266298E+02 3.136000E+02 2.003307E+02 2.502372E+02 3.804477E+02 1.266842E+02 3.186000E+02 2.001219E+02 2.499763E+02 3.806566E+02 1.265230E+02 3.236000E+02 1.998907E+02 2.496875E+02 3.808878E+02 1.263673E+02 3.286000E+02 1.988290E+02 2.483614E+02 3.819494E+02 1.264304E+02 3.336000E+02 1.985659E+02 2.480328E+02 3.822125E+02 1.262822E+02 3.386000E+02 1.982758E+02 2.476703E+02 3.825027E+02 1.261408E+02 3.436000E+02 1.971681E+02 2.462867E+02 3.836104E+02 1.262148E+02 3.486000E+02 1.968385E+02 2.458751E+02 3.839399E+02 1.260829E+02 3.536000E+02 1.964831E+02 2.454311E+02 3.842954E+02 1.259575E+02 3.586000E+02 1.961017E+02 2.449547E+02 3.846768E+02 1.258384E+02 3.636000E+02 1.956925E+02 2.444435E+02 3.850860E+02 1.257263E+02 3.686000E+02 1.952563E+02 2.438986E+02 3.855222E+02 1.256208E+02 3.736000E+02 1.947907E+02 2.433170E+02 3.859878E+02 1.255227E+02 3.786000E+02 1.942961E+02 2.426993E+02 3.864823E+02 1.254318E+02 3.836000E+02 1.937716E+02 2.420441E+02 3.870069E+02 1.253483E+02 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2D-9 Sheet 2 of 3 Revision 23 December 2016 Time(s) Break Path No. 1 Flow(1) Break Path No. 2 Flow(2) (lbm/s) Thousand (Btu/s) (lbm/s) Thousand (Btu/s) 3.886000E+02 1.932153E+02 2.413492E+02 3.875632E+02 1.252729E+02 3.936000E+02 1.926279E+02 2.406155E+02 3.881506E+02 1.252052E+02 3.986000E+02 1.920068E+02 2.398396E+02 3.887717E+02 1.251460E+02 4.036000E+02 1.914574E+02 2.391534E+02 3.893210E+02 1.250730E+02 4.086000E+02 1.909151E+02 2.384760E+02 3.898634E+02 1.250000E+02 4.136000E+02 1.903349E+02 2.377512E+02 3.904436E+02 1.249365E+02 4.186000E+02 1.897178E+02 2.369804E+02 3.910607E+02 1.248823E+02 4.236000E+02 1.890630E+02 2.361625E+02 3.917155E+02 1.248377E+02 4.286000E+02 1.890361E+02 2.361289E+02 3.917423E+02 1.246267E+02 4.336000E+02 1.882818E+02 2.351866E+02 3.924967E+02 1.246075E+02 4.386000E+02 1.874848E+02 2.341911E+02 3.932937E+02 1.245989E+02 4.436000E+02 1.872848E+02 2.339412E+02 3.934937E+02 1.244324E+02 4.486000E+02 1.870107E+02 2.335989E+02 3.937678E+02 1.242849E+02 4.536000E+02 1.860347E+02 2.323797E+02 3.947438E+02 1.243223E+02 4.586000E+02 1.856217E+02 2.318639E+02 3.951568E+02 1.242106E+02 4.636000E+02 1.851226E+02 2.312405E+02 3.956559E+02 1.241211E+02 4.686000E+02 1.845362E+02 2.305079E+02 3.962423E+02 1.240542E+02 4.736000E+02 1.838579E+02 2.296606E+02 3.969206E+02 1.240112E+02 4.786000E+02 1.836536E+02 2.294055E+02 3.971249E+02 1.238424E+02 4.836000E+02 1.827479E+02 2.282742E+02 3.980306E+02 1.238586E+02 4.886000E+02 1.822836E+02 2.276942E+02 3.984949E+02 1.237575E+02 4.936000E+02 1.816661E+02 2.269229E+02 3.991123E+02 1.236965E+02 4.986000E+02 1.814072E+02 2.265994E+02 3.993713E+02 1.235404E+02 5.036000E+02 1.804155E+02 2.253607E+02 4.003630E+02 1.235772E+02 5.086000E+02 1.802293E+02 2.251281E+02 4.005492E+02 1.234009E+02 5.136000E+02 1.792619E+02 2.239197E+02 4.015166E+02 1.234304E+02 5.186000E+02 1.789515E+02 2.235319E+02 4.018270E+02 1.232858E+02 5.236000E+02 1.782211E+02 2.226197E+02 4.025573E+02 1.232517E+02 5.286000E+02 1.774647E+02 2.216748E+02 4.033138E+02 1.232240E+02 5.336000E+02 1.769559E+02 2.210393E+02 4.038225E+02 1.231304E+02 5.386000E+02 1.764325E+02 2.203855E+02 4.043459E+02 1.230402E+02 5.436000E+02 1.755528E+02 2.192866E+02 4.052257E+02 1.230436E+02 5.486000E+02 1.748402E+02 2.183965E+02 4.059382E+02 1.230023E+02 5.536000E+02 1.744347E+02 2.178900E+02 4.063438E+02 1.228795E+02 5.586000E+02 1.736270E+02 2.168810E+02 4.071515E+02 1.228623E+02 5.636000E+02 1.729183E+02 2.159958E+02 4.078602E+02 1.228185E+02 8.055732E+02 1.729183E+02 2.159958E+02 4.078602E+02 1.228185E+02 8.056732E+02 8.997459E+01 1.114205E+02 4.908039E+02 1.415338E+02 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2D-9 Sheet 3 of 3 Revision 23 December 2016 Time(s) Break Path No. 1 Flow(1) Break Path No. 2 Flow(2) (lbm/s) Thousand (Btu/s) (lbm/s) Thousand (Btu/s) 8.086000E+02 8.991267E+01 1.113432E+02 4.908658E+02 1.413797E+02 1.677900E+03 8.991267E+01 1.113432E+02 4.908658E+02 1.413797E+02 1.678000E+03 7.760926E+01 9.597472E+01 3.388307E+02 1.647982E+02 1.714013E+03 7.760926E+01 9.597472E+01 3.388307E+02 1.647982E+02 1.714113E+03 7.522245E+01 8.655221E+01 3.412175E+02 7.583629E+01 2.000000E+03 7.210053E+01 8.296007E+01 3.443395E+02 7.639954E+01 2.000100E+03 7.209993E+01 8.295938E+01 3.443401E+02 7.598320E+01 2.500000E+03 6.909680E+01 7.950393E+01 3.473432E+02 7.652502E+01 2.500100E+03 6.909620E+01 7.950324E+01 3.473438E+02 7.610868E+01 3.000000E+03 6.609308E+01 7.604779E+01 3.503469E+02 7.665050E+01 3.000100E+03 6.609248E+01 7.604710E+01 3.503475E+02 7.577608E+01 3.500000E+03 6.308935E+01 7.259166E+01 3.533506E+02 7.631789E+01 3.500100E+03 6.308875E+01 7.259097E+01 3.533512E+02 7.523526E+01 3.600000E+03 6.248861E+01 7.190043E+01 3.539514E+02 7.534353E+01 Notes: 1. Mass and energy exiting from the steam-generator side of the break (path 1). 2. Mass and energy exiting from the pump-side of the break (path 2).
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 6.2D-10 LOCA MASS AND ENERGY RELEASE A NALYSIS CORE DECAY HEAT RATE Time (sec) Decay Heat Generation Rate (Btu/Btu) 10 0.053876 15 0.050401 20 0.048018 40 0.042401 60 0.039244 80 0.037065 100 0.035466 150 0.032724 200 0.030936 400 0.027078 600 0.024931 800 0.023389 1000 0.022156 1500 0.019921 2000 0.018315 4000 0.014781 6000 0.013040 8000 0.012000 10000 0.011262 15000 0.010097 20000 0.009350 40000 0.007778 60000 0.006958 80000 0.006424 100000 0.006021 150000 0.005323 200000 0.004847 400000 0.003770 600000 0.003201 800000 0.002834 1000000 0.002580
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 6.2D-11 DEPS SSPS CASE - MASS BALANCE Time(s) 0 26.2 26.2 208.51 805.67 1714.01 3600 Mass (thousand lbm)
Initial In RCS and ACC 746.11 746.11 746.11 746.11 746.11 746.11 746.11 Added Mass Pumped Injection 0 0 0 91.98 438.75 939.84 1745.78 Total Added 0 0 0 91.98 438.75 939.84 1745.78 *** Total Available ***
746.11 746.11 746.11 838.09 1184.86 1685.95 2491.88 Distribution Reactor Coolant 527.79 58.46 81.87 143.19 143.19 143.19 143.19 Accumulator 218.32 154.59 131.19 0 0 0 0 Total Contents 746.11 213.05 213.05 143.19 143.19 143.19 143.19 Effluent Break Flow 0 533.04 533.04 684.35 1031.12 1552.74 2338.14 ECCS Spill 0 0 0 0 0 0 0 Total Effluent 0 533.04 533.04 684.35 1031.12 1552.74 2338.14 *** Total Accountable ***
746.11 746.09 746.09 827.54 1174.31 1695.93 2481.33 *See Section 6.2D.3.1.4 for a description of the mass and energy balance tables. Some round-off and truncation exists.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 6.2D-12 DEPS SSPS FAILURE ENERGY BALANCE Time(s) Energy (million Btu)
Initial Energy In RCS, ACC, SG Added Energy Pumped Injection Decay Heat Heat from Secondary Total Added *** Total Available ***
Distribution Reactor Coolant Accumulator Core Stored Primary Metal Secondary Metal Steam Generator Total Contents Effluent Break Flow ECCS Spill Total Effluent *** Total Accountable ***
- See Section 6.2D.3.1.4 for a description of the mass and energy balance tables. Some round-off and truncation exists.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 6.2D-13 DOUBLE-ENDED HOT-LEG BREAK SEQUENCE OF EVENTS Time(s) Event Description 0.0 Break Occurs and Loss of Offsite Power is Assumed 1.1 Compensated Pressurizer Pressure for Reactor Trip (1859.7 psia) Reached and Turbine Trip Occurs 4.0 Low-Pressurizer Pressure SI Setpoint (1694.7 psia) Reached 12.9 Broken Loop Accumulator Begins Injecting Water 13.1 Intact Loop Accumulator Begins Injecting Water 23.2 End of Blowdown Phase 23.2 Transient Modeling Terminated
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 6.2D-15 DOUBLE-ENDED PUMP SUCTION BREAK SEQUENCE OF EVENTS (MINIMUM SAFEGUARDS)
Time(s) Event Description 0.0 Break Occurs and Loss of Offsite Power is Assumed 0.7 Containment Hi-1 Reached 1.2 Reactor Trip Occurs on Compensated Pressurizer Pressure Setpoint of 1,859.7 psia and SG Throttle Valves Closed 4.3 Low Pressurizer Pressure SI Setpoint = 1,694.7 psia Reached (SI begins after a 27-second delay and feedwater control valve starts to close) 7.9 Containment Hi-2 Reached 13.3 Main Feedwater Control Valve Closed 15.4 Broken-Loop Accumulator Begins Injecting Water 15.6 Intact-Loop Accumulator Begins Injecting Water 26.2 End of Blowdown Phase 46.3 Pumped Safety Injection Begins 48.0 Broken-Loop Accumulator Water Injection Ends 48.6 Intact-Loop Accumulator Water Injection Ends 71.0 CFCUs On 107.0 Containment Sprays Begin Injecting 208.5 End of Reflood for Minimum Safeguards Case 568.6 Mass and Energy Release Assumption: Broken-Loop SG Equilibration to 61.7 psia 805.7 Mass and Energy Release Assumption: Broken-Loop SG Equilibration to 51.7 psia 1553.1 Mass and Energy Release Assumption: Intact-Loop SG Equilibration to 61.7 psia 1,678.2 Cold Leg Recirculation Begins 1,714.0 Mass and Energy Release Assumption: Intact-Loop SG Equilibration to 41.7 psia 3,600.0 End of Sensible Heat Release from RCS and SGs 3,798.0 Containment Sprays Terminated 25,200.0 Switchover to Hot Leg Recirculation 10,400.000 Transient Terminated DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2D-17 Sheet 1 of 2 Revision 23 December 2016 DIABLO CANYON CONTAINMENT LOCA INTEGRITY ANALYSIS PARAMETERS Parameter Value Auxiliary Saltwater Temperature (°F) 64 RWST Water Temperature (°F) 90 Initial Containment Temperature (°F) 120 Initial Containment Pressure (psia) 16.0 Initial Relative Humidity (%) 18 Net Free Volume (ft
- 3) 2,550,000 Reactor Containment Fan Coolers Total CFCUs 5 Analysis Maximum 3 Analysis Minimum 2 Safety Injection Time (sec) 7.0 Delay Time (sec) Without Offsite Power 64.0 CCW Flow to the CFCUs (gpm)
During Injection During Recirculation 8,000 7,450 Containment Spray Pumps Total CSPs 2 Analysis Maximum 2 Analysis Minimum 1 Flowrate (gpm)
During Injection During Recirculation Table 6.2.D-18 0 Safety Injection Time (sec) 7.0 Spray Delay Time (sec) Without Offsite Power 100.0 Containment Spray Termination Time, (sec)
Minimum Safeguards Maximum Safeguards (1 CSP) Maximum Safeguards (2 CSPs) 3,798 3,018 1,824 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2D-17 Sheet 2 of 2 Revision 23 December 2016 Parameter Value ECCS Recirculation ECCS Cold-Leg Recirculation Switchover, sec Minimum Safeguards Maximum Safeguards (1 CSP) Maximum Safeguards (2 CSPs) 1,678 1,033 829 Containment ECCS Cold-Leg Recirculation Flow, (gpm)
Minimum Safeguards (1 RHR train) Maximum Safeguards (2 RHR trains) 3,252.3 8,082.4 ECCS Hot-Leg Recirculation Switchover, sec 25,200 Containment ECCS Hot-Leg Recirculation Flow, (gpm) Minimum Safeguards (1 RHR train)
Maximum Safeguards (2 RHR trains) 3,071.7 4,576.8 Component Cooling Water System Total CCW Heat Exchangers 2 Analysis Maximum 2 Analysis Minimum 1 CCW Flow Rate to RHR Heat Exchanger (gpm per available HX) 4,800 ASW Flow Rate to CCW Heat Exchanger (gpm per available HX) 10,300 CCW Misc. Heat Loads (MBTU/hr)
During Injection During Recirculation 1.0 2.0 CCW Flow Rate to Misc. Heat Loads (gpm)
During Injection During Recirculation 2,500 500
DCPP UNITS 1 & 2 FSAR UPDATE Revision 18 October 2008 TABLE 6.2D-18 CONTAINMENT SPRAY FLOW RATES AS A FUNCTION OF CONTAINMENT PRESSURE Containment Pressure (psig) 1 CSP Spray Flow Rate (gpm) 2 CSPs Spray Flow Rate (gpm) 0 3036 6142 10 2926 5922 20 2806 5692 30 2686 5442 40 2546 5182 47 2456 4992 DCPP UNITS 1 & 2 FSAR UPDATE Revision 18 October 2008 TABLE 6.2D-19 GOTHIC THERMAL CONDUCTOR MODELING No. Description Materials Surface Area (ft 2) Thickness (in) Initial Temp (°F) 1 Concrete Interior Walls Paint Concrete 79965 0.0075 12 120 2 Concrete Floor Paint Concrete 13012 0.0075 24 120 3 SS Fuel Transfer Tube Stainless Steel 8852 0.144 120 4 SS Structures Stainless Steel 857 0.654 120 5 CS Structures Paint Carbon Steel 48024 0.0075 0.0815 120 6 CS Structures Paint Carbon Steel 60941 0.0075 0.133 120 7 CS Lined Containment Concrete Shell Paint Carbon Steel HGap = 10 Concrete 90560 0.0075 0.375 0.0168 35.6007 120 8 CS Structures Paint Carbon Steel 42517 0.0075 0.567 120 9 CS Structures Paint Carbon Steel 56494 0.0075 0.738 120 10 CS Structures Paint Carbon Steel 31902 0.0075 1.355 120 11 CS SG Snubbers Paint Carbon Steel 522 0.0075 3.0 120 12 CS RCP Motors Paint Carbon Steel 1610 0.0075 6.99 200 DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 6.2D-20 MATERIAL PROPERTIES FROM THE GOTHIC MODEL (REFERENCE 15)
Material Thermal Conductivity (BTU/hr-ft-°F) Vol. Heat Capacity (BTU/ft 3-°F) Paint 0.2083 35.91 Carbon Steel 28 58.8 Air Gap 0.0148 0.018 Concrete 1.04 23.4 Stainless Steel 8.6 58.8
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 6.2D-21
SUMMARY
OF LOCA PEAK CONTAINMENT PRESSURE AND TEMPERATURES Break Location Peak Pressure (psig) Time (sec) Peak Air Temp (°F) Time (sec) Press @ 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (psig) Temp @ 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (°F) DEHL 41.4 22.5 261.8 22.4 - - DEPS min SI 39.8 23.5 259.3 23.5 8.5 165.0 Acceptance Criteria <47 (Containment Design Pressure) - <271 (Containment Design Temperature) - <50% of peak pressure -
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 6.2D-22 INITIAL CONTAINMENT CONDITIONS, FAN COOLER AND CONTAINMENT SPRAY PUMP ASSUMPTIONS Parameter Value Containment net free volume (ft
- 3) 2,550,000 Initial containment temperature (°F) 120.0 Initial containment pressure (psia) 16.0 Initial relative humidity (%)
18 Number of fan coolers
- All - Analysis maximum
- Containment safeguards failure 5 3 2 High containment pressure setpoint (psig) 5.0 Fan Cooler Actuation LOCA - 7.0 sec + 64 sec delay MSLB - 5.0 psig + 44 sec delay Containment fan cooler heat removal (MBTU/hr) vs. Component Cooling Water (CCW) Flow rate (gpm) assuming
T sat = 271°F and TCCW = 125°F (1) Flow Heat Removal 1000 59.12 1500 73.18 2000 82.47 2500 89.03 3000 93.93 3250 95.90 3500 97.73 Containment fan cooler air flowrate (ft 3/min per fan cooler) 34,000 (2) Number of spray pumps
- All
- Containment safeguards failure 2
1 High-high containment pressure setpoint (psig) 24.7 Containment Spray Actuation LOCA - 7.0 sec + 100 sec delay MSLB - 24.7 psig + 76 sec delay Containment spray flowrate (gpm) vs. containment pressure (psig) Press Flow 1 Pump Flow 2 Pumps 0 3036 6142 10 2926 5922 20 2806 5692 30 2686 5442 40 2546 5182 47 2456 4992 RWST/containment spray water temperature (°F) 90.0 Notes: 1. To conservatively model CFCU heat rate at 34,000 cfm, the heat rate data from 47,000 cfm was multiplied by 0.9600. 2. For CFCUs with SRC cooling coils the required flow to provide an equivalent heat rate is 37,000 cfm.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 6.2D-23 PEAK PRESSURES AND TEMPERATURES FOR CONTAINMENT RESPONSE TO MAIN STEAMLINE BREAKS Case Description Peak Pressure (psig @ sec)
Peak Temperature
(°F @ sec)
Break Initial Power Failure 1a 1.4 ft 2 DER 102% containment safeguards 29.55 @ 229.2 282.3 @ 223.5 2a 1.4 ft 2 DER 70% containment safeguards 30.04 @ 613.7 282.0 @ 235.3 3a 1.4 ft 2 DER 30% containment safeguards 32.21 @ 612.6 278.9 @ 248.0 4a 1.4 ft 2 DER 0% containment safeguards 32.60 @ 368.5 278.0 @ 256.7 1b 1.4 ft 2 DER 102% FRV 37.01 @ 464.1 281.3 @ 222.9 2b* 1.4 ft 2 DER 70% FRV 42.41 @ 607.1 281.2 @ 237.1 3b 1.4 ft 2 DER 30% FRV 39.99 @ 692.8 277.4 @ 263 4b 1.4 ft 2 DER 0% FRV 31.63 @ 413.4 276.0 @ 261.9 1c 1.4 ft 2 DER 102% MSIV/CV 31.29 @ 251.5 310.2 @ 26.5 2c 1.4 ft 2 DER 70% MSIV/CV 32.68 @ 294.7 310.7 @ 25.8 3c 1.4 ft 2 DER 30% MSIV/CV 33.17 @ 347.6 311.4 @ 24.9 4c 1.4 ft 2 DER 0% MSIV/CV 34.22 @ 417.3 312.4 @ 24.4 5c 0.73 ft 2 Split Break 102% MSIV/CV 33.52 @ 629.8 296.6 @ 118.7 6c 0.87 ft 2 Split Break 70% MSIV/CV 33.84 @ 665.9 302.4 @ 104.0 7c* 0.94 ft 2 Split Break 30% MSIV/CV 32.88 @ 719.6 303.6 @ 98.1 8c 0.90 ft 2 Split Break 0% MSIV/CV 30.82 @ 712.6 297.9 @ 108.8 *These cases were analyzed with a CFCU actuation delay of 44 seconds and a CS actuation delay of 76 seconds. The remainder of the cases documented in this table assumed a CFCU actuation delay of 38 seconds and a CS actuation delay of 74.5 seconds.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 6.2D-24 SEQUENCE OF EVENTS MAIN STEAMLINE BREAK, MFRV FAILURE, 70% POWER Event Time (sec) SI Low Steamline Pressure Setpoint Reached 0.051 AFW Initiation 0.051 Closure of Steamline CV on Faulted Loop 0.1 Reactor Trip - Start of Rod Motion 2.1 Faulted Loop MFRV Fully Closed Failed Open High Containment Pressure Setpoint Reached 5.7 SI Flow Starts 27.1 Fan Coolers Start 49.7 Faulted Loop Backup MFIV Fully Closed 64.1 SI Boron Reaches Core 148 High-High Containment Pressure Setpoint Reached 161.1 Containment Sprays Start 237.1 Accumulator Injection n/a AFW Re-aligned from Faulted SG 600 Peak Containment Pressure Occurs 607 Mass Release Terminated 630
DCPP UNITS 1 & 2 FSAR UPDATE FIGURE 6.2D-1 STEAMLINE BREAK MASS RELEASE TO CONTAINMENT 1.4 ft 2 DER. 70% POWER, FRV FAILURE UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008 DCPP UNITS 1 & 2 FSAR UPDATE FIGURE 6.2D-2 STEAMLINE BREAK ENTHALPY OF BREAK EFFLUENT 1.4 ft 2 DER. 70% POWER, FRV FAILURE UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008 DCPP UNITS 1 & 2 FSAR UPDATE Diablo Canyon Unit 2 LOCA Containment AnalysisDouble Ended Hot Leg BreaksContainment Pressure 0 5 10 15 20 25 30 35 40 45 1101 0 0Time (seconds)Pressure (psig)
FIGURE 6.2D-3 CONTAINMENT PRESSURE DOUBLE-ENDED HOT LEG BREAK UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008 DCPP UNITS 1 & 2 FSAR UPDATE Diablo Canyon Unit 2 LOCA Containment AnalysisDouble Ended Hot Leg BreakContainment Gas Temperature 120 140 160 180 200 220 240 260 280110100Time (seconds)Temperature (°F)
FIGURE 6.2D-4 CONTAINMENT TEMPERATURE DOUBLE-ENDED HOT LEG BREAK UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008 Revision 22 May 2015 DCPP UNITS 1 & 2 FSAR UPDATE Diablo Canyon Unit 2 LOCA Containment AnalysisDouble Ended Hot Leg BreakContainment Sump Temperature 120 140 160 180 200 220 240 260 280110100Time (seconds)Temperature (°F)
FIGURE 6.2D-5 CONTAINMENT SUMP TEMPERATURE DOUBLE-ENDED HOT LEG BREAK UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008 DCPP UNITS 1 & 2 FSAR UPDATEFSAR UPDATEFIGURE 6.2D-6 LOCA CONTAINMENT INTEGRITY DOUBLE ENDED PUMP SUCTION BREAK WITH EDG/SSPS SINGLE FAILURE 64 SECOND DELAY IN STARTING THE CFCUsUNITS 1 AND 2DIABLO CANYON SITE DCPP UNITS 1 & 2 FSAR UPDATEFSAR UPDATEFIGURE 6.2D-7 LOCA CONTAINMENT INTEGRITY DOUBLE ENDED PUMP SUCTION BREAK WITH EDG/SSPS SINGLE FAILURE 64 SECOND DELAY IN STARTING THE CFCUs UNITS 1 AND 2Diablo Canyon Site DCPP UNITS 1 & 2 FSAR UPDATEFIGURE 6.2D-9 FSAR UPDATE FIGURE 6.2D-9SLB 1.4 FT 2 DER at70% POWER, FRV SINGLEFAILURE CONTAINMENT PRESSUREUNITS 1 and 2Diablo Canyon Site DCPP UNITS 1 & 2 FSAR UPDATEFIGURE 6.2D-10FSAR UPDATEFIGURE6.2D-10 SLB 1.4 FT. 2 DER at 70% POWER, FRV SINGLEFAILURECONTAINMENT VAPORTEMPERATUREUnits 1 and 2Diablo Canyon Site