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{{#Wiki_filter:DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 RADIOACTIVE WASTE MANAGEMENT CONTENTS Section Title Page 11 RADIOACTIVE WASTE MANAGEMENT 11-1 11.1 SOURCE TERMS (Historical) 11.1-1 11.1.1 BASIC PHYSICAL DATA AND CONSTANTS (Historical) 11.1-2 11.1.2 DETERMINATION OF ACTIVITY INVENTORIES IN REACTOR CORE (Historical) 11.1-2 11.1.3 DETERMINATION OF INVENTORIES IN FUEL ROD GAPS (Historical) 11.1-3 11.1.4 DETERMINATION OF PRIMARY COOLANT ACTIVITIES (Historical) 11.1-4 11.1.5 DETERMINATION OF TRITIUM ACTIVITIES IN PRIMARY COOLANT (Historical) 11.1-5 11.1.5.1 Ternary Fissions - Cladding Diffusion (Historical) 11.1-5 11.1.5.2 Tritium Produced from Boron Reactions (Historical) 11.1-5 11.1.5.3 Tritium Produced from Lithium Reactions (Historical) 11.1-5 11.1.5.4 Control Rod Sources (Historical) 11.1-5 11.1.5.5 Tritium Production from Deuterium Reactions (Historical) 11.1-6 11.1.5.6 Total Tritium Sources in Coolant (Historical) 11.1-6 11.1.6 DETERMINATION OF SECONDARY SYSTEM ACTIVITIES (Historical) 11.1-6 11.
1.7 REFERENCES
11.1-6 11.2 LIQUID RADWASTE SYSTEM 11.2-1 11.2.1 DESIGN BASES 11.2-1 11.2.1.1 General Design Criterion 2, 1967 - Performance Standards 11.2-1 11.2.1.2 General Design Criterion 3, 1971 - Fire Protection 11.2-1 11.2.1.3 General Design Criterion 4, 1967 - Sharing of Systems 11.2-2 11.2.1.4 General Design Criterion 11, 1967 - Control Room 11.2-2 11.2.1.5 General Design Criterion 12, 1967 - Instrumentation and Control Systems 11.2-2 i Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.2.1.6 General Design Criterion 17, 1967 - Monitoring Radioactivity Releases 11.2-2 11.2.1.7 General Design Criterion 40, 1967 - Missile Protections 11.2-2 11.2.1.8 General Design Criterion 49, 1967 - Containment Design Basis 11.2-2 11.2.1.9 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment 11.2-2 11.2.1.10 General Design Criterion 56, 1971 - Primary Containment Isolation 11.2-3 11.2.1.11 General Design Criterion 68, 1967 - Fuel and Waste Storage Radiation Shielding 11.2-3 11.2.1.12 General Design Criterion 69, 1967 - Protection Against Radioactivity Release from Spent Fuel and Waste Storage 11.2-3 11.2.1.13 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment 11.2-3 11.2.1.14 Liquid Radwaste System Safety Function Requirements 11.2-3 11.2.1.15 10 CFR Part 20 - Standards for Protection Against Radiation 11.2-3 11.2.1.16 10 CFR 50.49 - Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants 11.2-3 11.2.1.17 10 CFR 50.55a(f) - Inservice Testing Requirements 11.2-4 11.2.1.18 10 CFR 50.55a(g) - Inservice Inspection Requirements 11.2-4 11.2.1.19 10 CFR Part 50, Appendix I - Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as is Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents 11.2-4 11.2.1.20 40 CFR Part 190 - Environmental Radiation Protection Standards for Nuclear Power Operations 11.2-4 11.2.1.21 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 11.2-4 11.2.1.22 Regulatory Guide 1.143, Revision 1, October 1979 - Design Guidance for Radioactive Waste Management Structures, Systems, and Components Installed In Light-Water-Cooled Nuclear Power Plants 11.2-4 11.2.1.23 NUREG-0737 (Item II.F.1), November 1980 - Clarification of TMI Action Plan Requirements 11.2-4 11.2.1.24 Generic Letter 96-06, September 1996 - Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions 11.2-5 11.2.2 SYSTEM DESCRIPTION 11.2-5 11.2.2.1 General 11.2-5 11.2.2.2 Liquid Radwaste System Operation 11.2-8 11.2.2.3 System Design 11.2-8 ii Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 CONTENTS (Continued) Section Title Page 11.2.2.4 Performance Data 11.2-9 11.2.2.5 Plant Releases 11.2-9 11.2.2.6 Dilution Factors 11.2-13 11.2.2.7 Calculated Doses 11.2-13 11.2.3 SAFETY EVALUATION 11.2-14 11.2.3.1 General Design Criterion 2, 1967 - Performance Standards 11.2-14 11.2.3.2 General Design Criterion 3, 1971 - Fire Protection 11.2-14 11.2.3.3 General Design Criterion 4, 1967 - Sharing of Systems 11.2-15 11.2.3.4 General Design Criterion 11, 1967 - Control Room 11.2-15 11.2.3.5 General Design Criterion 12, 1967 - Instrumentation and Control Systems 11.2-15 11.2.3.6 General Design Criterion 17, 1967 - Monitoring Radioactivity Releases 11.2-16 11.2.3.7 General Design Criterion 40, 1967 - Missile Protection 11.2-17 11.2.3.8 General Design Criterion 49, 1967 - Containment Design Basis 11.2-17 11.2.3.9 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment 11.2-17 11.2.3.10 General Design Criterion 56, 1971 - Primary Containment Isolation 11.2-17 11.2.3.11 General Design Criterion 68, 1967 - Fuel and Waste Storage Radiation Shielding 11.2-17 11.2.3.12 General Design Criterion 69, 1967 - Protection Against Radioactivity Release from Spent Fuel and Waste Storage 11.2-17 11.2.3.13 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment 11.2-21 11.2.3.14 Liquid Radwaste System Safety Function Requirements 11.2-24 11.2.3.15 10 CFR Part 20 - Standards for Protection Against Radiation 11.2-24 11.2.3.16 10 CFR 50.49 - Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants 11.2-25 11.2.3.17 10 CFR 50.55a(f) - Inservice Testing Requirements 11.2-25 11.2.3.18 10 CFR 50.55a(g) - Inservice Inspection Requirements 11.2-25 11.2.3.19 10 CFR Part 50, Appendix I - Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as is Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents 11.2-25 11.2.3.20 40 CFR Part 190 - Environmental Radiation Protection Standards for Nuclear Power Operations 11.2-26 11.2.3.21 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 11.2-26 11.2.3.22 Regulatory Guide 1.143, Revision 1, October 1979 - Design iii Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 CONTENTS (Continued) Section Title Page Guidance for Radioactive Waste Management Structures, Systems, and Components Installed In Light-Water-Cooled Nuclear Power Plants 11.2-26 11.2.3.23 NUREG-0737 (Item II.F.1), November 1980 - Clarification of TMI Action Plan Requirements 11.2-27 11.2.3.24 Generic Letter 96-06, September 1996 - Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions 11.2-27 11.2.4 TESTS AND INSPECTIONS 11.2-27 11.2.5 INSTRUMENTATION APPLICATIONS 11.2-27 11.
2.6 REFERENCES
11.2-27 11.
2.7 REFERENCES
DRAWINGS 11.2-28 11.3 GASEOUS RADWASTE SYSTEM 11.3-1 11.3.1 DESIGN BASES 11.3-1 11.3.1.1 General Design Criterion 2, 1967 - Performance Standards 11.3-1 11.3.1.2 General Design Criterion 3, 1971 - Fire Protection 11.3-1 11.3.1.3 General Design Criterion 4, 1967 - Sharing of Systems 11.3-1 11.3.1.4 General Design Criterion 11, 1967 - Control Room 11.3-1 11.3.1.5 General Design Criterion 12, 1967 - Instrumentation and Control Systems 11.3-1 11.3.1.6 General Design Criterion 17, 1967 - Monitoring Radioactivity Releases 11.3-1 11.3.1.7 General Design Criterion 18, 1967 - Monitoring Fuel and Waste Storage 11.3-2 11.3.1.8 General Design Criterion 68, 1967 - Fuel and Waste Storage Radiation Shielding 11.3-2 11.3.1.9 General Design Criterion 69, 1967 - Protection Against Radioactivity Release from Spent Fuel and Waste Storage 11.3-2 11.3.1.10 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment 11.3-2 11.3.1.11 10 CFR Part 20 - Standards for Protection Against Radiation 11.3-2 11.3.1.12 10 CFR Part 50 Appendix I - Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as is Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents 11.3-2 11.3.1.13 40 CFR Part 190 - Environmental Radiation Protection iv Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 CONTENTS (Continued) Section Title Page Standards for Nuclear Power Operations 11.3-2 11.3.1.14 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 11.3-3 11.3.1.15 Regulatory Guide 1.143, Revision 1, October 1979 - Design Guidance for Radioactive Waste Management Structures, Systems, and Components Installed In Light-Water-Cooled Nuclear Power Plants 11.3-3 11.3.1.16 NUREG-0737 (Items II.F.1 and III.D.1.1), November 1980 - Clarification of TMI Action Plan Requirements 11.3-3 11.3.2 SYSTEM DESCRIPTION 11.3-3 11.3.2.1 Gaseous Radwaste System Operation 11.3-5 11.3.2.2 System Design 11.3-5 11.3.2.3 Plant Releases 11.3-6 11.3.2.4 Dilution Factors 11.3-9 11.3.2.5 Doses 11.3-9 11.3.3 SAFETY EVALUATION 11.3-10 11.3.3.1 General Design Criterion 2, 1967 - Performance Standards 11.3-10 11.3.3.2 General Design Criterion 3, 1971 - Fire Protection 11.3-10 11.3.3.3 General Design Criterion 4, 1967 - Sharing of Systems 11.3-10 11.3.3.4 General Design Criterion 11, 1967 - Control Room 11.3-11 11.3.3.5 General Design Criterion 12, 1967 - Instrumentation and Control Systems 11.3-11 11.3.3.6 General Design Criterion 17, 1967 - Monitoring Radioactivity Releases 11.3-11 11.3.3.7 General Design Criterion 18, 1967 - Monitoring Fuel and Waste Storage 11.3-11 11.3.3.8 General Design Criterion 68, 1967 - Fuel and Waste Storage Radiation Shielding 11.3-12 11.3.3.9 General Design Criterion 69, 1967 - Protection Against Radioactivity Release from Spent Fuel and Waste Storage 11.3-12 11.3.3.10 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment 11.3-13 11.3.3.11 10 CFR Part 20 - Standards for Protection Against Radiation 11.3-14 11.3.3.12 10 CFR Part 50 Appendix I - Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as is Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents 11.3-14 11.3.3.13 40 CFR Part 190 - Environmental Radiation Protection Standards for Nuclear Power Operations 11.3-15 v Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 CONTENTS (Continued) Section Title Page 11.3.3.14 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 11.3-15 11.3.3.15 Regulatory Guide 1.143, Revision 1, October 1979 - Design Guidance for Radioactive Waste Management Structures, Systems, and Components Installed In Light-Water-Cooled Nuclear Power Plants 11.3-15 11.3.3.16 NUREG-0737 (Items II.F.1 and III.D.1.1), November 1980 - Clarification of TMI Action Plan Requirements 11.3-15 11.3.4 TESTS AND INSPECTIONS 11.3-16 11.3.5 INSTRUMENTATION APPLICATIONS 11.3-16 11.
3.6 REFERENCES
11.3-16 vi Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 CONTENTS (Continued) Section Title Page 11.4 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING SYSTEM 11.4-1 11.4.1 DESIGN BASES 11.4-1 11.4.1.1 General Design Criterion 2, 1967 - Performance Standards 11.4-1 11.4.1.2 General Design Criterion 3, 1971 - Fire Protection 11.4-1 11.4.1.3 General Design Criterion 4, 1967 - Sharing of Systems 11.4-1 11.4.1.4 General Design Criterion 11, 1967 - Control Room 11.4-1 11.4.1.5 General Design Criterion 12, 1967 - Instrumentation and Control Systems 11.4-1 11.4.1.6 General Design Criterion 16, 1967 - Monitoring Reactor Coolant Pressure Boundary 11.4-1 11.4.1.7 General Design Criterion 17, 1967 - Monitoring Radioactive Releases 11.4-2 11.4.1.8 General Design Criterion 18, 1967 - Monitoring Fuel and Waste Storage 11.4-2 11.4.1.9 General Design Criterion 19, 1971 - Control Room 11.4-2 11.4.1.10 General Design Criterion 21, 1967 - Single Failure Definition 11.4-2 11.4.1.11 General Design Criterion 40, 1067 - Missile Protection 11.4-2 11.4.1.12 General Design Criterion 49, 1967 - Containment Design Basis 11.4-2 11.4.1.13 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment 11.4-2 11.4.1.14 General Design Criterion 56, 1971 - Primary Containment Isolation 11.4-3 11.4.1.15 General Design Criterion 69, 1967 - Protection Against Radioactivity Release from Spent Fuel and Waste Storage 11.4-2 11.4.1.16 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment 11.4-2 11.4.1.17 Radiological Monitoring System Safety Function Requirements 11.4-2 11.4.1.18 10 CFR Part 20 - Standards for Protection Against Radiation 11.4-3 11.4.1.19 10 CFR 50.49 - Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants 11.4-3 11.4.1.20 10 CFR 50.68(b) - Criticality Accident Requirements 11.4-4 11.4.1.21 10 CFR Part 50, Appendix I - Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as is Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents 11.4-4 11.4.1.22 Safety Guide 13, March 1971 - Fuel Storage Facility Design Basis 11.4-4 11.4.1.23 Regulatory Guide 1.21, Revision 1, June 1974 - Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous vii Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 CONTENTS (Continued) Section Title Page Effluents from Light-Water-Cooled Nuclear Power Plants 11.4-4 11.4.1.24 Regulatory Guide 1.45, May 1973 - Reactor Coolant Pressure Boundary Leakage Detection Systems 11.4-4 11.4.1.25 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 11.4-5 11.4.1.26 NUREG-0737 (Items II.F.1, III.A.1.2, and III.D.1.1), November 1980 - Clarification of TMI Action Plan Requirements 11.4-5 11.4.2 SYSTEM DESCRIPTION 11.4-5 11.4.2.1 Continuous Monitoring 11.4-5 11.4.2.2 Sampling 11.4-20 11.4.3 SAFETY EVALUATION 11.4-21 11.4.3.1 General Design Criterion 2, 1967 - Performance Standards 11.4-21 11.4.3.2 General Design Criterion 3, 1971 - Fire Protection 11.4-21 11.4.3.3 General Design Criterion 4, 1967 - Sharing of Systems 11.4-21 11.4.3.4 General Design Criterion 11, 1967 - Control Room 11.4-21 11.4.3.5 General Design Criterion 12, 1967 - Instrumentation and Control Systems 11.4-22 11.4.3.6 General Design Criterion 16, 1967 - Monitoring Reactor Coolant Pressure Boundary 11.4-22 11.4.3.7 General Design Criterion 17, 1967 - Monitoring Radioactive Releases 11.4-22 11.4.3.8 General Design Criterion 18, 1967 - Monitoring Fuel and Waste Storage 11.4-22 11.4.3.9 General Design Criterion 19, 1971 - Control Room 11.4-22 11.4.3.10 General Design Criterion 21, 1967 - Single Failure Definition 11.4-23 11.4.3.11 General Design Criterion 40, 1967 - Missile Protection 11.4-23 11.4.3.12 General Design Criterion 49, 1967 - Containment Design Basis 11.4-23 11.4.3.13 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment 11.4-23 11.4.3.14 General Design Criterion 56, 1971 - Primary Containment Isolation 11.4-23 11.4.3.15 General Design Criterion 69, 1967 - Protection Against Radioactivity Release from Spent Fuel and Waste Storage 11.4-23 11.4.3.16 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment 11.4-24 11.4.3.17 Radiological Monitoring System Safety Function Requirements 11.4-24 11.4.3.18 10 CFR Part 20 - Standards for Protection Against Radiation 11.4-24 11.4.3.19 10 CFR 50.49 - Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants 11.4-24 viii Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 CONTENTS (Continued) Section Title Page 11.4.3.20 10 CFR 50.68(b) - Criticality Accident Requirements 11.4-24 11.4.3.21 10 CFR Part 50, Appendix I - Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as is Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents 11.4-25 11.4.3.22 Safety Guide 13, March 1971 - Fuel Storage Facility Design Basis 11.4-25 11.4.3.23 Regulatory Guide 1.21, Revision 1, June 1974 - Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants 11.4-25 11.4.3.24 Regulatory Guide 1.45, May 1973 - Reactor Coolant Pressure Boundary Leakage Detection Systems 11.4-25 11.4.3.25 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 11.4-25 11.4.3.26 NUREG-0737 (Items II.F.1, III.A.1.2, and III.D.1.1), November 1980 - Clarification of TMI Action Plan Requirements 11.4-26 11.4.4 TESTS AND INSPECTIONS 11.4-27 11.4.4.1 Alarm Setpoints 11.4-27 11.4.4.2 Definitions 11.4-27 11.4.4.3 Calibration Procedure 11.4-27 11.4.4.4 Test Frequencies 11.4-28 11.4.4.5 System Summary 11.4-28 11.4.5 INSTRUMENTATION APPLICATIONS 11.4-28 11.
4.6 REFERENCES
11.4-28 11.5 SOLID RADWASTE SYSTEM 11.5-1 11.5.1 DESIGN BASES 11.5-1 11.5.1.1 General Design Criterion 3, 1971 - Fire Protection 11.5-1 11.5.1.2 General Design Criterion 4, 1967 - Sharing of Systems 11.5-1 11.5.1.3 General Design Criterion 17, 1967 - Monitoring Radioactivity Releases 11.5-1 11.5.1.4 General Design Criterion 18, 1967 - Monitoring Fuel and Waste Storage 11.5-1 11.5.1.5 General Design Criterion 68, 1967 - Fuel and Waste Storage Radiation Shielding 11.5-1 11.5.1.6 General Design Criterion 70, 1967 - Control of Releases of ix Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 CONTENTS (Continued) Section Title Page Radioactivity to the Environment 11.5-1 11.5.1.7 10 CFR Part 20 - Standards for Protection Against Radiation 11.5-2 11.5.1.8 10 CFR 61.55 - Waste Classification 11.5-2 11.5.1.9 10 CFR Part 71 - Packaging and Transportation of Radioactive Material 11.5-2 11.5.1.10 40 CFR Part 190 - Environmental Radiation Protection Standards for Nuclear Power Operations 11.5-2 11.5.1.11 49 CFR Parts 171-178 - Subchapter C-Hazardous Materials Regulations 11.5-2 11.5.1.12 Regulatory Guide 1.143, Revision 1, October 1979 - Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed In Light-Water-Cooled Nuclear Power Plants 11.5-2 11.5.1.13 IE Bulletin 79-19, August 1979 - Packaging of Low-Level Radioactive Waste for Transport and Burial 11.5-2 11.5.1.14 10 CFR 50.48(c) - National Fire Protection Association Standard NFPA 805 11.5-3 11.5.2 SYSTEM DESCRIPTION 11.5-3 11.5.2.1 System Inputs 11.5-3 11.5.2.2 Components 11.5-3 11.5.2.3 Mixed Waste 11.5-6 11.5.2.4 Component Failures and System Malfunctions 11.5-6 11.5.2.5 Packaging 11.5-6 11.5.2.6 Storage Facilities 11.5-7 11.5.2.7 Shipment 11.5-8 11.5.3 SAFETY EVALUATION 11.5-8 11.5.3.1 General Design Criterion 3, 1971 - Fire Protection 11.5-8 11.5.3.2 General Design Criterion 4, 1967 - Sharing of Systems 11.5-8 11.5.3.3 General Design Criterion 17, 1967 - Monitoring Radioactivity Releases 11.5-8 11.5.3.4 General Design Criterion 18, 1967 - Monitoring Fuel and Waste Storage 11.5-8 11.5.3.5 General Design Criterion 68, 1967 - Fuel and Waste Storage Radiation Shielding 11.5-9 11.5.3.6 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment 11.5-9 11.5.3.7 10 CFR Part 20 - Standards for Protection Against Radiation 11.5-9 11.5.3.8 10 CFR 61.55 - Waste Classification 11.5-10 11.5.3.9 10 CFR Part 71 - Packaging and Transportation of Radioactive x Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 CONTENTS (Continued) Section Title Page Material 11.5-10 11.5.3.10 40 CFR Part 190 - Environmental Radiation Protection Standards for Nuclear Power Operations 11.5-10 11.5.3.11 49 CFR Parts 171-178 - Subchapter C-Hazardous Materials Regulations 11.5-10 11.5.3.12 Regulatory Guide 1.143, Revision 1, October 1979 - Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed In Light-Water-Cooled Nuclear Power Plants 11.5-10 11.5.3.13 IE Bulletin 79-19, August 1979 - Packaging of Low-Level Radioactive Waste for Transport and Burial 11.5-10 11.5.3.14 10 CFR 50.48(c) - National Fire Protection Association Standard NFPA 805 11.5-11 11.5.4 TESTS AND INSPECTIONS 11.5-11 11.5.5 INSTRUMENTATION APPLICATIONS 11.5-11 11.
5.6 REFERENCES
11.5-11 11.5.7 REFERENCE DRAWINGS 11.5-11 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM 11.6-1 11.6.1 DESIGN BASES 11.6-1 11.6.1.1 10 CFR Part 50, Appendix I - Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as is Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents 11.6-1 11.6.2 PROGRAM DESCRIPTION 11.6-1 11.6.2.1 Expected Background 11.6-1 11.6.2.2 Critical Pathways 11.6-2 11.6.2.3 Sampling Media, Location and Frequency 11.6-3 11.6.2.4 Analytical Sensitivity 11.6-3 11.6.2.5 Data Analysis and Presentation 11.6-4 11.6.2.6 Program Statistical Sensitivity 11.6-5 11.6.3 SAFETY EVALUATION 11.6-5 11.6.3.1 10 CFR Part 50, Appendix I - Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as is Reasonably Achievable" for Radioactive Material in xi Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 CONTENTS (Continued) Section Title Page Light-Water-Cooled Nuclear Power Reactor Effluents 11.6-5 11.
6.4 REFERENCES
11.6-5 xii Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 TABLES Table Title 11.0-1 Applicable Design Basis Criteria 11.1-1 Library of Physical Data for Isotopes (Historical) 11.1-2 Basic Assumptions for Core and Coolant Inventories for Design Basis Case (Historical) 11.1-2A Basic Assumptions for Core and 1% Fuel Defects Coolant Inventories 11.1-3 Basic Assumptions for Core and Coolant Inventories for Normal Operation Case (Historical) 11.1-4 Core Activity Inventories for Design Basis Case (Curies) (Historical) 11.1-5 Core Activity Inventories for Normal Operation Case (Curies) (Historical) 11.1-6 Basic Assumptions for Fuel Rod Gap Activities (Historical) 11.1-7 Activity in Fuel Rod Gaps (Historical) 11.1-8 Input Constants for Coolant Activities for Design Basis Case (Historical) 11.1-9 Input Constants for Coolant Activities for Normal Operation Case (Historical) 11.1-10 Basic Data for Corrosion Product Activities (Historical) 11.1-11 Primary Coolant Activities for Design Basis Case (Historical) 11.1.11A Updated Primary Coolant Activities Based on 3580 MWt, 600 EFPDs, 1% Fuel Defects 11.1-12 Primary Coolant Activities for Normal Operation Case (Historical) 11.1-13 Reactor Coolant Nitrogen-16 Activity (Historical) 11.1-14 Deposited Corrosion Product Activity in Steam Generator (Historical) xiii Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 TABLES (Continued) Table Title 11.1-15 Demineralizer and Evaporator Decontamination Factors (Historical)11.1-16 11.1.16 Production and Removals in Primary Coolant for Design Basis Case (Historical) 11.1-17 Production and Removals in Primary Coolant for Normal Operation Case (Historical) 11.1-18 Basic Assumptions for Pressurizer Activities (Historical) 11.1-19 Activity in Pressurizer for Design Basis Case (Historical) 11.1-20 Activity in Pressurizer for Normal Operation Case (Historical) 11.1-21 Basic Assumptions for Tritium Activity in Primary Coolant (Historical) 11.1-22 Tritium Activities in Primary Coolant (Historical) 11.1-23 Steam System Operating Conditions Assumed for Activity Analysis for Normal Operation Case (Historical) 11.1-24 Additional Secondary System Operating Parameters (Historical) 11.1-25 Steam Generator Partition Factors (Historical) 11.1-26 Total Additions and Removals of Activity in Each Steam Generator for Normal Operation Case (Curies) (Historical) 11.1-27 Equilibrium Activities and Concentrations in Each Steam Generator for Normal Operation Case (Historical) 11.1-28 Total Additions and Removals of Activity in the Condenser for Normal Operation Case (Curies) (Historical) 11.1-29 Equilibrium Activities and Concentrations in the Condenser for Normal Operation Case (Historical) 11.1-30 Total Additions and Removals of Activity in the Condenser Vapor Space for Normal Operation Case (Historical) xiv Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 TABLES (Continued) Table Title 11.1-31 Equilibrium Activities and Concentrations in the Condenser Vapor Space for Normal Operation Case (Historical) 11.2-1 Assumptions Used for Input Waste Streams and Activity Calculations (Historical) 11.2-2 Assumptions for Calculations of Activity Released from CVCS (Historical) 11.2-3 Activity Concentration Spectrum I Through V for Input Waste Sources, Design Basis Case (Historical) 11.2-4 Activity Concentration Spectrum I Through V for Input Waste Sources, Design Basis Case (Historical) 11.2-5 Annual Flow and Isotopic Spectra for Liquid Waste Inputs (Historical) 11.2-6 Isotopic Flows Through CVC System, Design Basis Case (Historical) 11.2-7 Isotopic Flows Through CVC System, Normal Operation Case (Historical) 11.2-8 Annual Flow and Activity Concentration of Process Streams for Design Basis Case (Historical) 11.2-9 Annual Flow and Activity Concentration of Process Streams for Normal Operation Case (Historical) 11.2-10 Equipment Design Summary Data - Liquid Radwaste System 11.2-11 Parameters Used in Tritium Analysis for Plant Water Sources (Historical) 11.2-12 Deleted in Revision 1 11.2-13 Calculated and Assumed Holdup Times for Liquid Waste System Tanks (Historical) 11.2-14 Estimated Annual Activity Release for Design Basis Case (One unit) (Historical) xv Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 TABLES (Continued) Table Title 11.2-15 Estimated Annual Activity Release for Normal Operation Case (One unit) (Historical) 11.2-16 Annual Flow and Activity Concentration of Process Streams for Steam Generator Blowdown System for Normal Operation Case (Historical) 11.2-17 Summary of Estimated Liquid Waste System Annual Waste Volumes for Units 1 and 2 (Historical) 11.2-18 Estimated Annual Liquid Effluent Release for Normal Operation Case with Anticipated Operational Occurrences (Historical) 11.2-19 Basic Assumptions for Liquid Pathways Exposures (Historical) 11.2-20 Bioaccumulation Factors (Historical) 11.2-21 Effluent Concentrations After Initial Dilution: Design Basis Case (Historical) 11.2-22 Effluent Concentrations After Initial Dilution: Normal Operation Case 11.2-23 Effluent Concentrations After Initial Dilution: Normal Operation with Anticipated Operational Occurrences (Historical) 11.2-24 Doses Resulting from Radioactive Releases in Liquid Wastes: Design Basis Case (mrem/yr) (Historical) 11.2-25 Doses Resulting from Radioactive Releases in Liquid Wastes: Normal Operation Case (mrem/yr) (Historical) 11.2-26 Doses Resulting from Radioactive Releases in Liquid Wastes: Normal Operational with Anticipated Operational Occurrences (mrem/yr) (Historical) 11.3-1 Equipment Design and Operating Parameters for Gaseous Radwaste System Units 1 and 2 11.3-2 Gaseous Waste System Release: Design Basis Case (Curies) (Historical) 11.3-3 Gaseous Waste System Release: Normal Operation Case (Curies) (Historical) xvi Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 TABLES (Continued) Table Title 11.3-4 Annual Gaseous Radwaste Flows (Historical) 11.3-5 Maximum Activity in Gas Decay Tank: Design Basis Case (Historical) 11.3-6 Maximum Activity in Gas Decay Tank: Normal Operation Case (Historical) 11.3-7 Activity in Volume Control Tank: Design Basis Case (Historical) 11.3-8 Activity in Volume Control Tank: Normal Operation Case (Historical) 11.3-9 Gaseous Releases due to Cold Shutdown and Startups (Historical) 11.3-10 Distances in Miles From DCPP Unit 1 Reactor Centerline to the Nearest Milk Cow, Meat Animal, Milk Goat, Residence, Vegetable Garden, and Site Boundary (Historical) 11.3-11 Estimates of Relative Concentration (/Q) at Locations Specified in Table 11.3-10 (Historical) 11.3-12 Estimates of Deposition (/Q) at Locations Specified in Table 11.3-10 (Historical) 11.3-13 Annual Average Atmosphere Activity Concentrations at Site Boundary for Design Basis Case (Historical) 11.3-14 Annual Average Atmospheric Activity Concentrations at Site Boundary for Normal Operation Case (Historical) 11.3-15 Offsite Doses for NW Sector at Distance 0.5 Mi: Design Basis Case (Historical) 11.3-16 Offsite Doses for NW Sector at Distance 3.6 Mi: Design Basis Case (Historical) 11.3-17 Offsite Doses for NNW Sector at Distance 0.5 Mi: Design Basis Case (Historical) 11.3-18 Offsite Doses for NNW Sector at Distance 1.5 Mi: Design Basis Case (Historical) xvii Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 TABLES (Continued) Table Title 11.3-19 Offsite Doses for NNW Sector at Distance 3.6 Mi: Design Basis Case (Historical) 11.3-20 Offsite Doses for N Sector at Distance 0.5 Mi: Design Basis Case (Historical) 11.3-21 Offsite Doses for NNE Sector at Distance 0.5 Mi: Design Basis Case (Historical) 11.3-22 Offsite Doses for NE Sector at Distance 0.5 Mi: Design Basis Case (Historical) 11.3-23 Offsite Doses for ENE Sector at Distance 0.7 Mi: Design Basis Case (Historical) 11.3-24 Offsite Doses for ENE Sector at Distance 4.5 Mi: Design Basis Case (Historical) 11.3-25 Offsite Doses for E Sector at Distance 1.0 Mi: Design Basis Case (Historical) 11.3-26 Offsite Doses for ESE Sector at Distance 1.0 Mi: Design Basis Case (Historical) 11.3-27 Offsite Doses for ESE Sector at Distance 3.7 Mi: Design Basis Case (Historical) 11.3-28 Offsite Doses for SE Sector at Distance 1.1 Mi: Design Basis Case (Historical) 11.3-29 Offsite Doses for SE Sector at Distance 3.7 Mi: Design Basis Case (Historical) 11.3-30 Offsite Doses for NW Sector at Distance 0.5 Mi: Normal Operation Case (Historical) 11.3-31 Offsite Doses for NW Sector at Distance 3.6 Mi: Normal Operation Case (Historical) 11.3-32 Offsite Doses for NNW Sector at Distance 0.5 Mi: Normal Operation Case (Historical) xviii Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 TABLES (Continued) Table Title 11.3-33 Offsite Doses for NNW Sector at Distance 1.5 Mi: Normal Operation Case (Historical) 11.3-34 Offsite Doses for NNW Sector at Distance 3.6 Mi: Normal Operation Case (Historical) 11.3-35 Offsite Doses for N Sector at Distance 0.5 Mi: Normal Operation Case (Historical) 11.3-36 Offsite Doses for NNE Sector at Distance 0.5 Mi: Normal Operation Case (Historical) 11.3-37 Offsite Doses for NE Sector at Distance 0.5 Mi: Normal Operation Case (Historical) 11.3-38 Offsite Doses for ENE Sector at Distance 0.7 Mi: Normal Operation Case (Historical) 11.3-39 Offsite Doses for ENE Sector at Distance 4.5 Mi: Normal Operation Case (Historical) 11.3-40 Offsite Doses for E Sector at Distance 1 Mi: Normal Operation Case (Historical) 11.3-41 Offsite Doses for ESE Sector at Distance 1 Mi: Normal Operation Case (Historical) 11.3-42 Offsite Doses for ESE Sector at Distance 3.7 Mi: Normal Operation Case (Historical) 11.3-43 Offsite Doses for SE Sector at Distance 1.1 Mi: Normal Operation Case (Historical) 11.3-44 Offsite Doses for SE Sector at Distance 3.7 Mi: Normal Operation Case (Historical) 11.4-1 Radiation Monitors and Readouts 11.4-2 Deleted in Revision 10 11.4-3 Radiation Monitor-Valve Control Operations xix Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 TABLES (Continued) Table Title 11.5-1 Solid Radwaste System Input Volumes 11.5-2 Activity in Radwaste System Demineralizers: Normal Operation Cases (Curies/Year) (Historical) 11.5-3 Deleted in Revision 6 11.5-4 Activity Collected in Radwaste Filter Cartridges at Time of Replacement: (Curies) (Historical) 11.5-5 Summary of Radwaste Materials Shipment 11.6-1 Deleted in Revision 23 11.6-2 Deleted in Revision 1 11.6-3 Deleted in Revision 11 11.6-4 Environmental Radiological Monitoring Program Summary (Historical) 11.6-5 Deleted in Revision 1 11.6-6 Deleted in Revision 1 11.6-7 Deleted in Revision 1 11.6-8 Deleted in Revision 1 11.6-9 Deleted in Revision 1 11.6-10 Deleted in Revision 11 11.6-11 Maximum Values for the Lower Limits of Detection (LLD) 11.6-12 Deleted in Revision 11 11.6-13 Estimated Relative Concentration 11.6-14 Estimated Depositions xx Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 FIGURES Figure Title 11.1-1 R. E. Ginna Plant Tritium Sources - Measured and Predicted (HISTORICAL) 11.2-1 Deleted in Revision 18 11.2-2 Liquid Waste Process Flow Diagram: Design Basis (HISTORICAL) 11.2-3 Liquid Waste Process Flow Diagram: Normal Operation (HISTORICAL) 11.2-4 Blowdown System Flow Diagram - Discharge Mode 11.2-5 Blowdown System Flow Diagram Recycle Path 11.2-6 Tritium Concentration in Water Versus Time (HISTORICAL) 11.2-7 Tritium Airborne Concentration in Fuel Handling Area Versus Time (HISTORICAL) 11.2-8 Tritium Airborne Concentration in Containment Versus Time (HISTORICAL) 11.2-9 Site Plot Plan - Radwaste Discharge 11.3-1 Deleted in Revision 1 11.3-2 Deleted in Revision 1 11.3-3 Deleted in Revision 1 11.3-4 Gaseous Waste Systems' Release Points 11.4-1 Radiation Monitoring System 11.5-1 Solid Radwaste System 11.5-2 Deleted in Revision 3 11.5-3 Spent Resin Flow Diagram 11.5-4 Location of Onsite Storage Facility xxi Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 FIGURES Figure Title 11.5-5(a) Solid Radwaste Storage Building 11.5-5A Deleted in Revision 11 11.5-6 Chemical and Volume Control System Displaying Filters 11.5-7 Spent Fuel Pool Displaying Filters 11.5-8 Liquid Radwaste System Displaying Filters 11.5-9 System for Capturing Expended Filter Cartridges 11.5-10 Cover of Filter Vessel 11.5-11 Spent Resin Storage Area 11.5-12 Load Out Station 11.6-1 Deleted in Revision 11 11.6-2 Deleted in Revision 11 11.6-3 Deleted in Revision 4 11.6-4 Deleted in Revision 11 11.6-5 Deleted in Revision 11 NOTE: (a) This figure corresponds to a controlled engineering drawing that is incorporated by reference into the FSAR Update. See Table 1.6-1 for the correlation between the FSAR Update figure number and the corresponding controlled engineering drawing number. xxii Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Chapter 11 RADIOACTIVE WASTE MANAGEMENT The purpose of this chapter is to provide a complete description and state the design objectives of the radioactive waste systems to demonstrate compliance with the general provisions of 10 CFR Part 20 and 10 CFR Part 50. Performance evaluation of the radioactive waste treatment systems is described. This chapter is divided into the following sections: 11.1 Source Terms 11.2 Liquid Radwaste System (LRS) 11.3 Gaseous Radwaste System (GRS) 11.4 Process and Effluent Radiological Monitoring System (RMS) 11.5 Solid Radwaste System (SRS) 11.6 Offsite Radiological Monitoring Program In the sections on liquid and gaseous radwaste systems, all significant release pathways of radioactive liquids and gases are identified and discussed, including those not directly associated with radwaste treatment systems; for example, blowdown system releases and steam leakage. The principles and guidelines used in the design, construction, and operation of the radioactive waste management systems described in Chapter 11 are specified in the individual sections of Chapter 11 and in Table 11.0-1. A pre-operation evaluation of the liquid and gaseous radwaste systems was performed to demonstrate the systems abilities to maintain releases within regulatory limits. To carry out an activity analysis of the plant that covers different combinations of basic operating parameters, two cases were selected: (1) Design Basis Case - source term analysis used as input into the design of plant systems and topical areas (e.g., shielding and radwaste) to ensure compliance with 10 CFR Part 20. Inputs and assumptions were conservative. (2) Normal Operation Case - source term analysis used for pre-operational evaluations to ensure compliance with the minimum equipment requirements of 10 CFR Part 50, Appendix I. Inputs and assumptions included anticipated operational occurrences and were more realistic than those for the design basis case. These cases do not represent the typical results of actual plant operation, and the operational basis and resulting values for either case are not license limits. They were 11-1 Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE developed for initial plant licensing and are not expected to be updated, therefore they are considered historical. HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED. For the Design Basis Case, the plant was assumed to have been operated for a full year at full thermal power of 3568 MWt with a capacity factor of 80 percent and a fuel defect level of 1 percent. The radwaste systems were assumed to be in operation as designed, and primary system leakage was assumed to be negligible. The complete set of other assumptions associated with this case are listed in Table 11.1-2 and discussed in detail in the following sections. For the Normal Operation Case the plant was assumed to have been operated for a full year at full power with a capacity factor of 80 percent and a fuel defect level of 0.12 percent. Coincident with this condition, it was assumed that there existed primary system leakage of 100 pounds per day to the secondary system, 1 percent of primary coolant noble gas inventory and 0.001 percent of primary coolant iodine inventory to the containment, and 160 pounds per day to the auxiliary building. These assumptions are in agreement with those provided in NUREG-0017, April 1976 for normal operation of a pressurized water reactor (PWR) (refer to Reference 9 of Section 11.1). The complete set of assumptions for this case is also discussed in the following sections. As a result of the analyses of these cases, the following conclusions can be drawn. Diablo Canyon Power Plant (DCPP) Unit 1 and Unit 2 can be operated under normal conditions, including the consideration of anticipated operational occurrences, in conformance with: (1) The general provisions of 10 CFR Part 20 and 10 CFR Part 50 (2) The annual dose limits established in 10 CFR Part 20 for the release of radioactive materials (3) The annual radiation dose limits specified in 10 CFR Part 50, Appendix I. The radioactive waste release values provided in this chapter are nominal values. Actual release values are reported to the NRC in the DCPP Annual Radioactive Effluent Release Report. 11-2 Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE 11.1 SOURCE TERMS This section describes the routine (or operational) source term and the pre-operational evaluation of source term for the Design Basis Case and the Normal Operation Case. Routine Source Term: Primary coolant concentrations, leakage rates, and effluent release rates based on the routine source term are used for evaluation of the LRS and GRS to determine whether these systems meet the dose design objectives of 10 CFR Part 50 Appendix I. The operational source term in the reactor coolant system (RCS) and supporting systems are monitored on a routine basis in accordance with plant Technical Specifications and plant-approved procedures. The information is readily available to site personnel for evaluating source term and trends. The station policy is to operate the plant with zero fuel defects to achieve a low source term in the primary coolant. The plant operating philosophy is to maintain leakage from the primary system well below Technical Specification limits. This operating philosophy is fundamental to minimizing the input of radioactive material into the LRS and the GRS, and thereby minimizes activity that may be released from the station. There have been periods when a fuel leak develops and the RCS Dose Equivalent I-131 specific activity (DEI) has been near the Technical Specification value of 1.0 Ci/gm for periods of time. The station has demonstrated that the LRS and GRS operating per station procedures have maintained plant releases at a fraction of Technical Specification and 10 CFR Part 20 limits. Routine operating RCS DEI is normally only a fraction of the Technical Specification value of 1.0 Ci/gm. The routine operating source term is much lower than the cases described below. Tritium is produced as part of the fission process of the reactor and its production is a direct function of power produced and capacity factor. Tritiated water is not removed in the treatment systems. Given that tritium has a 12.3 year half-life, essentially all the tritium produced is released from the plant via the LRS or through evaporation via the plant vent. HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED. Pre Operation Source Term Evaluations: 11.1-1 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE The pre-operational evaluation of radioactive materials produced and stored in the reactor system is reported and discussed in this section. These sources have been computed for two basic sets of plant operating conditions: the Design Basis Case and the Normal Operation Case. The complete isotopic source terms are presented in tabular form along with the basic assumptions used in the computations. The activities and concentrations were calculated with the EMERALD-NORMAL (Reference 8) digital computer program. A detailed discussion of the physical data and assumptions used is contained in the following paragraphs: 11.1.1 BASIC PHYSICAL DATA AND CONSTANTS The values of isotopic physical data used in the radiological effects analyses are listed in Table 11.1-1. The values of half-lives and fission yields were taken from the Meek and Rider report (Reference 1) and from Tobias (Reference 10) and are in general agreement with those in TID-14844 (Reference 2) and ORNL-2127 (Reference 3). The values for average beta energies are those provided in Perkins and King (Reference 4) and the average gamma energies are taken from Tobias. The values of decay constants were calculated from the half-lives with the standard formula. The fission yields were modified to account for plutonium buildup by using the values for uranium and plutonium fissioned during an equilibrium core cycle as follows: Yield (U + Pu) = [Fissions (U) x Yield (U)] + [Fissions (Pu) x Yield (Pu)] Fissions (U + Pu) The number of fissions per megawatt-second is taken to be 3.15 x 1016, which agrees well with the value of 3.2 x 1016 used in Reference 2. 11.1.2 DETERMINATION OF ACTIVITY INVENTORIES IN REACTOR CORE HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED. The EMERALD-NORMAL program was run for 11 months with zero initial activities, and the result was decayed one month. The resulting core activities (assuming one-third of the core was replaced at refueling) were then set equal to the initial activities and the process repeated. Taking one-third of the activities after the first cycle, plus one-third of the activities after the second cycle, gives an approximation to the initial activities for an equilibrium core cycle. The core inventories for the year's operating cycle were then computed using an 80 percent capacity factor to ensure a realistic inventory of Kr-85. The power level and other basic assumptions are provided in Tables 11.1-2 and 11.1-3. The resulting core inventories are listed in Tables 11.1-4 and 11.1-5. These calculated core inventories are in general agreement with those tabulated in TID-14844, with those listed in the Reactor Safety Study (WASH-1400) (Reference 12), and with those listed in the DCPP Preliminary Safety Analysis Report (Reference 5). 11.1-2 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE The actual maximum operating configuration is 21 months of operation, with a mixture of fuel with enrichments up to 5 percent, and with a maximum burnup of 50,000 MWD/MTU. The updated equilibrium isotopic core inventory of dose significant isotopes is presented in Table 15.5-77. A summary of key assumptions and a description of the method used to determine the updated core inventory is provided in Table 11.1-2A and Section 15.5.3.1, respectively. HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED. 11.1.3 DETERMINATION OF INVENTORIES IN FUEL ROD GAPS The computed gap activities are based on buildup in the fuel from the fission process and diffusion to the fuel rod gap at rates dependent on the operating temperature. For this analysis, the fuel pellets were divided into five concentric rings, each with a release rate dependent on the mean fuel temperature within that ring. The diffusing isotope is assumed present in the gas gap when it has diffused to the boundary of the outer ring. The core temperature distribution used in this analysis, based on hot channel factors of FH = 1.70 and Fq = 2.82, is presented in Table 11.1-6. The diffusion coefficient, D', for Xe and Kr in UO2, varies with temperature in accordance with the following expression: E 1 1 ( ) D'T D' (1673) exp R T 1673 (11.1-1) where: E = activation energy D' (1673 ) = diffusion coefficient at 1673 K = 1 x 10-11 sec-1 T = temperature in K R = gas constant This expression is valid for temperatures above 1100°C. Below this temperature, fission gas release occurs mainly by two temperature-independent phenomena, recoil and knock-out, and is predicted by using D' at 100°C. The value used for D' (1673 K), based on data at burnups greater than 1019 fission/cc, is used to account for possible fission gas release by other mechanisms and pellet cracking during irradiation. The diffusion coefficients for iodine isotopes are assumed to be the same as those for Xe and Kr (References 6 and 7). The resulting fractions of core activity present in the fuel rod gaps are listed in Table 11.1-7, along with the total inventories of activity in the gaps. 11.1-3 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE As part of the implementation of Alternative Source Terms to assess post-accident dose consequences to the public as well as habitability of the control room and technical support center, the isotopic inventories in fuel rod gaps have been updated as discussed in Section 15.5.3.3. The isotopic activity in the gap assumed for the Locked Rotor Accident and the Control Rod Ejection Accident are provided in Table 15.5-80. The isotopic activity in the gap assumed for the Fuel Handling Accident is presented in Table 15.5-47C. 11.1.4 DETERMINATION OF PRIMARY COOLANT ACTIVITIES The basic data and assumptions used in calculating the coolant concentrations for the Design Basis Case and the Normal Operation Case are provided in Tables 11.1-2, 11.1-3, 11.1-8, 11.1-9, and 11.1-10. The coolant concentrations and activities, listed in Tables 11.1-11, 11.1-12, and 11.1-13, are provided for the operating temperature. The activities of corrosion products deposited in the steam generator are provided in Table 11.1-14. The total amounts of activity produced and removed from the coolant during the operating period are provided in Tables 11.1-16 and 11.1-17. All models and equations, including parent-daughter production, purification terms, boron feed-bleed terms, and coolant leakage terms, are provided in detail in Reference 8, and are generally consistent with those provided in NUREG-0017, April 1976 (Reference 9). The demineralizer decontamination factors assumed are listed in Table 11.1-15. The basic assumptions and data used in calculating the activities in the pressurizer are listed in Table 11.1-18, and the calculated activities for the two cases are provided in Tables 11.1-19 and 11.1-20. The method used to develop the reactor coolant concentration associated with the updated equilibrium isotopic core inventory discussed in Section 11.1.2 is summarized in Section 15.5.3.2. The updated coolant concentrations with 1 percent fuel defects is presented in Table 11.1-11A. A summary of key assumptions used in developing the coolant concentrations is provided in Table 11.1-2A. 11.1.5 DETERMINATION OF TRITIUM ACTIVITIES IN PRIMARY COOLANT Tritium atoms are generated in the fuel at a rate of approximately 8 x 10-5 atoms per fission, or 1.05 x 10-2 curies/MWt/day. Any boron-bearing control rods in the core are a potential source of tritium. A direct source of tritium is the reaction of neutrons with dissolved boron in the reactor coolant. Boron is used in the reactor coolant for reactivity control. Neutron reactions with lithium are also a direct source of tritium. Lithium hydroxide is used for pH control. Figure 11.1-1 shows calculated versus measured tritium production in the reactor coolant for the R. E. Ginna plant. A 10 percent release from the fuel rods was assumed for the calculation. 11.1-4 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.1.5.1 Ternary Fissions - Cladding Diffusion With zirconium alloy cladding, approximately 10 percent of the tritium produced in the fuel will diffuse through the cladding into the coolant (Reference 11). 11.1.5.2 Tritium Produced from Boron Reactions The neutron reactions with boron that result in the production of tritium are: B10 (n, 2 ) T B10 (n, ) Li7 (n, n ) T B11 (n, T) Be9 B10 (n, d) Be9* (n, ) Li6 (n, )T Of the above reactions, only the first two contribute significantly to the tritium production. The B11 (n, T) Be9 reaction has a threshold of 14 MeV and a cross section of 5 mb. Since the neutrons produced at this energy result in a flux of less than 109 n/cm2-sec, the tritium produced from this reaction is negligible. The B10 (n,d) reaction may be neglected since Be9* has been found to be unstable. 11.1.5.3 Tritium Produced from Lithium Reactions The neutron reactions with lithium resulting in the production of tritium are: Li7 (n, n ) T Li6 (n, ) T Lithium hydroxide is used for pH adjustment of the reactor coolant. Lithium concentrations may reach a value of 6.0 ppm at the beginning-of-cycle. During normal plant operations the lithium is maintained within an operational band per plant procedure. At the end-of-cycle, the lithium concentrations may decrease to 0.0 ppm. This is accomplished by the addition of Li7 OH and by a cation demineralizer included in the chemical and volume control system. This demineralizer will remove any excess of lithium such as could be produced in the B10 (n, ) Li7 reaction. The Li6 (n,) T reaction is controlled by limiting the Li6 impurity in the Li7 OH used in the reactor coolant and by lithiating the demineralizers with 99.9 atom percent Li7. 11.1.5.4 Control Rod Sources In a fixed burnable poison rod, there are two primary sources of tritium generation: the B10 (n, 2 ) T and the B10 (n, ) Li7 (n, n ) T reactions. Unlike the coolant, where the Li7 level is controlled, there is a buildup of Li7 in the burnable poison rods. The burnable poison rods are required during the first year of operation only. During this time, the tritium production is 72 curies/pound B10. 11.1-5 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE The control rod materials used at DCPP are Ag-In-Cd; there are no tritium sources in these materials. 11.1.5.5 Tritium Production from Deuterium Reactions Since the fraction of naturally occurring deuterium in water is less than 0.0015, the tritium produced from this reaction is negligible (less than 1 curie per year). 11.1.5.6 Total Tritium Sources in Coolant A summary of the sources of tritium in the RCS are listed in Table 11.1-22, and all basic data and assumptions used are provided in Table 11.1-21. The calculated total tritium produced in the reactor plant, 1640 curies/year, agrees fairly well with the NUREG-0017, April 1976 value of 0.4 Ci/MWt/year, which gives 1427 curies/year, the difference being conservative. 11.1.6 DETERMINATION OF SECONDARY SYSTEM ACTIVITIES In order to estimate the potential plant releases as a result of secondary system leakage or discharges, during periods when significant activity exists in the steam system, the steam system activity levels have been determined. As discussed earlier, the range of possible combinations of plant operating conditions has been represented by making an activity analysis on the basis of 0.12 percent fuel defects, coincident with a leakage of 100 pounds per day to the secondary system. The projected plant releases for this condition are based on the assumption that these conditions persist throughout the full year. The steam system operating conditions assumed for the activity analysis are listed in Tables 11.1-23, 11.1-24, and 11.1-25, and the results of the analysis are provided in Tables 11.1-26 through 11.1-31. The equations and models for the activity balances and transport calculations are detailed in Reference 8, and they are generally consistent with those provided in NUREG-0017, April 1976. 11.
1.7 REFERENCES
- 1. M.E. Meek and B.F. Rider, Summary of Fission Product Yields for U-235, U-238, Pu-239, and Pu-241 at Thermal, Fission Spectrum, and 14 MeV Neutron Energies, Report Number APED-5398, March 1, 1968.
- 2. J.J. DiNunno, et al, Calculation of Distance Factors for Power and Test Reactor Sites, AEC Report Number TID-14844, March 23, 1962.
- 3. J.O. Blomeke and M.F. Todd, Uranium-235 Fission - Product Production as a Function of Thermal Neutron Flux, Irradiation Time, and Decay Time, AEC Report ORNL-2127, August 19, 1957.
- 4. J.F. Perkins and R.W. King, "Energy Release from the Decay of Fission Products," Nuclear Science and Engineering, 1958.
11.1-6 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE
- 5. Preliminary Safety Analysis Report, Nuclear Units Number 1 & 2, Diablo Canyon Site, Pacific Gas and Electric Company.
- 6. D.F. Toner and J.S. Scott, "Fission Product Release From UO2," Nuclear Safety, Vol. 3, No. 2, December 1961.
- 7. J. Belle, Uranium Dioxide: Properties and Nuclear Applications, Naval Reactors, DRD of USAEC, 1961.
- 8. S.G. Gillespie and W.K. Brunot, EMERALD NORMAL - A Program for the Calculation of Activity Releases and Doses from Normal Operation of a Pressurized Water Plant, Program Description and User's Manual, Pacific Gas and Electric Company, Revision 1, December 1974.
- 9. NUREG-0017, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors, USNRC, (PWR-GALE Code), April 1976.
- 10. A. Tobias, Data from the Calculation of Gamma Radiation Spectra and Beta Heating from Fission Products, (Revision 2), RD/B/M2453, Central Electricity Generating Board, England, October 1972.
- 11. WCAP 8253, Source Term Data for Westinghouse Pressurized Water Reactors, May 1974.
- 12. Reactor Safety Study (WASH-1400), An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants, Appendix VI, U.S. Nuclear Regulatory Commission, October 1975.
11.1-7 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.2 LIQUID RADWASTE SYSTEM The LRS collects and processes the radioactive liquid wastes generated from the primary side of the plant during operation. The LRS reduces the activity of these liquid radwastes to levels acceptable for discharge to the environment. The system is designed to minimize dose to plant personnel and the general public in accordance with applicable regulations. The LRS interfaces with the plants liquid drain systems as follows: (1) The equipment drain subsystem terminates at the inlet to the reactor coolant drain tank (RCDT) in containment or the miscellaneous equipment drain tank (MEDT) in the auxiliary building (refer to Section 9.3.3). The LRS processing equipment includes these tanks and associated downstream process piping and equipment. (2) The floor drain subsystem terminates at the inlet to the containment structure sumps in containment or the auxiliary building and residual heat removal (RHR) system sumps in the auxiliary building (refer to Section 9.3.3). The LRS processing equipment includes these sumps and associated downstream process piping and equipment. The portions of the LRS between and including the LRS containment isolation valves (CIVs) are PG&E Design Class I. The LRS CIVs are supplied by 125-Vdc, Class 1E power. The remainder of the LRS is PG&E Design Class II. Applicable codes and standards for process equipment used in the LRS are presented in Table 3.2-3. The seismic and quality group classifications for these components and associated piping are also described in Section 3.2.2. Three other major liquid radwaste streams that may be radiologically contaminated are included in this chapter, although they are not technically part of the LRS (refer to Section 11.2.3.13.2). 11.2.1 DESIGN BASES 11.2.1.1 General Design Criterion 2, 1967 - Performance Standards The PG&E Design Class I portion of the LRS is designed to withstand the effects of, or is protected against, natural phenomena such as earthquakes, tornadoes, flooding, winds, tsunamis, and other local site effects. 11.2.1.2 General Design Criterion 3, 1971 - Fire Protection The LRS is designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. 11.2-1 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.2.1.3 General Design Criterion 4, 1967 - Sharing of Systems The LRS is not shared by the DCPP units unless it is shown safety is not impaired by the sharing. 11.2.1.4 General Design Criterion 11, 1967 - Control Room The LRS is designed to or contains instrumentation and controls that support actions to maintain the safe operational status of the plant from the control room or from an alternate location if control room access is lost due to fire or other causes. 11.2.1.5 General Design Criterion 12, 1967 - Instrumentation and Control Systems Instrumentation and controls are provided as required to monitor and maintain the LRS variables within prescribed operating ranges. 11.2.1.6 General Design Criterion 17, 1967 - Monitoring Radioactivity Releases The LRS is designed to provide means for monitoring the facility effluent discharge paths, and the facility environs for radioactivity that could be released from normal operations, from anticipated transients, and from accident conditions. 11.2.1.7 General Design Criterion 40, 1967 - Missile Protections The engineered safety feature (ESF) containment isolation portion of the LRS is designed to be protected against dynamic effects and missiles that might result from plant equipment failures. 11.2.1.8 General Design Criterion 49, 1967 - Containment Design Basis The PG&E Design Class I portion of the LRS is designed so that the containment can accommodate, without exceeding the design leakage rate, pressures and temperatures resulting from the largest credible energy release following a loss-of-coolant accident (LOCA), including a considerable margin for effects from metal-water or other chemical reactions that could occur as a consequence of failure of emergency core cooling systems. 11.2.1.9 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment The LRS 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 capabilities to test 11.2-2 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits. 11.2.1.10 General Design Criterion 56, 1971 - Primary Containment Isolation The PG&E Design Class I portion of the LRS contains valves in piping that penetrate containment and that are connected directly to the containment atmosphere. To ensure containment integrity is maintained, each penetration contains one automatic isolation valve or one check valve inside containment and one automatic isolation valve outside containment. 11.2.1.11 General Design Criterion 68, 1967 - Fuel and Waste Storage Radiation Shielding The LRS is designed to provide shielding for radiation protection to meet the requirements of 10 CFR Part 20. 11.2.1.12 General Design Criterion 69, 1967 - Protection Against Radioactivity Release from Spent Fuel and Waste Storage The LRS is designed to provide containment of radioactive releases to the public environs as a result of an accident. 11.2.1.13 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment The LRS includes those means necessary to maintain control over the plant radioactive effluents during normal operation, including anticipated operational occurrences, and during accidents. 11.2.1.14 Liquid Radwaste System Safety Function Requirements (1) Protection from Jet Impingement - Inside Containment The PG&E Design Class I containment isolation portion of the LRS located inside containment is designed to be protected against the effects of jet impingement which may result from high energy pipe rupture. 11.2.1.15 10 CFR Part 20 - Standards for Protection Against Radiation The LRS supports the protection of personnel from radiation sources such that occupational doses and doses to individual members of the public are maintained below the annual limits prescribed in 10 CFR Part 20. 11.2.1.16 10 CFR 50.49 - Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants 11.2-3 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE The LRS components that require environmental qualification (EQ) are qualified to the requirements of 10 CFR 50.49. 11.2.1.17 10 CFR 50.55a(f) - Inservice Testing Requirements The LRS ASME code valves are tested to the requirements of 10 CFR 50.55a(f)(4) and 10 CFR 50.55a(f)(5) to the extent practical. 11.2.1.18 10 CFR 50.55a(g) - Inservice Inspection Requirements The LRS ASME components are inspected to the requirements of 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(g)(5) to the extent practical. 11.2.1.19 10 CFR Part 50, Appendix I - Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as is Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents The LRS is designed to support maintaining offsite annual doses from liquid effluents below the limits specified in 10 CFR Part 50, Appendix I. 11.2.1.20 40 CFR Part 190 - Environmental Radiation Protection Standards for Nuclear Power Operations The LRS supports the protection of members of the public from radiation sources from the uranium fuel cycle such that annual doses are maintained below the limits specified in 40 CFR Part 190. 11.2.1.21 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 LRS provides instrumentation to monitor system variables during and following an accident. 11.2.1.22 Regulatory Guide 1.143, Revision 1, October 1979 - Design Guidance for Radioactive Waste Management Structures, Systems, and Components Installed In Light-Water-Cooled Nuclear Power Plants The LRS design follows the guidance of Regulatory Guide 1.143, Revision 1 as it relates to the quality group classification of components. 11.2.1.23 NUREG-0737 (Item II.F.1), November 1980 - Clarification of TMI Action Plan Requirements 11.2-4 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Item II.F.1 - Additional Accident Monitoring Instrumentation: Position (5) - The LRS provides continuous instrumentation to monitor containment sump wide-range water level in the control room. 11.2.1.24 Generic Letter 96-06, September 1996 - Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions LRS piping has been evaluated for the issue of thermal overpressurization of isolated piping sections that could affect containment integrity during design-basis accident conditions, as described in Generic Letter 96-06, September 1996. 11.2.2 SYSTEM DESCRIPTION 11.2.2.1 General DCPP Unit 1 and Unit 2 share a common LRS, except for equipment located inside containment. A detailed piping and instrumentation schematic of the LRS is shown in Figure 3.2-19. The common LRS consists of the following five collection subsystems: (1) Equipment drain subsystem (refer to Section 9.3.3) (2) Floor drain subsystem (refer to Section 9.3.3) (3) Chemical drain subsystem (refer to Section 11.2.2.1.3) (4) Laundry and hot shower and laundry/distillate subsystem (refer to Section 11.2.2.1.4) (5) Demineralizer regenerant subsystem (refer to Section 11.2.2.1.5) These five collection subsystems are described in the following sections. The floor drain, chemical drain, laundry/distillate, laundry and hot shower, and demineralizer regenerant subsystems generally collect low activity liquid radwastes. The equipment drain subsystem collects liquids with variable activity levels. The demineralizer regenerant subsystem is used as backup for the floor drain and equipment drain subsystems. Following treatment, effluents from the LRS are released to the environment at the discharge structure via either units auxiliary saltwater (ASW) system (refer to Figure 11.2-9). The LRS releases are diluted in the ASW system and main circulating water system flows. Releases require positive operator action, are continuously monitored, and are automatically isolated in the event of a high radiation alarm or a power failure. 11.2-5 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE A major source of liquid radwaste is the RCS. The bulk of these wastes are processed and retained within the CVCS, with a portion being routed to the LRS. A piping and instrumentation schematic of the CVCS is shown in Figure 3.2-8. A complete description of the CVCS is included in Section 9.3.4. 11.2.2.1.1 Equipment Drain or Closed Drain Subsystem Closed drainage from equipment in the containment building flows to the RCDT as described in Section 9.3.3.2.1.1. When the RCDT reaches a preset liquid level, the wastes are automatically pumped to the liquid holdup tanks (LHUTs)for processing in the CVCS. The pumps may also be manually started prior to reaching the preset level. The wastes may also be pumped to the equipment drain receiver tanks or the refueling water storage tanks (RWSTs) when required. An integrating flow meter on the reactor coolant drain pump discharge reads out on the auxiliary building control board digital system, and RCDT high and low level alarms to the main annunciator are provided. Closed drainage from equipment in the auxiliary building is collected in the MEDT as described in Section 9.3.3.2.1.2. When the MEDT reaches a preset liquid level, the wastes are automatically pumped to the equipment drain receiver tanks. A high level alarm function is provided. Closed drain wastes being transferred to the equipment drain receiver tanks are routed to one of the two tanks. When that tank reaches its high level setpoint, incoming flow is automatically diverted to the second equipment drain receiver tank. The filled tanks are normally recirculated, sampled, and analyzed before further batch processing. A high and low level alarm function is provided for each tank. 11.2.2.1.2 Floor Drains and Open Drain Subsystem Inside containment floor drain wastes are collected in the containment sumps and the reactor cavity sump as described in Section 9.3.3.2.2.1. Integrating flow meters in the discharge lines from the reactor cavity sumps and containment sumps are provided to detect leakage from in-containment sources. Potentially contaminated auxiliary building floor drain wastes are collected in the auxiliary building sump as described in Section 9.3.3.2.2.2. The RHR compartments spills collect in the RHR sumps that are normally discharged to the floor drain receivers. 11.2-6 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE When a sump has filled to a preset level, the radwastes are automatically pumped to one of two floor drain receiver tanks. When a tank has reached its high level setpoint, sump flow will automatically be diverted to the second floor drain receiver tank. The filled tank is normally recirculated, sampled, and analyzed before further batch processing. A high level alarm function is provided for each sump. A high level and low level alarm function is provided for each floor drain receiver tank. 11.2.2.1.3 Chemical Drain Subsystem Chemical wastes are generated due to routine chemical and radiochemical sampling and analyses. Chemical wastes from both units drain by gravity to a divided chemical drain tank. The filled section is recirculated, sampled, and analyzed before discharge. 11.2.2.1.4 Laundry and Hot Shower, and Laundry/Distillate Subsystem Laundry, hot shower, and treated CVCS liquids are generally very low in activity. The laundry and hot shower wastes are generated by laundering contaminated protective clothing and by personnel decontamination. CVCS LHUT water is routed to the LRS when reuse in the reactor cavity or spent fuel pool is not possible. The hot shower wastes flow by gravity to one of the laundry and hot shower tanks. When one of the laundry and hot shower tanks is filled, the flow is manually diverted to the second tank. The filled tank is recirculated, sampled, and analyzed before further batch processing or discharge. The laundry waste will normally drain to one of the laundry/distillate tanks for discharge. Treated LHUT water may be drained to one of the laundry/distillate tanks or to one of the demineralizer regenerant receiver tanks for further treatment by the LRS. When one of the laundry/distillate tanks is filled, the flow is manually diverted to the second tank. The filled tank is recirculated, sampled, and analyzed before further batch processing or discharge. 11.2.2.1.5 Demineralizer Regenerant Subsystem The demineralizer regenerant subsystem consists of two 15,000 gallon demineralizer regenerant receivers (arranged in parallel) located adjacent to the equipment drain receivers in the auxiliary building. Originally, it was intended that regeneration wastes from the steam generator blowdown (SGBD) treatment system, deborating demineralizers, or evaporator distillate demineralizers were to be routed to these tanks, and neutralized by concentrated sulfuric acid or sodium hydroxide. The SGBD regeneration system never operated and changes in California environmental regulations halted neutralization of other regenerants in the demineralizer regenerant receivers. 11.2-7 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE The demineralizer regenerant receivers collect equipment or floor drain liquid and function as surge capacity for these subsystems. In addition, treated LHUT liquid can be drained to the demineralizer regenerant receivers for additional processing. 11.2.2.2 Liquid Radwaste System Operation The LRS is operated on a batch basis. When a floor drain receiver, equipment drain receiver, chemical drain, laundry/distillate tank, laundry and hot shower tank, or demineralizer regenerant receiver is filled, it is isolated to prevent accumulation of additional contaminated waste. Control interlocks prevent tanks from being simultaneously filled and discharged. The tanks are normally recirculated, sampled, and analyzed to determine whether additional treatment is required. Batches of equipment and floor drains are normally processed to reduce radioactivity concentrations prior to discharge. 11.2.2.2.1 Liquid Radwaste Processing Sub System Batches that require further treatment are processed through the radwaste media filters, ion exchangers, filters, and/or mobile liquid process systems. The radwaste media filters and ion exchangers are normally operated in series. Mobile liquid process systems may be used to augment LRS in-plant components. The mobile liquid process systems meet the requirements of the associated LRS in-plant components. Treated liquid is collected in the processed waste receivers. In addition, treated LHUT liquid can be routed to the processed waste receivers. These batches are then sampled and analyzed prior to discharge. 11.2.2.2.2 Liquid Radwaste Discharge Sub System Batches that contain sufficiently low quantities of radioactivity to meet discharge limits are treated by filtration and discharged to the outfall of the circulating water system via the ASW discharge. Circulating water flow is verified prior to initiating a release. Written operating procedures govern the mechanics of discharging liquid radwaste to the unrestricted area (Reference 8). As part of the procedures, records of plant water inventories, circulating water flow rates, and radwaste batch analysis data sheets are kept to ensure that the discharges are maintained below the applicable regulatory limits. 11.2.2.3 System Design The components of the LRS are listed in Table 11.2-10. A similar listing for the CVCS is provided in Table 9.3-6. Included are equipment size or capacity, applicable flowrate, material of construction, and design temperature and pressure. 11.2-8 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Applicable codes and standards for process equipment used in the LRS are presented in Table 3.2-3. Equivalent data for the CVCS are shown in Table 9.3-5. The seismic and quality group classifications for these components and associated piping are also described in Section 3.2.2. Radiological monitoring is discussed in detail in Section 11.4. The routing of LRS piping is strictly controlled by the design engineers and is specified on the piping drawings. Consequently, there are field-fabricated lines, but no field-routed lines. Lines that are field-fabricated have the following characteristics: (1) the lines are routed to minimize operator dose, all deviations from the specified routing require prior approval of the piping engineer, and as-built drawings are made showing final dimensions, (2) the sizes, schedules, materials, and code classes are specified on the piping drawings, (3) the field-fabricated piping is similar to shop-fabricated piping in design, quality assurance procedures, and inspection, (4) pipe hanger placement is not specified on piping drawings, but a maximum spacing between hangers is specified as a design standard, and (5) a design review is conducted in accordance with the quality assurance procedures described in Chapter 17 in the same manner as for shop-fabricated piping. 11.2.2.4 Performance Data LRS process equipment is evaluated for its effectiveness in removing radioactivity on a batch basis. Approved plant procedures govern concentrations of activity that must be processed and limits the activity in any individual batch prior to release. When the process equipment removal efficiency no longer produces water to meet these requirements, the filters, media, or ion exchange resin is replaced. 11.2.2.5 Plant Releases 11.2.2.5.1 Current Operational Releases Refer to Section 11.2.3.13.1 for a discussion of current operational releases from the LRS. HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED. 11.2-9 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.2.2.5.2 Pre-Operational Estimated Release Evaluation The following information provides historical perspective demonstrating that DCPP's anticipated operation would result in releases within applicable regulatory limits. This information does not necessarily reflect current operating conditions or practices. The estimated releases based on the original design were calculated with the EMERALD-NORMAL computer program, supplemented by hand calculations. Two process flow diagrams of the LRS are presented to depict the modes of operation for the two cases used for radiation release analysis: the Design Basis Case, and the Normal Operation Case. The flow diagram for the Design Base Case (refer to Figure 11.2-2) shows the waste stream sources and processing route of liquid radwaste for the assumptions of this case. The numbered waste input streams have their annual flow and isotopic spectra listed in Table 11.2-5. The numbered process streams are listed in Table 11.2-8, along with flows and isotopic concentrations. The flow diagram for the Normal Operation Case is shown in Figure 11.2-3. Flows and isotopic parameters for this case are listed in Tables 11.2-5 and 11.2-9. The detailed assumptions used in calculation of estimated activity release from the LRS are listed in Table 11.2-1. Tables 11.2-3 and 11.2-4 list the activity concentration spectrum for the input sources. A tabulation of the estimated annual release by isotope is provided in Tables 11.2-14 and 11.2-15 for the two cases. For the Normal Operation Case with anticipated operational occurrences, as defined in Regulatory Guide 1.112, April 1976 (Reference 7), an additional release of 0.15 Ci/yr per reactor with the same isotopic makeup as in Table 11.2-15 was assumed. This annual release is provided in Table 11.2-18 for one unit. Since an average of 2 days holdup time is available upstream of both the boric acid and waste treatment systems, no additional release, assuming the systems are out of service, was postulated. The detailed assumptions used in the calculation of estimated activity release from the CVCS to the LRS are listed in Table 11.2-2. A tabulation of the isotopic flows through the system components is provided in Tables 11.2-6 and 11.2-7, and the estimated annual releases to the LRS are provided in Tables 11.2-8 and 11.2-9 for the Design Basis and Normal Operation Cases. For purposes of estimating annual average plant radionuclide releases, approximately two-thirds of the boric acid evaporator distillate produced was assumed to be recycled to the primary water storage tank, and the rest routed to the LRS. This release (350,000 gallons/year for each unit) is primarily for tritium control purposes. A list of assumed decay times for system tanks is provided in Table 11.2-13. Demineralizer DFs are listed in Table 11.1-15. During conditions corresponding to the Design Basis Case, which assumes a 1 percent fuel defect level with negligible primary system leakage, the floor drain wastes will have 11.2-10 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE an approximate activity level of 0.0015 mCi/cc. These wastes are normally filtered and released to the main condenser circulating water discharge tunnel. During conditions corresponding to the Normal Operation Case, which assumes a 0.12 percent fuel defect level with primary system leakage, the activity level of the floor drain wastes is approximately 0.1 mCi/cc. Under this operating condition, the wastes would be processed through the filters and/or ion exchangers. The chemical drain wastes would have an approximate activity level of 9 x 10-4 Ci/cc during conditions corresponding to the Design Basis Case. During conditions corresponding to the Normal Operation Case, the wastes would be approximately 1 x 10-4 Ci/cc. During conditions corresponding to the assumptions of Design Basis Case, the approximate activity level in the processed effluent from the equipment drain receiver tanks would be 0.007 Ci/cc. The activity level of the processed effluent would be 0.002 Ci/cc during conditions corresponding to the Normal Operation Case. The SGBD system is depicted on two process flow diagrams that show the two paths of operation of the system and the assumptions used in the radiation release analyses. Figure 11.2-4 shows the discharge path with blowdown directed to the blowdown tank and the circulating water discharge structure. Figure 11.2-5 shows the recycle path with blowdown directed to the blowdown treatment system. For the Design Basis Case, it was assumed that the blowdown is processed via the discharge path because this case assumes no primary-to-secondary leakage and no activity is present in the secondary system. For the Normal Operation Case, it was assumed that the blowdown is processed only via the recycle path and the discharge path is isolated since the assumptions for this case result in significant levels of activity in the secondary system. Table 11.2-16 shows the total annual volumetric flows for one unit and the activity concentrations corresponding to the alphabetically labeled process streams in these two figures for the Normal Operation Case. The estimated annual activity releases from the turbine building sump for one unit are listed in Tables 11.2-14 and 11.2-15 for the Design Basis Case and the Normal Operation Case, respectively. Table 11.2-17 lists the total annual estimated volumes released from the LRS for two units for both cases and includes the turbine building sump discharge. The tritium concentration in the RCS is controlled by bleeding coolant from the RCS to the LRS via the CVCS (refer to Section 9.3.4). Other losses from the RCS are through radioactive decay and mixing of the primary coolant with refueling cavity water during refueling. The calculation of the tritium concentration in the various plant water sources was based on the following assumptions: (1) Tritium activity in the RCS is provided in Table 11.1-22 for the anticipated operational occurrences case. 11.2-11 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE (2) The total volume of water released from the RCS for the analysis of the Design Basis Case and the anticipated operational occurrences case is 350,000 gallons per year per unit. (3) The temperature of the spent fuel pool is constant at 100°F throughout the life of the plant except during refueling when the temperature rises to 125°F. The air above the spent fuel pool is 80°F at 70 percent relative humidity. (4) The water temperature of the refueling tunnel, when filled with borated water from the RWST, is 125°F. The air above the refueling tunnel is 80°F at 70 percent relative humidity. Mixing with 15 percent of the water from the spent fuel pool occurs during each refueling period. (5) During refueling, the containment purge fans are in continuous operation. (6) No water is lost from the spent fuel pool or RWST, except through evaporation. (7) Evaporation from the spent fuel pool is calculated by the equation (Reference 3): V 95.0 0.425 Va ( pw pa ) (11.2-1) A W where: V = loss rate from the water volume, lbm/hr A = exposed area of the water volume, ft2 W = latent heat of vaporization of the water, Btu/lbm Va = velocity of the air across the surface of the water, ft/min pw = vapor pressure of the water, in. of Hg pa = vapor pressure of the water in air, in. of Hg The important parameters for evaluating tritium losses and distribution in the plant are provided in Table 11.2-11. The resulting tritium concentrations in various plant areas are shown in Figures 11.2-6 through 11.2-8. It should be noted that the tritium concentrations plotted in these figures are yearly averages or, in the case of airborne concentrations during refueling, are averages during the refueling periods, based on a 1 year fuel cycle with 11 months operation and 1 month refueling. The tritium management procedures are designed to ensure that tritium airborne concentrations in all normally occupied plant areas are significantly below levels required to ensure compliance with 10 CFR Part 20, Subparts C and D. The restriction of primary coolant tritium concentration by releasing 11.2-12 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE demineralized water from the RCS is intended to reduce in-plant personnel radiation dose. The above analysis was performed to demonstrate that DCPP normal operation radiological effluents meet the criteria of 10 CFR Part 50, Appendix I. This analysis is conservative and bounds DCPP current operation. The analysis was performed prior to plant operation and will be modified only if a design change or operational change rendered it non-conservative. 11.2.2.6 Dilution Factors 11.2.2.6.1 Current Operational Doses The main condenser cooling water and the ASW system of Unit 1 and Unit 2 are used for dilution of released liquid radwastes. The dilution flow available per unit is 862,000 gallons per minute. Refer to Section 10.4.5 for a description of the circulating water system and Section 9.2.7 for a description of the ASW system. HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED. 11.2.2.6.2 Pre-Operational Dose Factors The pre-operational estimated activity concentrations listed in Tables 11.2-21 through 11.2-23 are for one unit's turbine building wastes and LRS diluted with one unit's annual circulating water flow. It should be noted that this method of calculation yields a maximum annual average concentration within the circulating water discharge structure. For the calculation of all internal and external doses from liquid effluent releases, a dilution factor of 5 was used from the point of release to the organism or dose point. This value was taken from Table A-1 of Regulatory Guide 1.109, March 1976 (Reference 5) as a conservative estimate of the dilution at the edge of the initial mixing zone for a high-velocity surface discharge. This factor is considered very conservative; experimental dye studies (Reference 2) have determined that the average dilution factor for fish, invertebrate, and sediment exposure to liquid effluents is 100. The dilution factor of 5 was confirmed by the NRC staff in the Final Environmental Statement for DCPP Unit 1 and Unit 2 (Reference 4). 11.2.2.7 Calculated Doses 11.2.2.7.1 Current Operation Doses Refer to Sections 11.2.3.15, 11.2.3.19, and 11.2.3.20 for discussions of current operational doses from the LRS. HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED. 11.2-13 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.2.2.7.2 Pre-Operation Estimated Doses Pre-operational radiation dose calculations are presented in Tables 11.2-24 through 11.2-26 for the Design Basis Case, the Normal Operation Case, and the Normal Operation Case with anticipated operational occurrences, respectively. These tables list the doses from one unit. The dose via water pathways was calculated with the EMERALD-NORMAL program, which uses liquid dose models and assumptions based on Regulatory Guide 1.109, March 1976. Pre-operational calculations included direct exposures through contact with water by swimming or by exposures in shoreline areas where minute quantities of radioactivity may be deposited. The usage factor for fish and invertebrate consumption and sediment exposure time were taken from Regulatory Guide 1.109, March 1976, as were the bioaccumulation factors for fish, invertebrates, and plants. The complete set of bioaccumulation factors used is listed in Table 11.2-20. A list of effluent concentrations after dilution is provided in Tables 11.2-21 through 11.2-23. The individual doses from fish consumption were based on rockfish caught and eaten by a sport fisherman, with a one day delay between the time of catch and consumption. The individual doses from invertebrate consumption were based on abalone, caught non-commercially and eaten with a one day delay between the time of catch and consumption. The possible doses from swimming or boating were based on the exposure periods listed in Regulatory Guide 1.109, March 1976. A summary of the dose assumptions for liquid pathways exposures is provided in Table 11.2-19. On the basis of the calculated estimates of radiation dose presented in Tables 11.2-24 through 11.2-26, it was concluded that under normal conditions, including the consideration of anticipated operational occurrences, the potential dose from liquid effluents from DCPP Unit 1 and Unit 2 would be well within the dose limits specified in 10 CFR Part 50, Appendix I. 11.2.3 SAFETY EVALUATION 11.2.3.1 General Design Criterion 2, 1967 - Performance Standards The PG&E Design Class I portion of the LRS is designed to perform its safety functions under the effects of earthquakes (refer to Section 3.7), winds and tornadoes (refer to Section 3.3), floods and tsunamis (refer to Section 3.4), and external missiles (refer to Section 3.5). 11.2.3.2 General Design Criterion 3, 1971 - Fire Protection The PG&E Design Class I LRS CIVs are designed to meet the requirements of 10 CFR 50.48(a) and (c) (refer to Section 9.5.1). 11.2-14 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.2.3.3 General Design Criterion 4, 1967 - Sharing of Systems Portions of the LRS outside containment are shared by Unit 1 and Unit 2. None of the LRS performs a safety function for either unit; therefore the sharing does not impair safety on either unit. 11.2.3.4 General Design Criterion 11, 1967 - Control Room The PG&E Design Class I LRS CIVs are provided with controls in the main control room consisting of control switches and position indication on the monitor light panels (refer to Section 6.1.2). If access to the control room is lost, these valves can be monitored and operated locally in the auxiliary building. LRS radiation indication and high radiation alarms are provided in the main control room (refer to Section 11.2.3.6). Signals from the RHR sump levels, RCDT level, containment structure sump levels, reactor cavity sump level, and turbine building sump level are provided to common system annunciator locations on the main control board to indicate abnormal conditions (refer to Section 11.2.3.5). 11.2.3.5 General Design Criterion 12, 1967 - Instrumentation and Control Systems The LRS is controlled from the auxiliary building control board digital system and the main control room panels where the following system instruments provide signals to alarm and warn of abnormal conditions. (1) Containment Structure Sumps and Reactor Cavity Sump (a) Three level switches are provided; one switch for high level alarm to the main annunciator and two switches that are spared in place (b) Level indicators are provided for both containment sumps and for the reactor cavity sump (c) An integrating flow meter on the containment sump pumps discharge line (2) RCDT (a) Level switches to initiate automatic pump starts on high level (b) Pump trips for low level (c) High and low level alarms to the main annunciator 11.2-15 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE (d) Local and remote level indication and a level controller to maintain a set level (e) An integrating flow meter on the RCDT pumps discharge line (3) MEDT and Auxiliary Building Sump (a) Level indicators are provided for both the MEDT and the auxiliary building sump (4) Equipment Drain Receivers and Floor Drain Receivers (a) Level switches for high and low level alarms to the main annunciator (b) Pump trips for low level (c) High level switches to actuate three-way control valves to transfer liquids from one tank to another (d) Level indicators to provide local and remote indication (5) Chemical Drain and Laundry Drain Tanks (a) Level switches for high and low alarm to the main annunciator (b) Level switches for transfer control (c) Level indicators for local and remote indication Refer to Section 11.2.3.6 for radiation monitoring instrumentation. 11.2.3.6 General Design Criterion 17, 1967 - Monitoring Radioactivity Releases Batch preparation and sampling is used to document LRS discharge to the environment. In addition, instrumentation is provided in the control room to monitor LRS radioactive releases. The liquid radwaste effluent monitor (R-18) continuously monitors discharges from the LRS. A high-radiation signal initiates automatic valve closure and flow is diverted to the equipment drain receiver tanks. The SGBD sample monitor (R-19) monitors the liquid phase of the secondary side of the SG for radioactivity. A high-radiation signal causes the isolation valves in the blowdown and sample lines, acting with the valve in the line from the blowdown tank to 11.2-16 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE the discharge structure, to close and the blowdown tank liquid effluent to be diverted to the equipment drain receiver tank. The SGBD tank liquid effluent monitor (R-23) continuously measures liquid effluent from the SGBD tank. A high-radiation signal isolates the blowdown discharge, and diverts blowdown tank liquid effluent to the equipment drain receiver tank. Refer to Section 11.4.2.1.2.1 and Table 11.4-1 for additional information. 11.2.3.7 General Design Criterion 40, 1967 - Missile Protection The provisions taken to protect the ESF containment isolation portion of the LRS from damage that might result from missiles and dynamic effects associated with equipment and high-energy pipe failures, respectively are discussed in Sections 3.5, 3.6, and 6.2.4. 11.2.3.8 General Design Criterion 49, 1967 - Containment Design Basis The LRS containment penetrations, 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 containment design leakage rates (refer to Sections 3.8.2.1.1.3 and 11.2.4). 11.2.3.9 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment The LRS CIVs required for containment closure are periodically tested for operability. Testing of the components required for the containment isolation system is discussed in Section 6.2.4. 11.2.3.10 General Design Criterion 56, 1971 - Primary Containment Isolation The PG&E Design Class I containment penetrations for the LRS are penetration group A or E (refer to Section 6.2.4). A description of the isolation valves and piping for each penetration is provided in Table 6.2-39. Group A and group E piping comply with the requirements of GDC 56, 1971. 11.2.3.11 General Design Criterion 68, 1967 - Fuel and Waste Storage Radiation Shielding Radiation shielding in the auxiliary building protects personnel working near waste storage facilities from doses in excess of 10 CFR Part 20 limits (refer to Section 12.1.7.3). 11.2.3.12 General Design Criterion 69, 1967 - Protection Against Radioactivity Release from Spent Fuel and Waste Storage 11.2-17 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE The PG&E Design Class I LRS CIVs are designed to ensure that the containment atmosphere will be isolated in the event of a release of radioactive material inside containment. The isolation time limits specified ensure that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses of a LOCA (refer to Section 6.2.4). In accordance with UFSAR Section 3.2.2.1.2, fluid systems and fluid system components that contain or may contain radioactive material, but whose failure would not result in calculated potential exposures in excess of 0.5 rem whole body (or its equivalent to parts of the body) at the site boundary, may be classified as PG&E Design Class II. As part of the acceptability of tanks that contain liquid waste that have a PG&E Design Class II classification, an evaluation was performed of the dose consequences at the site boundary of a Liquid Holdup Tank (LHUT) failure accident. Radioactive liquid waste holdup tanks are used as part of the chemical and volume control system (CVCS) to collect and permit decay of radioactive liquids drawn from the reactor primary coolant for reactivity control. The CVCS is described in detail in Section 9.3.4. Five liquid holdup tanks are provided for the two units to afford operating flexibility and allow one or more tanks to be isolated from the rest of the system for extended periods of time. The liquid processed through the holdup tanks contains dissolved fission and activation products, as well as radioactive noble gases mixed with nitrogen cover gas used in the tanks. The liquid holdup tanks are located in vaults which are Design Class I structures, so that in the event of a rupture or spill all liquids are retained in the vaults. The volume of holdup tank vaults is sufficient to contain the full contents of the holdup tank without spillage from the vaults. Any gases released from the liquid holdup tanks are collected by the auxiliary building ventilation system and discharged via the auxiliary building vent. In the evaluation of the liquid waste holdup tank rupture accident, the following fission product accumulation and release assumptions are used for the design basis case. (1) The reactor has been operating at full power with 1 percent defective fuel for an equilibrium core cycle. (2) A liquid holdup tank has been filled with primary coolant at a rate of 132 gpm, with credit for decay as the tank is filling. (3) The failure occurs immediately upon completion of the liquid transfer, releasing the entire contents of the tank to the LHUT vault. The assumption of the release of the contents of only a single tank is based on a design that 11.2-18 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE allows all liquid holdup tanks to be isolated from each other when they are in use. (4) All of the noble gases and varying amounts of the iodines are released, unfiltered, from the LHUT vault to the environment, via the plant vent. The release is treated as a ground level release. (5) All of the noble gases are exhausted to the environment over a 2-hour time period, or at the rate of the LHUT ventilation exhaust flow, whichever results in a more rapid release rate. (6) The iodine concentration in the LHUT vault air space and ventilation exhaust flow is based on an equilibrium iodine distribution balance between the spilled LHUT liquid and the air, and is dependent on the temperature, pH, and concentration of iodine in the spilled fluid. The iodines are released to the environment as they evolve out of the spilled liquid at the rate of the LHUT vault ventilation exhaust flow over a 30-day period. No liquids escape from the vaults during the accident. Computer code RADTRAD 3.03 (Reference 9) is used to calculate the whole body dose resulting from the LHUT rupture using the semi-infinite cloud submersion model, and using air submersion dose coefficients provided in Table III.1 of Federal Guidance Report 12 (Reference 10) and inhalation dose conversion factors provided in Table 2.1 of Federal Guidance Report 11 (Reference 11). The reactor coolant isotopic concentrations based on 1% fuel defects are provided in Table 11.1-11A. Atmospheric dispersion factors used in the analysis are given in Table 2.3-145. The dose consequences at both the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) due to airborne releases following a LHUT Rupture are estimated to be below the acceptance criteria of 0.5 rem whole body and 3 rem thyroid. A similar assessment was also performed for the Volume Control Tank (VCT) rupture. Though the VCT tank is PG&E Design Class I, the dose consequence analysis for the VCT tank rupture conservatively uses an acceptance criteria, the commitment provided in DCPP UFSAR Section 3.2.2.1.2 for acceptability of dose consequences at the site boundary for a failure in a PG&E Design Class II system. The VCT is used as part of the CVCS to collect the excess water released from the RCS during modes 1 through 6, which is not accommodated by the pressurizer. For a complete description of the CVCS and the VCT in all modes of operation refer to Section 9.3.4. The liquid processed through the volume control tank contains dissolved fission and activation products, as well as undissolved radioactive noble gases. A spray nozzle located inside the tank on the inlet line strips part of the nobles gases from the incoming liquid, and these gases are retained in the VCT vapor space. In addition, an 11.2-19 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE overpressure of hydrogen cover gas is provided for the tank to control the hydrogen concentration in the reactor coolant. The VCT is located in a vault which is a PG&E Design Class I structure, so that in the event of a rupture or spill all liquids are retained in the vault. The volume of the tank vault is sufficient to contain the full contents of the tank without spillage from the vault. Any gases released from the VCT are collected by the auxiliary building ventilation system and discharged via the auxiliary building vent. In the evaluation of the VCT rupture accident, the following fission product accumulation and release assumptions are used for the design basis case: (1) The reactor has been operating at full power with 1 percent defective fuel for an equilibrium core cycle. (2) The volume control tank contains its maximum equilibrium inventory of radioactivity at the time of the accident. The failure of the tank releases the entire tank contents to the VCT vault. (3) All of the noble gases and varying amounts of the iodines are released, unfiltered, from the VCT vault to the environment, via the plant vent. The release is treated as a ground level release. (4) All of the noble gases are exhausted to the environment over a 2-hour time period, or at the rate of the VCT ventilation exhaust flow, whichever results in a more rapid release rate. (5) The iodine concentration in the VCT vault air space and ventilation exhaust flow is based on an equilibrium iodine distribution balance between the spilled VCT liquid and the air, and is dependent on the temperature, pH, and concentration of iodine in the spilled fluid. The iodines are released to the environment as they evolve out of the spilled liquid at the rate of the VCT vault ventilation exhaust flow over a 30-day period. No liquids escape from the vaults during the accident. Computer code RADTRAD 3.03 (Reference 9) is used to calculate the whole body dose resulting from the VCT rupture using the semi-infinite cloud submersion model, and using air submersion dose coefficients provided in Table III.1 of Federal Guidance Report 12 (Reference 10) and inhalation dose conversion factors provided in Table 2.1 of Federal Guidance Report 11 (Reference 11). The reactor coolant isotopic concentrations based on 1% fuel defects are provided in Table 11.1-11A. Atmospheric dispersion factors used in the analysis are given in Table 2.3-145. The dose consequences at both the EAB and the LPZ due to airborne releases following a VCT Rupture are estimated to be below the acceptance criteria of 0.5 rem whole body and 3 rem thyroid. 11.2-20 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.2.3.13 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment 11.2.3.13.1 Current Operational Releases The actual releases from the plant site are summarized in the Annual Radioactive Effluent Release Report to the NRC. The report summarizes the liquid, gaseous, and solid radwastes that were released from the site for the past year and provides the calculated dose to the public, which is calculated using the offsite dose calculation manual (ODCM). Releases from DCPP have routinely demonstrated compliance within the general provisions of, and dose limits established in, 10 CFR Part 20 and 10 CFR Part 50. 11.2.3.13.2 Release Points The four major release pathways are as follows: (1) from the LRS to the discharge structure via either the Unit 1 or Unit 2 ASW discharge lines (refer to Figures 11.2-2, 11.2-3, and 11.2-9); (2) from the Unit 1 or Unit 2 SGBD: (a) via the SGBD tanks to the discharge structure via the respective unit's discharge conduit (refer to Figures 11.2-4 and 11.2-9); (b) via diversion from the SGBD recycle line to the main condenser circulating water discharge. (3) from the turbine building sump to the discharge structure via either the Unit 1 or Unit 2 turbine-generator building sump discharge line (refer to Figure 11.2-5). (4) from the condensate demineralizer regenerant system via either the high conductivity tank (HCT) or the low conductivity tank (LCT). Other minor discharge pathways exist. Those identified pathways, monitored radiologically, are included in the ODCM and/or implementing procedures. The LRS is described in Section 11.2.2. The other three main discharge pathways are described in Sections 11.2.3.13.2.1 through 11.2.3.13.2.3 below. 11.2.3.13.2.1 Turbine Building Drain System The concentration of radioactivity in the turbine building drains is expected to be low, even in the event of significant primary-to-secondary SG leakage. The radiation level and flow of liquid from the turbine building drains are monitored at the oily water 11.2-21 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE separator to verify that there are no unaccounted for or unexpected releases from the turbine building drains. If significant radioactivity is detected coming from the turbine building drains, the discharge can be routed for evaluation and disposition. The monitoring system is in conformance with Regulatory Guide 1.21, Revision 1 (Reference 8) and General Design Criterion 4, 1967. Turbine building sump wastes are normally released to the environment via each unit's circulating water discharge structure (refer to Figure 11.2-9). A detailed piping and instrumentation schematic of the turbine building sump systems is shown in Figure 3.2-27. 11.2.3.13.2.2 Steam Generator Blowdown System The SGBD system for each unit provides two processing paths. One path discharges blowdown flow to the environment via the SGBD tank and the circulating water discharge tunnel. The other path recycles blowdown flow to the main condenser circulating water discharge via the blowdown treatment system and/or the blowdown treatment bypass line. The recycle path can discharge a portion of blowdown flow to the discharge tunnel. Blowdown flow for each unit can be directed to either blowdown path alone, or to both paths simultaneously. During plant operation, SG water collects suspended and dissolved solids that are brought in by the feedwater. Water must be blown down from the SGs to maintain low solids concentration for efficient operation (Reference 1). The blowdown flow may be directed to either or both the discharge and recycle paths for a total blowdown flow of approximately 400 gpm per unit to maintain water chemistry and low solids concentration. If activity is detected at preset levels in the blowdown system, the blowdown will be automatically isolated from the discharge path and blowdown flow will be limited to the recycle path at up to 150 gpm per unit. Activity in the blowdown that is discharged to the circulating water tunnel is continuously recorded by R-23. In the event that activity is released via the SGBD tank vent, the quantity is calculated based on the continuous readings recorded by R-19. Activity released to the environment is calculated as described in the ODCM. The SGBD system in the recycle configuration is typically used to recover blowdown for return to the condensate system. The normal flow path for blowdown recycling is through the flash tank to the heat exchanger and to the main condenser. A prefilter and a mixed bed demineralizer, which are normally bypassed, are available as part of the system for use in the event of a primary-to-secondary leak. When used, the prefilter removes both radioactive and non-radioactive suspended solids and the mixed bed demineralizer removes both radioactive and non-radioactive dissolved anions and cations. Effluent from the filter and demineralizer set could be returned through a resin trap filter to the condensate. The mixed bed resin is replaced when exhausted. 11.2-22 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE A description of the design bases, on-line monitoring, isolation criteria, design codes, system description, safety evaluation, inspection, and testing of the SGBD system is provided in Section 10.4.8. 11.2.3.13.2.3 Condensate Demineralizer Regenerant Solution The condensate demineralizer regenerant solution is the waste created from regenerating anion and cation resin used to clean water in the main condenser. The resin is regenerated with sulfuric acid and sodium hydroxide. The HCT normally receives the initial rinse water that may contain most of the acid, caustic, and contaminants removed from the resin. The LCT normally receives cleaner water used as a final rinse of the resin. The concentration of radioactivity in both the HCT and the LCT is typically not detected, unless blowdown is isolated or if a primary-to-secondary leak is present. The HCT is discharged as a discrete batch. The tank pH is adjusted within an acceptable range consistent with national pollutant discharge elimination system (NPDES) and other state requirements. The tank is then recirculated, sampled and analyzed prior to issuing a discharge permit. The LCT is normally discharged as a discrete batch. The tank is recirculated, sampled and analyzed prior to issuing a discharge permit. In some cases, the LCT is allowed to discharge as a continuous discharge for a period of time. In this configuration, more than one sample is taken during the discharge period to ensure that the discharge remains within limits. 11.2.3.13.2.4 Typical Volumes Released The following tables provide an approximation of routine plant discharge volumes. These are not limiting volumes. Volumes can change as defined in the site procedures as long as 10 CFR Part 20, and 10 CFR Part 50, Appendix I criteria are maintained. Typical LRS Annual Release Volumes: Source UFSAR Section Nominal Quantity (gallons) Equipment Drains 11.2.2.1.1 400,000 to 500,000 Floor Drains 11.2.2.1.2 200,000 to 400,000 Treated CVCS 11.2.2.1; 470,000 to 750,000 Table 11.2-1 #27 Laundry & Hot 11.2.2.1.4 50,000 to 250,000 Shower Drains Chemical Drains 11.2.2.1.3 13,000 to 30,000 Sub Total 1.1 M to 1.9 M 11.2-23 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Typical Plant Annual Release Volumes: Pathways Nominal Quantity (gallons) LRS 1.1 M to 1.9 M U-1 SGBD 30M U-2 SGBD 30M Turbine Bldg Sump 15M U-1 Condensate Demineralizer Regenerant Tanks 5M U-2 Condensate Demineralizer Regenerant Tanks 5M Only trace amounts of activity are in the pathways that are not LRS. 11.2.3.14 Liquid Radwaste System Safety Function Requirements (1) Protection from Jet Impingement - Inside Containment The provisions taken to provide protection of the inside containment PG&E Design Class I portion of the LRS from the effects of jet impingement which may result from high energy pipe rupture are discussed in Section 3.6. 11.2.3.15 10 CFR Part 20 - Standards for Protection Against Radiation Occupational doses to plant personnel from the LRS are maintained as low as is reasonably achievable (ALARA) by: (1) Shielding in locations that contain LRS piping and equipment (refer to Section 12.1.2) (2) Heating, ventilation and air-conditioning (HVAC) systems that direct air flows for controlled discharge and provide filtration of radioactive materials (refer to Section 12.2.2) (3) Monitoring and alarming of potentially radioactive releases from the LRS (refer to Section 11.2.3.6) (4) Administrative controls for working in areas of potentially high radioactivity (refer to Section 17.5). Current operational offsite doses are calculated in accordance with the ODCM contained in site-approved procedures in accordance with 10 CFR Part 20 and appropriate U.S. Nuclear Regulatory Commission (NRC) regulatory guides. The only significant pathways to man from liquid radwastes are through food chains involving marine organisms. Liquid releases have been examined for the possible 11.2-24 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE effects on man and on aquatic organisms. For man, the pathways considered are the intake of aquatic foods that were grown within the radiological influence of the plant. Technical Specifications require programs for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents within the regulatory limits of 10 CFR Part 20. Doses are reported on quarterly and annual bases. The programs are implemented by plant procedures, and include remedial actions to be taken whenever the program limits are exceeded. 11.2.3.16 10 CFR 50.49 - Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants The LRS components required to function in harsh environments under accident conditions are qualified to the applicable environmental conditions to ensure that they perform their safety functions. These components include solenoid and air operated valves for the CIVs, limit switches for the CIVs, sump level transmitters, and temperature elements and are listed on the EQ Master List. Section 3.11 describes the DCPP EQ Program and the requirements for the environmental design of electrical and related mechanical equipment. 11.2.3.17 10 CFR 50.55a(f) - Inservice Testing Requirements The inservice testing (IST) requirements for the LRS are contained in the IST Program Plan. 11.2.3.18 10 CFR 50.55a(g) - Inservice Inspection Requirements The inservice inspection (ISI) requirements for the LRS are contained in the DCPP ISI Program Plan. 11.2.3.19 10 CFR Part 50, Appendix I - Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as is Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents The LRS maintains control over the releases of radioactive materials in effluents by: (1) Providing appropriate holdup capacity for retention of liquid effluents containing radioactive material (refer to Section 11.2.2.5.2) (2) Continuously monitoring the LRS for leakage (refer to Section 11.2.2.1) (3) Sampling all discharges prior to release to the environment and using acceptable dilution factors (refer to Section 11.2.2.1) 11.2-25 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE (4) Processing batches that require further treatment through the radwaste media filters, ion exchangers, filters, and/or mobile liquid process systems (refer to Section 11.2.2.2.1) Technical Specifications require programs for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents within the regulatory limits of 10 CFR Part 50, Appendix I. Doses are reported on quarterly and annual bases. The programs are implemented by plant procedures, and include remedial actions to be taken whenever the program limits are exceeded. 11.2.3.20 40 CFR Part 190 - Environmental Radiation Protection Standards for Nuclear Power Operations Current operational doses from all sources to members of the public are summarized in Annual Radioactive Effluent Release Reports and are demonstrated to be within the limits specified by 40 CFR Part 190, as implemented under 10 CFR 20.1301(e) (refer to Section 11.2.3.13.1). 11.2.3.21 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 LRS post-accident instrumentation for meeting Regulatory Guide 1.97, Revision 3, requirements consist of CIV position and containment sump wide-range water level (refer to Table 7.5-6). 11.2.3.22 Regulatory Guide 1.143, Revision 1, October 1979 - Design Guidance for Radioactive Waste Management Structures, Systems, and Components Installed In Light-Water-Cooled Nuclear Power Plants The LRS is PG&E Design Class II with the exception of the containment penetrations and associated valves which are PG&E Design Class I (refer to Section 11.2). The PG&E Design Class II LRS equipment and piping that are applicable to Regulatory Guide 1.143, Revision 1 are designated PG&E QA Class R, PG&E Piping Symbol H. 11.2.3.23 NUREG-0737 (Item II.F.1), November 1980 - Clarification of TMI Action Plan Requirements Item II.F.1 - Additional Accident Monitoring Instrumentation: Position (5) - Continuous instrumentation to monitor containment sump wide-range water level is provided in the control room (refer to Section 7.5.2.1.3 and Figures 7.5-1 and 7.5-1B). Instrument ranges and accuracies are provided in Table 7.5-4. 11.2-26 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.2.3.24 Generic Letter 96-06, September 1996 - Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions Valves FCV-500 and FCV-253, the inboard CIVs for penetrations 49 and 50, respectively, are installed with a hole drilled through the ball. These ball valves have a hollow ball that rotates between two valve seats. When the valve sees pressure, the ball is pushed into the valve away from the first valve seat and forms a tight seal against the second valve seat. The hole drilled through the ball allows for unidirectional pressure relief (one end of the hole sees the environment inside the pipe while the other end of the hole sees the inside of the valve body). This ensures that the valves meet the requirements of Generic Letter 96-06, September 1996. 11.2.4 TESTS AND INSPECTIONS Tests and inspections of the LRS are done in accordance with plant procedures. 11.2.5 INSTRUMENTATION APPLICATIONS Refer to Section 11.2.3.5 for a discussion of the instrumentation requirements and controls for the LRS. 11.
2.6 REFERENCES
- 1. A.B. Sisson, et al., Evaluation for Removal of Radionuclides from PWR Steam Generator Blowdowns, International Water Conference, November 2, 1971.
- 2. Preliminary Safety Analysis Report, Nuclear Unit Number 2, Diablo Canyon Site, Pacific Gas and Electric Company, Docket Number 50-323.
- 3. Handbook of Fan Engineering, 6th Ed., Buffalo Forge Company.
- 4. Diablo Canyon Final Environmental Statement, U.S. Atomic Energy Commission, May 1973, Docket Numbers 50-275 and 50-323.
- 5. Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, USNRC, March 1976.
- 6. EMERALD-NORMAL, a Program for the Calculation of Activity Releases and Potential Doses from the Normal Operation of a Pressurized Water Reactor Plant, Revision 2, Pacific Gas and Electric Company, July 1976.
11.2-27 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE
- 7. Regulatory Guide 1.112, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Light-Water Cooled Power Reactors, USNRC, April 1976.
- 8. Regulatory Guide 1.21, Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants, Revision 1, USNRC, June 1974.
- 9. RADTRAD 3.03 (GUI Mode Version), A Simplified Model for RADionuclide Transport and Removal And Dose Estimation, NUREG/CR-6604, Users Guide
- Supplement 2, October 2002.
- 10. K.F. Eckerman and J.C. Ryman, External Exposure to Radionuclides in Air, Water, and Soil, Federal Guidance Report 12, EPA-402-R-93-081, Environmental Protection Agency, 1993.
- 11. K.F. Eckerman et. al., Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, Federal Guidance Report 11, EPA-520/1-88-020, Environmental Protection Agency, 1988.
11.
2.7 REFERENCES
DRAWINGS Figures representing controlled engineering drawings are incorporated by reference and are identified in Table 1.6-1. The contents of the drawings are controlled by DCPP procedures. 11.2-28 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.3 GASEOUS RADWASTE SYSTEM The GRS provides controlled handling and disposal of gaseous radwastes generated during plant operation. The system is designed to minimize dose to plant personnel and the general public in accordance with applicable regulations. The GRS, including the gas analyzer package, is designed and fabricated as PG&E Design Class II. The equipment and piping code classifications are listed in Table 3.2-3. 11.3.1 DESIGN BASES 11.3.1.1 General Design Criterion 2, 1967 - Performance Standards The GRS is designed to withstand the effects of, or is protected against, the Design Earthquake (DE). 11.3.1.2 General Design Criterion 3, 1971 - Fire Protection The GRS is designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. 11.3.1.3 General Design Criterion 4, 1967 - Sharing of Systems The GRS is not shared by the DCPP units unless it is shown safety is not impaired by the sharing. 11.3.1.4 General Design Criterion 11, 1967 - Control Room The GRS is designed to or contains instrumentation and controls that support actions to maintain the safe operational status of the plant from the control room or from an alternate location if control room access is lost due to fire or other causes. 11.3.1.5 General Design Criterion 12, 1967 - Instrumentation and Control Systems Instrumentation and controls are provided as required to monitor and maintain the GRS variables within prescribed operating ranges. 11.3.1.6 General Design Criterion 17, 1967 - Monitoring Radioactivity Releases The GRS is designed to provide means for monitoring the facility effluent discharge paths and the facility environs for radioactivity that could be released from normal operations, from anticipated transients, and from accident conditions. 11.3-1 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.3.1.7 General Design Criterion 18, 1967 - Monitoring Fuel and Waste Storage The GRS is provided with monitoring and alarm instrumentation for conditions that might contribute to radiation exposures. 11.3.1.8 General Design Criterion 68, 1967 - Fuel and Waste Storage Radiation Shielding The GRS is designed to provide shielding for radiation protection to meet the requirements of 10 CFR Part 20. 11.3.1.9 General Design Criterion 69, 1967 - Protection Against Radioactivity Release from Spent Fuel and Waste Storage The GRS is designed to provide containment of radioactive releases to the public environs as a result of an accident. 11.3.1.10 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment The GRS includes those means necessary to maintain control over the plant gaseous radioactive effluents during normal operation, including anticipated operational occurrences. 11.3.1.11 10 CFR Part 20 - Standards for Protection Against Radiation The GRS supports the protection of personnel from radiation sources such that occupational doses and doses to individual members of the public are maintained below the annual limits prescribed in 10 CFR Part 20. 11.3.1.12 10 CFR Part 50 Appendix I - Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as is Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents The GRS is designed to support maintaining offsite annual doses from gaseous effluents below the limits specified in 10 CFR Part 50 Appendix I. In the event that half of those radiation dose levels are exceeded in any calendar quarter, DCPP initiates investigations into the cause and reports corrective actions within 30 days from the end of the quarter to the NRC. 11.3.1.13 40 CFR Part 190 - Environmental Radiation Protection Standards for Nuclear Power Operations The GRS supports the protection of members of the public from radiation sources from 11.3-2 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE the uranium fuel cycle such that annual doses are maintained below the limits specified in 40 CFR Part 190. 11.3.1.14 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 GRS provides instrumentation to monitor system variables during and following an accident. 11.3.1.15 Regulatory Guide 1.143, Revision 1, October 1979 - Design Guidance for Radioactive Waste Management Structures, Systems, and Components Installed In Light-Water-Cooled Nuclear Power Plants The GRS design follows the guidance of Regulatory Guide 1.143, Revision 1 as it relates to the seismic design and quality group classification of components. 11.3.1.16 NUREG-0737 (Items II.F.1 and III.D.1.1), November 1980 - Clarification of TMI Action Plan Requirements Item II.F.1 - Additional Accident Monitoring Instrumentation: Position (1) - The GRS is designed with noble gas effluent monitors that are installed with an extended range (ER) designed to function during accident conditions. Position (2) - The GRS is designed with provisions for sampling of plant effluents for post-accident releases of radioactive iodines and particulates and onsite laboratory capabilities. 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 GRS are periodically pressure leak tested and visually inspected for leakage into the building environment. 11.3.2 SYSTEM DESCRIPTION The GRS is designed to collect, store, monitor, and release plant waste gas streams that may contain radioactive noble gases, iodine, and radioactive particulate material. In addition to releases from the GRS, other releases of radioactive gases from the plant are possible under some operating conditions; for example, in periods during which primary coolant system leakage occurs. These additional pathways that can, under some plant operating conditions, release radioactive gases are: (1) Containment purge (refer to Section 9.4.5) 11.3-3 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE (2) Auxiliary building ventilation (refer to Section 9.4.2) (3) Fuel handling building ventilation (refer to Section 9.4.4) (4) Main condenser air ejector (refer to Section 10.4.1) (5) Gland steam condenser (refer to Section 10.4.3) (6) SGBD system (refer to Section 10.4.8) (7) 10 percent atmospheric dump valves (refer to Section 10.4.4) (8) Main steam and reheat relief valves (refer to Sections 10.2 and 10.3) (9) Solid radwaste storage building (RSB) ventilation (refer to Section 11.5) (10) Laundry facility ventilation (refer to Section 11.3.3.10) (11) Condenser vacuum pump (refer to Section 10.4.2) The GRS is designed to process radioactive gases consisting primarily of nitrogen and hydrogen with low levels of oxygen. The gases are collected by the vent header system from various primary and auxiliary systems. Radioactive or potentially radioactive gaseous wastes result from collection of excess cover gas in the LHUTs, degasification in the volume control tank, and cover gas displaced in the pressurizer relief tank and RCDT. A piping schematic for the GRS is shown in Figure 3.2-24. Each unit has a vent header network and surge tank. Gases collected by the vent header system are routed to the surge tank, which is equipped with a safety relief valve to prevent overpressurization. Each unit's surge tank feeds that unit's waste gas compressor and/or a shared spare compressor through a pressure control valve set to maintain constant compressor suction pressure. The system is designed such that the shared spare compressor will automatically start if the pressure in the surge tank rises above 3 psig. An oxygen monitor on the moisture separator discharge limits the concentration of oxygen that can be fed to the gas decay tanks. The monitor actuates an alarm at 2 percent oxygen concentration and trips the compressors at 4 percent. The compressor discharge is routed into a network of valves feeding the gas decay tanks. Based on downstream conditions and operator selection, the system controls the positioning of these valves. One tank will be filling, with one tank on standby, and one tank being used for cover gas. The system will automatically switch to the standby tank when the fill tank reaches 100 psig. Should this fail to happen, the system will alarm when the tank reaches 105 psig. 11.3-4 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE The gas decay tanks are provided for the holdup of radioactive gases prior to release to the environment. The size and number of gas decay tanks was selected to provide capability to hold gases for 45 days when necessary (to allow decay). The holdup time required is that which would result in releases that are in compliance with release rate and dose limits. Each gas decay tank may be operated as a cover gas supply for the LHUTs. Normal coolant letdown then displaces the gases back into the GRS. This process effectively increases the volume of storage available for gaseous holdup. The sampling system associated with the GRS is used to monitor oxygen content of the gases in the system and to collect grab samples for oxygen concentration analysis. Sample points exist in this system including all influent sources and each of the gas decay tanks. These sample points may be monitored via grab samples which may be taken from manual sample taps. The gas analyzer is equipped with a sample tap for taking bottled samples to undergo radiological testing. 11.3.2.1 Gaseous Radwaste System Operation The GRS handles gaseous discharge from the various sources. The volume of gases originates from displaced cover gas in the LHUTs. The gaseous flow is discharged to the vent header and routed to the surge tank. The suction pressure to the normally operating compressor is maintained at approximately 0.5 psig via a pressure regulator on the surge tank outlet. During normal operation, the waste gas compressor starts and stops based on surge tank pressure to maintain a nominal 1.1 psig to 2.0 psig in the surge tank. If a control malfunction occurs causing the compressor to continuously operate, a recycle valve would open at a nominal 0.9 psig pressure in the surge tank, returning gases from the moisture separator. This supplies gases back to the compressor suction preventing evacuation of the compressor suction piping. In addition, if the LHUT header pressure is at or below the setpoint (nominally 0.75 psig), a recycle valve from the gas decay tanks opens to maintain the cover gases in the LHUTs. A shared Unit 1 and Unit 2 spare compressor will start on a high surge tank pressure of a nominal 3 psig. The waste gas compressor discharges into a gas decay tank through a series of automatic valves. When the decay tank is filled to a nominal 100 psig, the inlet valve to the filled tank closes and the inlet valve to a standby tank opens. Should this fail to happen, a high pressure alarm sounds at a nominal 105 psig. The remaining tank is now positioned as the standby tank with the filled tank isolated for decay and release. 11.3.2.2 System Design The GRS, including the gas analyzer package, is designed and fabricated as PG&E Design Class II. The equipment and piping code classifications are listed in Table 3.2-
- 3. The design and operating parameters for the GRS equipment are shown in Table 11.3-1.
11.3-5 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.3.2.3 Plant Releases 11.3.2.3.1 Current Operational Releases Refer to Section 11.3.3.10 for a discussion of current operational releases from the GRS. 11.3.2.3.2 Pre-Operational Estimated Release Evaluation HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED. The following information provides historical perspective demonstrating that DCPP's anticipated operation would result in releases within applicable regulatory limits. This information does not necessarily reflect current operating conditions or practices. The potential release pathways for radioactive gases have been described in previous paragraphs. Table 11.3-2 lists the estimated annual releases for the Design Basis Case, and Table 11.3-3 lists the estimated annual releases for the Normal Operation Case. The assessments of total curies released via each of the various pathways are discussed below. All release calculations were performed using the EMERALD-NORMAL (Reference 1) computer code. Table 11.3-4 lists the estimated annual gaseous radwaste flows. Tables 11.3-5 through 11.3-8 list the estimated activities in the gas decay tanks and the volume control tank for the Design Basis and Normal Operation Cases. For the release from venting of the gas decay tanks, it was assumed that the full volume of primary coolant is degassed twice a year, and the quantities of radioactive gases released are listed in Table 11.3-2 for the Design Basis Case and in Table 11.3-3 for the Normal Operation Case. Containment venting releases are based on an assumed leakage of 1 percent of the primary coolant noble gas inventory and 0.001 percent of the primary coolant iodine inventory per day to the containment. It was assumed that the containment air is purged 24 times a year, and that the containment air is first circulated through the charcoal filters (refer to Section 9.4.5) if the plant has been operating with significant primary coolant leakage to the containment. This recirculation is assumed to reduce the airborne iodine content by 98 percent, based on a filter DF of 10 for iodine, a circulation flowrate of 24,000 cfm for 16 hours, a containment free volume of 2.6 x 106 cubic feet, and a mixing efficiency of 70 percent. The resulting releases of iodine and noble gases for the Normal Operation Case are provided in Table 11.3-3. In the event of leakage of primary coolant to the auxiliary building, the auxiliary building ventilation system can become a release pathway. In the calculation of release via this pathway, in the Normal Operation Case, it was assumed that primary coolant leakage is 160 pounds per day, with a partition factor of 1 for noble gases and 0.0075 for iodine, 11.3-6 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE with no iodine filtration in the ventilation system. The calculated releases via this pathway are provided in Table 11.3-3. Because of the cleanup of the spent fuel pool water, the very large iodine removal capability of the water, the long decay times for noble gases and iodines in fuel rods handled in the pool, low fuel temperatures, and opportunity for prior release of gases in the containment, the amounts of iodine and noble gases released from the spent fuel pool to the environment under normal conditions are small. In the calculation of the release via this pathway, it was assumed that one-third of a core of spent fuel with 100 hours decay is placed in the spent fuel pool at the beginning of an operating cycle. Radionuclides diffuse through defects in fuel elements when cold at a rate 105 less than at normal core operating temperature. A partition factor of 0.001 was assumed for iodine and 1 for noble gases in the pool. The fuel handling building ventilation charcoal filter system was assumed to be 90 percent efficient for iodine removal. The calculated releases via this pathway are provided in Table 11.3-3. Releases from the waste concentrator condenser vent were assumed to be negligible, and small amounts of gas that are released via this mechanism were included in the analysis by conservatively assuming that they are vented to the atmosphere before entering the GRS. During periods when primary-to-secondary leakage exists, some release of noble gases and iodine will occur via the main condenser air ejector. To calculate these releases, it was assumed that all noble gases are transferred directly to the condenser vapor space, where they are released through the ejector after some decay. The parameters used in the calculation are provided in Tables 11.1-23 through 11.1-25. These assumptions are consistent with those provided in NUREG-0017, April 1976 (Reference 2). Five percent of the iodine leakage from the primary to the secondary system is assumed to be in volatile form and to behave in a manner similar to a noble gas at SG operating temperatures. A partition factor of 0.15 is assumed for volatile iodine in the condenser, and zero for nonvolatile iodine, so the iodine release from this pathway is entirely in volatile form. The activity releases via this pathway are provided in Table 11.3-3. A small potential source of activity release exists from the gland steam condenser. Because of the small steam flowrate used, however, both the noble gas and the iodine release rates via this mechanism were assumed to be negligible compared with the air ejector releases, and this mechanism was not calculated separately. The SGBD tank vent was not considered to be a significant source of iodine release for the purposes of the offsite dose calculations since the blowdown tank was not expected to be used during periods when significant activity is present in the blowdown. Significant release via the 10 percent atmospheric dump valves and main steam and reheat relief valves was not expected under normal operating conditions. 11.3-7 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE A small potential source of gaseous activity release exists from the laundry/RSB and solid radwaste storage facility (SRSF) ventilation systems. However, because of the degassed nature of contaminated materials that enter these facilities, noble gas release rate via these pathways were assumed to be negligible compared with other listed pathways, and therefore were not included. Similarly, since iodine will not be gaseous, iodine releases via these pathways were not included. As a result of normal secondary system steam and water leakage, some iodines and noble gases will escape to the environment if significant activity exists in the secondary system. For the Normal Operation Case, a water leakage of 5 gpm was assumed, along with a steam leakage of 1700 lb/hr. The releases via this pathway are also listed in Table 11.3-3. During plant startup after a cold shutdown, small quantities of radioactive gases may be released from the vacuum pumps (when the condenser is pumped down) and from the cover gas of the LHUTs. Gases from the LHUTs are processed by the GRS, but the discharge from the condenser is vented directly to the atmosphere. Calculations were performed to evaluate the contribution to the total annual air dose at the site boundary in the NW sector of gases released from the condenser during startup. The results of these calculations are listed in Table 11.3-9. Additional assumptions used in the calculations are: (1) Initial secondary system activity equal to equilibrium levels with 0.12 percent fuel defects and 100 pounds per day primary-to-secondary leakage (Normal Operation Case). (2) An "equivalent downtime" was used that is equal to a step change from full power to cold shutdown, a 24-hour down-time, and a step change back to full power. (3) During shutdown, all noble gases were assumed to accumulate in the condenser vapor space; an effective partition factor of 0.15 is assumed for iodine between the secondary system water and the condenser vapor space for volatile iodine species, and zero for nonvolatile iodine. (4) All airborne condenser activity is immediately released to the environment on startup. All iodine released was assumed to be volatile, and therefore not to be absorbed into food pathways. The critical dose point is considered to be the site boundary in the NW sector. This analysis was performed to demonstrate that DCPP normal operation radiological effluents meet the criteria of 10 CFR Part 50 Appendix I. This analysis is conservative and bounds DCPP current operation. This analysis was performed prior to plant operation and would be modified only if a design change or operational change rendered it non-conservative. 11.3-8 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.3.2.4 Dilution Factors 11.3.2.4.1 Current Operational Dilution Factors The meteorological program is discussed in detail in Section 2.3. All current plant releases are calculated in the Annual Radioactive Effluent Release Report, using meteorological conditions derived from 5-year average meteorological data. 11.3.2.4.2 Pre-Operational Dilution Factors HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED. The pre-operational evaluation values of dilution factor (/Q) used in the calculation of annual average offsite radiation dose are provided in Table 11.3-11 for the locations specified in Table 11.3-10. The values of deposition rate (D/Q) used in the calculation of annual average offsite radiation dose are provided in Table 11.3-12 for the locations specified in Table 11.3-10. The D/Q values were derived from Figure 7 of Regulatory Guide 1.111, March 1976 (Reference 3) for a ground level release. The resulting pre-operational evaluation maximum offsite annual average atmospheric activity concentrations in each onshore sector are provided in Table 11.3-13 for the Design Basis Case and in Table 11.3-14 for the Normal Operation Case. 11.3.2.5 Doses 11.3.2.5.1 Current Operational Doses Refer to Section 11.3.3.11 for a discussion of current operational doses from the GRS. 11.3.2.5.2 Pre-Operational Doses HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED. Pre-operational dose calculations were performed at the critical distances for each age group and existing dose pathways in each onshore sector within 5 miles of Unit 1. The critical distances assumed in each sector are provided in Table 11.3-10. The milk cow and milk goat food pathways were not identified within 5 miles of the plant and were therefore not considered. The models used to calculate the offsite radiation doses are those discussed in Regulatory Guide 1.109, March 1976 (Reference 4) for a ground level release. The standard usage factors provided in Regulatory Guide 1.109, March 1976 were used, as well as the standard decay times, transfer factors, and dose conversion factors. Since a ground level release was assumed, the gamma total body and air doses were calculated using the semi-infinite cloud model described in Regulatory Guide 1.109, March 1976, which gives generally conservative results. 11.3-9 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE The results of the offsite dose calculations are presented in Tables 11.3-15 through 11.3-29 for the Design Base Case, and in Tables 11.3-30 through 11.3-44 for the Normal Operation Case. Actual results obtained from the environmental radiological monitoring program are discussed in Section 11.6. On the basis of the calculated estimates of radiation dose presented in Tables 11.3-15 through 11.3-44 and the results of the radiological monitoring program, it can be concluded that, under normal conditions, including consideration of anticipated operational occurrences, the potential dose from gaseous effluents from DCPP Unit 1 and Unit 2 will be well within the dose limits provided in 10 CFR Part 50, Appendix I. 11.3.3 SAFETY EVALUATION 11.3.3.1 General Design Criterion 2, 1967 - Performance Standards The GRS is designed to withstand the DE (refer to Section 11.3.3.15). This design protects the GRS, ensuring its design functions can be performed during the DE. The waste gas compressor seal water coolers are analyzed to meet DE, Double Design Earthquake (DDE) and Hosgri Earthquake (HE) as detailed in Section 9.2.2.3.1. 11.3.3.2 General Design Criterion 3, 1971 - Fire Protection The gases directed into the decay tanks being filled during waste gas compressor operation are continuously monitored for oxygen content via oxygen analyzer CEL 75 and 76, which alarm at 2 percent oxygen concentration. However, hydrogen concentration in the system is not monitored as its concentration is assumed at all times to be 4 percent (the flammability limit for hydrogen is 6 percent oxygen concentration) or more by volume for all plant operating modes. In this manner, the potential for explosive hydrogen/oxygen mixtures will be mitigated. The capability exists for diluting the gas with nitrogen (refer to Figure 3.2-24) using pneumatically operated valves controlled from the auxiliary building control board digital system. In the event that dilution with nitrogen is necessary, a grab sample could be taken for isotopic analyses from the gas decay tanks and waste gas sources as required by the plant Technical Specifications. 11.3.3.3 General Design Criterion 4, 1967 - Sharing of Systems One of the three waste gas compressors, the swing compressor, is shared between DCPP Unit 1 and Unit 2. Cross-tie valve FCV-417 allows any waste gas compressor outlet to be directed to either unit's waste gas decay tanks. The swing compressor and the waste gas storage tanks have no safety functions; therefore the sharing does not impair safety on either unit. 11.3-10 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.3.3.4 General Design Criterion 11, 1967 - Control Room Radiation indication and high radiation alarms for gas decay tank effluent are provided in the main control room. Operator actions in the event of high radiation indication are controlled by plant procedures. 11.3.3.5 General Design Criterion 12, 1967 - Instrumentation and Control Systems The GRS is controlled from the auxiliary building control board digital system. System instruments provide signals to alarm on the auxiliary building control board digital system to warn of abnormal conditions. 11.3.3.6 General Design Criterion 17, 1967 - Monitoring Radioactivity Releases Batch preparation and sampling is used to document GRS discharge to the environment. In addition, instrumentation is provided in the control room and on the auxiliary building control board digital system to monitor GRS radioactive releases. The gas decay tank discharge gas monitor (R-22) continuously monitors the gaseous discharge from the gas decay tanks. A high-radiation signal alarms on the main control board and on the auxiliary building control board digital system and closes the gas decay tanks discharge valves. Refer to Section 11.4.2.1.2.1 and Table 11.4-1 for additional information. 11.3.3.7 General Design Criterion 18, 1967 - Monitoring Fuel and Waste Storage Instrumentation is provided in the control room and on the auxiliary building control board digital system for conditions that might contribute to radiation exposures from the GRS. The auxiliary building control board digital system area monitor (R-10) continuously monitors for high radiation in the auxiliary building that may indicate a rupture of a LHUT or other storage tank. Indication is provided in the control room and on the auxiliary building control board digital system. The gas decay tank cubicle radiation monitors (R-41, R-42, and R-43) continuously monitor noble gas activity in gas decay tanks 1, 2, and 3, respectively. The detectors are located in compartments adjacent to the decay tanks and provide indication on the auxiliary building control board digital system. Refer to Section 11.4.2.1.2.1 and Table 11.4-1 for additional information. 11.3-11 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.3.3.8 General Design Criterion 68, 1967 - Fuel and Waste Storage Radiation Shielding Each waste gas decay tank is located in an individual shielded vault and is equipped with a safety relief valve that discharges to the plant vent. 11.3.3.9 General Design Criterion 69, 1967 - Protection Against Radioactivity Release from Spent Fuel and Waste Storage As discussed in Section 11.3.2.2, the GRS is designed and fabricated as PG&E Design Class II. In accordance with UFSAR Section 3.2.2.1.2, fluid systems and fluid system components that contain or may contain radioactive material, but whose failure would not result in calculated potential exposures in excess of 0.5 rem whole body (or its equivalent to parts of the body) at the site boundary, may be classified as PG&E Design Class II. As part of the acceptability of its PG&E Design Class II classification, an evaluation was performed of the dose consequences at the site boundary of a Gas Decay Tank (GDT) failure accident. Three gas decay tanks are provided for each unit to afford operating flexibility and allow one or more tanks to be isolated from the rest of the system for an extended period of time. Most of the gas stored in the decay tanks is nitrogen cover gas displaced from the liquid waste holdup tanks. The radioactive components are principally the noble gases krypton and xenon, the particulate daughters of some of the krypton and xenon isotopes, and trace quantities of halogens. In the evaluation of the GDT failure accident, the fission product accumulation and release assumptions for the tank rupture are consistent with those of NRC Safety Guide 24, March 1972 (Reference 5). These assumptions are: (1) The reactor has been operating at full power with 1 percent defective fuel and a shutdown to cold condition has been conducted at the end of an equilibrium core cycle. (2) All noble gases have been removed from the primary cooling system and volume control tank and transferred to the gas decay tank that is assumed to fail. Credit is taken for radioactive decay during transfer. (3) The failure occurs immediately on completion of the waste gas transfer, releasing the entire maximum contents of the tank to the auxiliary building and exhausted to the environment via the plant vent assuming a ground level release. The assumption of the release of the noble gas inventory from only a single tank is based on a design that allows all gas decay tanks to be isolated from each other when they are in use. 11.3-12 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE (4) All of the noble gases are exhausted from the auxiliary building over a 2-hour time period or at the rate of the ventilation exhaust flow whichever results in a more rapid release rate. Computer code RADTRAD 3.03 (Reference 6) is used to calculate the whole body dose resulting from the GDT rupture using the semi-infinite cloud submersion model and using air submersion dose coefficients provided in Table III.1 of Federal Guidance Report 12 (Reference 7). The reactor coolant isotopic concentrations based on 1% fuel defects are provided in Table 11.1-11A. Atmospheric dispersion factors used in the analysis are given in Table 2.3-145. The dose consequences at both the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) due to airborne releases following a GDT Rupture are estimated to be below the acceptance criteria of 0.5 rem whole body and 3 rem thyroid. The quantity of radioactive material contained in each gas decay tank is limited to less than or equal to 105 curies noble gases (considered as Xe-133 equivalent), to ensure that in the event of an uncontrolled release of the tank's contents, the resulting whole body exposure to a member of the public at the nearest site boundary will be less than 0.5 rem. 11.3.3.10 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment Each gas decay tank is equipped with a flow control valve connected to the plant vent. The discharge of each valve is routed into a common flow control valve that provides redundant means of isolation and requires manual operation to ensure no inadvertent venting may take place. Downstream of the common flow control valve is a radiation monitor that controls a downstream control valve. If the activity in the discharging waste gas exceeds its upper limit, the control valve closes, terminating the release. The final processing of waste gas prior to release to the atmosphere is by a high-efficiency particulate air (HEPA) filter located just downstream of the radiation control valve and just upstream of the plant vent. Most radwaste gas releases are routed to the plant vent, with the exception of any release from the 10 percent atmospheric dump valves, main steam relief valves, reheat valves, SRSF, RSB/laundry facility exhaust, chemical laboratories exhaust, turbine building exhaust, condenser vacuum pump exhaust, and any miscellaneous steam leakage. Locations of the release points in the plant are shown in Figure 11.3-4. The design exit velocity at the plant vent is 17 m/sec during normal plant operation. The actual gaseous releases are performed in accordance with approved plant procedures. The plant vent pathway is continuously monitored for noble gases, particulates, and iodines (refer to Section 11.3.3.12). 11.3-13 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE The containment atmosphere is sampled and evaluated prior to release via the plant vent pathway during power operations. During periods of maintenance and refueling outages most releases continue via the plant vent pathway. However the containment equipment hatch may be open at times during maintenance resulting in direct communication between the air inside containment and the outside air. If flow out of the containment equipment hatch is detected, containment air is evaluated per plant procedures to characterize the release for reporting. Tritium is also released via the plant vent system mainly due to evaporation from the spent fuel pool. 11.3.3.11 10 CFR Part 20 - Standards for Protection Against Radiation Occupational doses to plant personnel from the GRS are maintained ALARA by: (1) Shielding in locations that contain GRS equipment (refer to Section 12.1.2) (2) HVAC systems that direct air flows for controlled discharge and provide filtration of radioactive materials (refer to Section12.2.2) (3) Monitoring and alarming of potentially radioactive releases from the GRS (refer to Section 11.3.3.4) (4) Administrative controls for working in areas of potentially high radioactivity (refer to Section 17.5) Current operational offsite doses are calculated in accordance with the ODCM contained in site-approved procedures in accordance with 10 CFR Part 20 and appropriate NRC regulatory guides. Technical Specifications require programs for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents within the regulatory limits of 10 CFR Part 20. Doses are reported on quarterly and annual bases. The programs are implemented by plant procedures, and include remedial actions to be taken whenever the program limits are exceeded. 11.3.3.12 10 CFR Part 50 Appendix I - Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as is Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents The GRS maintains control over the releases of radioactive materials in effluents by: (1) Providing appropriate holdup capacity for retention of gaseous effluents containing radioactive material (refer to Section 11.3.2) 11.3-14 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE (2) Continuously monitoring the GRS for leakage (refer to Section 11.3.2) (3) Sampling all discharges prior to release to the environment and using acceptable dilution factors (refer to Section 11.3.2) Technical Specifications require programs for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents within the regulatory limits of 10 CFR Part 50, Appendix I. Doses are reported on quarterly and annual bases. The programs are implemented by plant procedures, and include remedial actions to be taken whenever the program limits are exceeded. 11.3.3.13 40 CFR Part 190 - Environmental Radiation Protection Standards for Nuclear Power Operations Current operational doses from all sources to members of the public are summarized in Annual Radioactive Effluent Release Reports. During preparation of the report, operational doses are compared to the respective 40 CFR Part 190 limit, as implemented under 10 CFR 20.1301(e) (refer to Section 11.3.3.11). 11.3.3.14 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 GRS post-accident instrumentation for meeting Regulatory Guide 1.97, Revision 3, requirements consist of radioactive gas holdup tank pressure, releases of airborne radioactive materials, and releases of radiation and radioactivity to the environs (refer to Table 7.5-6). 11.3.3.15 Regulatory Guide 1.143, Revision 1, October 1979 - Design Guidance for Radioactive Waste Management Structures, Systems, and Components Installed In Light-Water-Cooled Nuclear Power Plants The GRS is PG&E Design Class II (refer to Section 11.3). The PG&E Design Class II GRS equipment and piping that are applicable to Regulatory Guide 1.143, Revision 1 are designated PG&E QA Class R. The portions of the GRS that are intended to store or delay the release of gaseous radioactive waste are seismically supported to withstand the DE and are designated PG&E Piping Symbol F. The remaining portions of the GRS are designated PG&E Piping Symbol H or E, as applicable. 11.3.3.16 NUREG-0737 (Items II.F.1 and III.D.1.1), November 1980 - Clarification of TMI Action Plan Requirements Item II.F.1 - Additional Accident Monitoring Instrumentation: 11.3-15 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Position (1) - The GRS primarily discharges to the plant vent. ER noble gas effluent monitoring is installed in the plant vent and is designed to function during accident conditions (refer to Sections 7.5.2.3 and 11.4.2.1.2.1). Position (2) - Installed capability is provided in the plant to obtain samples of the particulate and iodine radioactivity concentrations that may be present in the gaseous effluent being discharged to the environment from the plant under accident and post-accident conditions. The technical support center (TSC) laboratory is available for onsite testing of the containment air samples (refer to Section 11.3.2). Item III.D.1.1 - Integrity of Systems Outside Containment Likely to Contain Radioactive Material for Pressurized-Water Reactors and Boiling-Water Reactors: The integrity of the GRS is continuously evaluated by monitoring the auxiliary building ventilation exhaust for levels of radiation that may indicate leakage from the GRS (refer to Section 5.2.3.26). 11.3.4 TESTS AND INSPECTIONS All GRS radiation and chemical monitors used in system evaluation are functionally tested and calibrated periodically to ensure the accuracy of measurements. The types of radiation monitors and locations are described in Section 11.4. Measurements are made on a continuous basis and records are maintained of the quantity of radioactive gases released. Comparison of results provides a check on the continuing performance of the waste gas systems. This approach proves effective in documenting deficiencies and their corrections. The routine radiation monitoring program also detects leakage from the GRS by detecting minute changes in the activity of air in the areas occupied by the system. Appropriate means are used to locate and correct any increase in leakage. 11.3.5 INSTRUMENTATION APPLICATIONS Refer to Section 11.3.3.5 for a discussion of the instrumentation requirements and controls for the GRS. 11.
3.6 REFERENCES
- 1. S.G. Gillespie and W.K. Brunot, EMERALD-NORMAL - A Program for the Calculation of Activity Releases and Doses from Normal Operation of a Pressurized Water Plant, Revision 1, Pacific Gas and Electric Company, December 1974.
- 2. NUREG-0017, Calculation of Releases of Radioactive Materials in Liquid and Gaseous Effluents from Pressurized Water Reactors (PWR-GALE code),
USNRC, April 1976. 11.3-16 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE
- 3. Regulatory Guide 1.111, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light Water-Cooled Reactors, USNRC, March 1976.
- 4. Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, USNRC, March 1976.
- 5. Safety Guide 24, Assumptions Used for Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure, USNRC, March 1972.
- 6. RADTRAD 3.03 (GUI Mode Version), A Simplified Model for RADionuclide Transport and Removal And Dose Estimation, NUREG/CR-6604, Users Guide
- Supplement 2, October 2002.
- 7. K.F Eckerman and J.C. Ryman, External Exposure to Radionuclides in Air, Water, and Soil, Federal Guidance Report 12, EPA-402-R-93-081, Environmental Protection Agency, 1993.
11.3-17 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.4 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING SYSTEM The RMS is designed to provide radioactivity measurements, recording capability, alarms, and/or automatic line isolation in order to control and/or process, the release of radioactive fluids in compliance with applicable regulations. In the event of an accident, the process and effluent RMS, in conjunction with the area RMS, provides information on the concentration and dispersion of radioactivity throughout the plant, thereby enabling operating personnel to evaluate the severity and mitigate the consequences of an accident. 11.4.1 DESIGN BASES 11.4.1.1 General Design Criterion 2, 1967 - Performance Standards Certain portions of the RMS are designed to withstand the effects of, or are protected against, natural phenomena, such as earthquakes, tornadoes, flooding, winds, tsunamis, or other local site effects. 11.4.1.2 General Design Criterion 3, 1971 - Fire Protection The PG&E Design Class I portions of the RMS are designed and located to minimize, consistent with other safety requirements, the probability and effects of fires and explosions. 11.4.1.3 General Design Criterion 4, 1967 - Sharing of Systems The RMS is not shared by the DCPP units unless it is shown safety is not impaired by the sharing. 11.4.1.4 General Design Criterion 11, 1967 - Control Room The RMS is designed to or contains instrumentation and controls that support actions to maintain the safe operational status of the plant from the control room. 11.4.1.5 General Design Criterion 12, 1967 - Instrumentation and Control Systems RMS instrumentation and controls are provided as required to monitor and maintain radiation levels within prescribed operating ranges. 11.4.1.6 General Design Criterion 16, 1967 - Monitoring Reactor Coolant Pressure Boundary The RMS provides means for monitoring the reactor coolant pressure boundary (RCPB) to detect leakage. 11.4-1 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.4.1.7 General Design Criterion 17, 1967 - Monitoring Radioactive Releases The RMS is designed to provide means for monitoring the containment atmosphere, the facility effluent discharge paths and the facility environs for radioactivity that could be released from normal operations, from anticipated transients, and from accident conditions. 11.4.1.8 General Design Criterion 18, 1967 - Monitoring Fuel and Waste Storage The RMS is provided with monitoring and alarm instrumentation for fuel and waste storage and handling areas for conditions that might contribute to radiation exposures. 11.4.1.9 General Design Criterion 19, 1971 - Control Room The RMS is designed to permit access and occupancy of the control room for operating the plant without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of design basis accidents. 11.4.1.10 General Design Criterion 21, 1967 - Single Failure Definition The portions of the RMS that support ESF functions are designed to perform their safety functions after sustaining a single failure. Multiple failures resulting from a single event are treated as a single failure. 11.4.1.11 General Design Criterion 40, 1967 - Missile Protection The ESF containment isolation portion of the RMS is designed to be protected against dynamic effects and missiles that might result from plant equipment failures. 11.4.1.12 General Design Criterion 49, 1967 - Containment Design Basis The RMS piping that penetrates the containment is designed to accommodate the pressures and temperatures resulting from the largest credible energy release following a LOCA without exceeding the design leakage rate. 11.4.1.13 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment The RMS 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 capabilities to test periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits. 11.4-2 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.4.1.14 General Design Criterion 56, 1971 - Primary Containment Isolation The PG&E Design Class I portion of the RMS contains valves in piping that penetrate containment and that are connected directly to the containment atmosphere. To ensure containment integrity is maintained, each penetration contains one automatic isolation valve or one check valve inside containment and one automatic isolation valve outside containment. 11.4.1.15 General Design Criterion 69, 1967 - Protection Against Radioactivity Release from Spent Fuel and Waste Storage The RMS is designed to support containment of radioactive releases from the spent fuel and waste storage areas to the public environs as a result of an accident. 11.4.1.16 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment The RMS is designed with provisions for maintaining control over the plants radioactive effluents such that the requirements of 10 CFR Part 20 are met for normal operation, and the requirements of 10 CFR 100.11 are met for accident conditions. 11.4.1.17 Radiological Monitoring System Safety Function Requirements (1) Protection from Jet Impingement - Inside Containment The PG&E Design Class I containment isolation portion of the RMS located inside containment is designed to be protected against the effects of jet impingement which may result from high energy pipe rupture. 11.4.1.18 10 CFR Part 20 - Standards for Protection Against Radiation The RMS supports the protection of personnel from radiation sources such that occupational doses and doses to individual members of the public are maintained below the annual limits prescribed in 10 CFR Part 20. 11.4.1.19 10 CFR 50.49 - Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants RMS components that require EQ are qualified to the requirements of 10 CFR 50.49. 11.4.1.20 10 CFR 50.68(b) - Criticality Accident Requirements The RMS is designed to support compliance with the applicable requirements of 10 CFR 50.68(b) to prevent inadvertent criticality. 11.4-3 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.4.1.21 10 CFR Part 50, Appendix I - Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as is Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents The RMS is designed to support maintaining offsite annual doses from effluents below the limits specified in 10 CFR Part 50, Appendix I. 11.4.1.22 Safety Guide 13, March 1971 - Fuel Storage Facility Design Basis Regulatory Position 7: The RMS includes monitors that alarm locally and in the control room if high radiation levels are experienced in the spent fuel pool area. The RMS also actuates the Automatic Iodine Removal Mode of the fuel handling building ventilation system (FHBVS). 11.4.1.23 Regulatory Guide 1.21, Revision 1, June 1974 - Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants The RMS provides radiological monitoring of all major and potentially significant paths for release of radioactive material during normal plant operation, including anticipated operational occurrences. 11.4.1.24 Regulatory Guide 1.45, May 1973 - Reactor Coolant Pressure Boundary Leakage Detection Systems Leakage detection systems are designed with acceptable methods to detect and identify the location of the source of RCPB leakage. The designs of the containment radioactivity monitors meet the requirements of Regulatory Guide 1.45, May 1973, with the following exceptions: (1) The containment air particulate and containment radiogas detectors are not constructed to withstand DDE accelerations; however, they are housed in a PG&E Design Class I structure and protected from external damage associated with a seismic event. Therefore, it is considered that these monitors can be returned to operational status within 36 hours of a DDE. (2) For low RCS activities, the containment atmosphere gaseous radioactivity monitors may not be capable of detecting a RCS leak of 1 gpm within one hour. 11.4-4 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.4.1.25 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 RMS provides instrumentation to monitor radiation levels and radioactivity concentrations in plant process streams, effluent paths, and areas during and following an accident. 11.4.1.26 NUREG-0737 (Items II.F.1, III.A.1.2, and III.D.1.1), November 1980 - Clarification of TMI Action Plan Requirements Item II.F.1 - Additional Accident Monitoring Instrumentation Position (1) - The RMS is designed with noble gas effluent monitors with an ER designed to function during accident conditions. Position (2) - The RMS is designed with provisions for sampling of plant effluents for post-accident releases of radioactive iodines and particulates. Position (3) - The RMS is designed with containment high-range radiation monitors. 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(f) - The TSC is provided with radiological 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. Item III.D.1.1 - Integrity of Systems Outside Containment Likely to Contain Radioactive Material for Pressurized-Water Reactors and Boiling Water Reactors: The RMS supports a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels. 11.4.2 SYSTEM DESCRIPTION 11.4.2.1 Continuous Monitoring 11.4.2.1.1 General Description The components of the RMS are designed for operation in the following ranges of conditions: (1) Temperature - An ambient temperature range of 40° to 120°F. 11.4-5 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE (2) Humidity - 0 to 95 percent relative humidity. (3) Pressure - Components in the auxiliary building and control room are designed for normal atmospheric pressure. Area monitoring system components inside the containment are designed to withstand containment test pressure. (4) Radiation - Process and area radiation monitors are of a non-saturating design so that they will register full-scale or over range if exposed to radiation levels up to 100 times full-scale indication. (5) Radiation monitoring equipment is designed and located such that radiation damage to electrical insulation and other materials will not affect their usefulness over the life of the plant. (6) The RMS is designed such that it can be checked, tested, and recalibrated as required. The only components of the RMS that are exposed to a wider range of conditions are located in the containment. This includes detectors and associated local alarm and indication equipment for the area-type monitoring channels there. Some of these, namely the low level-normal ops monitors are not expected to operate following a major LOCA. Post-accident high-range gamma monitors are used for post-accident situations. Most of the control room RMS equipment used for normal operation and anticipated operational occurrences are centralized in cabinets. Data loggers are provided in the RMS cabinets in the control room (refer to Table 11.4-1). Each monitoring channel on the data loggers is sequentially recorded. Equipment used solely for post-accident monitoring (PAM) is located in additional cabinets in the control room. The digital RMS equipment is located in a six-bay cabinet while the control room pressurization system (CRPS) monitors are in a separate control room rack. Sliding channel drawers are normally used to facilitate maintenance and allow replacement of units, assemblies, and entire channels if needed. It is possible to completely remove the various chassis from the cabinet after disconnecting the cable connectors from the rear of these units. Detector output is usually measured in either counts per minute (cpm), milliroentgens per hour (mR/hr), roentgens per hour (R/hr), or microcuries per cubic centimeter (Ci/cc). Each channel has a minimum range of three decades. Radiation monitors are listed in Tables 5.2-16 and 11.4-1. The iodine monitors are isotopic I-131 process monitors and read in cpm or Ci/cc. The RMS is divided into the following subsystems: 11.4-6 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE (1) The process RMS that monitors radiation levels in various plant process streams and effluent paths (2) The area RMS that monitors radioactivity in various areas within the plant The locations of all detectors with respect to plant equipment for Unit 1 are listed in Table 11.4-1. Unit 2 detectors are in corresponding locations. Piping sequence and locations of process detectors are found in appropriate piping schematics, shown as figures in Section 3.2. 11.4.2.1.2 Process Radiation Monitoring System 11.4.2.1.2.1 Description This system, as illustrated in Figure 11.4-1, consists of multiple channels that monitor radiation levels in various plant operating systems. The output from each channel detector except the digital radiation monitors is transmitted to the RMS cabinets where the radiation level is indicated on a meter and recorded by a multipoint recorder (refer to Table 11.4-1). Except for the air particulate/iodine/noble gas monitors in the TSC and the adjacent laboratory, the gas decay tank cubicles, and the SGBD overboard monitor, the RMS cabinets for most process radiation monitors are located in the control room. The RMS cabinets for the TSC and the adjacent laboratory are located in the computation center. High radiation level alarms are indicated on the RMS cabinets with annunciation on main annunciators. Except for the monitors in the TSC, the adjacent laboratory and some supplementary monitors, the main annunciator for process monitors is at the control board in the control room. The control board annunciator provides several windows that alarm for input channels (process or area) detecting high radiation. The main annunciator for the TSC and the adjacent laboratory is in the TSC panel annunciator in the computation center. Verification of which channel has alarmed is done at the RMS cabinets serving that annunciator (refer to Table 11.4-1). A tabulation of the process radiation monitoring channels is found in Table 11.4-1. The minimum sensitivity is based on a Co-60 background level of 2 mR/hr. A typical channel contains a completely integrated modular assembly that includes the following: (1) Log Level Amplifier Accepts detector pulses, performs a log integration (converts total pulse rate to a logarithmic analog signal), and amplifies the resulting output for suitable indicating and recording. 11.4-7 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE (2) Power Supplies Furnishes electrical power for the circuits, relays, alarm lights, and detectors. (3) Test-Calibration Circuitry Provides a pre-calibrated pulse signal to test channel electronics and a solenoid-operated radiation check source to verify channel operation. A common annunciator on the main control board indicates when a channel is in the test mode. (4) Radiation Level Meter Provides a dual scale calibrated logarithmically from 101 to 104, and 101 to 106 cpm. The wide-range level signal is also recorded by the data logger recorder. (5) Indicating Lights Indicate high-radiation alarms, tests, and circuit failures. A number of annunciator windows on the main control board are actuated either on high radiation signal from the channels or from any channel failure. A separate window is lit when channels are placed in the test mode. However, the digital radiation monitoring system and a few other monitors do not have this test alarm feature. (6) Bistable Circuits Two bistable circuits are provided: one to alarm on high radiation (actuation point may be set at any level within the range of the instruments), and one to alarm on loss of signal (circuit failure). (7) Check Source A remotely operated long half-life radiation check source is furnished for each channel. The check source simulates the radiation being monitored. The check source interaction is sufficient to cause an upscale indication. The main steam line radiation monitors and the CRPS radiation monitors have a fixed keep alive source mounted directly on the detector. It is used only to keep the channel out of a low fail condition. The process RMS consists of the radiation monitoring channels described in Items 1 through 28 below. (The prefix numbers, where used with channel identification, indicate monitors associated with Unit 1 or Unit 2; 0 indicates a shared monitor or no unit designation.) 11.4-8 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE The sample lines for the air particulate and gaseous radiation monitors (containment, RHR heat exchanger compartment exhaust, control room, TSC, laboratories, and plant vent) are designed and installed in accordance with the recommendations of Reference 2. The flow in each of the sample lines is turbulent. Particle deposition due to gravity and Brownian diffusion are assumed to be small since the horizontal runs of the sample lines are short and the sample velocity is high. Long-radius bends are used for all sample lines, including the inlet lines to the monitors, to preclude deposition due to extreme turns. Isokinetic probes are used wherever the sample is taken from a moving airstream. Deposition in the basically vertical sample line runs are assumed to be largely due to turbulent deposition. The following are process radiation monitors that do not support ESF functions. For those that do support ESF functions, refer to Section 11.4.2.1.4. (1) Containment - Air Particulate Monitor (1-R-11 and 2-R-11) This monitor is provided to measure air particulate gamma radioactivity in the containment. The sampler for this channel takes a continuous air sample from the containment atmosphere. The inlet line from inside the containment is routed through the containment penetration to the monitor, which is located adjacent to the penetration. The sample is monitored by a scintillation counter-filter paper detector assembly. The filter paper has a collection efficiency of approximately 99% for particles of 0.3 microns or larger. The particulate matter is collected on the filter paper's constantly moving surface and is viewed by a gamma scintillation detector. The sample is returned to the containment after it passes through the series-connected gas monitor. The pulse signal is transmitted to the RMS cabinets in the control room. For activity releases inside containment, the air particulate and gas monitors RE-11 and RE-12 will alarm in the control room. Lead shielding is provided to reduce the background level to where it does not interfere with the detector's sensitivity. (2) Containment - Radioactive Gas Monitor (1-R-12 and 2-R-12) This monitor is provided to measure gaseous beta-gamma radioactivity in the containment. The detector consists of a beta-gamma sensitive Geiger-Mueller (G-M) tube mounted in the monitor container. 11.4-9 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE This channel takes a continuous air sample from the containment atmosphere that passes through the air particulate monitor (1-R-11, 2-R-11), and then through the gas monitor assembly. The sample is circulated in a fixed volume where it is monitored by the radiation detector. The sample is then returned to the containment. Its output is transmitted to the RMS cabinets in the control room. For activity releases inside containment, the air particulate and gas monitors RE-11 and RE-12 will alarm in the control room. (3) RHR Heat Exchanger Compartment Exhaust Duct Air Particulate Detector Monitor (1-R-13 and 2-R-13) This monitor is provided to measure air particulate gamma radioactivity in the RHR heat exchanger compartments' exhaust ducts to detect a leaking recirculation loop component in the event of a LOCA. It operates in the same manner as the containment air particulate monitor. The sampler for this channel takes a continuous common air sample from the exhaust ducts of both RHR compartments. An isokinetic probe is installed in each RHR compartment exhaust duct. The monitor is located in proximity to the sample points. The sample is monitored by a scintillation counter-filter paper detector assembly. The sample is then returned to the exhaust ducts. Ducts may be sampled individually by use of a selector switch at the console. High radiation is annunciated at the main control board. (4) Plant Noble Gas Vent Monitor (1-R-14, 2-R-14) (1-R-14R, 2-R-14R) Each channel consists of a pressurized three liter volume monitored by a beta scintillation detector. RM-14 is part of the normal range (NR) skid. RM-14R is part of the redundant normal range (RNR) skid. Local indication for these channels is provided by the local radiation processors (LRPs) mounted on their respective skids. Remote indication for these channels is provided by the radiation display units (RDUs) for their respective skids. The RDUs are mounted in the RMS panels in the control room. 11.4-10 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE (5) Condenser Air Ejector Gas Monitor (1-R-15, 1-R-15R, 2-R-15, 2-R-15R) These channels monitor the discharge from the air ejector exhaust header of the condensers. Gaseous radiation is indicative of a primary-to-secondary system leak. The gas discharge is routed to the plant vent. Both RM-15 and RM-15R share the same detector skid. Local indication for these channels is provided by the wall-mounted LRPs associated with each detector. Remote indication for these channels is provided by RDUs. The RDUs are mounted in the RMS panels in the control room. High radiation is annunciated at the main control board. (6) Component Cooling Liquid Monitors (1-R-17A, 2-R-17A, 1-R-17B, 2-R-17B) These channels continuously monitor the component cooling water (CCW) system for radiation indicative of a leak of reactor coolant from the RCS and/or the RHR loop to the CCW. Each channel employs an off-line detector using a bypass line from CCW pump discharge to suction. Due to the discharge piping configuration, however, only one monitor is sampling flow representative of the bulk system when only one CCW heat exchanger is in service. A high-radiation-level signal initiates closure of the valve located in the component cooling surge tank vent line to prevent gaseous radiation release. Adequate lead shielding is provided to reduce the effect of background radiation so that it does not interfere with the detector's sensitivity. (7) Liquid Radwaste Effluent Monitor (0-R-18) This channel continuously monitors discharges from the LRS. Automatic valve closure is initiated to prevent further release after a high radiation level is indicated and alarmed, and flow is diverted to the equipment drain receiver tanks. A scintillation detector located in an in-line sampler monitors these effluent discharges. An alarm function is provided on the main control board and the auxiliary building control board digital system. Adequate lead shielding is provided to reduce the effect of background radiation so that it does not interfere with the detector's sensitivity. In addition, samples from the LRS batches are analyzed in the laboratory. (8) SGBD Sample Monitor (1-R-19, 2-R-19) This channel monitors the liquid phase of the secondary side of the SG for radioactivity (which would indicate a primary-to-secondary system leak) providing backup information to that of the condenser air ejector gas monitors. Blowdown samples from each of the four SGs are combined in a common header and the common sample is continuously monitored by a scintillation detector in an in-line sampler assembly. 11.4-11 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Adequate lead shielding is provided to reduce the effect of background radiation so that it does not interfere with the detector's sensitivity. High activity alarm indications are displayed locally and at the RMS cabinets, with annunciation at the main control board in the control room. If a high activity alarm occurs, isolation valves in the blowdown and sample lines acting with the valve in the line from the blowdown tank to the discharge structure will close and the blowdown tank liquid effluent will be diverted to the equipment drain receiver tank. Subsequent identification of the leaking SG would then be made by manual override of sample line isolation and drawing separate samples from each SG for analysis. (9) Gas Decay Tank Discharge Gas Monitor (1-R-22, 2-R-22) This channel monitors the gaseous discharge from the gas decay tanks. The detector consists of a G-M tube inserted into an in-line fixed volume container that includes adequate shielding to reduce the background radiation low enough not to interfere with the detector's sensitivity. This channel will alarm on the main control board and auxiliary building control board digital system and close the gas decay tanks discharge valve on a high radiation level signal. (10) Plant Vent Particulate Monitors (1-R-28, 2-R-28)(1-R-28R, 2-R-28R) Each channel consists of a fixed particulate filter monitored by a beta scintillation detector. RM-28 is part of the NR skid. RM-28R is part of the RNR skid. The sample for each skid is isokinetically drawn from the plant vent stack. Local indication for these channels is provided by the LRPs mounted on their respective skids. Remote indication for these channels is provided by the RDUs for their respective skids. The RDUs are mounted in the RMS panels in the control room. (11) Plant Vent Iodine Monitors (1-R-24, 2-R-24)(1-R-24R, 2-R-24R) Each channel consists of a charcoal cartridge filter monitored by a gamma scintillation detector. Iodine is discriminated using a single channel analyzer. RM-24 is part of the NR skid. RM-24R is part of the RNR skid. The sample for each skid is isokinetically drawn from the plant vent stack. Local indication for these channels is provided by the LRPs mounted on their respective skids. Remote indication for these channels is provided by the RDUs for their respective skids. The RDUs are mounted in the RMS panels in the control room. 11.4-12 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE (12) SGBD Tank Liquid Effluent Monitor (1-R-23, 2-R-23) This channel is provided to continuously measure liquid effluent from the blowdown tank. The channel employs a scintillation detector located in an off-line sample chamber. Adequate shielding is employed to reduce effects of background radiation. The count rate is handled in the same manner as for the basic RMS. Output is recorded in conjunction with, and parallel to, the recorded outputs of the flow elements related to the blowdown tank inputs and effluents at a local panel specifically provided for that function. Output is also recorded at the data logger in the RMS cabinets in the control room. Alarms are provided on the main control board for high and low radiation (instrument failure). A high-radiation signal isolates the blowdown discharge, and diverts blowdown tank liquid effluent to the equipment drain receiver tank. (13) High-Range Plant Vent Gas Monitor (1-R-29, 2-R-29) Post-accident Monitor This monitor measures high-range gross gamma radioactivity in the plant vent. The detector consists of a shielded ion chamber contiguous to the plant vent and mounted on an adjacent support structure. A control room readout, with an associated recording device, is provided on the PAM panel. Also provided on this panel are the high and low (instrument failure) radiation alarms. The high and fail radiation alarms are also provided on the main control board. The high alarm also alarms in the State of California Office of Emergency Services in Sacramento. (14) Deleted (15) ER Noble Gas Monitor (1-R-87, 2-R-87) The ER channel, RM-87, uses a beta scintillation detector operated in the current mode. The ER noble gas detector is less sensitive and the volume of the detection chamber is smaller than those of the NR noble gas channel. The ER chamber is not pressurized. The sample is isokinetically drawn off the plant vent stack at approximately 1/20th the rate of the sample for the NR and RNR skids. The chamber is downstream of two identical trains of particulate and iodine roughing filters/grab samplers. Alternating between the trains allows removing a grab sample while continuing to monitor the stack. The grab samplers are to be used for assessing post-accident releases of particulates and iodine using laboratory instruments. All of this equipment is mounted on the ER skid. Local indication for RM-87 is provided on the LRP for the NR Skid. 11.4-13 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Remote indication for RM-87 is provided on the RDU for the NR skid. The RDU is mounted in the RMS panels in the control room. Indication for RM-87 is on the same indicating channel used for RM-14. (16) TSC Air Supply Radioactive Particulate Monitor (0-R-66) This monitor is provided to measure air particulate gamma radioactivity in the ventilation air supply to the TSC. The isokinetic flow sampler for this channel takes a continuous air sample from the TSC ventilation air supply duct. (17) TSC Air Supply Noble Gas Monitor (0-R-67) This monitor is provided to measure the noble gas activity in the ventilation air supply to the TSC. The same isokinetic flow as described in item (16) above is used. (18) TSC Air Supply Iodine Radiation Monitor (0-R-82) This monitor is provided to measure the iodine activity in the ventilation air supply to the TSC. The same isokinetic flow as described in item (16) above is used. (19) Laboratory Adjacent to the TSC Radioactive Particulate Monitor (0-R-68) This monitor is provided to measure air particulate gamma radioactivity in the laboratory. The sample is drawn directly from the room. (20) Laboratory Adjacent to the TSC Noble Gas Monitor (0-R-69) This monitor is provided to measure the noble gas activity in the laboratory. The sample is drawn directly from the room. (21) Laboratory Adjacent to the TSC Iodine Radiation Monitor (0-R-83) This monitor is provided to measure the iodine activity in the laboratory. The sample is drawn directly from the room. (22) SRSF and RSB Ventilation Exhaust Air Particulate Samplers (0-RX-55, 0-RX-56) These samplers consist of in-line particulate filter assemblies that provide the capability to sample and subsequently assess (via laboratory analysis) the, concentrations of radioactive material present in the exhaust from the SRSF and RSB. RX-55 samples the exhaust from the SRSF, and RX-56 11.4-14 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE samples the exhaust from the laundry/respirator cleaning facility and RSB prior to the discharge of these effluent points to the environment. (23) Oily Water Separator Effluent Monitor (0-R-3) This channel continuously monitors the turbine building sump retention tank discharge into the oily water separator. (24) Deleted (25) Main Steam Line Activity Monitors (1-R-71 through 1-R-74 and 2-R-71 through 2-R-74) These monitors consist of gamma-sensitive G-M tubes and are provided to continuously monitor the main steam lines. The detectors are located next to each main steam line and measure the steam activity from the line's shine. (26) Gas Decay Tank Cubicle Radiation Monitors (1-R-41, 1-R-42, 1-R-43, 2-R-41, 2-R-42, 2-R-43) These monitors are for detecting noble gas activity in the gas decay tanks. The detectors are located in compartments adjacent to the decay tanks and provide indication on the auxiliary building control board digital system. (27) Solid Radwaste Inspection Station Radiation Monitors (0-R-84, 0-R-85) These radiation monitors are provided to permit the assessment of the contact (R-84) and one-meter (R-85) radiation dose rates being given off from material containers being prepared for storage and shipment as solid radioactive waste. The detectors are located at the decontamination/inspection station in the RSB. (28) Laundry Processing Room Radiation Monitoring Assembly (0-R-92) The laundry processing room radiation monitoring channel (R-92) extracts a sample from the air space in the laundry processing room and monitors this sample to determine the radioactivity concentration in the air. Samples are extracted by a sampling nozzle located in the room and returned to the laundry processing room. 11.4-15 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.4.2.1.2.2 Design Evaluation For ruptures or leaks in the waste processing system, plant area monitors and the vent stack monitor will alarm on an increase in radiation over a preset level. For cases where leaks are involved, the operator may control activity release by system isolation. For inadvertent releases relative to violation of administrative procedures, monitors provide alarms and the means for limiting radioactive releases. The gas decay tank discharge monitor will close the flow control valve in the waste decay tanks discharge line when the radiation level in the line exceeds a preset level. Where liquid waste releases are involved, the liquid radwaste discharge monitor trips shut a valve in the discharge line when the radioactivity in the discharge line exceeds a preset level and redirects the flow to the equipment drain receiver tanks. For SGBD releases, the blowdown effluent monitor and the blowdown sample monitor isolate the blowdown discharge and will divert the blowdown tank liquid effluent to the equipment drain receiver tanks. 11.4.2.1.3 Area Radiation Monitoring System 11.4.2.1.3.1 Description This system consists of multiple channels that monitor radiation levels in various areas of the plant. The system has low-range monitors for normal operation and high-range monitors for post-accident conditions. These monitors and their locations are listed in Table 11.4-1. The selection and location of the monitoring areas are based on multiple considerations, including occupancy status of various plant zones, potential for increase in background activity levels due to operations carried out in a particular location, and desirability of surveillance of infrequently visited areas. A typical channel of the area RMS consists of a fixed-position, gamma-sensitive G-M tube detector. The detector count rate is amplified, and its log count rate is displayed by the readout in the RMS cabinets. The radiation level is indicated locally at the detector and at the RMS cabinets and it is also recorded. Except for the area monitors in the TSC and the adjacent laboratory, most RMS cabinets are located in the control room (refer to Table 11.4-1). The RMS cabinet for the area monitors in the TSC and laboratory is located in the TSC computation center. High-radiation alarms are displayed on main annunciators, on the RMS cabinets, and at the detector location. The control room annunciator provides several windows that alarm for channels detecting high radiation, except for the monitors in the TSC and laboratory. The main annunciator for high radiation detected by the TSC and laboratory area monitors is the TSC HVAC annunciator located in the computation center. Verification of which 11.4-16 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE channel has alarmed is done at the RMS cabinets. Each channel contains a completely integrated modular assembly (refer to Section 11.4.2.1.2). The log level amplifier module amplifies the radiation level signal for indication and recording. The module also provides controls for actuation of the channel check source. A meter is mounted on the front of each readout module and is scaled to read logarithmically (refer to Table 11.4-1). A local meter, scaled logarithmically, is mounted at the detector assembly. Two mutually redundant high-range containment monitors RE-30 and RE-31 are provided for each unit, each consisting of a detector mounted inside the containment liner to mitigate the effects of local hot spots and to obtain the best "view" of the containment free volume. The units are powered from separate instrument power channels (refer to Section 8.3.1.1.5). Each detector is a hermetically sealed, stacked, parallel plate, three-terminal guarded ionization chamber, operated in the saturated mode. The detector and its special cable are environmentally qualified to IEEE 323-1974. Each readout has a range of 1 to 107 R/hr and has high alarm, failure alarm, logarithmic scale recorder, and electronic system and detector checks. For those radiation monitors that support ESF functions, refer to Section 11.4.2.1.4. 11.4.2.1.3.2 Design Evaluation Radiation detection instruments are located in areas of the plant that house equipment containing or processing radioactive materials. These instruments continually detect, display, and, as appropriate, record operating radiation levels. If the radiation level should rise above the channels pre-determined high alarm setpoint, an alarm is initiated in a control room. Local annunciation is provided at the detector, with the exception of the high-range containment detectors, to indicate high radiation levels to personnel in the area. The monitoring system is operated in conjunction with regular and special radiation surveys and with chemical and radiochemical analyses performed by the plant staff. Adequate information and warning is thereby provided for the continued safe operation of the plant and assurance that personnel dose does not exceed 10 CFR Part 20 limits. 11.4-17 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.4.2.1.4 Radiation Monitoring Channels that Support Engineered Safety Features 11.4.2.1.4.1 Description Radiation monitoring channels that support ESF functions include the following: (1) Main Control Room Air Intake Monitor (1-R-25, 1-R-26) (2-R-25, 2-R-26) These channels monitor air entering the normal air intakes for the control room ventilation system (CRVS) to support the ESF function to initiate Mode 4 CRVS operation. These monitors use a scintillation type general-purpose detector and have a range of 10-2 to 103 mR/hr. Each monitor has a control room readout module with instrument failure alarm. A high radiation signal provides control room annunciation and automatic actuation. (2) Containment Purge Exhaust (CPE) Monitors (1-R-44A, 2-R-44A)(1-R-44B, 2-R-44B) Each channel consists of a beta scintillation detector mounted to the side of the CPE duct. The detectors are mounted diametrically opposed on the 48-inch CPE duct. Their location is downstream of the CPE fan, E-3. Local indication for each channel is provided by the wall mounted LRPs associated with each detector. Remote indication for each channel is provided by the RDUs. The RDUs are mounted in the RMS panels in the control room. These monitors support an ESF function to close the containment ventilation isolation (CVI) valves in the case of high radioactivity exhausting the containment. (3) CRPS Ventilation Intake Air Monitor (1-R-51, 1-R-52) (2-R-53, 2-R-54) These channels monitor air entering the CRPS ducts and support the ESF function to shut down the operating CRPS train and automatically start the opposite units CRPS. These monitors use G-M tube-type general purpose detectors and have ranges of 10-2 to 104 mR/hr. Each monitor has a control room readout module with instrument failure alarm. A high radiation signal provides control room annunciation and automatic actuation. (4) Spent Fuel Pool Area Monitor (1-R-58, 2-R-58) 11.4-18 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE This channel is located near the spent fuel pool and supports the ESF function to transfer the FHBVS to the Automatic Iodine Removal Mode should the pre-determined setpoint be exceeded. This monitor uses a G-M detector and has an indicating range of 10-1 to 104 mR/hr. It provides continuous monitoring and indication with alert and high radiation level alarms in the main control room. Local audible and visual indicators are also provided. (5) New Fuel Storage Area Monitor (1-R-59, 2-R-59) This monitor is located near the new fuel storage area and supports the ESF function to transfer the FHBVS to the Automatic Iodine Removal Mode should the pre-determined setpoint be exceeded. This channel uses a G-M detector and has an indicating range of 10-1 to 104 mR/hr. It provides continuous monitoring and indication with alert and high radiation level alarms in the main control room. Local audible and visual indicators are also provided. 11.4.2.1.4.2 Design Evaluation An evaluation of instrumentation function relative to monitoring and controlling releases of radioactivity from various plant systems is discussed below. (1) Fuel Handling Inside Containment The air exhausted from the containment through the containment purge and exhaust lines is monitored by RM-44A and RM-44B. In the event that the pre-determined high alarm setpoint levels are exceeded, these radiation monitoring channels will initiate a signal that would cause the closure of the CVI valves and mitigate the consequences of the accident. (2) Fuel Handling Outside Containment Radiological conditions local to the spent and new fuel storage areas are monitored by RM-58 and RM-59, respectively. These monitors provide local and control room alarm annunciation, and will initiate a change in the mode of operation of the FHBVS to Automatic Iodine Removal Mode in the event that a high alarm setpoint is exceeded. An evaluation of instrumentation functions relative to monitoring and controlling radiological exposures to control room operators is discussed below. 11.4-19 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE (1) CRVS Normal Air Supply Monitoring The normal air intakes to the CRVS are monitored by RM-25 and RM-26. MODE 4 operation of the CRVS is automatically initiated upon detection of radioactive contaminants at the normal air intakes. (2) CRPS Air Supply Monitoring The CRPS air supply piping is monitored for radioactivity by RM-51 and RM-52 (Unit 1) and RM-53 and RM-54 (Unit 2). In the event an accident occurs in one unit, the system automatically selects the pressurization intake train of the opposite unit. With radiation detected at both pressurization intakes, one of the trains will start. However, the operator manually switches to the intake with lower airborne radioactivity. 11.4.2.2 Sampling 11.4.2.2.1 Basis for Selection of Sample Locations Locations for periodic sampling are based on the following: (1) Sampling of process fluids that contain radioactivity (2) Sampling of process fluids not normally radioactive that may become radioactively contaminated due to some component failure 11.4.2.2.2 Expected Composition and Concentration Because of the diversity of sources of sampled fluids, the activity levels are expected to range from negligible to the reactor coolant concentrations. Concentrated liquid samples may have specific activities higher than that of the RCS. 11.4.2.2.3 Quantity to Be Measured Samples expected to contain radioactivity are analyzed periodically as specified in the radiological monitoring and controls procedures and the chemical analysis procedures of the Plant Manual. 11.4.2.2.4 Sampling Frequency and Procedures Sampling frequency varies according to the sample being analyzed and previous activity level of the sample. The sampling frequency for effluents is specified in the radiological and monitoring procedures of the Plant Manual. The sampling frequency for non-effluent samples is specified in the chemical analysis procedures of the Plant Manual. 11.4-20 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.4.2.2.5 Analytical Procedures and Sensitivity Analytical procedures are in accordance with the Plant Chemistry Manual. The required sensitivities are specified in the radiological and monitoring procedures of the Plant Manual. Equipment used for radiation analysis is located near the primary chemical laboratory and the laboratory near the TSC. 11.4.2.2.6 Influence of Results on Plant Operations The Technical Specifications lists the appropriate radioactive contamination limits on the pertinent systems, as well as the required actions if the limits are exceeded. 11.4.3 SAFETY EVALUATION 11.4.3.1 General Design Criterion 2, 1967 - Performance Standards The PG&E Design Class I portions of the RMS are seismically qualified (refer to Section 3.7), except for those that are classified as Regulatory Guide 1.97, Revision 3, Category 2 which are not required to be seismically qualified (refer to Table 7.5-6). The CCW discharge radiation monitors, which are PG&E Design Class I, are seismically qualified for the purpose of maintaining CCW pressure boundary integrity only (refer to Section 3.2.2.5). The PG&E Design Class I portions of the RMS are protected from the effects of floods and tsunamis (refer to Section 3.4) and external missiles (refer to Section 3.5). With the exception of certain PG&E Design Class I portions of the RMS that are located in structures that are vulnerable to tornadoes, PG&E Design Class I RMS equipment are designed to perform their safety functions under the effects of winds and tornadoes (refer to Section 3.3). Those portions that are located in structures that are vulnerable to tornadoes are not required for plant safe shutdown (refer to Section 3.3.2). 11.4.3.2 General Design Criterion 3, 1971 - Fire Protection The RMS is designed to meet the requirements of 10 CFR 50.48(a) and (c) (refer to Section 9.5.1) 11.4.3.3 General Design Criterion 4, 1967 - Sharing of Systems Certain RMS components, as designated in Table 11.4-1, are common to both Unit 1 and Unit 2. All shared RMS components are PG&E Design Class II, perform no safety function, and are isolated from PG&E Design Class I components as described in Sections 7.1.3 and 8.3.1.1.5; therefore safety is not impaired by the sharing. 11.4-21 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.4.3.4 General Design Criterion 11, 1967 - Control Room The RMS provides indication and alarms in the control room on PAM panels PAM-1 and PAM-2; and on radiation monitoring racks RNRMA, RNRMB, RNRMC, RNRMD, RNRME, RNRMS1, RNRMS2, RNRMS3, RNRMS4, RNGFFD, and RCRM (refer to Table 11.4-1). 11.4.3.5 General Design Criterion 12, 1967 - Instrumentation and Control Systems The RMS provides indication and recording of radiation levels within plant areas and process streams (refer to Sections 11.4.2.1.2 and 11.4.2.1.3). In addition, the RMS provides signals that support automatic functions for controlling radiation within plant effluents (refer to Sections 11.4.2.1.2 through 11.4.2.1.4). 11.4.3.6 General Design Criterion 16, 1967 - Monitoring Reactor Coolant Pressure Boundary The RMS monitors the containment building for elevated airborne activity levels that may be indicative of RCPB leakage (refer to Sections 5.2.3.23.1.1 and 11.4.3.24). 11.4.3.7 General Design Criterion 17, 1967 - Monitoring Radioactive Releases The containment is continuously sampled for particulate and gaseous radioactivity during normal operations and anticipated transients (refer to Section 11.4.2.1.2.1). The ODCM identifies the RMS channels required to be operable during normal operations and anticipated transients that monitor all releases from the units, and the methodology to determine their respective alarm and trip setpoints. Refer to Section 11.4.3.25 for a discussion of the RMS channels required to monitor radioactivity under accident conditions. 11.4.3.8 General Design Criterion 18, 1967 - Monitoring Fuel and Waste Storage The fuel and waste storage and handling areas are provided with monitoring and alarm systems for radioactivity (refer to Sections 9.1.1.3.5, 9.1.2.3.5, 11.3.3.7, 11.4.2.1.4, and 11.5.3.4). 11.4.3.9 General Design Criterion 19, 1971 - Control Room CRPS operation is automatically initiated on a normal air intake radiation monitor signal (refer to Sections 6.4.1.3.6 and 9.4.1.3.7). Two radiation monitors are provided for each CRPS air intake. Evaluations of postaccident control room radiological exposures are presented in Section 15.5. 11.4-22 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.4.3.10 General Design Criterion 21, 1967 - Single Failure Definition The portions of the RMS that support the ESF functions of the FHBVS, the CRVS, and CVI are designed to provide redundant logic channels (refer to Section 11.4.2.1.4). These redundant channels are electrically isolated and physically separated (refer to Sections 7.1.3 and 8.3.1.4); therefore a single failure within a channel will not prevent the ESF support function of the RMS from being accomplished. 11.4.3.11 General Design Criterion 40, 1967 - Missile Protection The provisions taken to protect the ESF containment isolation portion of the RMS from damage that might result from missiles and dynamic effects associated with equipment and high-energy pipe failures, respectively are discussed in Sections 3.5, 3.6, and 6.2.4. 11.4.3.12 General Design Criterion 49, 1967 - Containment Design Basis The RMS containment penetrations, 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 containment design leakage rates. Refer to Sections 3.8.1.13 and 11.2.3.24 for additional details. 11.4.3.13 General Design Criterion 54, 1971 - Piping Systems Penetrating Containment The RMS CIVs required for containment closure are periodically tested for operability. Testing of the components required for the containment isolation system is discussed in Section 6.2.4. 11.4.3.14 General Design Criterion 56, 1971 - Primary Containment Isolation The PG&E Design Class I containment penetrations for the RMS are penetration Group A or E (refer to Section 6.2.4). A description of the isolation valves and piping for each penetration is provided in Table 6.2-39. Group A and Group E piping comply with the requirements of GDC 56, 1971. 11.4.3.15 General Design Criterion 69, 1967 - Protection Against Radioactivity Release from Spent Fuel and Waste Storage In the event of an accident in the fuel handling building, high radioactivity would be detected by the RMS, and the exhaust air would be diverted through charcoal filters (refer to Sections 9.1.2.3.8, 9.4.4.3.11, and 15.5.22). The RMS allows for early detection of a release from a waste gas decay tank, allowing operator action to terminate release (refer to Section 15.5.24). 11.4-23 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.4.3.16 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment The RMS provides annunciation and actuation signals to support the control of radioactivity releases such that offsite doses do not exceed the annual limits of 10 CFR Part 20 for normal operations and anticipated transients and the limits of 10 CFR 100.11 for accidents (refer to Section 11.4.2). 11.4.3.17 Radiological Monitoring System Safety Function Requirements (1) Protection from Jet Impingement - Inside Containment The provisions taken to provide protection of the inside containment PG&E Design Class I containment isolation portion of the RMS from the effects of jet impingement which may result from high energy pipe rupture are discussed in Section 3.6. 11.4.3.18 10 CFR Part 20 - Standards for Protection Against Radiation The RMS provides local and remote indications and is operated in conjunction with routine radiation surveys to provide adequate information of radiation levels in the plant. This supports the safe operation of the plant and the assurance that occupational doses to plant personnel are maintained below the annual limits prescribed in 10 CFR Part 20 (refer to Section 11.4.2.1.3.2). The RMS provides annunciation and actuation signals to support the control of radioactivity releases such that offsite doses to members of the public do not exceed the annual limits of 10 CFR Part 20 and 40 CFR Part 190 for normal operations and anticipated transients (refer to Sections 11.2, 11.3, and 11.5). 11.4.3.19 10 CFR 50.49 - Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants The Class 1E RMS components required to function in harsh environments under accident conditions are qualified to the applicable environmental conditions to ensure they will continue to perform their safety functions. Section 3.11 describes the DCPP EQ program and the requirements for the environmental design of the electrical and related mechanical equipment. The affected components are listed in the EQ Master List. 11.4.3.20 10 CFR 50.68(b) - Criticality Accident Requirements In accordance with 10 CFR 50.68(b)(6), radiation monitors are provided in storage and associated handling areas when fuel is present to detect excessive radiation levels and to initiate appropriate safety actions (refer to Sections 9.1.1.3.8 and 9.1.2.3.10). 11.4-24 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.4.3.21 10 CFR Part 50 Appendix I - Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as is Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents Effluent monitor setpoints are established to close isolation valves such that annual average releases of radioactive material in gaseous effluents remain within the dose values specified in 10 CFR Part 50 Appendix I. 11.4.3.22 Safety Guide 13, March 1971 - Fuel Storage Facility Design Basis Regulatory Position 7: Radiation levels are monitored locally in the spent fuel pool area. Signal inputs are provided to local alarms and control room annunciators, and to the FHBVS to initiate the automatic iodine removal mode of operation (refer to Sections 9.1.2.3.5 and 9.4.4.3.14). 11.4.3.23 Regulatory Guide 1.21, Revision 1, June 1974 - Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants The RMS continuously monitors all significant plant radiological effluents. A summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant is reported in the Annual Radioactive Effluent Release Report administered by the ODCM. 11.4.3.24 Regulatory Guide 1.45, May 1973 - Reactor Coolant Pressure Boundary Leakage Detection Systems The containment radioactivity monitors support the detection of RCPB leakage (refer to Sections 5.2.3.23.1.1 and 11.4.3.6). 11.4.3.25 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 RMS post-accident instrumentation for meeting Regulatory Guide 1.97, Revision 3, requirements consists of (refer to Table 7.5-6): (1) Condenser air removal system exhaust noble gas monitors (2) Containment area high range radiation monitors (3) Radiation exposure rate monitors (inside buildings or areas) 11.4-25 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE (4) Plant vent noble gas, particulates, and iodine monitors (5) Plant vent particulate filters installed on post-accident grab sampling equipment (6) Vent from SG safety relief valves or 10 percent atmospheric dump valves 11.4.3.26 NUREG-0737 (Items II.F.1, III.A.1.2, and III.D.1.1), November 1980 - Clarification of TMI Action Plan Requirements Item II.F.1 - Additional Accident Monitoring Instrumentation Position (1) - ER noble gas effluent monitoring is installed for the plant vent and is designed to function during accident conditions (refer to Sections 7.5.2.3 and 11.4.2.1.2.1). Position (2) - Installed capability is provided in the plant to obtain samples of the particulate and iodine radioactivity concentrations that may be present in the gaseous effluent being discharged to the environment from the plant under accident and postaccident conditions. The TSC laboratory is available for onsite testing of the plant vent air samples. Position (3) - High-range radiation monitoring is installed in the containment to monitor ambient gamma radiation following a LOCA (refer to Table 7.5-6). Item III.A.1.2 - Upgrade Emergency Support Facilities: NUREG-0737, Supplement 1, January 1983 provides the requirements for III.A.1.2. Section 8.2.1(f) - Radiation monitors with continuous indication and alarm capabilities are installed in the TSC. The ability to distinguish radioiodines at the required low concentration is provided by a dedicated monitor. The RMS provides warning of high radiation levels thereby, in conjunction with shielding and ventilation, assuring 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 (refer to Sections 6.4.2.3.4 and 12.1.7.8). Item III.D.1.1 - Integrity of Systems Outside Containment Likely to Contain Radioactive Material for Pressurized-Water Reactors and Boiling Water Reactors: During operation a routine radiation monitoring program has been implemented to detect changes in activity of air in the areas occupied by the GRS. If leakage is indicated by an increase in airborne activity, specific tank check methods are used to identify the location of the leak. For portions of the charging and letdown systems, excessive leakage into controlled areas is indicated by off normal radioactivity levels. 11.4-26 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.4.4 TESTS AND INSPECTIONS 11.4.4.1 Alarm Setpoints The alarm/trip setpoints for radioactive liquid and gaseous effluent radiation monitors (as defined in Technical Specifications and Equipment Control Guidelines) are determined in accordance with the methodology and parameters in the ODCM. The alarm/trip setpoints for all other process and area radiation monitors are established by administrative procedures and controlled in Volume 9B, Table T-IIC-2, "I&C RMS Data Book for Radiation Monitoring and Allied System Data," and are based on protection of public health and safety, plant personnel health and safety, and maintaining efficient plant operation. Table 11.4-3 lists those monitors that affect valve control operations, together with their effect. 11.4.4.2 Definitions Radiation Monitor Channel Functional Test - Injection of a simulated signal into the channel as close to the sensor as practicable to verify operability including alarm and/or trip functions. Radiation Monitor Channel Source Check - The qualitative assessment of channel response when the channel sensor is exposed to a radiological source. Other definitions are listed in Section 1 of the Technical Specifications. 11.4.4.3 Calibration Procedure Area and process monitors were initially calibrated by their original manufacturer. Response curves for each detector were provided with the instrument. These curves essentially relate detector performance to the energy spectrum that the detector would see in operation. 11.4.4.3.1 Area Monitors Based upon the requirements of the plant Technical Specifications, traceable radioactive sources are used to calibrate the area monitors. The monitors are also functionally checked periodically in accordance with the plant Technical Specifications. 11.4.4.3.2 Process Monitors For the process monitors the, detectors are calibrated with traceable radioactive sources on the frequency defined in the plant Technical Specifications or Equipment Control Guidelines (refer to Chapter 16). Further, the detector response is correlated to the results of analysis of the process stream with calibrated counting room equipment. 11.4-27 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.4.4.4 Test Frequencies Calibration and functional checks of the process and area monitors are performed at frequencies that are in accordance with the plant Technical Specifications and Equipment Control Guidelines. 11.4.4.5 System Summary It is concluded that the administrative controls imposed on the operator, combined with the RMS design, provide a high degree of assurance against accidental release of radioactivity to the environment. 11.4.5 INSTRUMENTATION APPLICATIONS Refer to Section 11.4.3.5 for a discussion of the instrumentation requirements and controls for the RMS. 11.
4.6 REFERENCES
- 1. Deleted in Revision 23.
- 2. ANSI N13.1-1969, American National Standard Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities.
11.4-28 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.5 SOLID RADWASTE SYSTEM The SRS is designed to process, package, and store the radioactive wastes generated by plant operations until they are shipped offsite to a licensed waste processing or disposal facility. The SRS is a PG&E Design Class II and III system. Figure 11.5-1 is a flow diagram of the SRS. 11.5.1 DESIGN BASES 11.5.1.1 General Design Criterion 3, 1971 - Fire Protection The SRS is designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. 11.5.1.2 General Design Criterion 4, 1967 - Sharing of Systems The SRS is not shared by the DCPP units unless it is shown safety is not impaired by the sharing. 11.5.1.3 General Design Criterion 17, 1967 - Monitoring Radioactivity Releases The SRS is designed to provide means for monitoring the effluent discharge paths, and the facility environs for radioactivity that could be released from normal operations, from anticipated transients, and from accident conditions. 11.5.1.4 General Design Criterion 18, 1967 - Monitoring Fuel and Waste Storage The SRS is provided with monitoring and alarm instrumentation for conditions that might contribute to radiation exposures. 11.5.1.5 General Design Criterion 68, 1967 - Fuel and Waste Storage Radiation Shielding The SRS is designed to provide shielding for radiation protection to meet the requirements of 10 CFR Part 20. 11.5.1.6 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment The SRS includes those means necessary to maintain control over the plant solid radioactive effluents during normal operation, including anticipated operational occurrences. 11.5-1 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.5.1.7 10 CFR Part 20 - Standards for Protection Against Radiation The SRS supports the protection of personnel from radiation sources such that occupational doses and doses to individual members of the public are maintained below the annual limits prescribed in 10 CFR Part 20. 11.5.1.8 10 CFR 61.55 - Waste Classification At DCPP, the classification of solid radioactive waste prior to shipment for disposal is in accordance with the requirements of 10 CFR 61.55. 11.5.1.9 10 CFR Part 71 - Packaging and Transportation of Radioactive Material The SRS ensures that packaging, preparation for shipment, and transportation of licensed material is accomplished in accordance with the requirements of 10 CFR Part 71. 11.5.1.10 40 CFR Part 190 - Environmental Radiation Protection Standards for Nuclear Power Operations The SRS supports the protection of members of the public from radiation sources from the uranium fuel cycle such that annual doses are maintained below the limits specified in 40 CFR Part 190. 11.5.1.11 49 CFR Parts 171-178 - Subchapter C-Hazardous Materials Regulations The SRS ensures the safe and secure transportation of hazardous materials in accordance with the requirements of 49 CFR Part 171 through 49 CFR Part 178. 11.5.1.12 Regulatory Guide 1.143, Revision 1, October 1979 - Design Guidance for Radioactive Waste Management Structures, Systems, and Components Installed In Light-Water-Cooled Nuclear Power Plants The SRS design follows the guidance of Regulatory Guide 1.143, Revision 1 as it relates to the quality group classification of components. 11.5.1.13 IE Bulletin 79-19, August 1979 - Packaging of Low-Level Radioactive Waste for Transport and Burial DCPP assures the safe transfer, packaging, and transport of low-level radioactive waste in accordance with IE Bulletin 79-19, August 1979. 11.5-2 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.5.1.14 10 CFR 50.48(c) - National Fire Protection Association Standard NFPA 805 The SRS is designed to meet the nuclear safety and radioactive release performance criteria of Section 1.5 of NFPA 805, 2001 Edition. 11.5.2 SYSTEM DESCRIPTION 11.5.2.1 System Inputs The SRS collects the following inputs for processing, packaging, and disposal: (1) Spent filter/ion exchange media (2) Spent ion exchange resin (3) Spent filter cartridges (4) Miscellaneous dry active wastes (i.e., contaminated paper, rags, clothing, tools, etc.) The SRS input from each of the plant solid radioactive waste streams is presented in Table 11.5-1. HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED. Tables 11.5-2 and 11.5-4 list the activities of the spent ion exchange resins, the spent filter media, and the spent cartridge filters for both the Normal Operation Case and the Design Basis Case. 11.5.2.2 Components The SRS has five major subsystems: (1) Spent resin processing system (2) Spent filter/ion exchange media processing system (3) Spent filter cartridge processing system (4) Mobile radwaste processing system (MRPS) (5) Dry active waste processing system The function of each of these subsystems is described in the following paragraphs. 11.5-3 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.5.2.2.1 Spent Resin Processing System The system for transferring spent resins from any of the ion exchangers to the spent resin storage tanks (SRSTs) consists of four separate headers connected to four eductors and discharge systems that permit the transfer of resin from any of the 30 ion exchanger units to either of two SRSTs. Pressurized air is used to transfer resin from either SRST to the loadout station (LS) to which the MRPS container is connected. Pressurized air can also be used to transfer resin from one SRST to the other. Two SRST eductors are provided to transfer resin from one SRST to the other. Figure 11.5-3 is a flow diagram of the spent resin processing system. The general layout of the system is shown in Figure 11.5-11. The SRSTs are located in shielded cells. All of the valves, instruments, and the discharge eductor are located in a separately shielded area (valve gallery) adjacent to the SRST cells. The spent resin LS is located on the east outside wall of the auxiliary building. A spent resin sampling system allows for the collection of grab samples as resins enter the SRSTs or while the spent resins are being transferred out of the SRSTs to the MRPS. All of the equipment associated with this system is considered potentially highly radioactive. None of the equipment, which is located behind shielding, will be approached either for operation or maintenance except under the direction of plant radiation protection personnel under the special work permit rules of the plant. 11.5.2.2.2 Spent Filter/Ion Exchange Media Processing System Pressurized air is used to transfer exhausted media from either of the two radwaste media filters to the LS to which the MRPS container is connected. 11.5.2.2.3 Spent Filter Cartridge Processing System This system is designed to remove and handle spent filter cartridges generated in the CVCS, spent fuel storage system, and LRS. The radioactively contaminated spent filter cartridges can be removed from the filter housing or vessels with the operator remaining behind shielding. The spent cartridges are transferred to storage or to the MRPS in shielded transfer casks. Transfer casks are provided to move underwater vacuum filters from the SFP to the MRPS. It is assumed that the whole change-out procedure takes one hour. This includes loading clean filter cartridges. Half an hour can be spent with the operator protected from radioactive spent cartridges by the transfer cask. For the remaining time, the operator is protected from the filter cartridges by concrete, steel, and/or lead shields. Figures 11.5-6 through 11.5-8 show the location of filters in the system. 11.5-4 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Figure 11.5-9 illustrates the system for capturing the filters using the grappling hook, cask, pulley, and mirrors. Figure 11.5-10 shows the cover of the filter vessel. 11.5.2.2.4 Mobile Radwaste Processing System The MRPS is a skid-mounted mobile radwaste dewatering/solidification system. The MRPS for media and resin is located on a concrete pad as shown in Figure 11.5-12 and in Bay 2 of the SRSF for filters. This space will accommodate the spillage of resins via concrete sloped to a drain within the area. The MRPS is operated on a batch basis to solidify concentrates, to dewater or solidify spent ion exchange or filtration media, and to encapsulate spent cartridge filters. Slurries from the media filter vessels are sluiced out to the MRPS and dewatered or solidified. Spent resin slurries are sluiced to the MRPS (refer to Section 11.5.2.5) from the SRST and dewatered or solidified. Filter cartridges are transferred to the MRPS container in a shielded spent filter transfer cask, if required. Waste concentrates, ion exchange media, filtration media, and cartridge filters are dewatered or solidified in accordance with the process control program detailed in the Plant Procedures Manual. Normally, containers are dewatered or solidified in processing shields. The containers are stored in the shields until shipment. The containers are transferred by mobile cranes into the shipping casks. Containers may be dewatered and/or solidified while in casks on the trailers by which they will be shipped. When processing is finished in shipping casks, the containers are shipped immediately so that no in-plant handling is required. Complete waste solidification or absence of free liquid prior to shipment is ensured by the implementation of a process control program consistent with the recommendations of NUREG-0472, March 1979 (Reference 1). For medium and high activity waste, level sensors monitor the levels in the waste containers and provide alarm signals to alert the MRPS operator to take action to prevent filling beyond preset levels. Low activity waste level may be monitored by sight by the MRPS operator. Potential waste container overflows are contained by the curbed processing pad, then flow into a sump that will return the spill to the radwaste system. 11.5.2.2.5 Dry Active Waste Processing System Potentially radioactive dry wastes are collected at appropriate locations throughout the plant, as dictated by the volume of the wastes generated during operation or maintenance. The wastes are then segregated, processed, and packaged. Compressible dry active wastes may be processed by compaction in either a drum or box compactor. During compaction, the airflow in the vicinity of the compactor is directed by the compactor exhaust fan through a HEPA filter before it is discharged. Large or highly radioactive components and equipment that have been contaminated during reactor operation and that are not amenable to compaction are handled either by 11.5-5 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE qualified plant personnel or by outside contractors specializing in radioactive materials handling, and the components and equipment are packaged in shipping containers of an appropriate size and design. 11.5.2.3 Mixed Waste Mixed waste is liquid or solid waste that is both hazardous and radioactive. Mixed waste is segregated and accumulated in drums in Bay 6 of the SRSF. Filled drums of mixed waste are placed in storage in Bay 5 of the SRSF. 11.5.2.4 Component Failures and System Malfunctions Analyses have been performed to evaluate potential dose to operating personnel should the SRSs malfunction or components fail. The components and systems considered most likely to fail are discussed below. During the transfer of spent resin to the SRSTs, a failure of the motive water pump could result in lines becoming clogged with resin. The lines may be cleaned out by starting up the second motive water pump and using the normal operating procedure. The CVCS resin transfer piping system includes cleanout flanges, which can be used if the lines cannot be cleaned by using the normal procedure. The cleanout flanges are located in areas that are shielded from the main resin transfer lines, and minimize the dose to operating personnel during a cleanout operation. During the transfer of spent resin from a demineralizer to an SRST, failure of a pneumatically operated valve on the outlet of the demineralizer would require special operator action. The valves have reach-rods extending through the shield wall, permitting manual operation of the valve. After flushing the demineralizer completely and allowing sufficient time for decay of any residual activity, the operating personnel would perform the required repair on the valve. A maximum dose of 800 mR is estimated to occur in the repair of a valve on a CVCS mixed bed demineralizer. As a backup in the event the pneumatically operated tank outlet valve fails during transfer of spent resin from one of the SRSTs, the valve can be controlled by manual operators extending through a second shield wall to the operating area. 11.5.2.5 Packaging Disposable mild steel liners are used for packaging dewatered, solidified, or encapsulated wastes. The typical liner sizes may range from 80 to 300 cubic feet. High -integrity containers (HICs) are also used for packaging dewatered or encapsulated wastes. Typical HIC sizes range from 75 to 200 cubic feet. Wet solid waste may be packaged for further offsite treatment, onsite storage, or offsite disposal. Dry active wastes may be packaged for further offsite treatment, onsite storage, or offsite disposal. For onsite storage, 55-gallon steel drums, 4 x 4 x 6 foot steel boxes, or 11.5-6 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE bags are typically used. Drums, boxes, and intermodal containers classified as Industrial Package Type 1 or Type 2 containers are utilized for transportation offsite, as applicable. 11.5.2.6 Storage Facilities Onsite storage for packaged wastes is provided by the SRSF and the RSB. These buildings are located east of the auxiliary building, as shown in Figure 11.5-4. Unpackaged activated metals and underwater vacuum cleaner filters are stored in the spent fuel pools. Disused radiation monitors may also be stored in the new fuel storage vaults. The SRSF provides a storage area for metal boxes, drums, and shielded filters. The SRSF can hold 580 drums or a combination of 65 boxes and 60 drums. The arrangement of the rooms in the SRSF is shown in Figure 11.5-5. A forklift is used for moving the containers into and within the storage area. Concrete walls provide shielding between the various vaults in the SRSF. Encapsulation or dewatering of filters may occur in the SRSF. Segregation and compaction of dry active waste is also performed in the SRSF. The RSB provides storage areas for 180 liners or HICs and for compacted dry active waste in 4 x 4 x 6 foot boxes or 55-gallon drums. The liner storage vaults are sized to accommodate 80 ft3 containers stacked 3 wide by 2 high. An overhead crane assembly is used for moving liners or HICs to their respective storage areas in the RSB. A shielded cask rail car is used to transport liners or HICs from the load-out area to the storage vaults. Encapsulation of filters and dewatering of resin may occur in the shielded cask rail car in the RSB truck bay. A liner-inspection/decontamination station is also provided for preparing containers for storage or shipping. The compacted dry active waste storage area (DAW vault) is sized to accommodate 522 boxes at 93 ft3 each, totaling 48,546 ft3 of storage. Drums can also be stored in this facility. A forklift is used for moving the containers into and within the DAW vault. The SGs and reactor vessel head assemblies (RVHAs) were removed from DCPP Unit 1 and Unit 2 during the SG and RVH replacement projects. These ten large components are temporarily stored in the steam generator storage facility (SGSF) specifically constructed for this purpose. The SGSF meets the radwaste storage requirements for temporary storage of the SGs and RVHAs until site decommissioning. The SGSF is designed to be used as a non-occupied mausoleum for the temporary storage of the SGs and RVHAs. No other radwaste storage is permitted within this facility. 11.5-7 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.5.2.7 Shipment The shipment of pre-packaged solid radwaste from the plant site to licensed waste processing or disposal facilities is contracted to firms licensed to transport radioactive material in accordance with applicable Department of Transportation (DOT) regulations. All shipping containers and transportation casks are in conformance with 49 CFR Part 171 through 49 CFR Part 178 and 10 CFR Part 71, as applicable. Table 11.5-5 summarizes the expected quantities to be shipped. 11.5.3 SAFETY EVALUATION 11.5.3.1 General Design Criterion 3, 1971 - Fire Protection The SRS is designed to meet the requirements of 10 CFR 50.48(a) and (c) (refer to Section 9.5.1). 11.5.3.2 General Design Criterion 4, 1967 - Sharing of Systems The SRS is shared between DCPP Unit 1 and Unit 2. Because the SRS has no safety functions, the sharing does not impair safety on either unit. 11.5.3.3 General Design Criterion 17, 1967 - Monitoring Radioactivity Releases Instrumentation is provided in the SRS buildings to monitor SRS radioactive releases. The RSB ventilation exhaust air particulate samplers (RX-55 and RX-56) are in-line particulate filter assemblies for sampling the exhaust from the SRSF and the RSB, respectively, prior to discharge to the environment. These filters provide the capability to sample and subsequently assess (via laboratory analysis) the concentrations of radioactive material present in the exhaust from the respective buildings. Refer to Section 11.4.2.1.2.1 and Table 11.4-1 for additional information. 11.5.3.4 General Design Criterion 18, 1967 - Monitoring Fuel and Waste Storage Instrumentation is provided in the SRS buildings for conditions that might contribute to radiation exposures from the SRS. The solid radwaste inspection station radiation monitors (R-84 and R-85) are area radiation monitors that permit the assessment of the radiation dose rates from waste containers being prepared for storage and shipment. The detectors are located at the decontamination/inspection station in the RSB. The radwaste storage monitor (R-90) is an area radiation monitor located in the truck bay of the RSB. 11.5-8 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE The laundry facility monitor (R-92) is a process radiation monitor located in the laundry processing room of the SRSF. Samples are extracted by a sampling nozzle located in the room and then returned to the laundry processing room. Refer to Section 11.4.2.1.2.1 and Table 11.4-1 for additional information. 11.5.3.5 General Design Criterion 68, 1967 - Fuel and Waste Storage Radiation Shielding Radiation shielding in the auxiliary building and the SRS buildings protects personnel working near systems containing radioactivity from doses in excess of 10 CFR Part 20 limits (refer to Sections 11.5.2.2 and 12.1.7.3). 11.5.3.6 General Design Criterion 70, 1967 - Control of Releases of Radioactivity to the Environment Solid radwastes are collected and contained in dedicated radwaste storage facilities. Control of releases of radioactivity to the environment from the SRS is provided by HEPA filters in the SRSF and RSB ventilation systems and monitoring of the SRS effluents (refer to Sections 11.5.2.2.5 and 11.5.3.3). Baling of dry wastes is carried out inside a closed dust shroud. For shipping purposes, wastes are packaged in containers that meet DOT requirements, and are shipped to licensed waste processing or disposal facilities in accordance with NRC and DOT regulations (refer to Sections 11.5.2.5, 11.5.2.7, 11.5.3.9, and 11.5.3.11). 11.5.3.7 10 CFR Part 20 - Standards for Protection Against Radiation Occupational doses to plant personnel from the SRS are maintained ALARA by: (1) Shielding in locations that contain SRS equipment (refer to Section 11.5.2) (2) HVAC systems that direct air flows for controlled discharge and provide filtration of radioactive materials (refer to Section12.2.2) (3) Monitoring of potentially radioactive releases from the SRS (refer to Sections 11.5.3.3 and 11.5.3.4) (4) Packaging that is in accordance with regulatory guidelines for radioactive material (refer to Section 11.5.2.5) (5) Protecting operating personnel in the event of a malfunction or component failure (refer to Section 11.5.2.4) 11.5-9 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE (6) Administrative controls for working in areas of potentially high radioactivity (refer to Section 17.5) 11.5.3.8 10 CFR 61.55 - Waste Classification Plant administrative controls ensure that the classification of solid radwastes is in conformance with the requirements of 10 CFR 61.55. 11.5.3.9 10 CFR Part 71 - Packaging and Transportation of Radioactive Material Plant administrative controls ensure that shipping containers and transportation casks for solid radwaste transportation to offsite waste processing or disposal facilities are in conformance with 10 CFR Part 71 (refer to Section 11.5.2.7). 11.5.3.10 40 CFR Part 190 - Environmental Radiation Protection Standards for Nuclear Power Operations Current operational doses from all sources to members of the public are summarized in Annual Radioactive Effluent Release Reports and are demonstrated to be within the limits specified by 40 CFR Part 190, as implemented under 10 CFR 20.1301(e). 11.5.3.11 49 CFR Parts 171-178 - Subchapter C-Hazardous Materials Regulations Plant administrative controls ensure that all shipping containers and transportation casks for solid radwaste transportation to offsite waste processing or disposal facilities are in conformance with 49 CFR Part 171 through 49 CFR Part 178 (refer to Section 11.5.2.7). 11.5.3.12 Regulatory Guide 1.143, Revision 1, October 1979 - Design Guidance for Radioactive Waste Management Structures, Systems, and Components Installed In Light-Water-Cooled Nuclear Power Plants The SRS is PG&E Design Class II and III (refer to Section 11.5). The PG&E Design Class II SRS equipment and piping that are used for processing radwaste liquids (spent resins) and are applicable to Regulatory Guide 1.143, Revision 1 are designated PG&E QA Class R, PG&E Piping Symbol H. 11.5.3.13 IE Bulletin 79-19, August 1979 - Packaging of Low-Level Radioactive Waste for Transport and Burial DCPP maintains copies of applicable NRC and DOT regulations for transportation of low-level radioactive materials, including those requirements for the offsite processing or disposal facilities. 11.5-10 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Designated personnel on site who are responsible for transportation of low-level radioactive materials and operation of the SRS are regularly trained on their respective responsibilities and on the governing NRC and DOT regulations. DCPP maintains an audit program to ensure continued compliance with applicable NRC and DOT regulations for the transfer, packaging, and transportation of low-level radioactive material. 11.5.3.14 10 CFR 50.48(c) - National Fire Protection Association Standard NFPA 805 The SRS is designed to meet the nuclear safety and radioactive release performance criteria of Section 1.5 of NFPA 805, 2001 Edition (refer to Section 9.5.1). 11.5.4 TESTS AND INSPECTIONS Tests and inspections of the SRS are done in accordance with plant procedures. 11.5.5 INSTRUMENTATION APPLICATIONS Refer to Sections 11.5.3.3 and 11.5.3.4 for discussions of the instrumentation requirements and controls for the SRS. 11.
5.6 REFERENCES
- 1. NUREG-0472, Radiological Effluent Technical Specifications for PWRs, Revision 3, USNRC, March 1979.
11.5.7 REFERENCE DRAWINGS Figures representing controlled engineering drawings are incorporated by reference and are identified in Table 1.6-1. The contents of the drawings are controlled by DCPP procedures. 11.5-11 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM The offsite radiological monitoring program complies with the requirements of the State of California Department of Health Services, Radiological Health Section, and the NRC. The monitoring program, required by the DCPP Technical Specifications, includes monitoring, sampling, analysis, and reporting, including performance of a Land Use Census and participation in an Interlaboratory Comparison Program. 11.6.1 DESIGN BASES 11.6.1.1 10 CFR Part 50, Appendix I - Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as is Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents The offsite radiological monitoring program provides data on measurable levels of radiation and radioactive materials in the environment in order to evaluate the relationship between quantities of radioactive materials released in effluents and resultant radiation doses to individuals from principal pathways of exposure. The program identifies changes in the use of unrestricted areas (e.g., for agricultural purposes) to permit modifications in monitoring programs for evaluating doses to individuals from principal pathways of exposure. 11.6.2 PROGRAM DESCRIPTION 11.6.2.1 Expected Background HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED. The 1984 results of the pre-operational monitoring program are shown in Table 11.6-4 and Reference 1. Table 11.6-4 summarizes measurements of external dose with thermoluminescent dosimeters (TLDs), gross beta, gamma isotopic, and I-131 activities in air samples and gamma isotopic and I-131 and/or tritium activities (as appropriate for particular samples) in marine and terrestrial samples. There were no known local man-made sources of radioactivity in the vicinity of DCPP; therefore, the variations shown in these tables are considered to be either natural variations or fallout from weapons testing. Gamma isotopic analyses were made of all marine and terrestrial samples. Only a few showed measurable activities above background. The results of all samples with detected activity during the pre-operational period of January 1, 1981, through March 31, 1984, are included in the Pre-operational Environmental Report (Reference 8). The data, presented on an annual basis in the Annual Radiological Environmental Operating Report, show local differences in activity as well as seasonal variations. 11.6-1 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE Terrestrial variations may be attributable to such factors as variation in the spatial distribution of radionuclides in the soil, the amount of rainfall, TLD locations in valleys as contrasted to hillsides, and secondary sources of airborne dust from such activities as construction or farming. These data and those for previous years serve as a baseline during plant operation. 11.6.2.2 Critical Pathways HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED. Based on the expected radiological releases from Unit 1 and Unit 2 (refer to Sections 11.2 and 11.3), and the tabulated estimates of dose, none of the releases were expected to significantly increase the total dose to man relative to natural background. Calculations showed which principal pathways for atmospheric releases will give the maximum doses. All doses through aquatic releases were expected to be negligible. The levels of radiation in environmental samples were expected to be very low and, for many isotopes, below the minimum detectable level, using the best techniques available at original plant licensing. For this reason, dose analyses are performed based principally on plant effluent data, with secondary analyses based on environmental data. For airborne releases, the ODCM is used with measured local meteorological data, measured release data of the gases and particulates, plus local demographic data, to estimate individual dose. From the gamma dosimeter stations for the direct radiation measurements, in general, the offsite stations are used to serve as reference points for natural background and manmade environmental radiation that is not associated with plant operations. Onsite and fenceline stations are used to measure dose from the plant. Therefore, direct radiation dose above background is obtained and compared to calculated doses. For radiological releases to the ocean, the ODCM is used in conjunction with effluent data to estimate dose from the consumption of aquatic foods grown within the radiological influence of the plant. Bioaccumulation factors for species in the vicinity of Diablo Cove are obtained from Regulatory Guide 1.109, Revision 1 (Reference 7). Aquatic food intake is based on the parameters provided in Reference 7 via the ODCM. Consideration is also given to any group that has unusually high per capita consumption. 11.6-2 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE 11.6.2.3 Sampling Media, Location and Frequency 11.6.2.3.1 Marine Samples The types of marine samples, the frequency of collection, and the sampling locations are presented in the ODCM. These samples were selected to represent various food products. 11.6.2.3.2 Terrestrial Samples Possible dose to man could result from atmospheric immersion and inhalation, and consumption of radionuclides deposited as particulates from the gaseous effluent of DCPP. To monitor the above pathways, various types of terrestrial samples are collected and analyzed. Air samples using particulate filters and iodine cartridges are taken continuously at a minimum of four sample locations. The sites were selected to provide data at downwind locations, major population centers, and areas that are not influenced by plant operations. It should be noted that eight of 16 sectors surrounding DCPP are located over water; therefore the five air sampling stations recommended by the Branch Technical Position for Radiological Environmental Monitoring Program (Revision 1, 1979) were reduced to four air sampling stations. Gamma dosimetry measurements are made at environmental monitoring stations using TLDs. The TLDs were selected because of their sensitivity and the ease of readout. Drinking water samples are collected from the domestic water system. Surface water samples are collected from the plant outfall. Samples of various foodstuffs produced in the area are also collected when available. The terrestrial sampling frequency reflects the areas that are most sensitive to changes in radioactive levels and in dose measurements. Thus, the airborne sampling is weekly, the TLD measurements are quarterly, and the terrestrial foods measurements are monthly, quarterly, or in season. 11.6.2.4 Analytical Sensitivity 11.6.2.4.1 Types of Analyses The types of radiological analyses performed on each sample are presented in the ODCM. The offsite radiological monitoring program emphasizes analyses for those radionuclides expected to be present in the DCPP effluent and those that will be the major contributors to dose to the public. The effluent from DCPP is expected to contain radionuclides whose identity and activity can be determined by gamma spectrometry. Thus, all samples are placed in a fixed 11.6-3 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE geometry and analyzed by gamma spectrometry. Other analysis techniques can be utilized as deemed necessary. 11.6.2.4.2 Measuring Equipment The equipment presently in use for the radiological monitoring program typically includes, but is not limited to: (1) Gas-flow proportional counter for gross beta analyses (2) High purity intrinsic germanium detectors (or equivalent) (3) TLDs for external dose measurements (4) Beta-gamma coincidence spectrometer (5) Liquid scintillation spectrometer 11.6.2.4.3 Sample Detection Sensitivity The ability to accurately determine the radioactivity in a sample is a function of many variables including the following: (a) sample size, (b) self-absorption in the sample, (c) detector counting efficiency, (d) counting time background count rate, (e) half-life of the isotope, (f) loss of radionuclides in sample preparation, and (g) ability to distinguish between isotopes with similar gamma emission energies. Consistent results are obtained by standardizing procedures that maintain as many of the above variables constant as practicable. 11.6.2.5 Data Analysis and Presentation The data acquired from the environmental monitoring program falls into the categories of: (1) Information on the distribution of radioactivity in lower trophic levels in the physical environs of DCPP (2) Information on external radiation in the vicinity of DCPP (3) Information on radionuclides in foodstuffs that may result in a dose to man In examining the distribution of radionuclides in the environment and lower trophic levels, comparisons are made to the preoperational data to determine if there are any biological or physical compartments in nature that are accumulating radioactivity. Similarly, external radioactivity measurements during plant operation are compared with the average and range of data obtained in the preoperational program. 11.6-4 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE If radionuclides due to plant effluents are found in foodstuffs, estimates of radiation dose are made that utilize the best estimates of food consumption. These dose calculations are compared with those based on plant emission data with the appropriate meteorological and aquatic dispersion models as discussed in Section 11.6.2.2. The data from the offsite radiological monitoring program are reported in the Annual Radiological Environmental Operating Report. The report includes the basic data on sampling locations, organisms collected, counting data, gross activity levels, identification of gamma emitting isotopes, and the associated counting errors. Tables 11.6-13 and 11.6-14 present estimated concentrations and depositions based on the monitoring program. 11.6.2.6 Program Statistical Sensitivity The activity in environmental samples is low after dilution and dispersion of radionuclides released by the power plant. For many isotopes, the radioactivity is below the lower limits of detection (LLD) that are listed in Table 11.6-11. With dose estimated from effluent data as shown in Sections 11.2 and 11.3, much lower dose levels can be estimated even though large errors may be introduced in the dispersion modeling. Thus, doses estimated using effluent data provide a more detailed definition of the dose increments due to the operation of DCPP than dose estimates calculated from the environmental measurements. Counting errors for effluent data and errors associated with the calculational models are used to determine the overall sensitivity of estimated dose. 11.6.3 SAFETY EVALUATION 11.6.3.1 10 CFR Part 50, Appendix I - Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as is Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents The ODCM provides data on measurable levels of radiation and radioactive materials in the environment. The data are reported in the Annual Radiological Environmental Operating Report. Annual land use censuses are conducted in the vicinity of DCPP, and reported in the Annual Radiological Environmental Operating Report, in order to provide compliance with 10 CFR 50 Part Appendix I, Section IV(B)(3). 11.
6.4 REFERENCES
- 1. 1984 Annual Environmental Radiological Report, Diablo Canyon Power Plant, Pacific Gas and Electric Company, San Ramon, CA, Report 411-85.123, 1985.
11.6-5 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UPDATE
- 2. Deleted in Revision 1.
- 3. Deleted in Revision 1.
- 4. Deleted in Revision 1.
- 5. Deleted in Revision 1.
- 6. Deleted in Revision 23.
- 7. Regulatory Guide 1.109, Revision 1, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, USNRC, October 1977.
- 8. Preoperational Radiological Environmental Report, Diablo Canyon Power Plant, Pacific Gas and Electric Company, San Ramon, CA, Report 411-84.530, 1985.
11.6-6 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.0-1 Sheet 1 of 4 APPLICABLE DESIGN BASIS CRITERIA CRITERION TITLE APPLICABILITY Radioactive Waste Management Source Liquid Gaseous Process and Solid Offsite Terms(a) Radwaste Radwaste Effluent Radwaste Radiological System System Radiological System Monitoring Monitoring Program System Section 11.1 11.2 11.3 11.4 11.5 11.6
- 1. General Design Criteria Criterion 2, 1967 Performance Standards X X X Criterion 3, 1971 Fire Protection X X X X Criterion 4, 1967 Sharing of Systems X X X X Criterion 11, 1967 Control Room X X X Criterion 12, 1967 Instrumentation and Control System X X X Monitoring Reactor Coolant Pressure Criterion 16, 1967 X Boundary Criterion 17, 1967 Monitoring Radioactivity Releases X X X X Criterion 18, 1967 Monitoring Fuel and Waste Storage X X X Criterion 19, 1971 Control Room X Criterion 21, 1967 Single Failure Definition X Criterion 49, 1967 Containment Design Basis X Criterion 54, 1971 Piping Systems Penetrating Containment X Criterion 56, 1971 Primary Containment Isolation X Fuel and Waste Storage Radiation Criterion 68, 1967 X X X Shielding Protection Against Radioactivity Release Criterion 69, 1967 X X X from Spent Fuel and Waste Storage Control of Releases of Radioactivity to the Criterion 70, 1967 X X X X Environment Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.0-1 Sheet 2 of 4 CRITERION TITLE APPLICABILITY Radioactive Waste Management Source Liquid Gaseous Process and Solid Offsite Terms(a) Radwaste Radwaste Effluent Radwaste Radiological System System Radiological System Monitoring Monitoring Program System Section 11.1 11.2 11.3 11.4 11.5 11.6
- 2. System Safety Function Requirements Protection from Missiles and Dynamic Effects X X
- 3. Code of Federal Regulations Standards for Protection Against 10 CFR Part 20 X X X X Radiation National Fire Protection Association 10 CFR 50.48(c) X Standard NFPA 805 Environmental Qualification of Electric 10 CFR 50.49 Equipment Important to Safety for X X Nuclear Power Plants 10 CFR 50.55a(f) Inservice Testing Requirements X 10 CFR 50.55a(g) Inservice Inspection Requirements X 10 CFR 50.68(b) Criticality Accident Requirements X Numerical Guides for Design Objectives and Limiting Conditions for Operation to 10 CFR Part 50, Meet the Criterion "As Low as is X X X X Appendix I Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents 10 CFR 61.55 Waste Classification X Packaging and Transportation of 10 CFR Part 71 X Radioactive Material Environmental Radiation Protection 40 CFR Part 190 X X X Standards for Nuclear Power Operations Subchapter C-Hazardous Materials 49 CFR Parts 171-178 X Regulations Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.0-1 Sheet 3 of 4 CRITERION TITLE APPLICABILITY Radioactive Waste Management Source Liquid Gaseous Process and Solid Offsite Terms(a) Radwaste Radwaste Effluent Radwaste Radiological System System Radiological System Monitoring Monitoring Program System Section 11.1 11.2 11.3 11.4 11.5 11.6
- 4. Atomic Energy Commission (AEC) Safety Guides Safety Guide 13, Fuel Storage Facility Design Basis X March 1971
- 5. Regulatory Guides Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Regulatory Guide 1.21, Releases of Radioactive Materials in X Revision 1, June 1974 Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants Regulatory Guide 1.45, Reactor Coolant Pressure Boundary X
May 1973 Leakage Detection Systems Instrumentation for Light-Water-Cooled Regulatory Guide 1.97, Nuclear Power Plants to Assess Plant X X X Revision 3, May 1983 and Environs Conditions During and Following an Accident Design Guidance for Radioactive Waste Regulatory Guide Management Systems, Structures, and 1.143, Revision 1, X X X Components Installed In Light-Water-October 1979 Cooled Nuclear Power Plants
- 6. NRC NUREGs NUREG-0737, Clarification of TMI Action Plan X X X November 1980 Requirements
- 7. NRC Generic Letters Assurance of Equipment Operability and Generic Letter 96-06, Containment Integrity During Design- X September 1996 Basis Accident Conditions Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.0-1 Sheet 4 of 4 CRITERION TITLE APPLICABILITY Radioactive Waste Management Source Liquid Gaseous Process and Solid Offsite Terms(a) Radwaste Radwaste Effluent Radwaste Radiological System System Radiological System Monitoring Monitoring Program System Section 11.1 11.2 11.3 11.4 11.5 11.6
- 8. Bulletins IE Bulletin 79-19, Packaging of Low-Level Radioactive X
August 1979 Waste for Transport and Burial (a) There are no design basis criteria applicable to the source terms Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-1 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED LIBRARY OF PHYSICAL DATA FOR ISOTOPES Number Nuclide Half-life, Yield Beta Energy, Gamma Energy, Decay Const., hours Fract MeV/Dis MeV/Dis hr-1 1 H-3 0.108E 06 0.800E-04 0.00620 0.0 0.642E-05 2 Cr-51 0.667E 03 0.0 0.00055 0.02900 0.104E-02 3 Mn-54 0.727E 04 0.0 0.00590 0.83500 0.953E-04 4 Fe-55 0.648E 02 0.0 0.0 0.00600 0.107E-01 5 Co-58 0.171E 04 0.0 0.03100 0.98100 0.405E-03 6 Fe-59 0.108E 04 0.0 0.12900 1.17000 0.642E-03 7 Co-60 0.461E 05 0.0 0.10400 2.49000 0.150E-04 8 Kr-83M 0.186E 01 0.470E-02 0.03900 0.00050 0.373E 00 9 Kr-85M 0.440E 01 0.103E-01 0.25200 0.16000 0.157E 00 10 Kr-85 0.941E 05 0.0 0.22100 0.00200 0.736E-05 11 Kr-87 0.127E 01 0.194E-01 1.34000 0.76400 0.546E 00 12 Kr-88 0.277E 01 0.279E-01 0.37200 2.03000 0.250E 00 13 Sr-89 0.123E 04 0.369E-01 0.55600 0.0 0.563E-03 14 Sr-90 0.245E 06 0.455E-01 0.16900 0.0 0.283E-05 15 Y-90 0.639E 02 0.0 0.91200 0.0 0.208E-01 16 Sr-91 0.972E 01 0.461E-01 0.62400 0.84000 0.713E-01 17 Y-91 0.147E 04 0.120E-02 0.59300 0.00400 0.471E-03 18 Sr-92 0.270E 01 0.453E-01 0.21400 1.29000 0.257E 00 19 Y-92 0.360E 01 0.390E-02 1.39000 0.48500 0.192E 00 20 Zr-95 0.157E 04 0.585E-01 0.11100 0.73900 0.441E-03 21 Nb-95 0.841E 03 0.136E-02 0.04500 0.76000 0.824E-03 22 Mo-99 0.680E 02 0.607E-01 0.40500 0.12600 0.102E-01 23 I-131 0.193E 03 0.319E-01 0.18300 0.39200 0.359E-02 24 Te-132 0.779E 02 0.464E-01 0.06100 0.23100 0.890E-02 25 I-132 0.240E 01 0.530E-03 0.48500 2.28000 0.289E 00 26 I-133 0.210E 02 0.620E-01 0.49300 0.62400 0.330E-01 27 Xe-133M 0.552E 02 0.0 0.20700 0.02100 0.126E-01 28 Xe-133 0.127E 03 0.0 0.15500 0.04500 0.546E-02 29 Cs-134 0.180E 05 0.410E-04 0.16800 1.57000 0.385E-04 30 I-134 0.866E 00 0.764E-01 0.94100 2.58000 0.800E 00 31 I-135 0.670E 01 0.600E-01 0.31600 1.56000 0.103E 00 32 Xe-135M 0.260E 00 0.0 0.10400 0.42100 0.267E 01 33 Xe-135 0.920E 01 0.313E-02 0.30400 0.26200 0.753E-01 34 Cs-136 0.312E 03 0.377E-03 0.11900 2.21000 0.222E-02 35 Cs-137 0.236E 06 0.633E-01 0.17300 0.56200 0.294E-05 36 Xe-138 0.233E 00 0.558E-01 0.5900 1.28000 0.297E 01 37 Ba-140 0.307E 03 0.596E-01 0.27400 0.21200 0.226E-02 38 La-140 0.401E 02 0.103E-02 0.43900 2.31000 0.173E-01 39 Ce-144 0.685E 04 0.485E-01 0.09300 0.01600 0.101E-03 40 Pr-144 0.292E 00 0.0 1.20000 0.06400 0.237E 01 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED BASIC ASSUMPTIONS FOR CORE AND COOLANT INVENTORIES FOR DESIGN BASIS CASE Reactor core thermal power, mw 3568.0 Duration of cycle, hr 8760.0 Capacity factor during period 0.800 Number of fissions per megawatt-second 0.315E17 Total mass of uranium in core, lb 1.97E5 Total mass of plutonium in core, lb 6.05E2 Reload uranium enrichment, percent 3.18 Reload mass of fissile plutonium, lb 0.0 Primary-to-secondary leakrate, gpm 0.0 Primary coolant leakage to containment, gpm 0.0 Primary coolant leakage to auxiliary building, gpm 0.0 Fraction of fuel with defective cladding 0.01 Weight of water in primary system, lb 5.66E5 Volume of water in primary system, gal. 9.40E4 Letdown flowrate, gpm 75.0 Capacity factor of primary cation demineralizer 0.1 Average shim bleed flowrate, gpm 1.0 Fraction of shim bleed flow discharged to environment 0.667 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-2A BASIC ASSUMPTIONS FOR UPDATED CORE AND 1% FUEL DEFECTS COOLANT INVENTORIES Reactor core thermal power, MWth 3580 Duration of cycle, months (based on EFPDs) 19 Capacity factor during period 1.0 Maximum Core Average Burnup 50 GWD/MTU Total mass of uranium in core, MTU 81.37 Number of Cycles 3 cycles in a 5 year period Uranium enrichment, percent 4.2 to 5.0 Primary-to-secondary leakrate, gpm 0.0 Primary coolant leakage to containment, gpm 0.0 Primary coolant leakage to auxiliary building, gpm 0.0 Fraction of fuel with defective cladding 0.01 Weight of water in primary system, lb 6.07E5 Volume of water in primary system, gal 7.37E4 Letdown flowrate, gpm 75.0 Capacity factor of primary cation demineralizer 0.1 Average shim bleed flowrate, gpm 1.0 DF for Mixed-bed demineralizer - Cs, Rb 1 - Others 10 DF for Cation-bed Demineralizer - Cs, Rbr 10 - Other 1 Fission product escape rate coefficient (sec -1) - Noble gases 6.5E-08 - I, Br, Rb, Cs 1.3E-08 - Mo 2.0E-09 - Te 1.0E-09 - Sr, Ba 1.0E-11 - All others 1.6E-12 Corrosion product concentrations in RCS (µCi/g) 3 times expected values from NUREG 0017 R1 Tritium Concentrations (µCi/g) 3.5 times expected values from NUREG 0017 R1 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-3 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED BASIC ASSUMPTIONS FOR CORE AND COOLANT INVENTORIES FOR NORMAL OPERATION CASE Reactor core thermal power, mw 3568.0 Duration of cycle, hr 8760.0 Capacity factor during period 0.800 Number of fissions per megawatt-second 0.315E17 Total mass of uranium in core, lb 1.97E5 Total mass of plutonium in core, lb 6.05E2 Reload uranium enrichment, percent 3.18 Reload mass of fissile plutonium, lb 0.0 Primary-to-secondary leakrate, gpm 0.0115 Primary coolant leakage to containment, gpm 0.0385 Primary coolant leakage to auxiliary building, gpm 0.0184 Fraction of fuel with defective cladding 0.0012 Weight of water in primary system, lb 5.66E5 Volume of water in primary system, gal. 9.40E4 Letdown flowrate, gpm 75.0 Capacity factor of primary cation demineralizer 0.1 Average shim bleed flowrate, gpm 1.0 Fraction of shim bleed flow discharged to environment 0.667 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-4 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED CORE ACTIVITY INVENTORIES FOR DESIGN BASIS CASE (CURIES) Nuclide Initial Act. Produced Decayed Lkge to Coolt. Inventory Equil Inven. Curies/Megawatt Cr-51 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Mn-54 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Fe-55 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Co-58 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Fe-59 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Co-60 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Kr-83M 0.0 0.4660E 11 0.4658E 11 0.2340E 06 0.1428E 08 0.1428E 08 0.4001E 04 Kr-85M 0.0 0.4317E 11 0.4314E 11 0.5127E 06 0.3129E 08 0.3129E 08 0.8769E 04 Kr-85 0.4007E 06 0.4639E 06 0.3932E 05 0.1249E 05 0.8128E 06 0.5461E 07 0.2278E 03 Kr-87 0.0 0.2817E 12 0.2816E 12 0.9662E 06 0.5893E 08 0.5893E 08 0.1652E 05 Kr-88 0.0 0.1857E 12 0.1856E 12 0.1389E 07 0.8475E 08 0.8475E 08 0.2375E 05 Sr-89 0.4946E 06 0.5532E 09 0.4910E 09 0.2510E 03 0.1116E 09 0.1121E 09 0.3129E 05 Sr-90 0.3073E 07 0.3425E 07 0.1173E 06 0.1194E 02 0.6380E 07 0.1382E 09 0.1786E 04 Y-90 0.3073E 07 0.4497E 09 0.4464E 09 0.1897E 01 0.6346E 07 0.1382E 09 0.1779E 04 Sr-91 0.4745E-14 0.8746E 11 0.8732E 11 0.3527E 03 0.1400E 09 0.1400E 09 0.3925E 05 Y-91 0.6767E 08 0.5924E 09 0.5177E 09 0.5060E 02 0.1424E 09 0.1437E 09 0.3992E 05 Sr-92 0.0 0.3094E 12 0.3093E 12 0.3470E 03 0.1376E 09 0.1376E 09 0.3857E 05 Y-92 0.0 0.2519E 12 0.2518E 12 0.6027E 02 0.1495E 09 0.1495E 09 0.4189E 05 Zr-95 0.8460E 08 0.6871E 09 0.5960E 09 0.6221E 02 0.1758E 09 0.1777E 09 0.4926E 05 Nb-95 0.1083E 09 0.1142E 10 0.1073E 10 0.6000E 02 0.1777E 09 0.1818E 09 0.4981E 05 Mo-99 0.8000E 05 0.1646E 11 0.1628E 11 0.9199E 05 0.1844E 09 0.1844E 09 0.5168E 05 I-131 0.4937E 07 0.3048E 10 0.2956E 10 0.3082E 06 0.9689E 08 0.9689E 08 0.2715E 05 Te-132 0.1553E 06 0.1098E 11 0.1084E 11 0.3510E 05 0.1409E 09 0.1409E 09 0.3950E 05 I-132 0.1602E 06 0.3560E 12 0.3559E 12 0.4614E 06 0.1426E 09 0.1426E 09 0.3995E 05 I-133 0.6024E-02 0.5444E 11 0.5425E 11 0.6155E 06 0.1883E 09 0.1883E 09 0.5278E 05 Xe-133M 0.1593E 04 0.4954E 09 0.4908E 09 0.7318E 05 0.4519E 07 0.4519E 07 0.1267E 04 Xe-133 0.3309E 07 0.8969E 10 0.8780E 10 0.3012E 07 0.1882E 09 0.1882E 09 0.5276E 05 Cs-134 0.1460E 01 0.3605E 07 0.1284E 07 0.1248E 05 0.3116E 07 0.3116E 07 0.8734E 03 I-134 0.0 0.1627E 13 0.1627E 13 0.7610E 06 0.2321E 09 0.2321E 09 0.6504E 05 I-135 0.0 0.1651E 12 0.1650E 12 0.5971E 06 0.1823E 09 0.1823E 09 0.5108E 05 Xe-135M 0.0 0.6164E 12 0.6163E 12 0.4329E 06 0.2643E 08 0.2643E 08 0.7407E 04 Xe-135 0.9747E-15 0.1264E 12 0.1838E 11 0.4567E 06 0.2789E 08 0.2789E 08 0.7816E 04 Cs-136 0.1542E 06 0.7228E 08 0.2129E 08 0.3588E 04 0.1145E 07 0.1145E 07 0.3209E 03 Cs-137 0.4430E 07 0.4946E 07 0.1753E 06 0.2235E 05 0.9173E 07 0.1658E 09 0.2571E 04 Xe-138 0.0 0.4416E 13 0.4416E 13 0.2779E 07 0.1695E 09 0.1695E 09 0.4751E 05 Ba-140 0.2375E 08 0.3580E 10 0.3423E 10 0.4367E 03 0.1810E 09 0.1610E 09 0.5074E 05 La-140 0.2732E 08 0.2668E 11 0.2652E 11 0.7071E 02 0.1642E 09 0.1842E 09 0.5162E 05 Ce-144 0.6129E 08 0.1306E 09 0.7999E 08 0.3644E 02 0.1119E 09 0.1473E 09 0.3135E 05 Pr-144 0.6127E 08 0.1877E 13 0.1876E 13 0.3643E 02 0.1119E 09 0.1473E 09 0.3135E 05 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-5 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED CORE ACTIVITY INVENTORIES FOR NORMAL OPERATION CASE (CURIES) Nuclide Initial Act. Produced Decayed Lkge to Coolt. Inventory Equil Inven. Curies/Megawatt Cr-51 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Mn-54 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Fe-55 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Co-58 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Fe-59 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Co-60 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Kr-83M 0.0 0.4660E 11 0.4658E 11 0.2809E 05 0.1428E 08 0.1428E 08 0.4001E 04 Kr-85M 0.0 0.4317E 11 0.4314E 11 0.6152E 05 0.3129E 08 0.3129E 08 0.8769E 04 Kr-85 0.4007E 06 0.4639E 06 0.3963E 05 0.1511E 04 0.8235E 06 0.6932E 07 0.2308E 03 Kr-87 0.0 0.2817E 12 0.2816E 12 0.1159E 06 0.5893E 08 0.5893E 08 0.1652E 05 Kr-88 0.0 0.1857E 12 0.1857E 12 0.1667E 06 0.8475E 08 0.8475E 08 0.2375E 05 Sr-89 0.4946E 08 0.5532E 09 0.4910E 09 0.3012E 02 0.1116E 09 0.1121E 09 0.3129E 05 Sr-90 0.3073E 07 0.3425E 07 0.1173E 06 0.1433E 01 0.6380E 07 0.1382E 09 0.1788E 04 Y-90 0.3073E 07 0.4497E 09 0.4465E 09 0.2276E 00 0.6346E 07 0.1382E 09 0.1779E 04 Sr-91 0.4745E-14 0.8746E 11 0.8732E 11 0.4233E 02 0.1400E 09 0.1400E 09 0.3925E 05 Y-91 0.6767E08 0.5924E 09 0.5177E 09 0.6072E 01 0.1424E 09 0.1437E 09 0.3992E 05 Sr-92 0.0 0.3094E 12 0.3093E 12 0.4164E 02 0.1376E 09 0.1376E 09 0.3857E 05 Y-92 0.0 0.2519E 12 0.2518E 12 0.7232E 01 0.1495E 09 0.1495E 09 0.4189E 05 Zr-95 0.8460E 08 0.6871E 09 0.5960E 09 0.7466E 01 0.1758E 09 0.1777E 09 0.4926E 05 Nb-95 0.1083E 09 0.1142E 10 0.1073E 10 0.7200E 01 0.1777E 09 0.1818E 09 0.4981E 05 Mo-99 0.8000E 05 0.1646E 11 0.1628E 11 0.1104E 05 0.1844E 09 0.1844E 09 0.5168E 05 I-131 0.4937E 07 0.3048E 10 0.2956E 10 0.3699E 05 0.9690E 08 0.9690E 08 0.2716E 05 Te-132 0.1553E 06 0.1098E 11 0.1084E 11 0.4212E 04 0.1409E 09 0.1409E 09 0.3950E 05 I-132 0.1602E 06 0.3560E 12 0.3559E 12 0.5537E 05 0.1426E 09 0.1426E 09 0.3995E 05 I-133 0.6024E 02 0.5444E 11 0.5425E 11 0.7387E 05 0.1883E 09 0.1883E 09 0.5278E 05 Xe-133M 0.1593E 04 0.4954E 09 0.4908E 09 0.8783E 04 0.4520E 07 0.4520E 07 0.1267E 04 Xe-133 0.3309E 07 0.8969E 10 0.8784E 10 0.3616E 06 0.1883E 09 0.1883E 09 0.5278E 05 Cs-134 0.1460E 01 0.3605E 07 0.1298E 07 0.1514E 04 0.3121E 07 0.3121E 07 0.8749E 03 I-134 0.0 0.1627E 13 0.1627E 13 0.9132E 05 0.2321E 09 0.2321E 09 0.6504E 05 I-135 0.0 0.1651E 12 0.1650E 12 0.7165E 05 0.1823E 09 0.1823E 09 0.5108E 05 Xe-135M 0.0 0.6164E 12 0.6163E 12 0.5195E 05 0.2643E 08 0.2643E 08 0.7407E 04 Xe-135 0.9747E-15 0.1264E 12 0.1838E 11 0.5481E 05 0.2789E 08 0.2789E 08 0.7816E 04 Cs-136 0.1542E 06 0.2228E 08 0.2129E 08 0.4307E 03 0.1145E 07 0.1145E 07 0.3210E 03 Cs-137 0.4430E 07 0.4946E 07 0.1755E 06 0.2686E 04 0.9197E 07 0.1887E 09 0.2578E 04 Xe-138 0.0 0.4416E 13 0.4416E 13 0.3335E 06 0.1695E 09 0.1695E 09 0.4751E 05 Ba-140 0.2375E 08 0.3580E 10 0.3423E 10 0.5240E 02 0.1810E 09 0.1810E 09 0.5074E 05 La-140 0.2732E 08 0.2668E 11 0.2652E 11 0.8486E 01 0.1842E 09 0.1842E 09 0.5162E 05 Ce-144 0.6129E 08 0.1306E 09 0.7999E 08 0.4372E 01 0.1119E 09 0.1473E 09 0.3135E 05 Pr-144 0.6127E 08 0.1877E 13 0.1876E 13 0.4372E 01 0.1119E 09 0.1473E 09 0.3135E 05 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-6 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED BASIC ASSUMPTIONS FOR FUEL ROD GAP INVENTORIES Percent of Core Fuel Within Given Fuel Temperature Temperature Range(a) Power, MWt Range, °F 0.0 0.1961 >3400 0.1 3.1373 3400 - 3200 0.3 10.3922 3200 - 3000 0.7 25.1 3000 - 2800 1.6 58.333 2800 - 2600 2.9 104.61 2600 - 2400 4.3 152.55 2400 - 2200 5.9 211.275 2200 - 2000 84.1 2999.02 <2000 (a) Based on hot channel factors of FH = 1.70 and Fq = 2.82. Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-7 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ACTIVITY IN FUEL ROD GAPS Nuclide Pellet Release Fraction Gap Inventory, Ci Kr-83M 0.000824 0.118E 05 Kr-85M 0.001240 0.388E 05 Kr-85 0.167000 0.138E 06 Kr-87 0.000668 0.394E 05 Kr-88 0.000998 0.846E 05 I-131 0.008220 0.797E 06 I-132 0.000901 0.128E 06 I-133 0.002710 0.510E 06 Xe-133M 0.004370 0.198E 05 Xe-133 0.006670 0.126E 07 I-134 0.000557 0.129E 06 I-135 0.001540 0.281E 06 Xe-135M 0.000303 0.801E 04 Xe-135 0.001800 0.502E 05 Xe-138 0.000316 0.536E 05 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-8 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED INPUT CONSTANTS FOR COOLANT ACTIVITIES FOR DESIGN BASIS CASE Fuel to Coolant, Fuel Escape Rate, Pur. Rate, P-S Leak Rate, Prim-Cont Lk Rate, Prim-Aux Lk Rate Nuclide Ci/hr sec-1 hr-1 hr-1 hr-1 hr-1 H-3 0.0 0.0 0.5897E-03 0.0 0.0 0.0 Cr-51 0.3010E-01 0.0 0.6034E-01 0.0 0.0 0.0 Mn-54 0.4820E-02 0.0 0.6034E-01 0.0 0.0 0.0 Fe-55 0.2890E-01 0.0 0.6034E-01 0.0 0.0 0.0 Co-58 0.2500E 00 0.0 0.6034E-01 0.0 0.0 0.0 Fe-59 0.1570E-01 0.0 0.6034E-01 0.0 0.0 0.0 Co-60 0.3090E-01 0.0 0.6034E-01 0.0 0.0 0.0 Kr-83M 0.3340E 02 0.6500E-07 0.8840E-03 0.0 0.0 0.0 Kr-85M 0.7316E 02 0.6500E-07 0.8840E-03 0.0 0.0 0.0 Kr-85 0.1426E 01 0.6500E-07 0.8840E-03 0.0 0.0 0.0 Kr-87 0.1379E 03 0.6500E-07 0.8840E-03 0.0 0.0 0.0 Kr-88 0.1982E 03 0.6500E-07 0.8840E-03 0.0 0.0 0.0 Sr-89 0.3582E-01 0.1000E-10 0.6034E-01 0.0 0.0 0.0 Sr-90 0.1704E-02 0.1000E-10 0.6034E-01 0.0 0.0 0.0 Y-90 0.2707E-03 0.1600E-11 0.8840E-03 0.0 0.0 0.0 Sr-91 0.5033E-01 0.1000E-10 0.6034E-01 0.0 0.0 0.0 Y-91 0.7220E-02 0.1600E-11 0.8840E-03 0.0 0.0 0.0 Sr-92 0.4952E-01 0.1000E-10 0.6034E-01 0.0 0.0 0.0 Y-92 0.8600E-02 0.1600E-11 0.8840E-03 0.0 0.0 0.0 Zr-95 0.8878E-02 0.1600E-11 0.6034E-01 0.0 0.0 0.0 Nb-95 0.8562E-02 0.1600E-11 0.6034E-01 0.0 0.0 0.0 Mo-99 0.1313E 02 0.2000E-08 0.8840E-03 0.0 0.0 0.0 I-131 0.4398E 02 0.1300E-07 0.5976E-01 0.0 0.0 0.0 Te-132 0.5009E 01 0.1000E-08 0.5976E-01 0.0 0.0 0.0 I-132 0.6584E 02 0.1300E-07 0.5976E-01 0.0 0.0 0.0 I-133 0.8783E 02 0.1300E-07 0.5576E-01 0.0 0.0 0.0 Xe-133M 0.1044E 02 0.6500E-07 0.8840E-03 0.0 0.0 0.0 Xe-133 0.4298E 03 0.6500E-07 0.8840E-03 0.0 0.0 0.0 Cs-134 0.1781E 01 0.1300E-07 0.3637E-01 0.0 0.0 0.0 I-134 0.1086E 03 0.1300E-07 0.5976E-01 0.0 0.0 0.0 I-135 0.8520E 02 0.1300E-07 0.5976E-01 0.0 0.0 0.0 Xe-135M 0.6177E 02 0.6500E-07 0.8840E-03 0.0 0.0 0.0 Xe-135 0.6517E 02 0.6500E-07 0.8840E-03 0.0 0.0 0.0 Cs-136 0.5120E 00 0.1300E-07 0.3637E-01 0.0 0.0 0.0 Cs-137 0.3189E 01 0.1300E-07 0.3637E-01 0.0 0.0 0.0 Xe-138 0.3988E 03 0.6500E-07 0.8840E-01 0.0 0.0 0.0 Ba-140 0.8231E-01 0.1000E-10 0.6034E-01 0.0 0.0 0.0 La-140 0.1009E-01 0.1000E-11 0.6034E-01 0.0 0.0 0.0 Ce-144 0.5199E-02 0.1600E-11 0.6034E-01 0.0 0.0 0.0 Pr-144 0.5199E-02 0.1600E-11 0.6034E-01 0.0 0.0 0.0 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-9 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED INPUT CONSTANTS FOR COOLANT ACTIVITIES FOR NORMAL OPERATION CASE Fuel to Coolant, Fuel Escape Rate, Pur. Rate, P-S Leak Rate, Prim-Cont Lk Rate, Prim-Aux Lk Rate Nuclide Ci/hr sec-1 hr-1 hr-1 hr-1 hr-1 H-3 0.0 0.0 0.5897E-03 0.7340E-05 0.2457E-04 0.1174E-04 Cr-51 0.3010E-01 0.0 0.6034E-01 0.7340E-05 0.2457E-04 0.1174E-04 Mn-54 0.4820E-02 0.0 0.6034E-01 0.7340E-05 0.2457E-04 0.1174E-04 Fe-55 0.2890E-01 0.0 0.6034E-01 0.7340E-05 0.2457E-04 0.1174E-04 Co-58 0.2500E 00 0.0 0.6034E-01 0.7340E-05 0.2457E-04 0.1174E-04 Fe-59 0.1570E-01 0.0 0.6034E-01 0.7340E-05 0.2457E-04 0.1174E-04 Co-60 0.3090E-01 0.0 0.6034E-01 0.7340E-05 0.2457E-04 0.1174E-04 Kr-83M 0.4008E 01 0.6500E-07 0.8840E-03 0.7340E-05 0.4167E-03 0.1174E-04 Kr-85M 0.8779E 01 0.6500E-07 0.8840E-03 0.7340E-05 0.4167E-03 0.1174E-04 Kr-85 0.1725E 00 0.6500E-07 0.8840E-03 0.73403-05 0.4167E-03 0.1174E-04 Kr-87 0.1854E 02 0.6500E-07 0.8840E-03 0.7340E-05 0.4167E-03 0.1174E-04 Kr-88 0.2379E 02 0.6500E-07 0.8840E-03 0.7340E-05 0.4167E-03 0.1174E-04 Sr-89 0.4298E-02 0.1000E-10 0.6034E-01 0.7340E-05 0.2457E-04 0.1174E-04 Sr-90 0.2045E-03 0.1000E-10 0.6034E-01 0.7340E-05 0.2457E-04 0.1174E-04 Y-90 0.3248E-04 0.1600E-11 0.8840E-03 0.7340E-05 0.2457E-04 0.1174E-04 Sr-91 0.6040E-02 0.1000E-10 0.6034E-01 0.7340E-05 0.2457E-04 0.1174E-04 Y-91 0.8664E-03 0.1600E-11 0.8840E-03 0.7340E-05 0.2457E-04 0.1174E-04 Sr-92 0.5942E-02 0.1000E-10 0.6034E-01 0.7340E-05 0.2457E-04 0.1174E-04 Y-92 0.1032E-02 0.1600E-11 0.8840E-03 0.7340E-05 0.2457E-04 0.1174E-04 Zr-95 0.1065E-02 0.1600E-11 0.6034E-01 0.7340E-05 0.2457E-04 0.1174E-04 Mb-95 0.1027E-02 0.1600E-11 0.6034E-01 0.7340E-05 0.2457E-04 0.1174E-04 Mb-99 0.1575E 01 0.2000E-08 0.8840E-03 0.7340E-05 0.2457E-04 0.1174E-04 I-131 0.5278E 01 0.1300E-07 0.5976E-01 0.7340E-05 0.2457E-04 0.1174E-04 Te-132 0.6011E 00 0.1000E-08 0.5976E-01 0.7340E-05 0.2457E-04 0.1174E-04 I-132 0.7901E 01 0.1300E-07 0.5976E-01 0.7340E-05 0.2457E-04 0.1174E-04 I-133 0.1054E 02 0.1300E-07 0.5976E-01 0.7340E-05 0.2457E-04 0.1174E-04 Me-133M 0.1253E 01 0.6500E-07 0.8840E-03 0.7340E-05 0.4167E-03 0.1174E-04 Mn-133 0.5160E 02 0.6500E-07 0.8840E-03 0.7340E-05 0.4167E-03 0.1174E-04 Cs-134 0.2161E 00 0.1300E-07 0.3637E-01 0.7340E-05 0.2457E-04 0.1174E-04 I-134 0.1303E 02 0.1300E-07 0.5976E-01 0.7340E-05 0.2457E-04 0.1174E-04 I-135 0.1022E 02 0.1300E-07 0.5976E-01 0.7340E-05 0.2457E-04 0.1174E-04 Xe-135M 0.7412E 01 0.6500E-07 0.8840E-03 0.7340E-05 0.4167E-03 0.1174E-04 Xe-135 0.7821E 01 0.6500E-07 0.8840E-03 0.7340E-05 0.4167E-03 0.1174E-04 Cs-136 0.6145E-01 0.1300E-07 0.3637E-01 0.7340E-05 0.2457E-04 0.1174E-04 Cs-137 0.3832E 00 0.1300E-07 0.3637E-01 0.7340E-05 0.2457E-04 0.1174E-04 Xe-138 0.4759E 02 0.6500E-07 0.8840E-03 0.7340E-05 0.4167E-03 0.1174E-04 Ba-140 0.7477E-02 0.1000E-10 0.6034E-01 0.7340E-05 0.2457E-04 0.1174E-04 La-140 0.1211E-02 0.1600E-11 0.6034E 01 0.73403-05 0.2457E-04 0.1174E-04 Ce-144 0.6239E-03 0.1600E-11 0.6034E-01 0.7340E-05 0.2457E-04 0.1174E-04 Pr-144 0.6239E-03 0.1600E-11 0.6034E-01 0.7340E-05 0.2457E-04 0.1174E-04 Note: The information presented in the above table was developed in support of the original license and is considered historical Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-10 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED BASIC DATA FOR CORROSION PRODUCT ACTIVITIES Core Wetted Areas, Effective, in2 Zirconium 9.42 x 106 Stainless steel 6.09 x 105 Inconel 1.01 x 106 Out-of-core Wetted Area, Inconel, in2 2.74 x 107 Coolant velocity, ft/sec Core 15.0 Steam generator 18.6 Nominal Base Metal Release Rates, mg/dm2-mo Zirconium 0.0 Stainless steel 0.5 Inconel 1.0 Coolant Crud Level, ppm 0.1 Permanent Crud Film, Nominal, mg/dm2 Incore 50 Out-of-core 50 Transient Crud Layer, Nominal, mg/dm2 Incore 50 Out-of-core 50 Total mass of metal in contact with primary coolant, lb 2.2 x 106 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-11 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED PRIMARY COOLANT ACTIVITIES FOR DESIGN BASIS CASE Nuclide Concentration, Ci/cc Activity, Ci H-3 0.7934E 00 0.2823E 03 Cr-51 0.1378E-02 0.4904E 00 Mn-54 0.2241E-03 0.7976E-01 Fe-55 0.1143E-02 0.4069E 00 Co-58 0.2600E-01 0.9250E-01 Fe-59 0.7236E-03 0.2575E 00 Co-60 0.1439E-02 0.5120E 00 Kr-83M 0.2513E 00 0.8943E 02 Kr-85M 0.1298E 01 0.4619E 03 Kr-85 0.4166E 01 0.1482E 04 Kr-87 0.7089E 00 0.2522E 03 Kr-88 0.2219E 01 0.7895E 03 Sr-89 0.1653E-02 0.5881E 00 Sr-90 0.7937E-04 0.2824E-01 Y-90 0.1382E-03 0.4919E-01 Sr-91 0.1075E-02 0.3824E 00 Y-91 0.1534E-01 0.5460E 01 Sr-92 0.4390E-03 0.1562E 00 Y-92 0.5619E-03 0.2000E 00 Zr-95 0.4105E-03 0.1461E 00 Nb-95 0.3989E-03 0.1420E 00 Mo-99 0.3331E 01 0.1185E 04 I-131 0.1951E 01 0.6941E 03 Te-132 0.2050E 00 0.7296E 02 I-132 0.7008E 00 0.2494E 03 I-133 0.2661E 01 0.9469E 03 Xe-133M 0.2243E 01 0.7983E 03 Xe-133 0.1947E 03 0.6927E 05 Cs-134 0.1375E 00 0.4893E 02 I-134 0.3549E 00 0.1263E 03 I-135 0.1467E 01 0.5221E 03 Xe-135M 0.2778E 00 0.9884E 02 Xe-135 0.3918E 01 0.1394E 04 Cs-136 0.3729E-01 0.1327E 02 Cs-137 0.2464E 00 0.8767E 02 Xe-138 0.3746E 00 0.1333E 03 Ba-140 0.2798E-02 0.9955E 00 La-140 0.9882E-03 0.3516E 00 Ce-144 0.2418E-03 0.8602E-01 Pr-144 0.2418E-03 0.8603E-01 Zn-65 0.8000E-02 0.2846E-01 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-11A UPDATED PRIMARY COOLANT ACTIVITIES (Based on 3580 MWt, 600 EFPDs, 1% Fuel Defects) Nuclide Activity Conc. Nuclide Activity Conc. Nuclide Activity Conc (µ/Ci/g) (µ/Ci/g) (µ/Ci/g) Kr-83m 3.77E-01 Y93 3.00E-04 Te134 2.92E-02 Kr-85m 1.36E+00 Y94 2.61E-05 Cs134m 5.29E-02 Kr-85 1.12E+01 Y95 1.11E-05 Cs134 7.39E+00 Kr-87 8.87E-01 Zr95 6.24E-04 Cs136 1.67E+00 Kr-88 2.49E+00 Zr97 3.87E-04 Cs137 4.21E+00 Kr-89 6.97E-02 Nb95m 6.97E-06 Cs138 9.99E-01 Xe-131m 4.06E+00 Nb95 6.30E-04 Cs139 8.96E-02 Xe-133m 4.18E+00 Nb97m 3.66E-04 Cs140 9.08E-03 Xe-133 2.81E+2 Nb97 4.12E-04 Cs142 1.07E-04 Xe-135m 8.14E-01 Mo99 7.65E-01 Ba137m 3.97E+00 Xe-135 7.89E+00 Mo101 2.05E-02 Ba139 7.85E-02 Xe-137 1.94E-01 Mo102 1.51E-02 Ba140 4.38E-03 Xe-138 6.54E-01 Mo105 7.67E-04 Ba141 1.21E-04 Tc99m 4.08E-01 Ba142 1.74E-04 Br83 7.20E-02 Tc101 2.00E-02 La140 1.43E-03 Br84 3.51E-02 Tc102 1.51E-02 La141 2.66E-04 Br85 3.67E-03 Tc105 7.85E-04 La142 2.34E-04 Br87 1.89E-03 Ru103 6.26E-04 La143 1.38E-05 I129 1.24E-07 Ru105 1.48E-04 Ce141 6.04E-04 I130 6.30E-02 Ru106 2.54E-04 Ce143 4.46E-04 I131 2.92E+00 Ru107 2.09E-06 Ce144 4.78E-04 I132 1.11E+00 Ru103m 5.69E-04 Ce145 2.08E-06 I133 4.30E+00 Rh105m 4.17E-05 Ce146 7.72E-06 I134 6.19E-01 Rh105 3.55E-04 Pr143 5.63E-04 I135 2.48E+00 Rh106 2.54E-04 Pr144 4.78E-04 I136 6.69E-03 Rh107 1.25E-05 Pr145 1.53E-04 Sn127 2.59E-06 Pr146 2.04E-05 Se81 5.84E-07 Sn128 5.16E-06 Nd147 2.42E-04 Se83 7.95E-07 Sn130 8.25E-07 Nd149 2.34E-05 Se84 4.58E-07 Sb127 3.16E-05 Nd151 1.72E-06 Rb86 6.40E-02 Sb128 2.89E-06 Pm147 1.19E-04 Rb88 2.54E+00 Sb129 3.92E-05 Pm149 2.12E-04 Rb89 1.41E-01 Sb130 2.75E-06 Pm151 6.02E-05 Rb90 1.11E-02 Sb131 1.16E-05 Sm151 6.11E-07 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-11A UPDATED PRIMARY COOLANT ACTIVITIES (Based on 3580 MWt, 600 EFPDs, 1% Fuel Defects) Nuclide Activity Conc. Nuclide Activity Conc. Nuclide Activity Conc (µ/Ci/g) (µ/Ci/g) (µ/Ci/g) Rb91 5.66E-03 Sb132 8.97E-07 Sm153 1.92E-04 Rb92 3.88E-04 Sb133 1.04E-06 Sr89 3.21E-03 Te125m 5.02E-04 Na24
- 1.41E-01 Sr90 2.30E-04 Te127m 3.50E-03 Cr51
- 9.30E-03 Sr91 1.36E-03 Te127 1.25E-02 Mn54
- 4.80E-03 Sr92 9.56E-04 Te129m 1.48E-02 Fe55
- 3.60E-03 Sr93 4.39E-05 Te129 1.64E-02 Fe59
- 9.00E-04 Sr94 7.48E-06 Te131m 3.71E-02 Co58
- 1.38E-02 Y90 6.32E-05 Te131 1.63E-02 Co60
- 1.59E-03 Y91m 7.46E-04 Te132 3.01E-01 Zn65
- 1.53E-03 Y91 4.59E-04 Te133m 1.92E-02 Np-239
- 6.60E-03 Y92 8.21E-04 Te133 8.00E-03
- Concentration values for corrosion products and Np-239 are assumed to be 3 times the expected values listed in NUREG-0017 R1.
Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-12 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED PRIMARY COOLANT ACTIVITIES FOR NORMAL OPERATION CASE Nuclide Concentration, Ci/cc Activity, Ci H-3 0.7377E 00 0.2625E 03 Cr-51 0.1377E-02 0.4901E 00 Mn-54 0.2240E-03 0.7970E-01 Fe-55 0.1143E-02 0.4066E 00 Co-58 0.1156E-01 0.4113E 01 Fe-59 0.7230E-03 0.2573E 00 Co-60 0.1438E-02 0.5116E 00 Kr-83M 0.3012E-01 0.1072E 02 Kr-85M 0.1553E 00 0.5528E 02 Kr-85 0.3579E 00 0.1273E 03 Kr-87 0.8500E-01 0.3025E 02 Kr-88 0.2658E 00 0.9458E 02 Sr-89 0.1982E-03 0.7052E-01 Sr-90 0.9517E-05 0.3387E-02 Y-90 0.1652E-04 0.5879E-02 Sr-91 0.1289E-03 0.4587E-01 Y-91 0.1784E-02 0.6347E 00 Sr-92 0.5267E-04 0.1874E-01 Y-92 0.6741E-04 0.2399E-01 Zr-95 0.4922E-04 0.1752E-01 Nb-95 0.4784E-04 0.1702E-01 Mo-99 0.3981E 00 0.1417E 03 I-131 0.2340E 00 0.8325E 02 Te-132 0.2459E-01 0.8749E 01 I-132 0.8408E-01 0.2992E 02 I-133 0.3192E 00 0.1136E 03 Xe-133M 0.2608E 00 0.9280E 02 Xe-133 0.2186E 02 0.7778E 04 Cs-134 0.1666E-01 0.5929E 01 I-134 0.4258E-01 0.1515E 02 I-135 0.1760E 00 0.6263E 02 Xe-135M 0.3332E-01 0.1186E 02 Xe-135 0.4674E 00 0.1663E 03 Cs-136 0.4470E-02 0.1591E 01 Cs-137 0.2958E-01 0.1052E 02 Xe-138 0.4495E-01 0.1599E 02 Ba-140 0.3355E-03 0.1194E 00 La-140 0.1185E-03 0.4215E-01 Ce-144 0.2899E-04 0.1032E-01 Pr-144 0.2899E-04 0.1032E-01 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-13 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED REACTOR COOLANT NITROGEN-16 ACTIVITY Location Activity, Ci/cc Core outlet 87 Reactor outlet nozzle 71 Steam generator inlet 67 Steam generator outlet 45 Reactor coolant pump inlet 43 Reactor coolant pump outlet 42 Reactor inlet nozzle 40 Core inlet 33 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-14 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED DEPOSITED CORROSION PRODUCT ACTIVITY IN STEAM GENERATOR Concentration, Ci/cm2 Operating Time, months Isotope 0 6 12 24 36 Mn-54 1.0 x 10-5 0.15 0.60 1.5 2.0 Mn-56 1.0 x 10-5 3.3 3.3 3.3 3.3 Co-58 1.0 x 10-2 4.5 10.2 11.0 11.0 Fe-59 1.0 x 10-4 1.4 3.0 3.0 3.0 Co-60 1.0 x 10-3 0.20 0.80 2.0 3.5 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-15 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED DEMINERALIZER AND EVAPORATOR DECONTAMINATION FACTORS Nuclide Primary Mxd Bed Primary Cation Letdown Mxd Bed Letdown Cation Letdown Anion BA Evap Feed Ion Exchangers(b) Waste Mxd. Beds(a) H-3 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 Cr-51 1.000E 01 1.000E 02 1.000E 02 1.000E 02 1.000E 00 1.000E 03 1.000E 03 Mn-54 1.000E 01 1.000E 02 1.000E 02 1.000E 02 1.000E 00 1.000E 03 1.000E 03 Fe-55 1.000E 01 1.000E 02 1.000E 02 1.000E 02 1.000E 00 1.000E 03 1.000E 03 Co-58 1.000E 01 1.000E 02 1.000E 02 1.000E 02 1.000E 00 1.000E 03 1.000E 03 Fe-59 1.000E 01 1.000E 02 1.000E 02 1.000E 02 1.000E 00 1.000E 03 1.000E 03 Co-60 1.000E 01 1.000E 02 1.000E 02 1.000E 02 1.000E 00 1.000E 03 1.000E 03 Kr-83M 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 Kr-85M 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 Kr-85 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 Kr-87 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 Kr-88 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 Sr-89 1.000E 01 1.000E 02 1.000E 02 1.000E 02 1.000E 00 1.000E 03 1.000E 03 Sr-90 1.000E 01 1.000E 02 1.000E 02 1.000E 02 1.000E 00 1.000E 03 1.000E 03 Y-90 1.000E 00 1.000E 00 1.000E 01 1.000E 00 1.000E 00 1.000E 03 1.000E 03 Sr-91 1.000E 01 1.000E 02 1.000E 02 1.000E 02 1.000E 00 1.000E 03 1.000E 03 Y-91 1.000E 00 1.000E 00 1.000E 01 1.000E 00 1.000E 00 1.000E 03 1.000E 03 Sr-92 1.000E 01 1.000E 02 1.000E 02 1.000E 02 1.000E 00 1.000E 03 1.000E 03 Y-92 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 03 1.000E 03 Zr-95 1.000E 01 1.000E 02 1.000E 02 1.000E 02 1.000E 00 1.000E 03 1.000E 03 Nb-95 1.000E 01 1.000E 02 1.000E 02 1.000E 02 1.000E 00 1.000E 03 1.000E 03 Mo-99 1.000E 00 1.000E 00 1.000E 02 1.000E 00 1.000E 00 1.000E 03 1.000E 03 I-131 1.000E 01 1.000E 00 1.000E 02 1.000E 00 1.000E 02 1.000E 02 1.000E 03 Te-132 1.000E 01 1.000E 00 1.000E 02 1.000E 00 1.000E 02 1.000E 03 1.000E 03 I-132 1.000E 01 1.000E 00 1.000E 02 1.000E 00 1.000E 02 1.000E 02 1.000E 03 I-133 1.000E 01 1.000E 00 1.000E 02 1.000E 00 1.000E 02 1.000E 02 1.000E 03 Xe-133M 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 Xe-133 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 Cs-134 2.000E 00 1.000E 01 2.000E 00 1.000E 01 1.000E 00 1.000E 03 2.000E 01 I-134 1.000E 01 1.000E 00 1.000E 02 1.000E 00 1.000E 02 1.000E 02 1.000E 03 I-135 1.000E 01 1.000E 00 1.000E 02 1.000E 00 1.000E 02 1.000E 02 1.000E 03 Xe-135M 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 Xe-135 1.000E 00 1.000E00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 Cs-136 2.000E 00 1.000E 01 2.000E 00 1.000E 01 1.000E 00 1.000E 03 2.000E 01 Cs-137 2.000E 00 1.000E 01 2.000E 00 1.000E 01 1.000E 00 1.000E 03 2.000E 01 Xe-138 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 1.000E 00 Ba-140 1.000E 01 1.000E 02 1.000E 02 1.000E 02 1.000E 00 1.000E 03 1.000E 03 La-140 1.000E 01 1.000E 02 1.000E 02 1.000E 02 1.000E 00 1.000E 03 1.000E 03 Ce-144 1.000E 01 1.000E 02 1.000E 02 1.000E 02 1.000E 00 1.000E 03 1.000E 03 Pr-144 1.000E 01 1.000E 02 1.000E 02 1.000E 02 1.000E 00 1.000E 03 1.000E 03 (a) Two waste mixed beds in series. (b) Boric Acid Evaporator has been abandoned in place. Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-16 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED PRODUCTION AND REMOVALS IN PRIMARY COOLANT FOR DESIGN BASIS CASE Nuclide Produced, Ci Decayed, Ci Cleaned Up, Ci Lked to Sec, Ci Lked to Cont, Ci Lked to Aux, Ci H-3 0.1642E 04 0.1463E 02 0.1345E 04 0.0 0.0 0.0 Cr-51 0.2637E 03 0.4455E 01 0.2070E 03 0.0 0.0 0.0 Mn-54 0.4222E 02 0.6647E-01 0.3366E 02 0.0 0.0 0.0 Fe-55 0.2532E 03 0.3805E 02 0.1718E 03 0.0 0.0 0.0 Co-58 0.2190E 04 0.1458E 02 0.1737E 04 0.0 0.0 0.0 Fe-59 0.1375E 03 0.1445E 01 0.1087E 03 0.0 0.0 0.0 Co-60 0.2707E 03 0.6729E-01 0.2161E 03 0.0 0.0 0.0 Kr-83M 0.2926E 06 0.2916E 06 0.5535E 03 0.0 0.0 0.0 Kr-85M 0.6409E 06 0.6359E 06 0.2855E 04 0.0 0.0 0.0 Kr-85 0.1250E 05 0.6654E 02 0.7987E 04 0.0 0.0 0.0 Kr-87 0.1208E 07 0.1205E 07 0.1562E 04 0.0 0.0 0.0 Kr-88 0.1736E 07 0.1728E 07 0.4885E 04 0.0 0.0 0.0 Sr-89 0.3138E 03 0.2897E 01 0.2482E 03 0.0 0.0 0.0 Sr-90 0.1493E 02 0.6984E-03 0.1192E 02 0.0 0.0 0.0 Y-90 0.5049E 01 0.4623E 01 0.3015E 00 0.0 0.0 0.0 Sr-91 0.4409E 03 0.2386E 03 0.1615E 03 0.0 0.0 0.0 Y-91 0.6483E 02 0.2065E 02 0.3098E 02 0.0 0.0 0.0 Sr-92 0.4338E 03 0.3511E 03 0.6602E 02 0.0 0.0 0.0 Y-92 0.3386E 03 0.3369E 03 0.1238E 01 0.0 0.0 0.0 Zr-95 0.7777E 02 0.5637E 00 0.6165E 02 0.0 0.0 0.0 Nb-95 0.7605E 02 0.1023E 01 0.5991E 02 0.0 0.0 0.0 Mo-99 0.1150E 06 0.1047E 06 0.7267E 04 0.0 0.0 0.0 I-131 0.3852E 06 0.2179E 05 0.2902E 06 0.0 0.0 0.0 Te-132 0.4388E 05 0.5676E 04 0.3050E 05 0.0 0.0 0.0 I-132 0.7610E 06 0.6303E 06 0.1044E 06 0.0 0.0 0.0 I-133 0.7694E 06 0.2734E 06 0.3961E 06 0.0 0.0 0.0 Xe-133M 0.9896E 05 0.8554E 05 0.4819E 04 0.0 0.0 0.0 Xe-133 0.3935E 07 0.3131E 07 0.4059E 06 0.0 0.0 0.0 Cs-134 0.1560E 05 0.1645E 02 0.1243E 05 0.0 0.0 0.0 I-134 0.9513E 06 0.8851E 06 0.5288E 05 0.0 0.0 0.0 I-135 0.7464E 06 0.4727E 06 0.2185E 06 0.0 0.0 0.0 Xe-135M 0.5840E 07 0.2304E 07 0.6113E 03 0.0 0.0 0.0 Xe-135 0.1519E 07 0.9150E 06 0.8591E 04 0.0 0.0 0.0 Cs-136 0.4485E 04 0.2574E 03 0.3372E 04 0.0 0.0 0.0 Cs-137 0.2793E 05 0.2248E 01 0.2227E 05 0.0 0.0 0.0 Xe-138 0.3474E 07 0.3473E 07 0.8258E 03 0.0 0.0 0.0 Ba-140 0.5458E 03 0.1965E 02 0.4202E 03 0.0 0.0 0.0 La-140 0.2388E 03 0.5309E 02 0.1483E 03 0.0 0.0 0.0 Ce-144 0.4554E 02 0.7609E-01 0.3631E 02 0.0 0.0 0.0 Pr-144 0.1831E 04 0.1785E 04 0.3631E 02 0.0 0.0 0.0 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-17 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED PRODUCTION AND REMOVALS IN PRIMARY COOLANT FOR NORMAL OPERATION CASE Nuclide Produced, Ci Decayed, Ci Cleaned Up, Ci Lked to Sec, Ci Lked to Cont, Ci Lked to Aux, Ci H-3 0.1642E 04 0.1384E 02 0.1271E 04 0.1583E 02 0.5299E 02 0.2532E 02 Cr-51 0.2637E 03 0.4452E 01 0.2068E 03 0.2516E-01 0.8424E-01 0.4026E-01 Mn-54 0.4222E 02 0.6643E-01 0.3364E 02 0.4092E-02 0.1370E-01 0.6547E-02 Fe-55 0.2532E 03 0.3803E 02 0.1717E 03 0.2088E-01 0.6991E-01 0.3341E-01 Co-58 0.2190E 04 0.1457E 02 0.1736E 04 0.2112E 00 0.7070E 00 0.3379E 00 Fe-59 0.1375E 03 0.1443E 01 0.1086E 03 0.1321E-01 0.4423E-01 0.2114E-01 Co-60 0.2707E 03 0.6725E-01 0.2159E 03 0.2627E-01 0.8794E-01 0.4203E-01 Kr-83M 0.3511E 05 0.3495E 05 0.6634E 02 0.5509E 00 0.3127E 02 0.8814E 00 Kr-85M 0.7691E 05 0.7610E 05 0.3417E 03 0.2837E 01 0.1611E 03 0.4540E 01 Kr-85 0.1512E 04 0.6269E 01 0.7526E 03 0.6249E 01 0.3547E 03 0.9998E 01 Kr-87 0.1449E 06 0.1445E 06 0.1873E 03 0.1555E 01 0.8826E 02 0.2488E 01 Kr-88 0.2084E 06 0.2070E 06 0.5852E 03 0.4859E 01 0.2758E 03 0.7774E 01 Sr-89 0.3765E 02 0.3474E 00 0.2976E 02 0.3621E-02 0.1212E-01 0.5794E-02 Sr-90 0.1791E 01 0.8376E-04 0.1429E 01 0.1739E-03 0.5821E-03 0.2782E-03 Y-90 0.6057E 00 0.5525E 00 0.3603E-01 0.2992E-03 0.1002E-02 0.4787E-03 Sr-91 0.5291E 02 0.2862E 02 0.1938E 02 0.2357E-02 0.7892E-02 0.3772E-02 Y-91 0.7779E 01 0.2407E 01 0.3611E 01 0.2999E-01 0.1004E 00 0.4798E-01 Sr-92 0.5205E 02 0.4212E 02 0.7922E 01 0.9637E-03 0.3226E-02 0.1542E-02 Y-92 0.4063E 02 0.4041E 02 0.1485E 00 0.1233E-02 0.4127E-02 0.1973E-02 Zr-95 0.9332E 01 0.6760E-01 0.7392E 01 0.8993E-03 0.3011E-02 0.1439E-02 Nb-95 0.9126E 01 0.1226E 00 0.7184E 01 0.8740E-03 0.2926E-02 0.1398E-02 Mo-99 0.1380E 05 0.1252E 05 0.8687E 03 0.7213E 01 0.2415E 02 0.1154E 02 I-131 0.4623E 05 0.2614E 04 0.3480E 05 0.4275E 01 0.1431E 02 0.6840E 01 Te-132 0.5285E 04 0.6807E 03 0.3658E 04 0.4493E 00 0.1504E 01 0.7189E 00 I-132 0.9131E 05 0.7562E 05 0.1252E 05 0.1538E 01 0.5148E 01 0.2461E 01 I-133 0.9233E 05 0.3279E 05 0.4751E 05 0.5835E 01 0.1954E 02 0.9336E 01 Xe-133M 0.1188E 05 0.9952E 04 0.5607E 03 0.4655E 01 0.2643E 03 0.7448E 01 Xe-133 0.4722E 06 0.3629E 06 0.4574E 05 0.3798E 03 0.2156E 05 0.6077E 03 Cs-134 0.1093E 04 0.1993E 01 0.1506E 04 0.3040E 00 0.1018E 01 0.4865E 00 I-134 0.1142E 06 0.1062E 06 0.6345E 04 0.7793E 00 0.2609E 01 0.1247E 01 I-135 0.8958E 05 0.5671E 05 0.2621E 05 0.3220E 01 0.1078E 02 0.5152E 01 Xe-135M 0.7006E 06 0.2764E 06 0.7333E 02 0.6089E 00 0.3456E 02 0.9742E 00 Xe-135 0.1823E 06 0.1092E 06 0.1025E 04 0.8511E 01 0.4831E 03 0.1362E 02 Cs-136 0.5383E 03 0.3086E 02 0.4042E 03 0.8158E-01 0.2731E 00 0.1305E 00 Cs-137 0.3357E 04 0.2699E 00 0.2674E 04 0.5397E 00 0.1807E 01 0.8635E 00 Xe-138 0.4168E 06 0.4167E 06 0.9908E 02 0.8227E 00 0.4670E 02 0.1316E 01 Ba-140 0.6550E 02 0.2356E 01 0.5038E 02 0.6130E-02 0.2052E-01 0.9807E-02 La-140 0.2865E 02 0.6365E 01 0.1778E 02 0.2163E-02 0.7241E-02 0.3461E-02 Ce-144 0.5465E 01 0.9124E-02 0.4354E 01 0.5296E-03 0.1773E-02 0.8474E-03 Pr-144 0.2195E 03 0.2141E 03 0.4354E 01 0.5297E-03 0.1773E-02 0.8474E-03 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-18 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED BASIC ASSUMPTIONS FOR PRESSURIZER ACTIVITIES Pressurizer liquid volume, gal. 8080 Pressurizer vapor volume, gal. 5400 Flowrate, gpm 1 Stripping fraction for noble gases 1 Stripping fraction for other isotopes 0.0 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-19 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ACTIVITY IN PRESSURIZER FOR DESIGN BASIS CASE Liquid Phase Steam Phase Activity, Concent, Activity, Concent Nuclide Ci Ci/cc Nuclide Ci Ci/cc H-3 0.224E 02 0.731E 00 Kr- 83M 0.510E-01 0.250E-02 Cr-51 0.369E-01 0.121E-02 Kr- 85M 0.623E 00 0.305E-01 Mn-54 0.676E-02 0.221E-03 Kr- 85 0.186E 04 0.913E 02 Fe-55 0.143E-01 0.468E-03 Kr- 87 0.983E-01 0.482E-02 Co-58 0.335E 00 0.109E-01 Kr- 88 0.671E 00 0.329E-01 Fe-59 0.203E-01 0.665E-03 Xe-133M 0.135E 02 0.664E 00 Co-60 0.438E-01 0.143E-02 Xe-133 0.258E 04 0.127E 03 Kr-83M 0.0 0.0 Xe-135M 0.443E 00 0.217E-01 Kr-85M 0.0 0.0 Xe-135 0.693E 01 0.340E 00 Kr-85 0.0 0.0 Xe-138 0.954E 02 0.468E-03 Kr-87 0.0 0.0 Kr-88 0.0 0.0 Sr-89 0.469E-01 0.153E-02 Sr-90 0.242E-02 0.792E-04 Y-90 0.314E-02 0.103E-03 Sr-91 0.310E-02 0.101E-03 Y-91 0.404E 00 0.132E-01 Pressurizer Deposited Activity Sr-92 0.377E-03 0.123E-04 Y-92 0.100E-02 0.327E-04 Nuclide Activity, Ci/c2 Zr-95 0.118E-01 0.387E-03 Nb-95 0.121E-01 0.397E-03 Cr-51 9.80E-02 Mo-99 0.425E 02 0.139E 01 Mn-54 1.50E-01 I-131 0.401E 02 0.131E 01 Mn-56 2.20E-02 Te-132 0.285E 01 0.931E-01 Co-58 3.80E 00 I-132 0.331E 01 0.108E 00 Co-60 1.60E-01 I-133 0.149E 02 0.488E 00 Fe-59 1.40E-01 Xe-133M 0.0 0.0 Xe-133 0.0 0.0 Cs-134 0.417E 01 0.136E 00 I-134 0.998E-01 0.326E-02 I-135 0.300E 01 0.982E-01 Xe-135M 0.0 0.0 Xe-135 0.0 0.0 Cs-136 0.875E 00 0.286E-01 Cs-137 0.751E 01 0.246E 00 Xe-138 0.0 0.0 Ba-140 0.655E-01 0.214E-02 La-140 0.549E-01 0.179E-02 Ce-144 0.728E-02 0.238E-03 Pr-144 0.728E-02 0.238E-03 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-20 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ACTIVITY IN PRESSURIZER FOR NORMAL OPERATION CASE Liquid Phase Steam Phase Nuclide Activity, Concent, Activity, Concent Ci/cc Ci Ci/cc Nuclide Ci H-3 0.211E 02 0.691E 00 Kr- 83M 0.612E-02 0.300E-03 Cr-51 0.349E-01 0.121E-02 Kr- 85M 0.745E-01 0.365E-02 Mn-54 0.675E-02 0.221E-03 Kr- 85 0.175E 03 0.860E 01 Fe-55 0.143E-01 0.468E-03 Kr- 87 0.118E-01 0.578E-03 Co-58 0.335E 00 0.109E-01 Kr- 88 0.803E-01 0.394E-02 Fe-59 0.203E-01 0.664E-03 Xe-133M 0.158E-01 0.773E-01 Co-60 0.438E-01 0.143E-02 Xe-133 0.291E 03 0.143E 02 Kr-83M 0.0 0.0 Xe-135M 0.532E-01 0.261E-02 Kr-85M 0.0 0.0 Xe-135 0.829E 00 0.406E-01 Kr-85 0.0 0.0 Xe-138 0.114E-02 0.561E-04 Kr-87 0.0 0.0 Kr-88 0.0 0.0 Sr-89 0.562E-02 0.184E-03 Sr-90 0.290E-03 0.950E-05 Y-90 0.376E-03 0.123E-04 Sr-91 0.272E-03 0.121E-04 Y-91 0.471E-01 0.154E-02 Pressurizer Deposited Activity Sr-92 0.453E-04 0.148E-05 Y-92 0.120E-03 0.393E-05 Nuclide Activity mCi/c2 Zr-95 0.142E-02 0.464E-04 Nb-95 0.146E-02 0.476E-04 Cr-51 9.80E-02 Mo-99 0.508E 01 0.166E 00 Mn-54 1.50E-01 I-131 0.481E 01 0.157E 00 Mn-56 2.20E-02 Te-132 0.342E 00 0.112E-01 Co-58 3.80E 00 I-132 0.397E 00 0.130E-01 Co-60 1.60E-01 I-133 0.179E 01 0.586E-01 Fe-59 1.40E-01 Xe-133M 0.0 0.0 Xe-133 0.0 0.0 Cs-134 0.505E 00 0.165E-01 I-134 0.120E-01 0.391E-03 I-135 0.360E 00 0.118E-01 Xe-135M 0.0 0.0 Xe-135 0.0 0.0 Cs-136 0.105E 00 0.343E-02 Cs-137 0.901E 00 0.295E-01 Xe-138 0.0 0.0 Ba-140 0.785E-02 0.257E-03 La-140 0.658E-02 0.215E-03 Ce-144 0.873E-03 0.285E-04 Pr-144 0.873E-03 0.285E-04 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-21 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED BASIC ASSUMPTIONS FOR TRITIUM ACTIVITY IN PRIMARY COOLANT
- 1. Core Thermal Power 3568 MWt
- 2. Plant Load Factor 0.8
- 3. Core Volume 1153 ft3
- 4. Core Volume Fractions
- a. UO2 0.3052
- b. Zr + SS 0.1000
- c. H2O 0.5948
- 5. Initial Reactor Coolant Boron Level
- a. Initial cycle 840 ppm
- b. Equilibrium cycle 1200 ppm
- 6. Reactor Coolant Volume 12,560 ft3
- 7. Reactor Coolant Transport Times
- a. Incore 0.77 sec
- b. Out-of-core 10.87 sec
- 8. Reactor Coolant Peak Lithium Level (99% pure Li7) 2.2 ppm
- 9. Core Average Neutron Fluxes, n/cm2-sec
- a. E > 6 MeV 2.91 x 1012
- b. E > 5 MeV 7.90 x 1012
- c. 3 MeV E 6 MeV 2.26 x 1013
- d. 1 MeV E 5 MeV 5.31 x 1013
- e. E < 0.625 eV 2.26 x 1013
- 10. Neutron Reaction Cross Sections
- a. B10 (n, 2) T: (1 MeV E 5 MeV) = 31.6 mb (spectrum weighted)
(E > 5 MeV) = 75 mb
- b. Li7 (n, n V) T: (3 MeV E 6 MeV) = 39.1 mb (spectrum weighted)
(E > 6 MeV) = 400 mb
- 11. Fraction of Ternary Tritium Diffusing Through Zirconium Cladding
- a. Design value 0.30
- b. Expected value 0.01 Note: The information presented in the above table was developed in support of the original license and is considered historical.
Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-22 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED TRITIUM ACTIVITY IN PRIMARY COOLANT Tritium Source Core Activity, Ci Coolant Activity, Ci Total Produced in Coolant, Ci Cycle = 8760 Hours Ternary fissions 0.196E05 0.121E03 1092.76 Burnable poison rods 0.0 0.0 0.0 Control rods 0.0 0.0 0.0 Boron shim control 0.0 0.270E02 442.40 Lithium-7 reaction 0.0 0.100E01 9.05 Lithium-6 reaction 0.0 0.105E02 94.97 Deuterium reaction 0.0 0.294E00 2.66 Total 0.196E05 0.160E03 1641.84 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-23 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED STEAM SYSTEM OPERATING CONDITIONS ASSUMED FOR ACTIVITY ANALYSIS FOR NORMAL OPERATION CASE Parameter Value Steam Gen. 1 Steam Gen. 2 Steam Gen. 3 Steam Gen. 4 Steam flowrate to condenser from unit (lb/hr) 3804300.0 3804300.0 3804300.0 3804300.0 Feedwater flowrate to unit (lb/hr) 3804725.0 3804725.0 3804725.0 3804725.0 Blowdown flowrate from unit (lb/hr) 17647.5 17647.5 17647.5 17647.5 Steam venting flowrate from unit (lb/hr) 0.0 0.0 0.0 0.0 Total weight of steam vented from unit during period-lb 0.0 0.0 0.0 0.0 Weight of water in unit (lb) 81500.0 81500.0 81500.0 81500.0 Volume of water in unit (ft3) 1680.4 1680.4 1680.4 1680.4 Density of water in unit (lb/ft3) 48.5 48.5 48.5 48.5 Fraction of primary to secondary leakage to this unit 0.25 0.25 0.25 0.25 Leakage flowrate to this unit (computed) (gal./min) 0.0029 0.0029 0.0029 0.0029 Total gallons of prim. cool. leaked into unit in period 1208.8 1208.8 1208.8 1208.8 Total steam flowrate of condenser (lb/hr) 15217200.0 Total condensate flowrate from condenser (lb/hr) 15221402.0 Weight of water in condenser (lb) 1700000.0 Volume of water in condenser (ft3) 27243.58 Density of water in condenser (lb/ft3) 62.39 Total primary to secondary leak rate (gal./min) 0.0115 Water leakage rate from secondary system (gal./min) 5.0 Steam leakage rate from secondary system (lb/hr) 1700.0 Flowrate of water to blowdown cleanup system (lb/hr) 51763.78 Capacity factor of blowdown cleanup system 1.0 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-24 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ADDITIONAL SECONDARY SYSTEM OPERATING PARAMETERS Mass of water in one steam generator, lb 81,500 Mass of steam in one steam generator, lb 7,200 Secondary side operating temperature, °F 519 Steam generator blowdown tank capacity, ft3 641 Air ejector flowrate - rated, scfm 25
- expected average, scfm 2.5 Total mass of water in secondary system, lb 1,000,000 (minus condenser)
Total mass of steam in secondary system, lb 61,200 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-25 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED STEAM GENERATOR PARTITION FACTORS(a) Nuclide St.Gen.1 St.Gen.2 St.Gen.3 St.Gen.4 H-3 1.0000 1.0000 1.0000 1.0000 Cr-51 0.0007 0.0007 0.0007 0.0007 Mn-54 0.0007 0.0007 0.0007 0.0007 Fe-55 0.0007 0.0007 0.0007 0.0007 Co-58 0.0007 0.0007 0.0007 0.0007 Fe-59 0.0007 0.0007 0.0007 0.0007 Co-60 0.0007 0.0007 0.0007 0.0007 Kr-83M 1.0000 1.0000 1.0000 1.0000 Kr-85M 1.0000 1.0000 1.0000 1.0000 Kr-85 1.0000 1.0000 1.0000 1.0000 Kr-87 1.0000 1.0000 1.0000 1.0000 Kr-88 1.0000 1.0000 1.0000 1.0000 Sr-89 0.0007 0.0007 0.0007 0.0007 Sr-90 0.0007 0.0007 0.0007 0.0007 Y-90 0.0007 0.0007 0.0007 0.0007 Sr-91 0.0007 0.0007 0.0007 0.0007 Y-91 0.0007 0.0007 0.0007 0.0007 Sr-92 0.0007 0.0007 0.0007 0.0007 Y-92 0.0007 0.0007 0.0007 0.0007 Zr-95 0.0007 0.0007 0.0007 0.0007 Nb-95 0.0007 0.0007 0.0007 0.0007 Mo-99 0.0007 0.0007 0.0007 0.0007 I-131 0.0065 0.0065 0.0065 0.0065 Te-132 0.0007 0.0007 0.0007 0.0007 I-132 0.0065 0.0065 0.0065 0.0065 I-133 0.0065 0.0065 0.0065 0.0065 Xe-133M 1.0000 1.0000 1.0000 1.0000 Xe-133 1.0000 1.0000 1.0000 1.0000 Cs-134 0.0007 0.0007 0.0007 0.0007 I-134 0.0065 0.0065 0.0065 0.0065 I-135 0.0065 0.0065 0.0065 0.0065 Xe-135M 1.0000 1.0000 1.0000 1.0000 Xe-135 1.0000 1.0000 1.0000 1.0000 Cs-136 0.0007 0.0007 0.0007 0.0007 Cs-137 0.0007 0.0007 0.0007 0.0007 Xe-138 1.0000 1.0000 1.0000 1.0000 Ba-140 0.0007 0.0007 0.0007 0.0007 La-140 0.0007 0.0007 0.0007 0.0007 Ce-144 0.0007 0.0007 0.0007 0.0007 Pr-144 0.0007 0.0007 0.0007 0.0007 (a) Iodine partition factors are for nonvolatile species only and include partitioning in moisture separators. Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-26 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED TOTAL ADDITIONS AND REMOVALS OF ACTIVITY IN EACH STEAM GENERATOR FOR NORMAL OPERATION CASE (CURIES) Nuclide Leakage in From Feedwater To Main Stm To Blowdown Vented Decayed Leaked H-3 0.3007E 01 0.6511E 04 0.6513E 04 0.0 0.0 0.8954E-03 0.7276E 00 Cr-51 0.5976E-02 0.5983E-03 0.8659E-03 0.5681E-02 0.0 0.2753E-04 0.9674E-07 Mn-54 0.9719E-03 0.9793E-04 0.1414E-03 0.9280E-03 0.0 0.4126E-06 0.1580E-07 Fe-55 0.4960E-02 0.4651E-03 0.6877E-03 0.4512E-02 0.0 0.2251E-03 0.7683E-07 Co-58 0.5015E-01 0.5043E-02 0.7288E-02 0.4782E-01 0.0 0.9040E-04 0.8142E-04 Fe-59 0.3137E-02 0.3150E-03 0.4554E-03 0.2988E-02 0.0 0.8944E-05 0.5088E-07 Co-60 0.6239E-02 0.6290E-03 0.9083E-03 0.5959E-02 0.0 0.4179E-06 0.1015E-06 Sr-89 0.8600E-03 0.8638E-04 0.1249E-03 0.8193E-03 0.0 0.2153E-05 0.1395E-07 Sr-90 0.4130E-04 0.4164E-05 0.6013E-05 0.3945E-04 0.0 0.5205E-09 0.6717E-09 Y-90 0.7106E-04 0.1159E-04 0.1160E-04 0.6919E-04 0.0 0.3850E-05 0.1296E-08 Sr-91 0.5599E-03 0.3580E-04 0.6114E-04 0.4011E-03 0.0 0.1334E-03 0.6831E-08 Y-91 0.7122E-02 0.1191E-02 0.1191E-02 0.7105E-02 0.0 0.1719E-04 0.1331E-06 Sr-92 0.2289E-03 0.4779E-05 0.1516E-04 0.9944E-04 0.0 0.1191E-03 0.1693E-08 Y-92 0.2928E-03 0.3159E-04 0.3218E-04 0.1919E-03 0.0 0.1896E-03 0.3595E-08 Zr-95 0.2136E-03 0.2147E-04 0.3103E-04 0.2036E-03 0.0 0.4192E-06 0.3467E-08 Nb-95 0.2076E-03 0.3162E-04 0.3163E-04 0.2075E-03 0.0 0.7978E-06 0.3534E-08 Mo-99 0.1713E 01 0.1612E 00 0.2381E 00 0.1562E 01 0.0 0.7425E-01 0.2660E-04 I-131 0.1015E 01 0.1372E 01 0.1418E 01 0.9519E 00 0.0 0.1678E-01 0.1584E-03 Te-132 0.1067E 00 0.1013E-01 0.1492E-01 0.9786E-01 0.0 0.4061E-02 0.1666E-05 I-132 0.3652E 00 0.2914E 00 0.3006E 00 0.2018E 00 0.0 0.2861E 00 0.3358E-04 I-133 0.1386E 01 0.1620E 01 0.1689E 01 0.1133E 01 0.0 0.1837E 00 0.1887E-03 Cs-134 0.7721E-01 0.2177E-01 0.2178E-01 0.7217E-01 0.0 0.2567E-04 0.2433E-05 I-134 0.1851E 00 0.4043E-01 0.5234E-01 0.3513E-01 0.0 0.1380E 00 0.5847E-05 I-135 0.7647E 00 0.6716E 00 0.7138E 00 0.4791E 00 0.0 0.2433E 00 0.7974E-04 Cs-136 0.1938E-01 0.4603E-02 0.5473E-02 0.1813E-01 0.0 0.3720E-03 0.6114E-06 Cs-137 0.1282E 00 0.3120E-01 0.3695E-01 0.1224E 00 0.0 0.3320E-05 0.4128E-05 Ba-140 0.1456E-02 0.1445E-03 0.2097E-03 0.1976E-02 0.0 0.1449E-04 0.2343E-07 La-140 0.5137E-03 0.8819E-04 0.8811E-04 0.5781E-03 0.0 0.4660E-04 0.9843E-08 Ce-144 0.1258E-03 0.1267E-04 0.1831E-04 0.1201E-03 0.0 0.5668E-07 0.2045E-08 Pr-144 0.1258E-03 0.1717E-04 0.1836E 04 0.1205E-03 0.0 0.1334E-02 0.2051E-08 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-27 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED EQUILIBRIUM ACTIVITIES AND CONCENTRATIONS IN EACH STEAM GENERATOR FOR NORMAL OPERATION CASE Nuclide Activity, Ci Concentration, Ci/cc H-3 0.1991E-01 0.4184E-03 Cr-51 0.3782E-05 0.7947E-07 Mn-54 0.6177E-06 0.1298E-07 Fe-55 0.3003E-05 0.6312E-07 Co-58 0.3183E-04 0.6689E-06 Fe-59 0.1989E-05 0.4180E-07 Co-60 0.3967E-05 0.8336E-07 Sr-89 0.5454E-06 0.1146E-07 Sr-90 0.2626E-07 0.5518E-09 Y-90 0.4584E-07 0.9633E-09 Sr-91 0.2670E-06 0.5611E-08 Y-91 0.4683E-05 0.9842E-07 Sr-92 0.6619E-07 0.1391E-08 Y-92 0.1331E-06 0.2797E-08 Zr-95 0.1355E-06 0.2848E-08 Nb-95 0.1381E-06 0.2903E-08 Mo-99 0.1030E-02 0.2164E-04 I-131 0.6670E-03 0.1402E-04 Te-132 0.6514E-04 0.1369E-05 I-132 0.1414E-03 0.2971E-05 I-133 0.7942E-03 0.1669E-04 Cs-134 0.9512E-04 0.1999E-05 I-134 0.2462E-04 0.5173E-06 I-135 0.3357E-03 0.7055E-05 Cs-136 0.2390E-04 0.5023E-06 Cs-137 0.1614E-03 0.3391E-05 Ba-140 0.9160E-06 0.1925E-07 La-140 0.3848E-06 0.8087E-08 Ce-144 0.7994E-07 0.1680E-08 Pr-144 0.8019E-07 0.1685E-08 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-28 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED TOTAL ADDITIONS AND REMOVALS OF ACTIVITY IN THE CONDENSER FOR NORMAL OPERATION CASE (CURIES) Nuclide From Mainsteam To Feedwater Decayed Water Leakage H-3 0.2605E-05 0.2605E 05 0.1867E-01 0.4283E 01 Cr-51 0.3464E-02 0.2394E-02 0.2777E-06 0.3935E-06 Mn-54 0.5658E-03 0.3918E-03 0.4171E-08 0.6442E-07 Fe-55 0.2751E-02 0.1861E-02 0.2223E-05 0.3060E-06 Co-58 0.2915E-01 0.2017E-01 0.9131E-06 0.3317E-05 Fe-59 0.1822E-02 0.1260E-02 0.9030E-07 0.2072E-06 Co-60 0.3633E-02 0.2516E-02 0.4225E-08 0.4137E-06 Sr-89 0.4995E-03 0.3456E-03 0.2174E-07 0.5682E-07 Sr-90 0.2405E-04 0.1666E-04 0.5262E-11 0.2739E-08 Y-90 0.4640E-04 0.4636E-04 0.5615E-07 0.7622E-08 Sr-91 0.2446E-03 0.1432E-03 0.1140E-05 0.2355E-07 Y-91 0.4765E-02 0.4764E-02 0.2508E-06 0.7833E-04 Sr-92 0.6063E-04 0.1912E-04 0.5481E-06 0.3144E-08 Y-92 0.1287E-03 0.1264E-03 0.2717E-05 0.2078E-07 Zr-95 0.1241E-03 0.8590E-04 0.4235E-08 0.1412E-07 Nb-95 0.1265E-03 0.1265E-03 0.1164E-07 0.2080E-07 Mo-99 0.9523E 00 0.6449E 00 0.7340E-03 0.1060E-03 I-131 0.5673E 01 0.5488E 01 0.2201E-02 0.9023E-03 Te-132 0.5967E-01 0.4052E-01 0.4026E-04 0.6663E-05 I-132 0.1202E 01 0.1166E 01 0.3760E-01 0.1917E-03 I-133 0.6755E 01 0.6482E 01 0.2389E-01 0.1066E-02 Cs-134 0.8715E-01 0.8711E-01 0.3746E-06 0.1432E-04 I-134 0.2094E 00 0.1618E 00 0.1446E-01 0.2660E-04 I-135 0.2855E 01 0.2687E 01 0.3104E-01 0.4418E-03 Cs-136 0.2189E-01 0.1842E-01 0.4569E-05 0.3028E-05 Cs-137 0.1478E 00 0.1248E 00 0.4094E-07 0.2052E-04 Ba-140 0.8390E-03 0.5782E-03 0.1458E-06 0.9507E-07 La-140 0.3524E-03 0.3528E-03 0.4610E-06 0.5801E-07 Ce-144 0.7322E-04 0.5070E-04 0.5729E-09 0.8337E-08 Pr-144 0.7345E-04 0.6860E-04 0.1020E-04 0.1129E-07 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-29 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED EQUILIBRIUM ACTIVITIES AND CONCENTRATIONS IN THE CONDENSER FOR NORMAL OPERATION CASE Nuclide Activity, Ci Concentration, Ci/cc H-3 0.4151E 00 0.5381E-03 Cr-51 0.3815E-07 0.4945E-10 Mn-54 0.6244E-08 0.8093E-11 Fe-55 0.2966E-07 0.3844E-10 Co-58 0.3215E-06 0.4168E-09 Fe-59 0.2008E-07 0.2603E-10 Co-60 0.4010E-07 0.5198E-10 Sr-89 0.5507E-08 0.7139E-11 Sr-90 0.2655E-09 0.3441E-12 Y-90 0.6685E-09 0.8665E-12 Sr-91 0.2282E-08 0.2959E-11 Y-91 0.6835E-07 0.8859E-10 Sr-92 0.3047E-09 0.3950E-12 Y-92 0.1908E-08 0.2473E-11 Zr-95 0.1369E-08 0.1774E-11 Nb-95 0.2016E-08 0.2613E-11 Mo-99 0.1013E-04 0.1314E-07 I-131 0.8746E-04 0.1134E-06 Te-132 0.6458E-06 0.8372E-09 I-132 0.1858E-04 0.2409E-07 I-133 0.1033E-03 0.1339E-06 Cs-134 0.1388E-05 0.1800E-08 I-134 0.2578E-05 0.3342E-08 I-135 0.4282E-04 0.5550E-07 Cs-136 0.2935E-06 0.3805E-09 Cs-137 0.1989E-05 0.2579E-08 Ba-140 0.9215E-08 0.1195E-10 La-140 0.5623E-08 0.7289E-11 Ce-144 0.8081E-09 0.1047E-11 Pr-144 0.1094E-08 0.1419E-11 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-30 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED TOTAL ADDITIONS AND REMOVALS OF ACTIVITY IN THE CONDENSER VAPOR SPACE FOR NORMAL OPERATION CASE Inleak Rate, Tot. Inleakage, Vent Rate, Vent Rate, Tot. Vented, Decayed, Nuclide Ci/hr Ci hr-1 Ci/hr Ci Ci Kr-83 0.7861E-04 0.5509E 00 0.1650E 00 0.2413E-04 0.1690E 00 0.3817E 00 Kr-85M 0.4049E-03 0.2837E 01 0.1650E 00 0.2071E-03 0.1451E 01 0.1385E 01 Kr-85 0.7133E-03 0.6249E 01 0.1650E 00 0.7133E-03 0.6244E 01 0.2787E-03 Kr-87 0.2219E-03 0.1555E 01 0.1650E 00 0.5151E-04 0.3610E 00 0.1194E 01 Kr-88 0.6933E-03 0.4859E 01 0.1650E 00 0.2755E-03 0.1930E 01 0.2927E 01 I-131 0.4575E-05 0.3206E-01 0.1650E 00 0.4477E-05 0.3136E-01 0.6824E-03 I-132 0.1646E-05 0.1153E-01 0.1650E 00 0.5985E-06 0.4193E-02 0.7338E-02 I-133 0.6245E-05 0.4376E-01 0.1650E 00 0.5204E-05 0.3645E-01 0.7290E-02 Xe-133M 0.6653E-03 0.4662E 01 0.1650E 00 0.6182E-03 0.4330E 01 0.3294E 00 Xe-133 0.5421E-01 0.3799E 03 0.1650E 00 0.5248E-01 0.3675E 03 0.1215E 02 I-134 0.8341E-06 0.5845E-02 0.1650E 00 0.1426E-06 0.9991E-03 0.4845E-02 I-135 0.3446E-05 0.2415E-01 0.1650E 00 0.2118E-05 0.1484E-01 0.9301E-02 Xe-135M 0.6224E-03 0.4362E 01 0.1650E 00 0.3628E-04 0.2543E 00 0.4107E 01 Xe-135 0.1304E-02 0.9136E 01 0.1650E 00 0.8951E-03 0.6270E 01 0.2862E 01 Xe-138 0.1174E-03 0.8227E 00 0.1650E 00 0.6170E-05 0.4324E-01 0.7794E 00 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.1-31 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED EQUILIBRIUM ACTIVITIES AND CONCENTRATIONS IN THE CONDENSER VAPOR SPACE FOR NORMAL OPERATION CASE Nuclide Activity, Ci Concentration, mCi/cc Kr-83M 0.1462E-03 0.5164E-07 Kr-85M 0.1255E-02 0.4434E-06 Kr-85 0.4323E-02 0.1527E-05 Kr-87 0.3122E-03 0.1103E-06 Kr-88 0.1670E-02 0.5897E-06 I-131 0.2697E-04 0.9524E-08 I-132 0.3618E-05 0.1278E-08 I-133 0.3141E-04 0.1109E-07 Xe-133M 0.3746E-02 0.1323E-05 Xe-133 0.3180E 00 0.1123E-03 I-134 0.8637E-06 0.3050E-09 I-135 0.1281E-04 0.4522E-08 Xe-135M 0.2194E-03 0.7747E-07 Xe-135 0.5419E-02 0.1914E-05 Xe-138 0.3740E-04 0.1321E-07 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-1 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Sheet 1 of 2 ASSUMPTIONS USED FOR INPUT WASTE STREAMS AND ACTIVITY CALCULATIONS BASED ON ORIGINAL SYSTEM DESIGN
- 1. Each stream of liquid waste input is categorized by one of the following isotopic concentration spectra, which are shown for the two analyses cases in Tables 11.2-3 and 11.2-4:
Spectrum I - Degassed primary coolant Spectrum II - Degassed primary coolant with 48 hours of decay Spectrum III - 1.0% of degassed primary coolant with 48 hours of decay Spectrum IV - 0.01% of degassed primary coolant with 48 hours of decay Spectrum V - 0.001% of primary coolant with 48 hours of decay; principally component cooling water.
- 2. Heat exchangers are periodically drained producing 25 gallons/inch inlet piping/year for the Design Basis Case and 1.5 times the above for the Normal Operation Case.
- 3. Pumps are periodically drained producing 10 gallons/inch suction/year for the Design Basis Case and 1.5 times the above for the Normal Operation Case.
- 4. Pump baseplate leakage of 0.10 gallons/day for the Design Basis Case and 1.5 times the above for the Normal Operation Case
- 5. Tanks are periodically drained producing 100 gallons/5 feet diameter/year for the Design Basis Case and 1.5 times the above for the Normal Operation Case.
- 6. Filter cartridge replacement three times per year, producing 15 gallons/replacement for the Design Basis Case and 1.5 times the above volume for the Normal Operation Case.
- 7. Sampling produces 4 gallons/sample (2 gallons/sample to laboratory drain and 2 gallons/sample due to line purging, 7 samples/day/unit).
- 8. Valve stem leakage of 10 cc/day for the Design Basis Case and 1.5 times the above for the Normal Operation Case.
- 9. No waste from relief valve discharges for the Design Basis Case and 50 gallons/year for the Normal Operation Case.
- 10. Laundry waste of 3 washloads per week producing 210 gallons/washload for the Design Basis Case and 1.5 times the above volume for the Normal Operation Case.
- 11. Personnel decontamination showers of 4 showers/day producing 5 gallons/shower for the Design Basis Case and 1.5 times the above volume for the Normal Operation Case.
- 12. Personnel handwashes of 25 per day producing 0.5 gallons/wash for the Design Basis Case and 1.5 times the above volume for the Normal Operation Case.
- 13. Periodic system piping drains of 200 gallons/year/system/unit for the Design Basis Case and 1.5 times the above volume for the Normal Operation Case.
Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-1 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Sheet 2 of 2
- 14. Containment fan cooler drains of 1,000 gallons/year/unit for the Design Basis Case and 1.5 times the above volume for the Normal Operation Case.
- 15. Each reactor coolant pump 3 seal leaks 100 cc/hour for the Design Basis Case and the Normal Operation Case.
- 16. Spent resin loadout area washdown of 200 gallons/year for the Design Basis Case and 1.5 times the above for the Normal Operation Case.
- 17. Demineralizer overflow produces 500 gallons/backwash and resin replacement of one demineralizer per set annually for the Design Basis Case and the Normal Operation Case.
- 18. Reactor coolant drain tank wastes are processed through the CVCS for the Design Basis Case and 500 gallons/year/unit to the liquid waste system for the Normal Operation Case.
- 19. Miscellaneous floor drain leakage of 5,000 gallons/year for the Design Basis Case and 1.5 times the above volume for the Normal Operation Case.
- 20. Miscellaneous leakage of 160 pounds/day/unit of Spectrum I wastes to the auxiliary building for the Normal Operation Case.
- 21. Miscellaneous leakage of 40 gallons/day/unit of Spectrum I wastes to the containment building for the Normal Operation Case.
- 22. RHR pump leakage of 0.25 gallons/pump for the Normal Operation Case.
- 23. Waste will be allowed 48 hours of decay before processing from the chemical drain, laundry and hot shower, miscellaneous equipment drain, and reactor coolant drain tanks. (This includes fill time and discharge time.)
- 24. Waste will be allowed 336 hours of decay in the floor drain receiver and equipment drain receiver tanks before processing. (This includes one-half of the fill time plus one-half of the discharge time.)
- 25. Waste will be allowed 24 hours of decay in the waste condensate tank before processing.
- 26. No decay credit is taken for sumps.
- 27. For tritium control, 350,000 gallons/unit will be discharged annually.
- 28. Primary-to-secondary steam generator leakage of 100 lb/day of primary coolant for the Normal Operation Case.
Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 &2 FSAR UPDATE TABLE 11.2-2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ASSUMPTIONS FOR CALCULATIONS OF ACTIVITY RELEASED FROM CVCS BASED ON ORIGINAL SYSTEM DESIGN Average Shim Bleed Flowrate, gpm 1.0 Capacity of Liquid Holdup Tank, gal. 83000 DFs for CVC Demineralizers (Given in Table 11.1-15) DF for Boric Acid Evaporator (a) - iodine 102
- other nuclides 103 Capacity of Monitor Tank, gal. 25000 Fraction of Boric Acid Evaporator Distillate Recycled (a) 0.333 (a) Boric acid evaporator has been abandoned in place Note: The information presented in the above table was developed in support of the original license and is considered historical.
Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-3 Sheet 1 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ACTIVITY CONCENTRATION SPECTRUM I THROUGH V FOR INPUT WASTE SOURCES, DESIGN BASIS CASE (Ci/cc)(a) Nuclide Spectrum I Spectrum II Spectrum III Spectrum IV Spectrum V H-3 0.11E+01 0.11E+01 0.11E-01 0.11E-03 0.11E-04 Cr-51 0.19E-02 0.19E-02 0.19E-04 0.19E-06 0.19E-07 Mn-54 0.31E-03 0.30E-03 0.30E-05 0.30E-07 0.30E-08 Fe-55 0.15E-02 0.92E-03 0.92E-05 0.92E-07 0.92E-08 Co-58 0.17E-01 0.16E-01 0.16E-03 0.16E-05 0.16E-06 Fe-59 0.10E-02 0.97E-03 0.97E-05 0.97E-07 0.97E-08 Co-60 0.19E-02 0.19E-02 0.19E-04 0.19E-06 0.19E-07 Kr-83M 0.35E+00 0.0 0.0 0.0 0.0 Kr-85M 0.18E+01 0.0 0.0 0.0 0.0 Kr-85 0.58E+01 0.0 0.0 0.0 0.0 Kr-87 0.99E+00 0.0 0.0 0.0 0.0 Kr-88 0.31E+01 0.0 0.0 0.0 0.0 Sr-89 0.24E-02 0.23E-02 0.23E-04 0.23E-06 0.23E-07 Sr-90 0.11E-03 0.11E-03 0.11E-05 0.11E-07 0.11E-08 Y-90 0.19E-04 0.16E-04 0.16E-06 0.16E-08 0.16E-09 Sr-91 0.15E-02 0.50E-04 0.50E-06 0.50E-08 0.50E-09 Y-91 0.21E-02 0.20E-02 0.20E-04 0.20E-06 0.20E-07 Sr-92 0.61E-03 0.27E-08 0.27E-10 0.27E-12 0.27E-13 Y-92 0.78E-04 0.25E-07 0.25E-09 0.25E-11 0.25E-12 Zr-95 0.57E-03 0.56E-03 0.56E-05 0.56E-07 0.56E-08 Nb-95 0.56E-03 0.56E-03 0.56E-05 0.56E-07 0.56E-08 Mo-99 0.46E-01 0.28E-01 0.28E-03 0.28E-05 0.28E-06 I-131 0.28E+01 0.23E+01 0.23E-01 0.23E-03 0.23E-04 Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-3 Sheet 2 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Nuclide Spectrum I Spectrum II Spectrum III Spectrum IV Spectrum V Te-132 0.29E+00 0.19E+00 0.19E-02 0.19E-04 0.19E-05 I-132 0.97E+00 0.20E+00 0.20E-02 0.20E-04 0.20E-05 I-133 0.38E+01 0.77E+00 0.77E-02 0.77E-04 0.77E-05 Xe-133M 0.31E+01 0.0 0.0 0.0 0.0 Xe-133 0.20E+03 0.0 0.0 0.0 0.0 Cs-134 0.19E+00 0.19E+00 0.19E-02 0.19E-04 0.19E-05 I-134 0.49E+00 0.10E-16 0.10E-18 0.10E-20 0.10E-21 I-135 0.21E+01 0.15E-01 0.15E-03 0.15E-05 0.15E-06 Xe-135M 0.39E+00 0.0 0.0 0.0 0.0 Xe-135 0.54E+01 0.0 0.0 0.0 0.0 Cs-136 0.51E-01 0.46E-01 0.46E-03 0.46E-05 0.46E-06 Cs-137 0.35E+00 0.35E+00 0.35E-02 0.35E-04 0.35E-05 Xe-138 0.51E+00 0.0 0.0 0.0 0.0 Ba-140 0.39E-02 0.35E-02 0.35E-04 0.35E-06 0.35E-07 La-140 0.14E-02 0.27E-02 0.27E-04 0.27E-06 0.27E-07 Ce-144 0.33E-03 0.33E-03 0.33E-05 0.33E-07 0.33E-08 Pr-144 0.33E-03 0.33E-03 0.33E-05 0.33E-07 0.33E-08 (a) A plateout decontamination factor of 10 for yttrium and 100 for molybdenum is assumed for all waste streams. Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-4 Sheet 1 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ACTIVITY CONCENTRATION SPECTRUM I THROUGH V FOR INPUT WASTE SOURCES, DESIGN BASIS CASE (Ci/cc)(a) Nuclide Spectrum I Spectrum II Spectrum III Spectrum IV Spectrum V H-3 0.11E+01 0.10E+01 0.10E-01 0.10E-03 0.10E-04 Cr-51 0.19E-02 0.19E-02 0.19E-04 0.19E-06 0.19E-07 Mn-54 0.31E-03 0.30E-03 0.30E-05 0.30E-07 0.30E-08 Fe-55 0.15E-02 0.92E-03 0.92E-05 0.92E-07 0.92E-08 Co-58 0.17E-01 0.16E-01 0.16E-03 0.16E-05 0.16E-06 Fe-59 0.10E-02 0.97E-03 0.97E-05 0.97E-07 0.97E-08 Co-60 0.19E-02 0.19E-02 0.19E-04 0.19E-06 0.19E-07 Kr-83M 0.42E-01 0.0 0.0 0.0 0.0 Kr-85M 0.22E+00 0.0 0.0 0.0 0.0 Kr-85 0.67E+00 0.0 0.0 0.0 0.0 Kr-87 0.12E+00 0.0 0.0 0.0 0.0 Kr-88 0.38E+00 0.0 0.0 0.0 0.0 Sr-89 0.28E-03 0.27E-03 0.27E-05 0.27E-07 0.23E-08 Sr-90 0.13E-04 0.13E-04 0.13E-06 0.13E-08 0.13E-09 Y-90 0.24E-05 0.19E-05 0.19E-07 0.19E-09 0.19E-10 Sr-91 0.18E-03 0.59E-07 0.59E-07 0.59E-09 0.59E-10 Y-91 0.25E-03 0.24E-03 0.24E-05 0.24E-07 0.24E-08 Sr-92 0.74E-04 0.33E-09 0.33E-11 0.33E-13 0.33E-14 Y-92 0.93E-05 0.30E-08 0.30E-10 0.30E-12 0.30E-13 Zr-95 0.68E-04 0.67E-04 0.67E-06 0.67E-08 0.67E-09 Nb-95 0.67E-04 0.67E-04 0.67E-06 0.67E-08 0.67E-09 Mo-99 0.56E-02 0.34E-02 0.34E-04 0.34E-06 0.34E-07 I-131 0.32E+00 0.27E+00 0.27E-02 0.27E-04 0.27E-05 Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-4 Sheet 2 of 2 Nuclide Spectrum I Spectrum II Spectrum III Spectrum IV Spectrum V Te-132 0.35E-01 0.23E+01 0.23E-03 0.23E-05 0.23E-06 I-132 0.12E+00 0.23E+01 0.23E-03 0.23E-05 0.23E-06 I-133 0.44E+00 0.91E+01 0.91E-03 0.91E-05 0.91E-06 Xe-133M 0.38E+00 0.0 0.0 0.0 0.0 Xe-133 0.32E+02 0.0 0.0 0.0 0.0 Cs-134 0.24E+01 0.24E+01 0.24E-03 0.24E-05 0.24E-06 I-134 0.60E+01 0.12E-17 0.12E-19 0.12E-21 0.12E-22 I-135 0.25E+00 0.17E-02 0.17E-04 0.17E-06 0.17E-07 Xe-135M 0.46E+01 0.0 0.0 0.0 0.0 Xe-135 0.65E+00 0.0 0.0 0.0 0.0 Cs-136 0.63E-02 0.56E-02 0.56E-04 0.56E-06 0.56E-07 Cs-137 0.42E+01 0.42E+01 0.42E-03 0.42E-05 0.42E-06 Xe-138 0.63E+01 0.0 0.0 0.0 0.0 Ba-140 0.47E-03 0.42E-03 0.42E-05 0.42E-07 0.42E-08 La-140 0.17E-03 0.32E-03 0.32E-05 0.32E-07 0.32E-08 Ce-144 0.40E-04 0.40E-04 0.40E-06 0.40E-08 0.40E-09 Pr-144 0.40E-04 0.40E-04 0.40E-06 0.40E-08 0.40E-09 (a) A plateout decontamination factor of 10 for yttrium and 100 for molybdenum is assumed for all waste streams. Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-5 Sheet 1 of 4 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ANNUAL FLOW AND ISOTOPIC SPECTRA FOR LIQUID WASTE INPUTS THIS TABLE APPLIES TO THE ESTIMATED RELEASES AND RADIOLOGICAL CONSEQUENCES CALCULATED, BASED ON ORIGINAL DESIGN (REFER TO FIGURES 11.2-2 and 11.2-3) Design Normal Basis Operation Stream Case Case Concentration Number(a) Stream Identification Flow(b) Flow(b) Spectrum 1A Safety injection tank drain 80 120 IV 1B Safety injection tank drain 80 120 IV 2A Safety injection system piping drains 200 300 IV 2B Safety injection system piping drains 200 300 IV 3A Spray additive tank relief and drain 140 260 III 3B Spray additive tank relief and drain 140 260 II 4A Containment spray pump drains 200 300 II 4B Containment spray pump drains 200 300 III 5A Component cooling water (CCW) surge tank relief valve 0 50 V 5B CCW surge tank relief valve 0 50 V 6A RHR heat exchanger (CCW) drain 600 900 V 6B RHR heat exchanger (CCW) drain 600 900 V 7A Waste gas compressor seal cooler relief valve (CCW) 0 100 V 7B Waste gas compressor seal cooler relief valve (CCW) 0 50 V 8A Sample sink drains 2,920 4,380 IV 8B Sample sink drains 2,920 4,380 IV (c) Chemical drain tank drain and overflow - - - (c) Laundry and hot shower tank drain and overflow - - - (c) Radwaste filter drains - - - (c) Floor drain receiver tank drain and overflow - - - 9 Spent resin loadout area drain 200 300 IV 10A Charging pump (CCP1) baseplate drain 135 165 III 10B Charging pump (CCP2) baseplate drain 135 165 III 11A Chemical mixing tank drain 20 30 IV 11B Chemical mixing tank drain 20 30 IV 12A (f) Boric acid and concentrates filter drain 90 135 IV 12B (f) Boric acid and concentrates filter drain 90 135 IV 13 (f) Concentrates holding tank drain and overflow 120 180 IV 14 (f) Concentrates holding tank pump drains 20 30 IV 15A Boric acid tank drain and overflow 400 600 IV 15B Boric acid tank drain and overflow 400 600 IV 16 Batching tank drain and overflow 100 150 IV 17A (f) Boric acid evaporator heat exchanger drains 300 450 IV 17B (f) Boric acid evaporator heat exchanger drains 300 450 IV Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-5 Sheet 2 of 4 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ANNUAL FLOW AND ISOTOPIC SPECTRA FOR LIQUID WASTE INPUTS THIS TABLE APPLIES TO THE ESTIMATED RELEASES AND RADIOLOGICAL CONSEQUENCES CALCULATED, BASED ON ORIGINAL DESIGN (REFER TO FIGURES 11.2-2 and 11.2-3) Design Normal Basis Operation Stream Case Case Concentration Number(a) Stream Identification Flow(b) Flow(b) Spectrum 18A (f) Boric acid evaporator pump drains 80 120 IV 18B (f) Boric acid evaporator pump drains 80 120 IV 19A Monitor tank drain and overflow 800 1,200 IV 19B Monitor tank drain and overflow 800 1,200 IV 20 Miscellaneous leakage (Units 1 and 2) 14,600 I 21 Miscellaneous floor drains (Units 1 and 2) 5,000 7,500 IV 22A Containment fan cooler drain 1,000 1,500 IV 22B Containment fan cooler drain 1,000 1,500 IV 23A Reactor coolant pump labyrinth seal relief valve (CCW) 0 50 V 23B Reactor coolant pump labyrinth seal relief valve (CCW) 0 50 V 24A Reactor coolant pump thermal barrier relief (CCW) - 50 V 24B Reactor coolant pump thermal - barrier relief (CCW) - 50 V 25A Biological shield plate relief (CCW) - 50 V 25B Biological shield plate relief (CCW) 50 V 26A Reactor coolant pump upper and lower bearing relief (CCW) - 50 V 06B Reactor coolant pump upper and lower bearing relief (CCW) - 50 V 27A Reactor vessel support coolers relief (CCW) - 50 V 27B Reactor vessel support coolers relief (CCW) - 50 V 28A Reactor coolant pump No. 3 seal 925 1,400 III 28B Reactor coolant pump No. 3 seal 925 1,400 III 29A Excess letdown heat exchanger relief (CCW) - 50 V 29B Excess letdown heat exchanger relief (CCW) - 50 V 30A Miscellaneous equipment leakages - 11,700 I 30B Miscellaneous equipment leakages - 11,700 I 31A Reactor coolant drain tank relief valve - 50 II 31B Reactor coolant drain tank relief valve - 50 II 33A Miscellaneous pump leakage 70 10 III 33B Miscellaneous pump leakage 70 110 III 34 Laboratory drains 10,220 15,330 IV 35 Laundry drains 32,760 49,140 V 36 Personnel decontamination 11,860 16,340 V 37A CVCS valve leakoffs 6 9 II Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-5 Sheet 3 of 4 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ANNUAL FLOW AND ISOTOPIC SPECTRA FOR LIQUID WASTE INPUTS THIS TABLE APPLIES TO THE ESTIMATED RELEASES AND RADIOLOGICAL CONSEQUENCES CALCULATED, BASED ON ORIGINAL DESIGN (REFER TO FIGURES 11.2-2 and 11.2-3) Design Normal Basis Operation Stream Case Case Concentration Number(a) Stream Identification Flow(b) Flow(b) Spectrum 37B CVCS valve leakoffs 6 9 II 38A Letdown heat exchanger tube side drain 75 110 II 38B Letdown heat exchanger tube side drain 75 110 II 39A Volume control tank drain 150 225 II 39B Volume control tank drain 150 225 II 40A Charging pump (CCP1) drains 140 210 II 40B Charging pump (CCP2) drains 140 210 II 41A Reactor coolant filter 45 60 II 41B Reactor coolant filter 45 60 II 42A Seal water injection filter 90 135 II 42B Seal water injection filter 90 135 II 43A Seal water filter 45 60 II 43B Seal water filter 45 60 II 44A Ion exchange filter 45 60 IV 44B Ion exchange filter 45 60 IV 45A Liquid holdup tank drain 1,050 1,575 II 45B Liquid holdup tank drain 700 1,050 II 46A CVCS miscellaneous piping drains 200 300 II 46B CVCS miscellaneous piping drains 200 300 II 47A RHR heat exchanger drain 700 1,050 II 47B RHR heat exchanger drain 700 1,050 II 48A RHR pump drain 280 420 II 48B RHR pump drain 280 420 II 49A RHR piping drains 200 300 II 49B RHR piping drains 200 300 II 50A Spent fuel pit cooling system piping drains 200 300 III 50B Spent fuel pit cooling system piping drains 200 300 III 51A Spent fuel pit cooling system filter drains 225 340 III 51B Spent fuel pit cooling system filter drains 225 340 III 52A Containment sump pump discharge piping drain 200 300 III 52B Containment sump pump discharge piping drain 200 300 III (c) Equipment drain receiver tank drain and overflow - - - (c) Equipment drain receiver tank pump drain - - - (c) Waste Concentrator Condensate tank drain and overflow - - - Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-5 Sheet 4 of 4 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ANNUAL FLOW AND ISOTOPIC SPECTRA FOR LIQUID WASTE INPUTS THIS TABLE APPLIES TO THE ESTIMATED RELEASES AND RADIOLOGICAL CONSEQUENCES CALCULATED, BASED ON ORIGINAL DESIGN (REFER TO FIGURES 11.2-2 and 11.2-3) Design Normal Basis Operation Stream Case Case Concentration Number(a) Stream Identification Flow(b) Flow(b) Spectrum (c) Waste Concentrator Condensate tank pump drain - - - (c) Radwaste concentrator heat exchanger drains - - - (c) Radwaste concentrator pump drains - - - (c) Waste concentrates tank drain and overflow - - - 53 Spent resin motive water pump drains 60 90 III 54A Waste gas moisture separator drain 40 60 III 54B Waste gas moisture separator drain 20 30 III 55A Waste gas decay tank drain 420 630 III 55B Waste gas decay tank drain 420 630 III 56A Waste gas compressor inlet piping drain 200 300 III 56B Waste gas compressor inlet piping drain 200 300 III 57A Sample sink drains 2,190 3,290 III 57B Sample sink drains 2,190 3,290 III 58A CVCS demineralizer overflow 5,500 5,500 III 58B CVCS demineralizer overflow 5,500 5,500 III 59A Steam generator blowdown tank drain (d) (d) (d) 59B Steam generator blowdown tank drain (d) (d) (d) (c) Spent resin tank overflow - - - 60A Waste regenerant tank discharge - 121,700 (e) 60B Waste regenerant tank discharge - 121,700 (e) (a) The letters "A" and "B" on the stream numbers refer to Units 1 and 2 inputs, respectively. (b) Annual estimated flow in gallons per year. (c) These streams are intra-system leakages and are not counted as input. (d) The steam generator blowdown tank will have significant activity only when significant primary to secondary leakage occurs. (e) Calculated from secondary system activity concentrations and blowdown flowrate. (f) Equipment is abandoned in place and no longer in use. Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-6 Sheet 1 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ISOTOPIC FLOWS THROUGH CVC SYSTEM, DESIGN BASIS CASE Liquid Holdup Tank BA Evap Feed Ion Exchangers Monitor Tank Inflow, Outflow, Inflow, Liq Outflow, Inflow, To Wst Cond Tank, Nuclide Ci/hr Ci/hr Ci/hr Ci/hr Ci/hr Ci/hr H-3 0.230E-00 0.229E-00 0.229E-00 0.223E-00 0.223E-00 0.149E-00 Cr-51 0.318E-05 0.188E-05 0.188E-09 0.188E-12 0.188E-12 0.124E-12 Mn-54 0.517E-06 0.492E-06 0.492E-10 0.492E-13 0.492E-13 0.328E-13 Fe-55 0.264E-05 0.120E-07 0.120E-11 0.120E-14 0.120E-14 0.705E-15 Co-58 0.267E-04 0.217E-04 0.217E-08 0.217E-11 0.217E-11 0.144E-11 Fe-59 0.167E-05 0.121E-05 0.121E-09 0.121E-12 0.121E-12 0.799E-13 Co-60 0.332E-05 0.329E-05 0.329E-09 0.329E-12 0.329E-12 0.219E-12 Kr-83 0.211E-01 0.0 0.0 0.0 0.0 0.0 Kr-85M 0.109E-00 0.365E-35 0.365E-35 0.0 0.0 0.0 Kr-85 0.243E-00 0.242E-00 0.242E-00 0.0 0.0 0.0 Kr-87 0.594E-01 0.0 0.0 0.0 0.0 0.0 Kr-88 0.186E-00 0.323E-55 0.323E-55 0.0 0.0 0.0 Sr-89 0.381E-05 0.287E-05 0.287E-09 0.287E-12 0.287E-12 0.190E-12 Sr-90 0.183E-06 0.183E-06 0.183E-10 0.183E-13 0.183E-13 0.122E-13 Y-90 0.344E-04 0.327E-06 0.327E-06 0.327E-09 0.327E-09 0.192E-10 Sr-91 0.248E-05 0.615E-21 0.615E-25 0.615E-28 0.615E-28 0.174E-28 Y-91 0.354E-02 0.279E-02 0.279E-02 0.279E-05 0.279E-05 0.185E-06 Sr-92 0.101E-05 0.669E-62 0.669E-66 0.669E-69 0.669E-69 0.205E-70 Y-92 0.141E-03 0.106E-45 0.106E-45 0.106E-48 0.106E-48 0.700E-51 Zr-95 0.946E-06 0.757E-06 0.757E-10 0.757E-13 0.757E-13 0.502E-13 Nb-95 0.919E-06 0.893E-06 0.893E-10 0.893E-13 0.893E-13 0.595E-13 Mo-99 0.830E-00 0.488E-02 0.488E-02 0.488E-05 0.488E-05 0.288E-07 I-131 0.490E-01 0.802E-02 0.802E-04 0.802E-06 0.802E-08 0.513E-08 Te-132 0.515E-02 0.582E-04 0.582E-06 0.582E-09 0.582E-11 0.349E-11 Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-6 Sheet 2 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Liquid Holdup Tank BA Evap Feed Ion Exchangers Monitor Tank Inflow, Outflow, Inflow, Liq Outflow, Inflow, To Wst Cond Tank, Nuclide Ci/hr Ci/hr Ci/hr Ci/hr Ci/hr Ci/hr I-132 0.176E-01 0.600E-04 0.600E-06 0.600E-08 0.600E-10 0.472E-11 I-133 0.669E-01 0.400E-08 0.400E-10 0.400E-12 0.400E-14 0.180E-14 Xe-133M 0.183E-00 0.329E-03 0.329E-03 0.0 0.0 0.0 Xe-133 0.154E-02 0.997E-00 0.997E-00 0.0 0.0 0.0 Cs-134 0.908E-02 0.890E-02 0.445E-03 0.445E-06 0.445E-06 0.297E-06 I-134 0.893E-02 0.0 0.0 0.0 0.0 0.0 I-135 0.369E-01 0.846E-24 0.846E-26 0.846E-28 0.846E-30 0.163E-30 Xe-135M 0.233E-01 0.128E-24 0.128E-24 0.0 0.0 0.0 Xe-135 0.327E-00 0.139E-16 0.139E-16 0.0 0.0 0.0 Cs-136 0.246E-02 0.804E-03 0.402E-04 0.402E-07 0.402E-07 0.261E-07 Cs-137 0.163E-01 0.162E-01 0.812E-03 0.812E-06 0.812E-06 0.542E-06 Xe-138 0.314E-01 0.0 0.0 0.0 0.0 0.0 Ba-140 0.645E-05 0.207E-05 0.207E-09 0.207E-12 0.207E-12 0.134E-12 La-140 0.228E-05 0.238E-05 0.238E-09 0.238E-12 0.238E-12 0.154E-12 Ce-144 0.557E-06 0.529E-06 0.529E-10 0.529E-13 0.529E-13 0.353E-13 Pr-144 0.557E-06 0.529E-06 0.529E-10 0.529E-13 0.529E-13 0.353E-13 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-7 Sheet 1 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ISOTOPIC FLOWS THROUGH CVC SYSTEM, NORMAL OPERATION CASE Liquid Holdup Tank BA Evap Feed Ion Exchangers Monitor Tank Inflow, Outflow, Inflow, Liq Outflow, Inflow, To Wst Cond Tank, Nuclide Ci/hr Ci/hr Ci/hr Ci/hr Ci/hr Ci/hr H-3 0.218E 00 0.217E 00 0.217E 00 0.211E 00 0.211E 00 0.140E 00 Cr-51 0.317E-05 0.188E-05 0.188E-09 0.188E-12 0.188E-12 0.124E-12 Mn-54 0.516E-06 0.492E-06 0.492E-10 0.492E-13 0.492E-13 0.328E-13 Fe-55 0.263E-05 0.120E-07 0.120E-11 0.120E-14 0.120E-14 0.705E-15 Co-58 0.266E-04 0.217E-04 0.217E-08 0.217E-11 0.217E-11 0.144E-11 Fe-59 0.167E-05 0.121E-05 0.121E-09 0.121E-12 0.121E-12 0.790E-13 Co-60 0.331E-05 0.329E-05 0.329E-09 0.329E-12 0.329E-12 0.219E-12 Kr-83M 0.252E-02 0.0 0.0 0.0 0.0 0.0 Kr-85M 0.130E-01 0.436E-36 0.436E-36 0.0 0.0 0.0 Kr-85 0.229E-01 0.228E-01 0.228E-01 0.0 0.0 0.0 Kr-87 0.713E-02 0.0 0.0 0.0 0.0 0.0 Kr-88 0.223E-01 0.386E-56 0.386E-56 0.0 0.0 0.0 Sr-89 0.457E-06 0.344E-06 0.344E-10 0.344E-13 0.344E-13 0.228E-13 Sr-90 0.219E-07 0.219E-07 0.219E-11 0.219E-14 0.219E-14 0.146E-14 Y-90 0.411E-05 0.392E-07 0.392E-07 0.392E-10 0.392E-10 0.230E-11 Sr-91 0.297E-06 0.737E-22 0.737E-26 0.737E-29 0.737E-29 0.209E-29 Y-91 0.412E-03 0.325E-03 0.325E-03 0.325E-06 0.325E-06 0.216E-07 Sr-92 0.122E-06 0.803E-63 0.803E-67 0.803E-70 0.803E-70 0.246E-71 Y-92 0.169E-04 0.127E-46 0.127E-46 0.127E-49 0.127E-49 0.840E-52 Zr-95 0.113E-06 0.908E-07 0.908E-11 0.908E-14 0.908E-14 0.602E-14 Nb-95 0.110E-06 0.107E-06 0.107E-10 0.107E-13 0.107E-13 0.713E-14 M0-99 0.992E-01 0.583E-03 0.583E-03 0.583E-06 0.583E-06 0.344E-08 I-131 0.588E-02 0.962E-03 0.962E-05 0.962E-07 0.962E-09 0.615E-09 Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-7 Sheet 2 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Liquid Holdup Tank BA Evap Feed Ion Exchangers Monitor Tank Inflow, Outflow, Inflow, Liq Outflow, Inflow, To Wst Cond Tank, Nuclide Ci/hr Ci/hr Ci/hr Ci/hr Ci/hr Ci/hr Te-132 0.618E-03 0.698E-05 0.698E-07 0.698E-10 0.698E-12 0.418E-12 I-132 0.211E-02 0.720E-05 0.720E-07 0.720E-09 0.720E-11 0.567E-12 I-133 0.802E-02 0.480E-09 0.480E-11 0.480E-13 0.480E-15 0.215E-15 Xe-133M 0.213E-01 0.383E-04 0.383E-04 0.0 0.0 0.0 Xe-133 0.174E 01 0.112E 00 0.112E 00 0.0 0.0 0.0 Cs-134 0.110E-02 0.108E-02 0.539E-04 0.539E-07 0.539E-07 0.360E-07 I-134 0.107E-02 0.0 0.0 0.0 0.0 0.0 I-135 0.443E-02 0.101E-24 0.101E-26 0.101E-28 0.101E-30 0.196E-31 Xe-135M 0.279E-02 0.153E-25 0.153E-25 0.0 0.0 0.0 Xe-135 0.390E-01 0.166E-17 0.166E-17 0.0 0.0 0.0 Cs-136 0.295E-03 0.964E-04 0.482E-05 0.482E-08 0.482E-08 0.313E-08 Cs-137 0.195E-02 0.195E-02 0.975E-04 0.975E-07 0.975E-07 0.650E-07 Xe-138 0.377E-02 0.0 0.0 0.0 0.0 0.0 Ba-140 0.773E-06 0.248E-06 0.248E-10 0.248E-13 0.248E-13 0.161E-13 La-140 0.273E-06 0.285E-06 0.285E-10 0.285E-13 0.285E-13 0.185E-13 Ce-144 0.668E-07 0.635E-07 0.835E-11 0.635E-14 0.635E-14 0.423E-14 Pr-144 0.668E-07 0.635E-07 0.635E-11 0.635E-14 0.635E-14 0.423E-14 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-8 Sheet 1 of 5 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ANNUAL FLOW AND ACTIVITY CONCENTRATION OF PROCESS STREAMS FOR DESIGN BASIS CASE Stream Number(a) 1 2 3 4 5 Annual Flow, gal./yr 3850 17370 0.0 140 44620 Nuclide Concentration, Ci/cc H-3 0.533E-02 0.697E-03 0.0 0.110E-01 0.110E-04 Cr-51 0.899E-05 0.118E-05 0.0 0.185E-04 0.176E-07 Mn-54 0.148E-05 0.193E-06 0.0 0.304E-05 0.303E-08 Fe-55 0.444E-04 0.581E-06 0.0 0.915E-05 0.548E-08 Co-58 0.795E-04 0.104E-04 0.0 0.164E-03 0.160E-06 Fe-59 0.471E-05 0.616E-06 0.0 0.970E-05 0.941E-08 Co-60 0.945E-05 0.124E-05 0.0 0.194E-04 0.194E-07 Sr-89 0.112E-04 0.146E-05 0.0 0.230E-04 0.224E-07 Sr-90 0.533E-06 0.697E-07 0.0 0.110E-05 0.110E-08 Y-90 0.778E-07 0.102E-07 0.0 0.160E-06 0.541E-09 Sr-91 0.242E-06 0.317E-07 0.0 0.499E-06 0.163E-10 Y-91 0.991E-05 0.130E-05 0.0 0.204E-04 0.199E-07 Sr-92 0.133E-10 0.173E-11 0.0 0.273E-10 0.122E-18 Y-92 0.119E-09 0.156E-10 0.0 0.245E-09 0.314E-16 Zr-95 0.271E-05 0.354E-06 0.0 0.558E-05 0.546E-08 Nb-95 0.270E-05 0.353E-06 0.0 0.556E-05 0.556E-08 M0-99 0.137E-03 0.179E-04 0.0 0.281E-03 0.172E-06 I-131 0.114E-01 0.149E-02 0.0 0.234E-01 0.197E-04 Te-132 0.925E-03 0.121E-03 0.0 0.190E-02 0.124E-05 I-132 0.954E-03 0.125E-03 0.0 0.197E-02 0.128E-05 I-133 0.374E-02 0.489E-03 0.0 0.770E-02 0.158E-05 Cs-134 0.943E-03 0.123E-03 0.0 0.194E-02 0.194E-05 I-134 0.492E-19 0.643E-20 0.0 0.101E-18 0.211E-38 I-135 0.707E-04 0.924E-05 0.0 0.146E-03 0.102E-08 Cs-136 0.225E-03 0.294E-04 0.0 0.462E-03 0.416E-06 Cs-137 0.169E-02 0.221E-03 0.0 0.347E-02 0.347E-05 Ba-140 0.170E-04 0.222E-05 0.0 0.349E-04 0.313E-07 La-140 0.129E-04 0.169E-05 0.0 0.266E-04 0.301E-07 Ce-144 0.161E-05 0.211E-06 0.0 0.332E-05 0.330E-08 Pr-144 0.161E-05 0.211E-06 0.0 0.332E-05 0.330E-08 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-8 Sheet 2 of 5 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Stream Number(a) 6 7 8 9 10 Annual Flow, gal./yr 10220 21360 76200 0.0 12690 Nuclide Concentration, Ci/cc H-3 0.110E-03 0.160E-02 0.469E-03 0.0 0.491E+00 Cr-51 0.176E-06 0.190E-05 0.567E-06 0.0 0.029E-03 Mn-54 0.303E-07 0.430E-06 0.126E-06 0.0 0.136E-03 Fe-55 0.548E-07 0.367E-07 0.208E-07 0.0 0.414E-03 Co-58 0.160E-05 0.208E-04 0.614E-05 0.0 0.733E-02 Fe-59 0.941E-07 0.114E-05 0.338E-06 0.0 0.435E-03 Co-60 0.194E-06 0.282E-05 0.828E-06 0.0 0.871E-03 Sr-89 0.224E-06 0.277E-05 0.821E-06 0.0 0.103E-02 Sr-90 0.110E-07 0.160E-06 0.469E-07 0.0 0.492E-04 Y-90 0.541E-08 0.156E-06 0.449E-07 0.0 0.720E-05 Sr-91 0.163E-09 0.287E-17 0.314E-10 0.0 0.328E-04 Y-91 0.199E-06 0.254E-05 0.750E-06 0.0 0.913E-03 Sr-92 0.122E-17 0.0 0.235E-16 0.0 0.434E-05 Y-92 0.314E-15 0.0 0.605E-16 0.0 0.563E-06 Zr-95 0.546E-07 0.701E-06 0.207E-06 0.0 0.250E-03 Nb-95 0.556E-07 0.797E-06 0.234E-06 0.0 0.249E-03 Mo-99 0.172E-05 0.134E-05 0.707E-06 0.0 0.127E-01 I-131 0.197E-03 0.102E-02 0.324E-03 0.0 0.105E+01 Te-132 0.124E-04 0.140E-04 0.631E-05 0.0 0.860E-01 I-132 0.128E-04 0.144E-04 0.651E-05 0.0 0.935E-01 I-133 0.158E-04 0.172E-07 0.305E-05 0.0 0.366E+00 Cs-134 0.194E-04 0.279E-03 0.821E-04 0.0 0.870E-01 I-134 0.211E-37 0.0 0.406E-38 0.0 0.345E-02 I-135 0.102E-07 0.171E-19 0.196E-08 0.0 0.212E-01 Cs-136 0.416E-05 0.319E-04 0.976E-05 0.0 0.207E-01 Cs-137 0.347E-04 0.506E-03 0.149E-03 0.0 0.156E+00 Ba-140 0.313E-06 0.238E-05 0.729E-06 0.0 0.157E-02 La-140 0.301E-06 0.274E-05 0.825E-06 0.0 0.118E-02 Ce-144 0.330E-07 0.468E-06 0.137E-06 0.0 0.149E-03 Pr-144 0.330E-07 0.468E-06 0.137E-06 0.0 0.149E-03 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-8 Sheet 3 of 5 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Stream Number(a) 58A&B 59A&B 60A&B 11 12 Annual Flow, gal./yr 11000 0.0 0.0 23990 0.0 Nuclide Concentration, Ci/cc H-3 0.110E-01 0.0 0.0 0.264E+00 0.0 Cr-51 0.185E-04 0.0 0.0 0.315E-03 0.0 Mn-54 0.304E-05 0.0 0.0 0.712E-04 0.0 Fe-55 0.915E-05 0.0 0.0 0.614E-05 0.0 Co-58 0.164E-03 0.0 0.0 0.345E-02 0.0 Fe-59 0.970E-05 0.0 0.0 0.189E-03 0.0 Co-60 0.194E-04 0.0 0.0 0.467E-03 0.0 Sr-89 0.230E-04 0.0 0.0 0.460E-03 0.0 Sr-90 0.110E-05 0.0 0.0 0.265E-04 0.0 Y-90 0.160E-06 0.0 0.0 0.259E-04 0.0 Sr-91 0.499E-06 0.0 0.0 0.694E-15 0.0 Y-91 0.204E-04 0.0 0.0 0.420E-03 0.0 Sr-92 0.273E-10 0.0 0.0 0.0 0.0 Y-92 0.245E-09 0.0 0.0 0.0 0.0 Zr-95 0.558E-05 0.0 0.0 0.116E-03 0.0 Nb-95 0.556E-05 0.0 0.0 0.132E-03 0.0 Mo-99 0.281E-03 0.0 0.0 0.223E-03 0.0 I-131 0.234E-01 0.0 0.0 0.170E+00 0.0 Te-132 0.190E-02 0.0 0.0 0.233E-02 0.0 I-132 0.197E-02 0.0 0.0 0.241E-02 0.0 I-133 0.770E-02 0.0 0.0 0.301E-05 0.0 Cs-134 0.194E-02 0.0 0.0 0.463E-01 0.0 I-134 0.101E-18 0.0 0.0 0.0 0.0 I-135 0.146E-03 0.0 0.0 0.910E-17 0.0 Cs-136 0.462E-03 0.0 0.0 0.530E-02 0.0 Cs-137 0.347E-02 0.0 0.0 0.838E-01 0.0 Ba-140 0.349E-04 0.0 0.0 0.396E-03 0.0 La-140 0.266E-04 0.0 0.0 0.454E-03 0.0 Ce-144 0.332E-05 0.0 0.0 0.775E-04 0.0 Pr-144 0.332E-05 0.0 0.0 0.775E-04 0.0 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-8 Sheet 4 of 5 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Stream Number(a) 13 14 15 16 17 Annual Flow, gal./yr 0.0 23990 700,000 723,990 800,190 Nuclide Concentration, Ci/cc H-3 0.0 0.264E+00 0.101E+01 0.985E-00 0.891E+00 Cr-51 0.0 0.315E-06 0.806E-12 0.104E-07 0.634E-07 Mn-54 0.0 0.712E-07 0.217E-12 0.236E-08 0.141E-07 Fe-55 0.0 0.614E-08 0.468E-14 0.203E-09 0.216E-08 Co-58 0.0 0.345E-05 0.957E-11 0.114E-06 0.688E-06 Fe-59 0.0 0.189E-06 0.503E-12 0.626E-08 0.379E-07 Co-60 0.0 0.467E-06 0.146E-11 0.155E-07 0.929E-07 Sr-89 0.0 0.460E-06 0.126E-11 0.152E-07 0.919E-07 Sr-90 0.0 0.265E-07 0.806E-13 0.878E-09 0.526E-08 Y-90 0.0 0.259E-07 0.126E-09 0.980E-09 0.516E-08 Sr-91 0.0 0.694E-18 0.116E-27 0.0 0.299E-11 Y-91 0.0 0.420E-06 0.121E-05 0.118E-05 0.114E-05 Sr-92 0.0 0.0 0.136E-69 0.0 0.224E-17 Y-92 0.0 0.0 0.463E-50 0.0 0.576E-17 Zr-95 0.0 0.116E-06 0.332E-12 0.384E-08 0.232E-07 Nb-95 0.0 0.132E-06 0.393E-12 0.437E-08 0.262E-07 Mo-99 0.0 0.223E-06 0.191E-06 0.192E-06 0.241E-06 I-131 0.0 0.170E-03 0.337E-07 0.567E-05 0.360E-04 Te-132 0.0 0.233E-05 0.232E-10 0.772E-07 0.673E-06 I-132 0.0 0.241E-05 0.312E-10 0.799E-07 0.690E-06 I-133 0.0 0.301E-08 0.121E-13 0.997E-10 0.291E-06 Cs-134 0.0 0.232E-02 0.196E-05 0.788E-04 0.791E-04 I-134 0.0 0.0 0.0 0.0 0.0 I-135 0.0 0.910E-20 0.187E-09 0.0 0.187E-09 Cs-136 0.0 0.265E-03 0.171E-05 0.898E-05 0.903E-05 Cs-137 0.0 0.419E-02 0.357E-05 0.142E-03 0.143E-03 Ba-140 0.0 0.396E-06 0.906E-12 0.131E-07 0.183E-07 La-140 0.0 0.454E-06 0.101E-11 0.150E-07 0.921E-07 Ce-144 0.0 0.775E-07 0.232E-12 0.257E-08 0.154E-07 Pr-144 0.0 0.776E-07 0.232E-12 0.257E-08 0.154E-07 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-8 Sheet 5 of 5 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Stream Number(a) 18 19 Annual Flow, gal./yr 800,190 (Refer to Section 11.2.3.13.2.4) Nuclide Concentration, Ci/cc H-3 0.891E+00 0.774E-06 Cr-51 0.634E-07 0.551E-13 Mn-54 0.141E-07 0.123E-13 Fe-55 0.216E-08 0.188E-14 Co-58 0.688E-06 0.598E-12 Fe-59 0.379E-07 0.329E-13 Co-60 0.929E-07 0.807E-13 Sr-89 0.919E-07 0.799E-13 Sr-90 0.526E-08 0.457E-14 Y-90 0.516E-08 0.448E-14 Sr-91 0.299E-11 0.260E-17 Y-91 0.114E-05 0.991E-12 Sr-92 0.224E-19 0.195E-23 Y-92 0.576E-17 0.501E-23 Zr-95 0.232E-07 0.202E-13 Nb-95 0.262E-07 0.228E-13 Mo-99 0.241E-06 0.209E-12 I-131 0.360E-04 0.313E-10 Te-132 0.673E-06 0.585E-12 I-132 0.690E-06 0.600E-12 I-133 0.291E-06 0.253E-12 Cs-134 0.791E-04 0.687E-10 I-134 0.0 0.0 I-135 0.187E-09 0.164E-15 Cs-136 0.903E-05 0.785E-11 Cs-137 0.143E-03 0.124E-09 Ba-140 0.813E-07 0.706E-13 La-140 0.921E-07 0.800E-13 Ce-144 0.154E-07 0.134E-13 Pr-144 0.154E-07 0.134E-13 (a) Refer to Figure 11.2-2 for waste stream number identification. Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-9 Sheet 1 of 5 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ANNUAL FLOW AND ACTIVITY CONCENTRATION OF PROCESS STREAMS FOR NORMAL OPERATION CASE Stream Number(a) 1 2 3 4 5 Annual Flow, gal./yr 29800 37530 100 220 65480 Nuclide Concentration, Ci/cc H-3 0.009E+00 0.307E+00 0.103E+01 0.103E-01 0.103E-04 Cr-51 0.153E-02 0.582E-03 0.185E-02 0.185E-04 0.176E-07 Mn-54 0.240E-03 0.914E-04 0.304E-03 0.304E-05 0.303E-08 Fe-55 0.120E-02 0.457E-03 0.915E-03 0.915E-05 0.548E-08 Co-58 0.131E-01 0.499E-02 0.164E-01 0.164E-03 0.160E-06 Fe-59 0.787E-03 0.299E-03 0.970E-03 0.970E-05 0.941E-08 Co-60 0.153E-02 0.582E-03 0.194E-02 0.194E-04 0.194E-07 Sr-89 0.219E-03 0.831E-04 0.271E-03 0.271E-05 0.263E-08 Sr-90 0.104E-04 0.395E-05 0.132E-04 0.132E-06 0.132E-09 Y-90 0.186E-05 0.706E-06 0.194E-05 0.194E-07 0.651E-10 Sr-91 0.142E-03 0.539E-04 0.590E-05 0.590E-07 0.193E-11 Y-91 0.197E-03 0.748E-04 0.245E-03 0.245E-05 0.239E-08 Sr-92 0.578E-04 0.220E-04 0.329E-09 0.329E-11 0.147E-19 Y-92 0.731E-05 0.278E-05 0.295E-08 0.295E-10 0.378E-17 Zr-95 0.535E-04 0.204E-04 0.667E-04 0.667E-06 0.653E-09 Nb-95 0.525E-04 0.199E-04 0.667E-04 0.667E-06 0.667E-09 Mo-99 0.437E-02 0.166E-02 0.341E-02 0.341E-04 0.209E-07 I-131 0.251E+00 0.955E-01 0.269E+00 0.269E-02 0.226E-05 Te-132 0.273E-01 0.104E-01 0.227E-01 0.227E-03 0.148E-06 I-132 0.917E-01 0.349E-01 0.234E-01 0.234E-03 0.153E-06 I-133 0.349E+00 0.133E+00 0.913E-01 0.913E-01 0.187E-06 Cs-134 0.186E-01 0.706E-02 0.236E-01 0.236E-03 0.235E-06 I-134 0.469E-01 0.178E-01 0.124E-17 0.124E-19 0.259E-39 I-135 0.196E+00 0.747E-01 0.175E-02 0.175E-04 0.122E-09 Cs-136 0.492E-02 0.187E-02 0.562E-02 0.562E-04 0.505E-07 Cs-137 0.328E-01 0.125E-01 0.417E-01 0.417E-03 0.417E-06 Ba-140 0.372E-03 0.141E-03 0.424E-03 0.424E-05 0.381E-08 La-140 0.131E-03 0.499E-04 0.323E-03 0.323E-05 0.366E-08 Ce-144 0.317E-04 0.120E-04 0.401E-04 0.401E-06 0.399E-09 Pr-144 0.317E-04 0.120E-04 0.401E-04 0.401E-04 0.399E-09 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-9 Sheet 2 of 5 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Stream Number(a) 6 7 8 9 10 Annual Flow, gal./yr 15330 67650 80100 1000 19000 Nuclide Concentration, Ci/cc H-3 0.103E-03 0.527E+00 0.278E-04 0.103E-01 0.459E+00 Cr-51 0.176E-06 0.705E-03 0.477E-07 0.195E-02 0.828E-03 Mn-54 0.303E-07 0.152E-03 0.820E-08 0.306E-03 0.136E-03 Fe-55 0.548E-07 0.216E-04 0.148E-07 0.153E-02 0.413E-03 Co-58 0.160E-05 0.748E-02 0.434E-06 0.167E-01 0.731E-02 Fe-59 0.941E-07 0.414E-03 0.255E-07 0.100E-02 0.434E-03 Co-60 0.194E-06 0.994E-03 0.526E-07 0.195E-02 0.869E-03 Sr-89 0.263E-07 0.118E-03 0.713E-08 0.278E-03 0.121E-03 Sr-90 0.132E-08 0.678E-05 0.357E-09 0.132E-04 0.590E-05 Y-90 0.651E-09 0.663E-05 0.176E-09 0.236E-05 0.869E-06 Sr-91 0.193E-10 0.365E-14 0.521E-11 0.181E-03 0.374E-05 Y-91 0.239E-07 0.110E-03 0.648E-08 0.250E-03 0.109E-03 Sr-92 0.147E-18 0.0 0.397E-19 0.737E-04 0.465E-06 Y-92 0.378E-16 0.0 0.102E-16 0.931E-05 0.601E-07 Zr-95 0.653E-08 0.302E-04 0.177E-08 0.681E-04 0.298E-04 Nb-95 0.667E-08 0.338E-04 0.181E-08 0.667E-04 0.298E-04 Mo-99 0.209E-06 0.929E-04 0.566E-07 0.556E-02 0.154E-02 I-131 0.226E-04 0.491E-01 0.613E-05 0.320E+00 0.121E+00 Te-132 0.148E-05 0.897E-03 0.400E-06 0.347E-01 0.102E-01 I-132 0.153E-05 0.926E-03 0.413E-06 0.117E+00 0.110E-01 I-133 0.187E-05 0.348E-05 0.507E-06 0.445E+00 0.430E-01 Cs-134 0.235E-05 0.120E-01 0.637E-06 0.236E-01 0.105E-01 I-134 0.259E-38 0.0 0.701E-39 0.598E-01 0.377E-03 I-135 0.122E-08 0.103E-15 0.330E-09 0.250E+00 0.235E-02 Cs-136 0.505E-06 0.152E-02 0.137E-06 0.625E-02 0.252E-02 Cs-137 0.417E-05 0.214E-01 0.113E-05 0.417E-01 0.186E-01 Ba-140 0.381E-07 0.114E-03 0.103E-07 0.473E-03 0.190E-03 La-140 0.366E-07 0.130E-03 0.991E-08 0.167E-03 0.143E-03 Ce-144 0.399E-08 0.200E-04 0.108E-08 0.403E-04 0.179E-04 Pr-144 0.399E-08 0.200E-04 0.108E-08 0.403E-04 0.179E-04 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-9 Sheet 3 of 5 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Stream Number(a) 58A&B 59A&B 60A&B 11 12 Annual Flow, gal./yr 11000 300 0 98650 0 Nuclide Concentration, Ci/cc H-3 0.1832-01 0.0 0.0 0.461E+00 0.0 Cr-51 0.185E-04 0.0 0.0 0.469E-03 0.0 Mn-54 0.304E-05 0.0 0.0 0.130E-03 0.0 Fe-55 0.915E-05 0.0 0.0 0.305E-05 0.0 Co-58 0.164E-05 0.0 0.0 0.587E-02 0.0 Fe-59 0.970E-05 0.0 0.0 0.305E-03 0.0 Co-60 0.194E-04 0.0 0.0 0.867E-03 0.0 Sr-89 0.271E-05 0.0 0.0 0.889E-04 0.0 Sr-90 0.132E-06 0.0 0.0 0.593E-05 0.0 Y-90 0.194E-07 0.0 0.0 0.590E-05 0.0 Sr-91 0.590E-07 0.0 0.0 0.101E-15 0.1 Y-91 0.245E-05 0.0 0.0 0.849E-04 0.0 Sr-92 0.329E-11 0.0 0.0 0.0 0.0 Y-92 0.295E-10 0.0 0.0 0.0 0.0 Zr-95 0.667E-06 0.0 0.0 0.234E-04 0.0 Nb-95 0.667E-06 0.0 0.0 0.286E-04 0.0 Mo-99 0.341E-04 0.0 0.0 0.137E-04 0.0 I-131 0.269E-02 0.0 0.0 0.181E-01 0.0 Te-132 0.227E-03 0.0 0.0 0.149E-03 0.0 I-132 0.234E-03 0.0 0.0 0.154E-03 0.0 I-133 0.913E-03 0.0 0.0 0.197E-06 0.0 Cs-134 0.236E-03 0.0 0.0 0.104E-01 0.0 I-134 0.124E-19 0.0 0.0 0.0 0.0 I-135 0.175E-04 0.0 0.0 0.241E-17 0.0 Cs-136 0.562E-04 0.0 0.0 0.758E-03 0.0 Cs-137 0.417E-03 0.0 0.0 0.187E-01 0.0 Ba-140 0.424E-05 0.0 0.0 0.561E-04 0.0 La-140 0.323E-05 0.0 0.0 0.645E-04 0.0 Ce-144 0.401E-06 0.0 0.0 0.170E-04 0.0 Pr-144 0.401E-06 0.0 0.0 0.170E-04 0.0 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-9 Sheet 4 of 5 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Stream Number(a) 13 14 15 16 17 Annual Flow, gal./yr 0 98650 7000000 798650 878950 Nuclide Concentration, Ci/cc H-3 0.0 0.461E+00 0.906E+00 0.851E+00 0.773E+00 Cr-51 0.0 0.469E-06 0.806-12 0.579E-07 0.570E-07 Mn-54 0.0 0.130E-06 0.217E-12 0.161E-07 0.154E-07 Fe-55 0.0 0.305E-08 0.468E-14 0.377E-09 0.169E-08 Co-58 0.0 0.587E-05 0.957E-11 0.725E-06 0.698E-06 Fe-59 0.0 0.305E-06 0.503E-12 0.377E-07 0.366E-07 Co-60 0.0 0.867E-06 0.146E-11 0.107E-06 0.102E-06 Sr-89 0.0 0.889E-07 0.151E-12 0.110E-07 0.106E-07 Sr-90 0.0 0.593E-08 0.957E-14 0.732E-09 0.698E-09 Y-90 0.0 0.590E-08 0.151E-10 0.742E-09 0.690E-09 Sr-91 0.0 0.101E-18 0.136E-28 0.125E-19 0.475E-12 Y-91 0.0 0.849E-07 0.141E-06 0.134E-06 0.122E-06 Sr-92 0.0 0.0 0.161E-70 0.0 0.0 Y-92 0.0 0.0 0.554E-51 0.0 0.0 Zr-95 0.0 0.234E-07 0.398E-13 0.289E-08 0.279E-08 Nb-95 0.0 0.286E-07 0.473E-13 0.353E-09 0.337E-09 Mo-99 0.0 0.137E-07 0.227E-07 0.216E-07 0.248E-07 I-131 0.0 0.181E-04 0.408E-08 0.224E-05 0.259E-05 Te-132 0.0 0.149E-06 0.277E-11 0.184E-07 0.532E-07 I-132 0.0 0.154E-06 0.373E-11 0.190E-07 0.549E-07 I-133 0.0 0.197E-09 0.141E-14 0.235E-10 0.462E-07 Cs-134 0.0 0.520E-03 0.237E-06 0.644E-04 0.586E-04 I-134 0.0 0.0 0.0 0.0 0.0 I-135 0.0 0.241E-20 0.131E-30 0.0 0.301E-10 Cs-136 0.0 0.379E-04 0.206E-07 0.470E-05 0.428E-05 Cs-137 0.0 0.935E-03 0.428E-06 0.116E-03 0.106E-03 Ba-140 0.0 0.561E-07 0.106E-12 0.693E-08 0.724E-08 La-140 0.0 0.654E-07 0.121E-12 0.797E-08 0.815E-08 Ce-144 0.0 0.170E-07 0.282E-13 0.210E-08 0.201E-08 Pr-144 0.0 0.170E-07 0.282E-13 0.210E-08 0.201E-08 Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-9 Sheet 5 of 5 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Stream Number(a) 18 19 Annual Flow, gal./yr 878750 (Refer to Section 11.2.3.13.2.4) Nuclide Concentration, Ci/cc H-3 0.773E+00 0.738E-06 Cr-51 0.570E-07 0.543E-13 Mn-54 0.154E-07 0.147E-13 Fe-55 0.169E-08 0.161E-14 Co-58 0.698E-06 0.666E-12 Fe-59 0.366E-07 0.349E-13 Co-60 0.102E-06 0.973E-13 Sr-89 0.106E-07 0.101E-13 Sr-90 0.698E-09 0.635E-15 Y-90 0.690E-09 0.658E-15 Sr-91 0.475E-12 0.453E-18 Y-91 0.122E-06 0.116E-12 Sr-92 0.0 0.0 Y-92 0.0 0.0 Zr-95 0.279E-08 0.266E-14 Nb-95 0.337E-09 0.306E-15 Mo-99 0.248E-07 0.237E-13 I-131 0.259E-05 0.247E-12 Te-132 0.532E-07 0.508E-13 I-132 0.549E-07 0.524E-13 I-133 0.462E-07 0.441E-13 Cs-134 0.586E-04 0.559E-10 I-134 0.0 0.0 I-135 0.301E-10 0.287E-16 Cs-136 0.428E-05 0.408E-11 Cs-137 0.106E-03 0.101E-09 Ba-140 0.724E-08 0.691E-14 La-140 0.815E-08 0.778E-14 Ce-144 0.201E-08 0.192E-14 Pr-144 0.201E-08 0.192E-14 (a) Refer to Figure 11.2-3 for waste stream number identification. Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-10 Sheet 1 of 3 EQUIPMENT DESIGN
SUMMARY
DATA LIQUID RADWASTE SYSTEM Volume Pressure, Temperature, Tank Quantity Type gal psig °F Material Reactor coolant drain 1(f) Horiz 400 25 267 SS Laundry and hot shower 2(a) Vert 1,000 5 300 CS Chemical drain 1(a) Horiz 1,000 0 150 SS Aux. Building sump 1(a) - 7,300 0 120 SS Misc. equip. drain 1(a) - 5,500 0 180 SS Processed waste 2(a) Vert 15,000 0 180 SS receiver Floor drain receiver 2(a) Vert 15,000 0 180 SS Equipment drain 2(a) Vert 15,000 0 180 SS receiver Demineralizer 2(a) Vert 15,000 0 180 (b) regenerant receiver Laundry/distillate 2(a) Vert 25,000 0 150 SS Containment structure 2(f) NA 700 0 140 (g) sump Reactor cavity 1(f) NA 300 0 140 (g) sump RHR pump 1(f) NA 500 0 180 (g) room sump Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-10 Sheet 2 of 3 Flow, Head, Pressure, Temp, Pump Qty. Type gpm ft psig °F Mtl.(d) Reactor coolant 2(f) Vert 150 170 150 300 SS drain tank Cent. Chemical drain 1(a) Horiz 20 122 150 180 SS Cent.(c) Laundry and hot 2(a) Horiz 20 122 150 180 SS shower drain Cent.(c) Misc. equip. 2(a) Vert 50 45 150 180 CI(h) drain tank Cent.(c) Floor drain 2(a) Vert. 50 300 150 180 SS receiver Cent.(c) Equipment drain 2(a) Vert 50 300 150 180 SS receiver Cent.(c) Processed waste 2(a) Vert 50 122 150 180 SS receiver Cent.(c) Containment 4(f) Vert 50 45 150 140 CI(h) structure Cent.(c) sump Demineralizer 2(a) Vert 50 300 150 180 SS regenerant Cent.(c) receiver Reactor cavity 2(f) Horiz 30 75 150 140 CI(h) sump Cent.(e) RHR pump room 4(f) Vert 50 40 150 180 CI(h) sumps Cent Aux. Building 2(a) Vert 50 45 150 180 CI(h) sump Cent. Laundry/distillate 2(a) Horiz 150 200 150 115 SS Cent. Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-10 Sheet 3 of 3 Miscellaneous Quantity Capacity Type Radwaste filters 5(a) 50 gpm Cartridge Media filters 2(a) 50 gpm Media bed Ion exchangers 2(a) 50 gpm Bead resin Spent Resin Transfer 2(a) 120 gpm Cartridge Filters (a) Equipment common to Units 1 and 2 (b) Carbon steel with a neoprene lining (c) Mechanical seal provided (d) Wetted surfaces only (e) Deep well jet pump (f) Per unit (g) Concrete (h) Cast iron (CI) Ni Resist Type D2 (original material). Replacement materials may vary. Revision 24 September 2018
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-11 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED PARAMETERS USED IN TRITIUM ANALYSIS FOR PLANT WATER SOURCES Parameter Value Primary system volume, gal. 94,000 Volume of water in primary water storage tank, gal. 200,000 Volume of water in refueling water storage tank, gal. 450,000 Volume of water in spent fuel pool, gal. 442,000 Percent of mixing of spent fuel pool water with refueling canal during refueling 15 Evaporative loss from spent fuel pool during operation, gal./day 500 Evaporative loss from spent fuel pool during refueling, gal./day 1,360 Evaporative loss from refueling canal during refueling, gal./day 485 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-13 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED CALCULATED AND ASSUMED HOLDUP TIMES FOR LIQUID WASTE SYSTEM TANKS Fill Release Assumed Decay Capacity, Time, Time, Time, Tank Qty. gal. Days Days Days Hours Liquid holdup 5(a) 83,000 46.1 3.07 21.0 504.0 Monitor 2 25,000 0.93 0.093 0.5 12.0 Waste Condensate 1 15,000 2.23 0.56 1.0 24.0 Reactor coolant drain 1 400 116.8 7.4E-4 2.0 48.0 Laundry and hot shower 2(a) 1,000 3.6 0.03 2.0 48.0 Chemical drain 1(a) 1,000 7.6 0.014 2.0 48.0 Miscellaneous equipment drain 1(a) 20,000 123.0 0.3 2.0 48.0 Equipment drain receiver 1 15,000 42.5 0.56 14.0 336.0 Floor drain receiver 1 15,000 33.7 0.08 14.0 336.0 Waste regenerant 1 15,000 - - 18.0 432.0 (a) Equipment common to both units. Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-14 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ESTIMATED ANNUAL ACTIVITY RELEASE FOR DESIGN BASIS CASE (ONE UNIT) Turbine Bldg Condensate Polishing Radwaste System, Sump, System, One Unit Total, Nuclide Ci/yr Ci/yr Ci/yr Ci/yr H-3 0.135E+04 0.0 0.0 0.135E+04 Cr-51 0.960E-04 0.0 0.0 0.960E-04 Mn-54 0.213E-04 0.0 0.0 0.213E-04 Fe-55 0.327E-05 0.0 0.0 0.327E-05 Co-58 0.104E-02 0.0 0.0 0.104E-02 Fe-59 0.574E-04 0.0 0.0 0.574E-04 Co-60 0.141E-03 0.0 0.0 0.141E-03 Sr-89 0.139E-03 0.0 0.0 0.139E-03 Sr-90 0.797E-03 0.0 0.0 0.797E-03 Y-90 0.781E-05 0.0 0.0 0.781E-05 Sr-91 0.453E-08 0.0 0.0 0.453E-08 Y-91 0.173E-02 0.0 0.0 0.173E-02 Sr-92 0.339E-14 0.0 0.0 0.339E-14 Y-92 0.872E-14 0.0 0.0 0.872E-14 Zr-95 0.351E-04 0.0 0.0 0.351E-04 Mb-95 0.397E-04 0.0 0.0 0.397E-04 Mo-99 0.365E-03 0.0 0.0 0.365E-03 I-131 0.545E-01 0.0 0.0 0.545E-01 Te-132 0.102E-02 0.0 0.0 0.102E-02 I-132 0.104E-02 0.0 0.0 0.104E-02 I-133 0.441E-03 0.0 0.0 0.441E-03 Cs-134 0.120E-02 0.0 0.0 0.120E-02 I-134 0.0 0.0 0.0 0.0 I-135 0.284E-06 0.0 0.0 0.284E-06 Cs-136 0.137E-01 0.0 0.0 0.137E-01 Cs-137 0.216E+00 0.0 0.0 0.216E+00 Ba-140 0.123E-03 0.0 0.0 0.123E-03 La-140 0.139E-03 0.0 0.0 0.139E-03 Ce-144 0.233E-04 0.0 0.0 0.233E-04 Pr-144 0.233E-04 0.0 0.0 0.233E-04 Total (excluding H-3) 0.411E+00 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-15 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ESTIMATED ANNUAL ACTIVITY RELEASE FOR NORMAL OPERATION CASE (ONE UNIT) Turbine Bldg Condensate Polishing Radwaste System, Sump, System(a), One Unit Total, Nuclide Curies/yr Curies/yr Curies/yr Curies/yr H-3 0.129E+04 0.430E+01 0.0 0.129E+04 Cr-51 0.948E-04 0.390E-06 0.106E-02 0.126E-02 Mn-54 0.256E-04 0.645E-07 0.246E-03 0.297E-03 Fe-55 0.281E-05 0.302E-06 0.120E-03 0.788E-03 Co-58 0.116E-02 0.334E-05 0.113E-01 0.136E-01 Fe-59 0.609E-04 0.207E-06 0.644E-03 0.771E-03 Co-60 0.170E-03 0.414E-06 0.163E-02 0.197E-02 Sr-89 0.176E-04 0.565E-07 0.183E-03 0.220E-03 Sr-90 0.111E-05 0.271E-08 0.108E-04 0.130E-04 Y-90 0.115E-05 0.764E-08 0.751E-05 0.948E-05 Sr-91 0.790E-09 0.239E-07 0.607E-07 0.934E-07 Y-91 0.203E-03 0.780E-06 0.232E-02 0.276E-02 Sr-92 0.0 0.310E-08 0.0 0.339E-08 Y-92 0.0 0.207E-07 0.0 0.226E-07 Zr-95 0.464E-05 0.143E-07 0.473E-04 0.568E-04 Nb-95 0.530E-06 0.207E-07 0.720E-04 0.794E-04 Mo-99 0.412E-04 0.103E-03 0.356E-01 0.391E-01 I-131 0.431E-02 0.895E-03 0.114E+01 0.125E+01 Te-132 0.885E-04 0.668E-05 0.288E-02 0.325E-02 I-132 0.913E-04 0.191E-03 0.0 0.309E-03 I-133 0.768E-04 0.103E-02 0.371E-01 0.418E-01 Cs-134 0.975E-03 0.143E-04 0.310E-01 0.141E+00 I-134 0.0 0.263E-04 0.0 0.288E-04 I-135 0.501E-07 0.446E-03 0.162E-03 0.488E-03 Cs-136 0.712E-02 0.302E-05 0.312E-02 0.112E-01 Cs-137 0.176E-00 0.207E-04 0.452E-01 0.242E+00 Ba-140 0.120E-04 0.955E-07 0.174E-04 0.323E-04 La-140 0.136E-04 0.501E-07 0.403E-04 0.114E-03 Ce-144 0.334E-05 0.796E-08 0.317E-04 0.383E-04 Pr-144 0.334E-05 0.111E-07 0.317E-04 0.383E-04 Total (excluding H-3) 0.160E+01 (a) Resin regenerant discharge. Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-16 Sheet 1 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ANNUAL FLOW AND ACTIVITY CONCENTRATION OF PROCESS STREAMS FOR STEAM GENERATOR BLOWDOWN SYSTEM FOR NORMAL OPERATION CASE BASED ON ORIGINAL SYSTEM DESIGN Stream Number A B C D E Annual Flow, gal./yr 76,300,000 0.0 42,865,000 42,865,000 0.0 Nuclide Concentration, Ci/cc H-3 0.420E-03 0.0 0.303E-03 0.303E-03 0.0 Cr-51 0.790E-07 0.0 0.102E-06 0.102E-08 0.0 Mn-54 0.130E-07 0.0 0.168E-07 0.168E-09 0.0 Fe-55 0.630E-07 0.0 0.013E-07 0.813E-09 0.0 Co-58 0.670E-06 0.0 0.864E-06 0.864E-08 0.0 Fe-59 0.420E-07 0.0 0.542E-07 0.542E-07 0.0 Co-60 0.830E-07 0.0 0.107E-06 0.107E-08 0.0 Sr-89 0.110E-07 0.0 0.142E-07 0.142E-09 0.0 Sr-90 0.550E-09 0.0 0.709E-09 0.709E-11 0.0 Y-90 0.960E-09 0.0 0.124E-08 0.124E-08 0.0 Sr-91 0.560E-08 0.0 0.722E-08 0.722E-10 0.0 Y-91 0.980E-07 0.0 0.126E-06 0.126E-06 0.0 Sr-92 0.140E-08 0.0 0.181E-08 0.181E-10 0.0 Y-92 0.280E-08 0.0 0.361E-08 0.361E-08 0.0 Zr-95 0.280E-08 0.0 0.361E-08 0.361E-10 0.0 Nb-95 0.290E-08 0.0 0.374E-08 0.374E-10 0.0 Mo-99 0.220E-04 0.0 0.284E-04 0.284E-04 0.0 I-131 0.140E-04 0.0 0.172E-04 0.172E-06 0.0 Te-132 0.140E-05 0.0 0.181E-05 0.181E-07 0.0 I-132 0.300E-05 0.0 0.368E-05 0.368E-07 0.0 I-133 0.170E-04 0.0 0.208E-04 0.208E-06 0.0 Cs-134 0.200E-05 0.0 0.258E-05 0.129E-05 0.0 I-134 0.520E-06 0.0 0.637E-06 0.637E-08 0.0 I-135 0.710E-05 0.0 0.870E-05 0.870E-07 0.0 Cs-136 0.500E-06 0.0 0.645E-06 0.322E-06 0.0 Cs-137 0.340E-05 0.0 0.439E-05 0.219E-05 0.0 Ba-140 0.190E-07 0.0 0.245E-07 0.245E-09 0.0 La-140 0.810E-08 0.0 0.104E-07 0.104E-09 0.0 Ce-144 0.170E-08 0.0 0.219E-08 0.219E-10 0.0 Pr-144 0.170E-08 0.0 0.219E-08 0.219E-10 0.0 Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-16 Sheet 2 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Stream Number F G H I Annual Flow, gal./yr 0.0 42,865,000 0.0 2,600,00 Nuclide Concentration, Ci/cc H-3 0.0 0.303E-03 0.0 0.540E-03 Cr-51 0.0 0.102E-08 0.0 0.490E-10 Mn-54 0.0 0.168E-09 0.0 0.810E-11 Fe-55 0.0 0.813E-09 0.0 0.380E-10 Co-58 0.0 0.864E-08 0.0 0.420E-09 Fe-59 0.0 0.542E-09 0.0 0.260E-10 Co-60 0.0 0.107E-08 0.0 0.520E-10 Sr-89 0.0 0.142E-09 0.0 0.710E-11 Sr-90 0.0 0.709E-11 0.0 0.340E-12 Y-90 0.0 0.124E-08 0.0 0.870E-12 Sr-91 0.0 0.722E-10 0.0 0.300E-11 Y-91 0.0 0.126E-06 0.0 0.890E-10 Sr-92 0.0 0.181E-10 0.0 0.390E-12 Y-92 0.0 0.361E-08 0.0 0.250E-11 Zr-95 0.0 0.361E-10 0.0 0.180E-11 Nb-95 0.0 0.374E-10 0.0 0.260E-11 Mo-99 0.0 0.284E-04 0.0 0.130E-07 I-131 0.0 0.172E-06 0.0 0.110E-06 Te-132 0.0 0.101E-07 0.0 0.840E-09 I-132 0.0 0.368E-07 0.0 0.240E-07 I-133 0.0 0.208E-06 0.0 0.130E-06 Cs-134 0.0 0.129E-05 0.0 0.180E-08 I-134 0.0 0.637E-08 0.0 0.330E-08 I-135 0.0 0.870E-07 0.0 0.560E-07 Cs-136 0.0 0.322E-06 0.0 0.380E-09 Cs-137 0.0 0.219E-05 0.0 0.260E-08 Ba-140 0.0 0.245E-09 0.0 0.120E-10 La-140 0.0 0.104E-09 0.0 0.730E-11 Ce-144 0.0 0.219E-10 0.0 0.100E-11 Pr-144 0.0 0.219E-10 0.0 0.140E-11 (a) Refer to Figures 11.2-4 and 11.2-5 for stream number identification. Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-17 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED
SUMMARY
OF ESTIMATED LIQUID WASTE SYSTEM ANNUAL WASTE VOLUMES FOR UNITS 1 AND 2 Total Annual Total Annual Volume, Gallons Volume, Gallons Stream (Design Basis (Normal Number(a) Case) Operation Case) Laundry, showers, handwashers 5 44,620 65,480 Chemical laboratory drains 6 10,220 15,330 Floor drain subsystem 7 21,360 -(b) Equipment drain subsystem 11 23,990 98,650 Steam generator blowdown F&H - -(c) treatment system CPS regenerant wastes - - 2,360,000 Turbine-generator building sump 1 - 5,200,000 CVCS (tritium control) 15 700,000 700,000 Total plant discharge for two units 800,190 8,439,460 (a) Refer to Figures 11.2-2, 11.2-3, and 11.2-5. (b) Floor drain volume included in equipment drain volume (stream 11) for treatment in normal operation case. (c) Analysis assumed no resin regeneration for steam generator blowdown treatment system. Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-18 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ESTIMATED ANNUAL LIQUID EFFLUENT RELEASE FOR NORMAL OPERATION CASE WITH ANTICIPATED OPERATIONAL OCCURRENCES (ONE UNIT) Nuclide Release, curies H-3 1.29 E+03 Cr-51 1.26 E-03 Mn-54 2.97 E-04 Fe-55 7.88 E-04 Co-58 1.36 E-02 Fe-59 7.71 E-04 Co-60 1.97 E-03 Sr-89 2.20 E-04 Sr-90 1.30 E-05 Y-90 9.48 E-06 Sr-91 9.34 E-08 Y-91 2.76 E-03 Sr-92 3.39 E-09 Y-92 2.26 E-08 Zr-95 5.68 E-05 Nb-95 7.94 E-05 Mo-99 3.91 E-02 I-131 1.25 E00 Te-132 3.25 E-03 I-132 3.09 E-04 I-133 4.18 E-02 Cs-134 1.44 E-01 I-134 2.88 E-05 I-135 4.88 E-04 Cs-136 1.12 E-02 Cs-137 2.42 E-01 Ba-140 3.23 E-05 La-140 1.14 E-04 Ce-144 3.83 E-05 Pr-144 3.83 E-05 Total (excluding H-3) 1.75 E00 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-19 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED BASIC ASSUMPTIONS FOR LIQUID PATHWAYS EXPOSURES Plant dilution flow 876000 gpm Length of plant cycle 8760.00 hr Biological and Environmental Parameters Receiving body of water Pacific Ocean Receiving water type Ocean Dilution factor from discharge to swimming water 5.0E 00 Dilution factor from discharge to drinking water 5.0E 00 Dilution factor from discharge to fish 5.0E 00 Dilution factor from discharge to invertebrates 5.0E 00 Dilution factor from discharge to aquatic plants 5.0E 00 Decay time from environment to water = 5.0E-01 Fish = 1.0E 00 Invertebrates = 1.0E 00 Aquatic Plants = 1.0E 00 Consumption by Man, days Decay time from discharge to sediment, days = 1.00E 00 Accumulation time for sediment activity, days = 1.10E 04 Food Consumption Rates, kg/yr Exposure Time Hours/Year Age Group Water Fish Invertebrates Aquatic Plants Swimming Shore Adult 0.0 2.10E 01 5.00E 00 0.0 5.20E 01 1.20E 01 Teenager 0.0 1.60E 01 3.80E 00 0.0 5.20E 01 6.70E 01 Child 0.0 6.90E 00 1.70E 00 0.0 2.90E 01 1.40E 01 Infant 0.0 0.0 0.0 0.0 0.0 0.0 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-20 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED BIOACCUMULATION FACTORS Nuclide Fish Inverteb. Aq. Plants H-3 9.00E-01 9.30E-01 9.30E-01 Cr-51 4.00E 02 2.00E 03 2.00E 03 Mn-54 5.50E 02 4.00E 02 5.50E 03 Fe-55 3.00E 03 2.00E 04 7.30E 02 Co-58 1.00E 02 1.00E 03 1.00E 03 Fe-59 3.00E 03 2.00E 04 7.30E 02 Co-60 1.00E 02 1.00E 03 1.00E 03 Sr-89 2.00E 00 2.00E 01 1.00E 01 Sr-90 2.00E 00 2.00E 01 1.00E 01 Y-90 2.50E 01 1.00E 03 5.00E 03 Sr-91 2.00E 00 2.00E 01 1.00E 01 Y-91 2.50E 01 1.00E 03 5.00E 03 Sr-92 2.00E 00 2.00E 01 1.00E 01 Y-92 2.50E 01 1.00E 03 5.00E 03 Zr-95 2.00E 02 8.00E 01 1.00E 03 Nb-95 3.00E 04 1.00E 02 5.00E 02 Mo-99 1.00E 01 1.00E 01 1.00E 01 I-131 1.00E 01 5.00E 01 1.00E 03 Te-132 1.00E 01 1.00E 02 1.00E 03 I-132 1.00E 01 5.00E 01 1.00E 03 I-133 1.00E 01 5.00E 01 1.00E 03 Cs-134 4.00E 01 2.50E 01 5.00E 01 I-134 1.00E 01 5.00E 01 1.00E 03 I-135 1.00E 01 5.00E 01 1.00E 03 Cs-136 4.00E 01 2.50E 01 5.00E 01 Cs-137 4.00E 01 2.50E 01 5.00E 01 Ba-140 1.00E 01 1.00E 02 5.00E 02 La-140 2.50E 01 1.00E 03 5.00E 03 Ce-144 1.00E 01 6.00E 02 6.00E 02 Pr-144 2.50E 01 1.00E 03 5.00E 03 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-21 Sheet 1 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED EFFLUENT CONCENTRATIONS AFTER INITIAL DILUTION: DESIGN BASIS CASE Nuclide Release Rate(a), Ci/yr Average Yearly Concentration in Discharge, Ci/cc H-3 1.35E+03 7.75E-07 Cr-51 9.60E-05 5.51E-14 Mn-54 2.13E-05 1.22E-14 Fe-55 3.27E-06 1.88E-15 Co-58 1.04E-03 5.97E-13 Fe-59 5.74E-05 3.29E-14 Co-60 1.41E-04 8.09E-14 Sr-89 1.39E-04 7.98E-14 Sr-90 7.97E-06 4.57E-15 Y-90 7.81E-06 4.48E-15 Sr-91 4.53E-09 2.60E-18 Y-91 1.73E-03 9.93E-13 Sr-92 3.39E-15 1.95E-24 Y-92 8.72E-15 5.00E-24 Zr-95 3.51E-05 2.01E-14 Nb-95 3.97E-05 2.28E-14 Mo-99 3.65E-04 2.09E-13 I-131 5.45E-02 3.13E-1 Te-132 1.02E-03 5.85E-13 I-132 1.09E-03 5.97E-13 I-133 4.41E-04 2.53E-13 Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UDPATE TABLE 11.2-21 Sheet 2 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Nuclide Release Rate(a), Ci/yr Average Yearly Concentration in Discharge, Ci/cc Cs-134 1.20E-01 6.89E-11 I-134 0.0 0.0 I-135 2.84E-07 1.63E-16 Cs-136 1.37E-02 7.86E-12 Cs-137 2.16E-01 1.24E-10 Ba-140 1.23E-04 7.06E-14 La-140 1.39E-04 7.98E-14 Ce-144 2.33E-05 1.34E-14 Pr-144 2.33E-05 1.34E-14 Totals 1.35E+03 7.75E-07 Totals excluding H-3 4.11E-01 2.36E-10 (a) One unit Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.2-22 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED EFFLUENT CONCENTRATIONS AFTER INITIAL DILUTION: NORMAL OPERATION CASE Average Yearly Release Concentration Rate(a), in Discharge, Nuclide Ci/yr Ci/cc H-3 1.29E+03 7.42E-07 Cr-51 1.16E-03 6.66E-13 Mn-54 2.72E-04 1.56E-13 Fe-55 1.23E-04 7.06E-14 Co-58 1.25E-02 7.17E-12 Fe-59 7.05E-04 4.05E-13 Co-60 1.80E-03 1.13E-12 Sr-89 2.01E-04 1.15E-13 Sr-90 1.19E-05 6.83E-15 Y-90 8.67E-06 4.97E-15 Sr-91 8.54E-08 4.90E-17 Y-91 2.52E-03 1.45E-12 Sr-92 0.31E-08 1.78E-18 Y-92 0.21E-07 1.20E-17 Zr-95 5.20E-05 2.98E-14 Nb-95 7.26E-05 4.17E-14 Mo-99 3.57E-02 2.05E-11 I-131 1.15E-00 6.60E-10 I-132 2.82E-04 1.62E-13 I-133 3.82E-02 2.19E-11 Cs-134 1.29E-01 7.40E-11 I-134 0.26E-04 1.49E-14 I-135 6.08E-04 3.49E-13 Cs-136 1.02E-02 5.85E-12 Cs-137 2.21E-01 1.27E-10 Ba-140 2.95E-05 1.69E-14 La-140 1.04E-04 5.97E-14 Ce-144 3.51E-05 2.01E-14 Pr-144 3.51E-05 2.01E-14 Totals 1.29E+03 7.42E-07 Totals excluding H-3 1.60E+00 9.24E-10 (a) One unit Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.2-23 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED EFFLUENT CONCENTRATIONS AFTER INITIAL DILUTION: NORMAL OPERATION WITH ANTICIPATED OPERATIONAL OCCURRENCES Average Yearly Release Concentration Rate(a), in Discharge, Nuclide Ci/yr Ci/cc H-3 1.29E+03 7.42E-07 Cr-51 1.26E-03 7.23E-13 Mn-54 2.99E-04 1.70E-13 Fe-55 7.88E-04 4.52E-13 Co-58 1.36E-02 7.80E-12 Fe-59 7.71E-04 4.42E-13 Co-60 1.97E-03 1.13E-12 Sr-89 2.20E-04 1.26E-13 Sr-90 1.30E-05 7.46E-15 Y-90 9.48E-06 5.44E-15 Sr-91 9.34E-08 5.36E-17 Y-91 2.76E-03 1.58E-12 Sr-92 3.39E-09 1.95E-18 Y-92 2.26E-08 1.30E-17 Zr-95 5.68E-05 3.26E-14 Nb-95 7.94E-05 4.56E-14 Mo-99 3.91E-02 2.24E-11 I-131 1.25E+00 7.17E-10 Te-132 3.25E-03 1.86E-12 I-132 3.09E-04 1.77E-13 I-133 4.18E-02 2.40E-10 Cs-134 1.41E-01 8.09E-11 I-134 2.88E-05 1.65E-14 I-135 4.88E-04 2.80E-13 Cs-136 1.12E-02 6.43E-12 Cs-137 2.42E-01 1.39E-10 Ba-140 3.23E-05 1.85E-14 La-140 1.14E-04 6.54E-14 Ce-144 3.83E-05 2.20E-14 Pr-144 3.83E-05 2.20E-14 Totals 1.29E+03 7.43E-07 Totals excluding H-3 1.75E+00 2.64E-03 (a) One unit Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.2-24 Sheet 1 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED DOSES RESULTING FROM RADIOACTIVE RELEASES IN LIQUID WASTES: DESIGN BASIS CASE (mrem/yr) Age Group = Adult Exposure Pathway Whole Body Skin Bone GI Tract Thyroid Lung Kidney Liver Drinking water 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Consumption of fish 7.22E-3 0.0 2.39E-3 4.30E-4 2.56E-3 7.51E-4 1.66E-3 4.33E-3 Consumption of invertebrates 5.17E-4 0.0 3.65E-4 2.09E-4 2.76E-3 1.43E-4 2.90E-4 6.89E-4 Consumption of aquatic plants 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Exposure to contaminated sediment 4.18E-4 4.87E-4 0.0 0.0 0.0 0.0 0.0 0.0 Swimming in water 4.67E-6 5.39E-5 0.0 0.0 0.0 0.0 0.0 0.0 Total 4.16E-4 5.41E-4 2.76E-3 6.39E-4 5.31E-3 8.94E-4 1.95E-3 5.02E-3 Age Group = Teenager Drinking water 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Consumption of fish 1.89E-3 0.0 2.53E-3 3.27E-4 2.34E-3 7.63E-4 1.62E-3 4.38E-3 Consumption of invertebrates 3.12E-4 0.0 3.85E-4 1.61E-4 2.56E-3 1.38E-4 2.77E-4 6.87E-4 Consumption of aquatic plants 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Exposure to contaminated sediment 2.33E-3 2.72E-3 0.0 0.0 0.0 0.0 0.0 0.0 Swimming in water 4.67E-6 5.39E-5 0.0 0.0 0.0 0.0 0.0 0.0 Total 4.54E-3 2.77E-3 2.91E-3 4.88E-4 4.89E-3 9.01E-4 1.90E-3 5.06E-3 Age Group = Child Drinking water 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Consumption of fish 8.42E-4 0.0 3.14E-3 2.30E-4 2.37E-3 6.11E-4 1.37E-3 3.83E-3 Consumption of invertebrates 1.58E-4 0.0 4.96E-4 9.23E-5 2.72E-3 1.15E-4 2.43E-4 6.22E-4 Consumption of aquatic plants 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Exposure to contaminated sediment 4.87E-4 5.68E-4 0.0 0.0 0.0 0.0 0.0 0.0 Swimming in water 2.60E-6 3.01E-5 0.0 0.0 0.0 0.0 0.0 0.0 Total 1.49E-3 5.98E-4 3.64E-3 3.22E-4 5.09E-3 7.27E-4 1.61E-3 4.45E-3 Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.2-24 Sheet 2 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Age Group = Infant Exposure Pathway Whole Body Skin Bone GI Tract Thyroid Lung Kidney Liver Drinking water 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Consumption of fish 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Consumption of invertebrates 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Consumption to aquatic plants 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Exposure to contaminated sediment 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Swimming in water 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Total 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.2-25 Sheet 1 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED DOSES RESULTING FROM RADIOACTIVE RELEASES IN LIQUID WASTES: NORMAL OPERATION CASE (mrem/yr) Age Group = Adult Exposure Pathway Whole Body Skin Bone GI Tract Thyroid Lung Kidney Liver Drinking water 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Consumption of fish 3.49E-3 0.0 2.61E-3 8.36E-4 4.77E-2 7.74E-4 1.94E-3 4.77E-3 Consumption of invertebrates 6.74E-4 0.0 5.48E-4 7.66E-4 5.65E-2 1.72E-4 5.79E-4 1.11E-3 Consumption of aquatic plants 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Exposure to contaminated sediment 4.36E-4 5.09E-4 0.0 0.0 0.0 0.0 0.0 0.0 Swimming in water 1.03E-5 5.88E-5 0.0 0.0 0.0 0.0 0.0 0.0 Total 4.56E-3 5.68E-4 3.16E-3 1.60E-3 1.04E-1 9.46E-4 2.52E-3 5.88E-3 Age Group = Teenager Drinking water 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Consumption of fish 2.06E-3 0.0 2.76E-3 5.83E-4 4.45E-2 7.96E-4 1.92E-3 4.75E-3 Consumption of invertebrates 4.61E-4 0.0 5.78E-4 5.05E-4 5.26E-2 1.73E-4 5.81E-4 9.82E-4 Consumption of aquatic plants 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Exposure to contaminated sediment 2.44E-3 2.84E-3 0.0 0.0 0.0 0.0 0.0 0.0 Swimming in water 1.03E-5 5.88E-5 0.0 0.0 0.0 0.0 0.0 0.0 Total 4.97E-3 2.90E-3 3.34E-3 1.09E-3 9.71E-2 9.69E-4 2.50E-3 5.73E-3 Age Group = Child Drinking water 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Consumption of fish 9.64E-4 0.0 3.43E-3 3.19E-4 4.59E-2 6.37E-4 1.636-3 4.16E-3 Consumption of invertebrates 3.07E-4 0.0 7.49E-4 2.23E-4 5.63E-2 1.45E-4 5.17E-4 8.95E-4 Consumption of aquatic plants 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Exposure to contaminated sediment 5.09E-4 5.94E-4 0.0 0.0 0.0 0.0 0.0 0.0 Swimming in water 5.74E-6 3.28E-5 0.0 0.0 0.0 0.0 0.0 0.0 Total 1.79E-3 6.27E-4 4.18E-3 5.42E-4 1.02E-1 7.82E-4 2.15E-3 5.05E-3 Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.2-25 Sheet 2 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Age Group = Infant Exposure Pathway Whole Body Skin Bone GI Tract Thyroid Lung Kidney Liver Drinking water 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Consumption of fish 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Consumption of invertebrates 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Consumption of aquatic plants 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Exposure to contaminated sediment 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Swimming in water 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Total 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.2-26 Sheet 1 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED DOSES RESULTING FROM RADIOACTIVE RELEASES IN LIQUID WASTES: NORMAL OPERATIONAL WITH ANTICIPATED OPERATIONAL OCCURRENCES (mrem/yr) Age Group = Adult Exposure Pathway Whole Body Skin Bone GI Tract Thyroid Lung Kidney Liver Drinking water 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Consumption of fish 3.74E-3 0.0 2.86E-3 8.86E-4 5.20E-2 8.20E-4 2.10E-3 5.20E-3 Consumption of invertebrates 7.32E-4 0.0 6.00E-4 8.30E-4 6.16E-2 1.82E-4 6.26E-4 1.21E-3 Consumption of aquatic plants 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Exposure to contaminated sediment 4.78E-4 5.58E-4 0.0 0.0 0.0 0.0 0.0 0.0 Swimming in water 1.13E-5 5.99E-5 0.0 0.0 0.0 0.0 0.0 0.0 Total 4.96E-3 6.18E-4 3.46E-3 1.72E-3 1.14E-1 1.00E-1 2.73E-3 6.41E-3 Age Group = Teenager Drinking water 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Consumption of fish 2.23E-3 0.0 3.02E-3 6.17E-4 4.85E-2 8.51E-4 2.08E-3 5.18E-3 Consumption of invertebrates 4.99E-4 0.0 6.32E-4 5.47E-4 5.74E-2 1.85E-4 6.31E-4 1.07E-3 Consumption of aquatic plants 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Exposure to contaminated sediment 2.67E-3 3.11E-3 0.0 0.0 0.0 0.0 0.0 0.0 Swimming in water 1.13E-5 5.99E-5 0.0 0.0 0.0 0.0 0.0 0.0 Total 5.41E-3 3.17E-3 3.66E-3 1.16E-3 1.06E-1 1.04E-3 2.71E-3 6.25E-3 Age Group = Child Drinking water 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Consumption of fish 1.04E-3 0.0 3.76E-3 3.32E-4 5.01E-2 6.80E-4 1.77E-3 4.54E-3 Consumption of invertebrates 3.31E-4 0.0 8.19E-4 2.40E-4 6.15E-2 1.54E-4 5.61E-4 9.75E-4 Consumption of aquatic plants 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Exposure to containment sediment 5.58E-4 6.51E-4 0.0 0.0 0.0 0.0 0.0 0.0 Swimming in water 6.28E-6 3.34E-5 0.0 0.0 0.0 0.0 0.0 0.0 Total 1.95E-3 6.84E-4 4.58E-3 5.71E-4 1.12E-1 8.34E-4 2.33E-4 5.51E-3 Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.2-26 Sheet 2 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Age Group = Infant Exposure Pathway Whole Body Skin Bone GI Tract Thyroid Lung Kidney Liver Drinking water 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Consumption of fish 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Consumption of invertebrates 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Consumption of aquatic plants 0.0 -- 0.0 0.0 0.0 0.0 0.0 0.0 Exposure to contaminated sediment 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Swimming in water 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Total 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-1 Sheet 1 of 2 EQUIPMENT DESIGN AND OPERATING PARAMETERS FOR GASEOUS RADWASTE SYSTEM, UNITS 1 AND 2
- 1. Waste Gas Compressor Number used: 3; 1 each unit, 1 shared Type: horizontal, centrifugal compressor Temperature: 70F - 130F Inlet pressure: 0.5 psig - 2.0 psig Capacity: 40 cfm at inlet pressure 2.0 psig and discharge pressure 110 psig Cooling water rate: 42.5 gpm Driver: 25 hp
- 2. Surge Tank Number used: 2; 1 each unit Type: horizontal Size: 18' x 1' Volume: 14 ft3 Design pressure: 405 psig/Design temperature: 650F Operating maximum pressure: 10 psig Material: ASTM A106 Carbon Steel (PG&E pipe specification K2)
- 3. Gas Decay Tank Number used: 6; 3 each unit Type: vertical Size: 13' x 8' Volume: 760 ft3 Design temperature: 150F Revision 22 May 2015
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-1 Sheet 2 of 2
- 3. Gas Decay Tank (continued)
Design pressure: 150 psig Operating maximum pressure: 105 psig Material: SA285C, carbon steel Design as per ASME Boiler and Pressure Vessel Code, Section III, Class C
- 4. Waste Gas Analyzer Number used: 2; 1 each unit Oxygen analyzer:
Range: 0-5% (+/-2% of full range) Alarm: 2% oxygen Hydrogen analyzer: Range: 0-5% (+/-2% of full range) 0-50% (+/-2% of full range) 0-100% (+/-2% of full range) Alarm: 3.5% hydrogen Sample channel: 16
- 5. Discharge filter Number used: 2; 1 each unit Type: HEPA filter Size: 8" x 8" x 5-7/8" Efficiency: 99.97% on 0.3 micron particles Capacity: 55 cfm at 1" water gauge differential Revision 22 May 2015
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED GASEOUS WASTE SYSTEM RELEASE: DESIGN BASIS CASE (CURIES) Gas Decay Steam Condenser Auxiliary Spent Secondary Tank Containment System Offgas Building Fuel Pool System Water Total Nuclide Venting Venting Leakage Venting Venting Release Leakage Release Kr- 83M 0.0 0.0 0.0 0.0 0.0 0.197E-19 --- 0.197E-19 Kr- 85M 0.0 0.0 0.0 0.0 0.0 0.224E-09 --- 0.224E-09 Kr- 85 0.505E 04 0.0 0.0 0.0 0.0 0.538E-01 --- 0.505E 04 Kr- 87 0.0 0.0 0.0 0.0 0.0 0.169E-26 --- 0.169E-26 Kr- 88 0.0 0.0 0.0 0.0 0.0 0.360E-13 --- 0.360E-13 Xe-133M 0.210E-02 0.0 0.0 0.0 0.0 0.127E-02 --- 0.337E-02 Xe-133 0.411E 03 0.0 0.0 0.0 0.0 0.189E 00 --- 0.411E 03 Xe-135M 0.482E-46 0.0 0.0 0.0 0.0 0.668E-07 --- 0.668E-07 Xe-135 0.262E-31 0.0 0.0 0.0 0.0 0.276E-06 --- 0.276E-04 Xe-138 0.0 0.0 0.0 0.0 0.0 0.0 --- 0.0 I -131 --- 0.0 0.0 0.0 0.0 0.112E-04 0.0 0.112E-04 I -132 --- 0.0 0.0 0.0 0.0 0.167E-05 0.0 0.167E-05 I -133 --- 0.0 0.0 0.0 0.0 0.194E-06 0.0 0.194E-06 I -134 --- 0.0 0.0 0.0 0.0 0.310E-08 0.0 0.310E-08 I -135 --- 0.0 0.0 0.0 0.0 0.177E-07 0.0 0.177E-07 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-3 Sheet 1 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED GASEOUS WASTE SYSTEM RELEASE: NORMAL OPERATION CASE (CURIES) Gas Decay Condenser Auxiliary Spent Steam Secondary Tank Containment Offgas Building Fuel Pool System System Water Total Nuclide Venting Venting Venting Venting Release Leakage Leakage Release Kr - 83 0.0 0.287E 00 0.169E 00 0.110E-01 0.236E-20 0.615E-04 --- 0.156E 01 Kr - 85M 0.0 0.350E 01 0.145E-01 0.567E 01 0.269E-10 0.317E-03 --- 0.106E 02 Kr - 85 0.439E 03 0.443E 03 0.624E 01 0.100E 02 0.654E-02 0.558E-03 --- 0.898E 03 Kr - 87 0.0 0.554E 00 0.361E 00 0.311E 01 0.203E-27 0.174E-03 --- 0.402E 01 Kr - 88 0.0 0.378E 01 0.193E 01 0.972E 01 0.432E-14 0.543E-03 --- 0.154E 02 Xe - 133M 0.244E-03 0.721E 02 0.433E 01 0.931E 01 0.152E-03 0.521E-03 --- 0.858E 02 Xe - 133 0.434E 02 0.118E 05 0.367E 03 0.760E 03 0.227E-01 0.424E-01 --- 0.129E 05 Xe - 135M 0.577E-47 0.920E 00 0.254E 00 0.122E 01 0.001E-08 0.486E-03 --- 0.239E 01 Xe - 135 0.313E-32 0.280E 02 0.626E-01 0.170E 02 0.331E-05 0.102E-02 --- 0.513E 02 Xe - 138 0.0 0.530E-01 0.432E-01 0.165E 01 0.0 0.919E-04 --- 0.174E 01 I - 131 --- 0.336E-01 0.312E-01 0.637E-01 0.859E-05 0.630E-03 0.897E-06 0.129E 00 I - 132 --- 0.188E 00 0.418E-02 0.230E-01 0.752E-06 0.134E-03 0.191E-06 0.215E 00 I - 133 --- 0.685E-02 0.363E-01 0.872E-01 0.165E-06 0.751E-03 0.106E-05 0.131E 00 I - 134 --- 0.378E-04 0.999E-03 0.117E-01 0.310E-08 0.234E-04 0.266E-07 0.127E-01 I - 135 --- 0.121E-02 0.148E-01 0.482E-01 0.177E-07 0.318E-03 0.441E-06 0.645E-01 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-3 Sheet 2 of 2 Total Nuclide Release H-3 0.33E 03 Cr - 51 0.0 Mn - 54 0.44E-01 Fe - 55 0.0 Co - 58 0.15E-00 Fe - 59 0.15E-01 Co - 60 0.68E-01 Sr - 89 0.33E-02 Sr - 90 0.56E-03 Y - 90 0.0 Sr - 91 0.0 Y - 91 0.0 Sr - 92 0.0 Y - 92 0.0 Zr - 95 0.0 Nb - 95 0.0 Mo - 99 0.0 Te - 132 0.0 Cs - 134 0.44E-01 Cs - 136 0.0 Cs - 137 0.75E-01 Ba - 140 0.0 La - 140 0.0 Ce - 144 0.0 Pr - 144 0.0 C - 14 0.80E 01 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-4 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ANNUAL GASEOUS RADWASTE FLOWS (Standard Cubic Feet Per Year) Source Flow Liquid holdup tanks: Displaced 60,300 Recycled 15,300 45,000 Volume control tank 19,000 Boric acid gas stripper(b) 7,400 Reactor coolant drain tank 250 Pressurizer relief tank 350 Nitrogen added to gas decay tanks 28,000 Total annual discharge 100,000(a) (a) Assumes: 1. Two cold shutdowns per year
- 2. Base loaded plant operation
- 3. Hydrogen controlled to less than 4%
(b) Equipment is abandoned in place and no longer in use. Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-5 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED MAXIMUM ACTIVITY IN GAS DECAY TANK: DESIGN BASIS CASE Nuclide Activity, Ci Concentration, Ci/cc Kr-83M 0.894E 02 0.421E 01 Kr-85M 0.462E 03 0.217E 02 Kr-85 0.160E 04 0.754E 02 Kr-87 0.252E 03 0.119E 02 Kr-88 0.790E 03 0.372E 02 Xe-133M 0.798E 03 0.376E 02 Xe-133 0.693E 05 0.326E 04 Xe-135M 0.988E 02 0.465E 01 Xe-135 0.139E 04 0.656E 02 Xe-138 0.133E 03 0.628E 01 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-6 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED MAXIMUM ACTIVITY IN GAS DECAY TANK: NORMAL OPERATION CASE Nuclide Activity, Ci Concentration, Ci/cc Kr-83M 0.107E 02 0.505E 00 Kr-85M 0.553E 02 0.260E 01 Kr-85 0.130E 03 0.612E 01 Kr-87 0.302E 02 0.142E 01 Kr-88 0.946E 02 0.445E 01 Xe-133M 0.928E 02 0.437E 01 Xe-133 0.778E 04 0.366E 03 Xe-135M 0.119E 02 0.558E 00 Xe-135 0.166E 03 0.783E 01 Xe-138 0.160E 02 0.753E 00 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-7 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ACTIVITY IN VOLUME CONTROL TANK: DESIGN BASIS CASE Liquid Phase Vapor Phase Nuclide Activity, Ci Concentration, Ci/cc Activity, Ci Concentration, Ci/cc Total Act, Ci H- 3 0.497E 01 0.732E 00 0.0 0.0 0.497E 01 Cr- 51 0.935E-02 0.138E-02 0.0 0.0 0.935E-02 Mn- 54 0.152E-02 0.224E-03 0.0 0.0 0.152E-02 Fe- 55 0.773E-02 0.114E-02 0.0 0.0 0.773E-02 Co- 58 0.785E-01 0.115E-01 0.0 0.0 0.785E-01 Fe- 59 0.491E-02 0.722E-03 0.0 0.0 0.491E-02 Co- 60 0.976E-02 0.144E-02 0.0 0.0 0.976E-02 Kr- 83M 0.0 0.0 0.495E 00 0.109E 00 0.495E 00 Kr- 85M 0.0 0.0 0.276E 01 0.609E 00 0.276E 01 Kr- 85 0.0 0.0 0.657E 01 0.145E 01 0.657E 01 Kr- 87 0.0 0.0 0.132E 01 0.291E 00 0.132E 01 Kr- 88 0.0 0.0 0.456E 01 0.101E 01 0.456E 01 Sr- 89 0.112E-01 0.165E-02 0.0 0.0 0.112E-01 Sr- 90 0.538E-03 0.792E-04 0.0 0.0 0.538E-03 Y - 90 0.928E-03 0.137E-03 0.0 0.0 0.928E-03 Sr- 91 0.710E-02 0.104E-02 0.0 0.0 0.710E-02 Y - 91 0.955E-01 0.140E-01 0.0 0.0 0.955E-01 Sr- 92 0.271E-02 0.398E-03 0.0 0.0 0.271E-02 Y - 92 0.374E-02 0.550E-03 0.0 0.0 0.374E-02 Zr- 95 0.278E-02 0.410E-03 0.0 0.0 0.278E-02 Nb- 95 0.271E-02 0.398E-03 0.0 0.0 0.271E-02 Mo- 99 0.223E 02 0.328E 01 0.0 0.0 0.223E 02 I -131 0.132E 02 0.194E 01 0.0 0.0 0.132E 02 Te-132 0.139E 01 0.204E 00 0.0 0.0 0.139E 01 I -132 0.441E 01 0.649E 00 0.0 0.0 0.441E 01 I -133 0.178E 02 0.262E 01 0.0 0.0 0.178E 02 Xe-133M 0.213E-02 0.314E-03 0.493E 01 0.109E 01 0.493E 01 Xe-133 0.378E-01 0.556E-02 0.416E 03 0.919E 02 0.416E 03 Cs-134 0.932E 00 0.137E 00 0.0 0.0 0.932E 00 I -134 0.183E 01 0.269E 00 0.0 0.0 0.183E 01 I -135 0.957E 01 0.141E 01 0.0 0.0 0.957E 01 Xe-135M 0.715E 00 0.105E 00 0.304E 00 0.672E-01 0.102E 01 Xe-135 0.260E 00 0.382E-01 0.858E 01 0.189E 01 0.884E 01 Cs-136 0.252E 00 0.371E 01 0.0 0.0 0.252E 00 Cs-137 0.167E 01 0.246E 00 0.0 0.0 0.167E 01 Xe-138 0.0 0.0 0.388E 00 0.857E-01 0.388E 00 Ba-140 0.190E-01 0.279E-02 0.0 0.0 0.190E-01 La-140 0.678E-02 0.998E-03 0.0 0.0 0.678E-02 Ce-144 0.164E-02 0.241E-03 0.0 0.0 0.164E-02 Pr-144 0.164E-02 0.241E-03 0.0 0.0 0.164E-02 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-8 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ACTIVITY IN VOLUME CONTROL TANK: NORMAL OPERATION CASE Liquid Phase Vapor Phase Nuclide Activity, Ci Concentration, Ci/cc Activity, Ci Concentration, Ci/cc Total Act, Ci H- 3 0.470E 01 0.692E 00 0.0 0.0 0.470E 01 CR- 51 0.934E-02 0.137E-02 0.0 0.0 0.934E-02 Mn- 54 0.152E-02 0.224E-03 0.0 0.0 0.152E-02 Fe- 55 0.772E-02 0.114E-02 0.0 0.0 0.772E-02 Co- 58 0.784E-01 0.115E-01 0.0 0.0 0.784E-01 Fe- 59 0.490E-02 0.772E-03 0.0 0.0 0.490E-02 Co- 60 0.975E-02 0.144E-02 0.0 0.0 0.975E-02 Kr- 83M 0.0 0.0 0.594E-01 0.131E-01 0.594E-01 Kr- 85M 0.0 0.0 0.330E 00 0.729E-01 0.330E 00 Kr- 85 0.0 0.0 0.619E 00 0.137E 00 0.619E 00 Kr- 87 0.0 0.0 0.158E 00 0.349E-01 0.158E 00 Kr- 88 0.0 0.0 0.547E 00 0.121E 00 0.547E 00 Sr- 89 0.134E-02 0.198E-03 0.0 0.0 0.134E-02 Sr- 90 0.646E-04 0.950E-05 0.0 0.0 0.646E-04 Y - 90 0.111E-03 0.163E-04 0.0 0.0 0.111E-03 Sr- 91 0.851E-03 0.125E-03 0.0 0.0 0.851E-03 Y - 91 0.111E-01 0.164E-02 0.0 0.0 0.111E-01 Sr- 92 0.325E-03 0.478E-04 0.0 0.0 0.325E-03 Y - 92 0.448E-03 0.660E-04 0.0 0.0 0.448E-03 Zr- 95 0.334E-03 0.491E-04 0.0 0.0 0.334E-03 Nb- 95 0.325E-03 0.477E-04 0.0 0.0 0.325E-03 Mo- 99 0.267E 01 0.392E 00 0.0 0.0 0.267E 01 I -131 0.158E 01 0.233E 00 0.0 0.0 0.158E 01 Te-132 0.166E 00 0.245E-01 0.0 0.0 0.166E 00 I -132 0.529E 00 0.779E-01 0.0 0.0 0.529E 00 I -133 0.214E 01 0.315E 00 0.0 0.0 0.214E 01 Xe-133M 0.256E-03 0.376E-04 0.573E 00 0.127E 00 0.574E 00 Xe-133 0.453E-02 0.667E-03 0.469E 02 0.104E 02 0.469E 02 Cs-134 0.113E 00 0.166E-01 0.0 0.0 0.113E 00 I -134 0.219E 00 0.323E-01 0.0 0.0 0.219E 00 I -135 0.115E 01 0.169E 00 0.0 0.0 0.115E 01 Xe-135M 0.858E-01 0.126E-01 0.365E-01 0.806E-02 0.122E 00 Xe-135 0.311E-01 0.458E-02 0.102E 01 0.226E 00 0.105E 01 Cs-136 0.303E-01 0.445E-02 0.0 0.0 0.303E-01 Cs-137 0.200E 00 0.295E-01 0.0 0.0 0.200E 00 Xe-138 0.00 0.0 0.466E-01 0.103E-01 0.446E-01 Ba-140 0.227E-02 0.335E-03 0.0 0.0 0.227E-02 La-140 0.813E-03 0.120E-03 0.0 0.0 0.813E-03 Ce-144 0.197E-03 0.289E-04 0.0 0.0 0.197E-03 Pr-144 0.197E-03 0.289E-04 0.0 0.0 0.197E-03 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-9 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED GASEOUS RELEASES DUE TO COLD SHUTDOWN AND STARTUPS (Released from Condenser) Fraction of Released, Total Annual Nuclide Ci Release Kr-83m 1.89E-8 1.21E-8 Kr-85m 2.91E-5 2.75E-6 Kr-85 4.32E-3 4.75E-6 Kr-87 6.36E-10 1.58E-10 Kr-88 4.14E-6 2.69E-7 I-131(a) 4.32E-5 3.32E-4 I-132(a) 7.69E-9 2.79E-7 I-133(a) 2.51E-5 1.90E-4 Xe-133m 2.80E-3 3.29E-5 Xe-133 2.80E-1 2.17E-5 I-134(a) 7.34E-15 5.73E-13 I-135(a) 1.98E-6 3.06E-5 Xe-135m 1.89E-5 1.20E-5 Xe-135 1.20E-3 2.63E-5 Xe-138 0.0 0.0 Totals 2.89E-1 6.54E-4 (a) Volatile form only. Notes: Beta air dose at 0.5 miles NW of plant = 1.55E-5 mrem/yr Gamma air dose at 0.5 miles NW of plant = 5.12E-6 mrem/yr Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-10 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED DISTANCES IN MILES FROM DCPP UNIT 1 REACTOR CENTERLINE TO THE NEAREST MILK COW, MEAT ANIMAL, MILK GOAT, RESIDENCE, VEGETABLE GARDEN, AND SITE BOUNDARY(c) 22-1/2° Radial Sectors(a) Nearest NW NNW N NNE NE ENE E ESE SE Milk cow None(b) None None None None None None None None Meat animal 0.5 0.5 0.5 0.5 0.5 0.7 1.0 1.0 1.1 Milk goat None None None None None None None None None Residence None 1.5 None None None 4.5 None None None Vegetable garden 3.6 3.6 None None None None None 3.7 3.7 Site boundary 0.5 0.5 0.5 0.5 0.5 0.7 1.0 1.0 1.1 (a) Sectors not shown contain no land beyond the site boundary, other than islets not used for the purposes indicated in this table. (b) None within 5 miles, typical of other places where "None" is used. (c) This table presents pre-operational data. Current operational data reside in the ODCM. Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-11 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ESTIMATES OF RELATIVE CONCENTRATION (/Q)(a) AT LOCATIONS SPECIFIED IN TABLE 11.3-10(b) 22-1/2° Radial Sectors Nearest NW NNW N NNE NE ENE E ESE SE Milk cow None None None None None None None None None Meat animal 1.58X10-6 8.67X10-7 4.93X10-7 2.44X10-7 1.62X10-7 9.18X10-8 1.07X10-7 5.20X10-7 1.32X10-6 Milk goat None None None None None None None None None Residence None 3.30X10-7 None None None 1.40X10-8 None None None Vegetable 1.50X10-7 1.50X10-7 None None None None None 1.00X10-7 1.00X10-7 garden Site boundary 1.58X10-6 8.67X10-7 4.93X10-7 2.44X10-7 1.62X10-7 9.18X10-8 1.07X10-7 5.20X10-7 1.32X10-6 (a) In units of seconds per cubic meter. (b) This table presents pre-operational data. Current operational data reside in the ODCM. Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-12 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ESTIMATES OF DEPOSITION (D/Q)(a) AT LOCATIONS SPECIFIED IN TABLE 11.3-10(b) 22-1/2° Radial Sectors Nearest NW NNW N NNE NE ENE E ESE SE Milk cow None None None None None None None None None Meat animal 2.54X10-8 1.33X10-8 5.50X10-9 3.27X10-9 4.13X10-9 1.77X10-9 1.65X10-9 8.47X10-9 2.90X10-8 Milk goat None None None None None None None None None Residence None 2.08X10-9 None None None 6.49X10-11 None None None Vegetable 8.13X10-10 4.27X10-10 None None None None None 7.91X10-10 3.20X10-9 garden Site boundary 2.54X10-8 1.33X10-8 5.50X10-9 3.27X10-9 4.13X10-9 1.77X10-9 1.65X10-9 8.47X10-9 2.90X10-8 (a) In units of meters-2, includes sector width and frequency of winds in each sector. (b) This table presents pre-operational data. Current operational data reside in the ODCM. Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-13 Sheet 1 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ANNUAL AVERAGE ATMOSPHERIC ACTIVITY CONCENTRATIONS AT SITE BOUNDARY FOR DESIGN BASIS CASE (Ci/cc) Sector Nuclide NW NNW N NNE NE ENE E ESE SE I-135 7.633E-22 4.189E-22 2.382E-22 1.179E-22 7.826E-23 4.277E-23 4.806E-23 2.336E-22 5.864E-22 H-3 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Cr-51 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Mn-54 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Fe-55 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Co-58 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Fe-59 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Co-60 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Kr-83M 9.868E-34 5.415E-34 3.079E-34 1.524E-34 1.012E-34 5.734E-35 6.683E-35 3.248E-34 8.245E-34 Kr-85M 1.121E-23 6.153E-24 3.499E-24 1.732E-24 1.150E-24 6.515E-25 7.594E-25 3.691E-24 9.369E-24 Kr-85 2.528E-10 1.387E-10 7.887E-11 3.903E-11 2.592E-11 1.469E-11 1.712E-11 8.318E-11 2.112E-10 Kr-87 8.455E-41 4.639E-41 2.638E-41 1.306E-41 8.669E-42 4.912E-42 5.726E-42 2.783E-41 7.063E-41 Kr-88 1.805E-27 9.905E-28 5.632E-28 2.788E-28 1.851E-28 1.049E-28 1.222E-28 5.941E-28 1.508E-27 Sr-89 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Sr-90 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Y-90 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Sr-91 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Y-91 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Sr-92 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Y-92 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Zr-95 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Nb-95 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Mo-99 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-131 4.819E-19 2.644E-19 1.504E-19 7.442E-20 4.941E-20 2.700E-20 3.034E-20 1.474E-19 3.702E-19 Te-132 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-132 7.214E-20 3.959E-20 2.251E-20 1.114E-20 7.397E-21 4.042E-21 4.542E-21 2.207E-20 5.542E-20 I-133 8.355E-21 4.585E-21 2.607E-21 1.290E-21 8.567E-22 4.681E-22 5.260E-22 2.556E-21 6.419E-21 Xe-133M 1.688E-16 9.265E-17 5.268E-17 2.607E-17 1.731E-17 9.810E-10 1.143E-17 5.557E-17 1.411E-16 Xe-133 2.058E-11 1.129E-11 6.420E-12 3.178E-12 2.110E-12 1.196E-12 1.393E-12 6.772E-12 1.719E-11 Cs-134 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-134 1.336E-22 7.332E-23 4.169E-23 2.064E-23 1.370E-23 7.486E-24 8.413E-24 4.088E-23 1.027E-22 Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-13 Sheet 2 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Sector Nuclide NW NNW N NNE NE ENE E ESE SE Xe-135M 3.345E-21 1.835E-21 1.044E-21 5.165E-22 3.429E-22 1.943E-22 2.265E-22 1.101E-21 2.794E-21 Xe-135 1.382E-18 7.582E-19 4.311E-19 2.134E-19 1.417E-19 8.028E-20 9.357E-20 4.547E-19 1.154E-18 Cs-136 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Cs-137 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Xe-138 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Ba-140 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 La-140 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Ce-144 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Pr-144 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 C-14 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-14 Sheet 1 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ANNUAL AVERAGE ATMOSPHERIC ACTIVITY CONCENTRATIONS AT SITE BOUNDARY FOR NORMAL OPERATION CASE (Ci/cc) Sector Nuclide NW NNW N NNE NE ENE E ESE SE H-3 1.401E-11 7.687E-12 4.371E-12 2.163E-12 1.436E-12 7.848E-13 8.820E-13 4.286E-12 1.076E-11 Cr-51 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Mn-54 1.918-15 1.053E-15 5.985E-14 2.967E-16 1.967E-16 1.075E-16 1.208E-16 5.869E-16 1.474E-15 Fe-55 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Co-58 4.466E-15 3.548E-15 2.017E-15 9.985E-16 6.629E-16 3.622E-16 4.071E-16 1.978E-15 4.967E-15 Fe-59 6.466E-16 3.548E-16 2.017E-16 9.985E-17 6.629E-17 3.622E-17 4.071E-17 1.978E-16 4.967E-16 Co-60 2.931E-15 1.608E-5 9.146E-16 4.526E-16 3.005E-16 1.642E-16 1.845E-16 8.968E-16 2.252E-15 Kr-83M 7.805E-14 4.283E-14 2.435E-14 1.205E-14 8.002E-15 4.535E-15 5.286E-15 2.569E-14 6.521E-14 Kr-85M 5.323E-13 2.921E-13 1.661E-13 8.221E-14 5.458E-14 3.093E-14 3.605E-14 1.752E-13 4.447E-13 Kr-85 4.558E-11 2.501E-11 1.422E-11 7.039E-12 4.674E-12 2.648E-12 3.087E-12 1.500E-11 3.808E-11 Kr-87 2.016E-13 1.104E-13 6.290E-14 3.113E-14 2.067E-14 1.171E-14 1.365E-14 6.635E-14 1.684E-13 Kr-88 7.725E-13 4.239E-13 2.410E-13 1.193E-13 7.921E-14 4.488E-14 5.232E-14 2.542E-13 6.454E-13 Sr-89 1.422E-16 7.805E-17 4.438E-17 2.197E-17 1.458E-17 7.969E-18 8.955E-10 4.352E-17 1.093E-16 Sr-90 2.414E-17 1.325E-17 7.532E-18 3.720E-18 2.475E-18 1.352E-18 1.520E-18 7.386E-18 1.854E-17 Y-90 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Sr-91 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Y-91 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Sr-92 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Y-92 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Zr-95 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Nb-95 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Mo-99 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-131 5.601E-15 3.073E-15 1.748E-15 8.649E-16 5.742E-16 3.138E-16 3.526E-16 1.714E-15 4.303E-15 Te-132 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-132 1.190E-15 6.529E-16 3.712E-16 1.837E-16 1.220E-16 6.666E-17 7.491E-17 3.640E-14 9.141E-16 I-133 5.673E-15 3.113E-15 1.770E-15 8.761E-16 5.816E-16 3.178E-16 3.572E-17 1.736E-15 4.358E-15 Xe-133M 4.259E-12 2.337E-12 1.329E-12 6.577E-13 4.367E-13 2.475E-13 2.884E-13 1.402E-12 3.558E-12 Xe-133 6.468E-10 3.549E-10 2.018E-10 9.989E-11 6.632E-11 3.758E-11 4.381E-11 2.129E-10 5.404E-10 Cs-134 1.918E-15 1.053E-15 5.985E-16 2.962E-16 1.967E-16 1.075E-16 1.208E-16 5.869E-16 1.474E-15 I-134 5.496E-16 3.016E-16 1.715E-16 8.487E-17 5.635E-17 3.079E-17 3.460E-17 1.682E-16 4.222E-16 I-135 2.787E-15 1.529E-15 8.697E-16 4.304E-16 2.858E-16 1.562E-16 1.755E-16 8.528E-16 2.141E-15 Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-14 Sheet 2 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Sector Nuclide NW NNW N NNE NE ENE E ESE SE Xe-135M 7.857E-14 4.311E-14 2.452E-14 1.213E-14 8.056E-15 4.565E-15 5.321E-15 2.586E-14 6.564E-14 Xe-135 2.287E-12 1.255E-12 7.135E-13 3.532E-13 2.345E-13 1.329E-13 1.549E-13 7.526E-13 1.911E-12 Cs-136 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Cs-137 3.254E-15 1.786E-15 1.015E-15 5.026E-16 3.337E-16 1.823E-16 2.049E-16 9.957E-16 2.500E-15 Xe-138 8.728E-14 4.789E-14 2.723E-14 1.348E-14 8.948E-15 5.071E-15 5.910E-15 2.872E-14 7.291E-14 Ba-140 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 La-140 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Ce-144 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Pr-144 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 C-14 3.448E-13 1.892E-13 1.076E-13 5.325E-14 3.536E-14 1.932E-14 2.171E-14 1.055E-13 2.649E-13 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-15 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR NW SECTOR AT DISTANCE 0.5 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) DESIGN BASIS CASE Dose Adult Teen Child Infant Ingestion bone 4.406E-08 3.486E-08 6.434E-08 0.0 Ingestion liver 6.312E-08 4.925E-08 6.592E-08 0.0 Ingestion whole body 3.611E-08 2.935E-08 4.974E-08 0.0 Ingestion thyroid 2.065E-05 1.421E-05 2.143E-05 0.0 Ingestion kidney 1.080E-07 6.383E-08 4.026E-08 0.0 Ingestion lung 0.0 0.0 0.0 0.0 Ingestion GI 1.663E-08 9.324E-09 5.645E-09 0.0 Total bone 4.406E-08 3.486E-08 6.434E-08 0.0 Total liver 6.312E-08 4.925E-08 6.592E-08 0.0 Total whole body 3.611E-08 2.935E-08 4.974E-08 0.0 Total thyroid 2.065E-05 1.421E-05 2.143E-05 0.0 Total kidney 1.080E-07 6.383E-08 4.026E-08 0.0 Total lung 0.0 0.0 0.0 0.0 Total GI 1.663E-08 9.324E-09 5.645E-09 0.0 Gamma air 5.145E-01 5.145E-01 5.145E-01 5.145E-01 Beta air 1.161E-02 1.161E-02 1.161E-02 1.161E-02 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-16 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR NW SECTOR AT DISTANCE 3.6 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) DESIGN BASIS CASE Dose Adult Teen Child Infant Ingestion bone 5.578E-10 9.049E-10 2.186E-09 0.0 Ingestion liver 7.992E-10 1.279E-09 2.289E-09 0.0 Ingestion whole body 4.573E-10 7.619E-10 1.690E-09 0.0 Ingestion thyroid 2.615E-07 3.688E-07 7.281E-07 0.0 Ingestion kidney 1.368E-09 1.657E-09 1.368E-09 0.0 Ingestion lung 0.0 0.0 0.0 0.0 Ingestion GI 2.105E-10 2.421E-10 1.918E-10 0.0 Total bone 5.578E-10 9.049E-10 2.186E-09 0.0 Total liver 7.992E-10 1.279E-09 2.239E-09 0.0 Total whole body 4.573E-10 7.619E-10 1.690E-09 0.0 Total thyroid 2.615E-07 3.688E-07 7.281E-07 0.0 Total kidney 1.368E-09 1.657E-09 1.368E-09 0.0 Total lung 0.0 0.0 0.0 0.0 Total GI 2.105E-10 2.421E-10 1.918E-10 0.0 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-17 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR NNW SECTOR AT DISTANCE 0.5 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) DESIGN BASIS CASE Dose Adult Teen Child Infant Ingestion bone 2.312E-08 1.829E-08 3.377E-08 0.0 Ingestion liver 3.313E-08 2.585E-08 3.460E-08 0.0 Ingestion whole body 1.895E-08 1.540E-08 2.610E-08 0.0 Ingestion thyroid 1.084E-05 7.455E-06 1.125E-05 0.0 Ingestion kidney 5.669E-08 3.350E-08 2.113E-08 0.0 Ingestion lung 0.0 0.0 0.0 0.0 Ingestion GI 8.726E-09 4.894E-09 2.962E-09 0.0 Total bone 2.312E-08 1.829E-08 3.377E-08 0.0 Total liver 3.313E-08 2.585E-08 3.460E-08 0.0 Total whole body 1.895E-08 1.540E-08 2.610E-08 0.0 Total thyroid 1.084E-05 7.455E-06 1.125E-05 0.0 Total kidney 5.669E-08 3.350E-08 2.113E-08 0.0 Total lung 0.0 0.0 0.0 0.0 Total GI 8.726E-09 4.894E-09 2.962E-09 0.0 Gamma air 2.823E-01 2.823E-01 2.823E-01 2.823E-01 Beta air 6.371E-03 6.371E-03 6.371E-03 6.371E-03 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-18 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR NNW SECTOR AT DISTANCE 1.5 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) DESIGN BASIS CASE Dose Adult Teen Child Infant Inhalation bone 2.077E-09 1.937E-09 2.985E-09 4.420E-09 Inhalation liver 2.971E-09 2.732E-09 3.047E-09 5.218E-09 Inhalation whole body 1.687E-09 1.617E-09 2.294E-09 3.049E-09 Inhalation thyroid 9.794E-07 7.988E-07 1.009E-06 1.722E-06 Inhalation kidney 5.094E-09 3.559E-09 1.884E-09 1.326E-09 Inhalation lung 0.0 0.0 0.0 0.0 Inhalation GI 5.289E-10 3.526E-10 1.810E-10 1.360E-10 External whole body 1.692E-03 1.692E-03 1.692E-03 1.692E-03 External skin 7.394E-02 7.394E-02 7.394E-02 7.394E-02 Total bone 1.692E-03 1.692E-03 1.692E-03 1.692E-03 Total liver 1.692E-03 1.692E-03 1.692E-03 1.692E-03 Total whole body 1.692E-03 1.692E-03 1.692E-03 1.692E-03 Total thyroid 1.693E-03 1.693E-03 1.693E-03 1.694E-03 Total kidney 1.692E-03 1.692E-03 1.692E-03 1.692E-03 Total lung 1.692E-03 1.692E-03 1.692E-03 1.692E-03 Total GI 1.692E-03 1.692E-03 1.692E-03 1.692E-03 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-19 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR NNW SECTOR AT DISTANCE 3.6 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) DESIGN BASIS CASE Dose Adult Teen Child Infant Ingestion bone 2.926E-10 4.746E-10 1.146E-09 0.0 Ingestion liver 4.192E-10 6.706E-10 1.175E-09 0.0 Ingestion whole body 2.398E-10 3.996E-10 8.862E-10 0.0 Ingestion thyroid 1.371E-07 1.934E-07 3.819E-07 0.0 Ingestion kidney 7.174E-10 8.692E-10 7.174E-10 0.0 Ingestion lung 0.0 0.0 0.0 0.0 Ingestion GI 1.104E-10 1.270E-10 1.006E-10 0.0 Total bone 2.926E-10 4.746E-10 1.146E-09 0.0 Total liver 4.192E-10 6.706E-10 1.175E-09 0.0 Total whole body 2.398E-10 3.996E-10 8.862E-10 0.0 Total thyroid 1.371E-07 1.934E-07 3.819E-07 0.0 Total kidney 7.174E-10 8.692E-10 7.174E-10 0.0 Total lung 0.0 0.0 0.0 0.0 Total GI 1.104E-10 1.270E-10 1.006E-10 0.0 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-20 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR N SECTOR AT DISTANCE 0.5 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) DESIGN BASIS CASE Dose Adult Teen Child Infant Ingestion bone 9.551E-09 7.557E-09 1.395E-08 0.0 Ingestion liver 1.368E-08 1.068E-08 1.429E-08 0.0 Ingestion whole body 7.829E-09 6.363E-09 1.078E-08 0.0 Ingestion thyroid 4.477E-06 3.080E-06 4.647E-06 0.0 Ingestion kidney 2.342E-08 1.384E-08 8.729E-09 0.0 Ingestion lung 0.0 0.0 0.0 0.0 Ingestion GI 3.605E-09 2.022E-09 1.224E-09 0.0 Total bone 9.551E-09 7.557E-09 1.395E-08 0.0 Total liver 1.368E-08 1.068E-08 1.429E-08 0.0 Total whole body 7.829E-09 6.363E-09 1.078E-08 0.0 Total thyroid 4.477E-06 3.080E-06 4.647E-06 0.0 Total kidney 2.342E-08 1.384E-08 8.729E-09 0.0 Total lung 0.0 0.0 0.0 0.0 Total GI 3.605E-09 2.022E-09 1.224E-09 0.0 Gamma air 1.605E-01 1.605E-01 1.605E-01 1.605E-01 Beta air 3.623E-03 3.623E-03 3.623E-03 3.623E-03 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-21 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR NNE SECTOR AT DISTANCE 0.5 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) DESIGN BASIS CASE Dose Adult Teen Child Infant Ingestion bone 5.688E-09 4.500E-09 8.306E-09 0.0 Ingestion liver 8.148E-09 6.358E-09 8.510E-09 0.0 Ingestion whole body 4.662E-09 3.789E-09 6.421E-09 0.0 Ingestion thyroid 2.666E-06 1.834E-06 2.767E-06 0.0 Ingestion kidney 1.395E-08 8.240E-09 5.198E-09 0.0 Ingestion lung 0.0 0.0 0.0 0.0 Ingestion GI 2.146E-09 1.204E-09 7.287E-10 0.0 Total bone 5.688E-09 4.500E-09 8.306E-09 0.0 Total liver 8.148E-09 6.358E-09 8.510E-09 0.0 Total whole body 4.662E-09 3.789E-09 6.421E-09 0.0 Total thyroid 2.666E-06 1.834E-06 2.767E-06 0.0 Total kidney 1.395E-08 8.240E-09 5.198E-09 0.0 Total lung 0.0 0.0 0.0 0.0 Total GI 2.146E-09 1.204E-09 7.287E-10 0.0 Gamma air 7.945E-02 7.945E-02 7.945E-02 7.945E-02 Beta air 1.793E-03 1.793E-03 1.793E-03 1.793E-03 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-22 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR NE SECTOR AT DISTANCE 0.5 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) DESIGN BASIS CASE Dose Adult Teen Child Infant Ingestion bone 7.178E-09 5.679E-09 1.048E-08 0.0 Ingestion liver 1.028E-08 8.024E-09 1.074E-08 0.0 Ingestion whole body 5.884E-09 4.782E-09 8.104E-09 0.0 Ingestion thyroid 3.365E-06 2.315E-06 3.492E-06 0.0 Ingestion kidney 1.760E-08 1.040E-08 6.560E-09 0.0 Ingestion lung 0.0 0.0 0.0 0.0 Ingestion GI 2.709E-09 1.519E-09 9.197E-10 0.0 Total bone 7.178E-09 5.679E-09 1.048E-08 0.0 Total liver 1.028E-08 8.024E-09 1.074E-08 0.0 Total whole body 5.884E-09 4.782E-09 8.104E-09 0.0 Total thyroid 3.365E-06 2.315E-06 3.492E-06 0.0 Total kidney 1.760E-08 1.040E-08 6.560E-09 0.0 Total lung 0.0 0.0 0.0 0.0 Total GI 2.709E-09 1.519E-09 9.197E-10 0.0 Gamma air 5.275E-02 5.275E-02 5.275E-02 5.275E-02 Beta air 1.190E-03 1.190E-03 1.190E-03 1.190E-03 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-23 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR ENE SECTOR AT DISTANCE 0.7 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) DESIGN BASIS CASE Dose Adult Teen Child Infant Ingestion bone 3.068E-09 2.427E-09 4.481E-09 0.0 Ingestion liver 4.395E-09 3.430E-09 4.591E-09 0.0 Ingestion whole body 2.515E-09 2.044E-09 3.464E-09 0.0 Ingestion thyroid 1.438E-06 9.892E-07 1.493E-06 0.0 Ingestion kidney 7.522E-09 4.445E-09 2.804E-09 0.0 Ingestion lung 0.0 0.0 0.0 0.0 Ingestion GI 1.158E-09 6.493E-10 3.931E-10 0.0 Total bone 3.068E-09 2.427E-09 4.481E-09 0.0 Total liver 4.395E-09 3.430E-09 4.591E-09 0.0 Total whole body 2.515E-09 2.044E-09 3.464E-09 0.0 Total thyroid 1.438E-06 9.892E-07 1.493E-06 0.0 Total kidney 7.522E-09 4.445E-09 2.804E-09 0.0 Total lung 0.0 0.0 0.0 0.0 Total GI 1.158E-09 6.493E-10 3.931E-10 0.0 Gamma air 2.989E-02 2.989E-02 2.989E-02 2.989E-02 Beta air 6.746E-04 6.746E-04 6.746E-04 6.746E-04 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-24 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR ENE SECTOR AT DISTANCE 4.5 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) DESIGN BASIS CASE Dose Adult Teen Child Infant Inhalation bone 7.466E-11 6.962E-11 1.073E-10 1.589E-10 Inhalation liver 1.068E-10 9.819E-11 1.095E-10 1.876E-10 Inhalation whole body 6.065E-11 5.813E-11 8.247E-11 1.096E-10 Inhalation thyroid 3.521E-08 2.872E-08 3.628E-08 6.191E-08 Inhalation kidney 1.831E-10 1.279E-10 6.772E-11 4.766E-11 Inhalation lung 0.0 0.0 0.0 0.0 Inhalation GI 1.901E-11 1.267E-11 6.505E-12 4.890E-12 External whole body 7.179E-05 7.179E-05 7.179E-05 7.179E-05 External skin 3.137E-03 3.137E-03 3.137E-03 3.137E-03 Total bone 7.179E-05 7.179E-05 7.179E-05 7.179E-05 Total liver 7.179E-05 7.179E-05 7.179E-05 7.179E-05 Total whole body 7.179E-05 7.179E-05 7.179E-05 7.179E-05 Total thyroid 7.183E-05 7.182E-05 7.183E-05 7.185E-05 Total kidney 7.179E-05 7.179E-05 7.179E-05 7.179E-05 Total lung 7.179E-05 7.179E-05 7.179E-05 7.179E-05 Total GI 7.179E-05 7.179E-05 7.179E-05 7.179E-05 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-25 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR E SECTOR AT DISTANCE 1.0 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) DESIGN BASIS CASE Dose Adult Teen Child Infant Ingestion bone 2.867E-09 2.268E-09 4.186E-09 0.0 Ingestion liver 4.107E-09 3.204E-09 4.289E-09 0.0 Ingestion whole body 2.350E-09 1.910E-09 3.236E-09 0.0 Ingestion thyroid 1.344E-06 9.243E-07 1.395E-06 0.0 Ingestion kidney 7.028E-09 4.153E-09 2.620E-09 0.0 Ingestion lung 0.0 0.0 0.0 0.0 Ingestion GI 1.082E-09 6.067E-10 3.673E-10 0.0 Total bone 2.867E-09 2.268E-09 4.186E-09 0.0 Total liver 4.107E-09 3.204E-09 4.289E-09 0.0 Total whole body 2.350E-09 1.910E-09 3.236E-09 0.0 Total thyroid 1.344E-06 9.243E-07 1.395E-06 0.0 Total kidney 7.028E-09 4.153E-09 2.620E-09 0.0 Total lung 0.0 0.0 0.0 0.0 Total GI 1.082E-09 6.067E-10 3.673E-10 0.0 Gamma air 3.484E-02 3.484E-02 3.484E-02 3.484E-02 Beta air 7.863E-04 7.863E-04 7.863E-04 7.863E-04 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-26 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR ESE SECTOR AT DISTANCE 1.0 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) DESIGN BASIS CASE Dose Adult Teen Child Infant Ingestion bone 1.472E-08 1.165E-08 2.150E-08 0.0 Ingestion liver 2.109E-08 1.646E-08 2.203E-08 0.0 Ingestion whole body 1.207E-08 9.807E-09 1.662E-08 0.0 Ingestion thyroid 6.900E-06 4.746E-06 7.162E-06 0.0 Ingestion kidney 3.609E-08 2.133E-08 1.345E-08 0.0 Ingestion lung 0.0 0.0 0.0 0.0 Ingestion GI 5.555E-09 3.116E-09 1.886E-09 0.0 Total bone 1.472E-08 1.165E-08 2.150E-08 0.0 Total liver 2.109E-08 1.646E-08 2.203E-08 0.0 Total whole body 1.207E-08 9.807E-09 1.662E-08 0.0 Total thyroid 6.900E-06 4.746E-06 7.162E-06 0.0 Total kidney 3.609E-08 2.133E-08 1.345E-08 0.0 Total lung 0.0 0.0 0.0 0.0 Total GI 5.555E-09 3.116E-09 1.886E-09 0.0 Gamma air 1.693E-01 1.693E-01 1.693E-01 1.693E-01 Beta air 3.821E-03 3.821E-03 3.821E-03 3.821E-03 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-27 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR ESE SECTOR AT DISTANCE 3.7 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) DESIGN BASIS CASE Dose Adult Teen Child Infant Ingestion bone 5.431E-10 8.810E-10 2.128E-09 0.0 Ingestion liver 7.781E-10 1.245E-09 2.180E-09 0.0 Ingestion whole body 4.452E-10 7.418E-10 1.645E-09 0.0 Ingestion thyroid 2.546E-07 3.590E-07 7.089E-07 0.0 Ingestion kidney 1.332E-09 1.613E-09 1.332E-09 0.0 Ingestion lung 0.0 0.0 0.0 0.0 Ingestion GI 2.050E-10 2.357E-10 1.867E-10 0.0 Total bone 5.431E-10 8.810E-10 2.128E-09 0.0 Total liver 7.781E-10 1.245E-09 2.180E-09 0.0 Total whole body 4.452E-10 7.418E-10 1.645E-09 0.0 Total thyroid 2.546E-07 3.590E-07 7.089E-07 0.0 Total kidney 1.332E-09 1.613E-09 1.332E-09 0.0 Total lung 0.0 0.0 0.0 0.0 Total GI 2.050E-10 2.357E-10 1.867E-10 0.0 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-28 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR SE SECTOR AT DISTANCE 1.1 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) DESIGN BASIS CASE Dose Adult Teen Child Infant Ingestion bone 5.046E-08 3.992E-08 7.369E-08 0.0 Ingestion liver 7.229E-08 5.641E-08 7.550E-08 0.0 Ingestion whole body 4.136E-08 3.361E-08 5.696E-08 0.0 Ingestion thyroid 2.365E-05 1.627E-05 2.455E-05 0.0 Ingestion kidney 1.237E-07 7.311E-08 4.611E-08 0.0 Ingestion lung 0.0 0.0 0.0 0.0 Ingestion GI 1.904E-08 1.068E-08 6.465E-09 0.0 Total bone 5.046E-08 3.992E-08 7.369E-08 0.0 Total liver 7.229E-08 5.641E-08 7.550E-08 0.0 Total whole body 4.136E-08 3.361E-08 5.696E-08 0.0 Total thyroid 2.365E-05 1.627E-05 2.455E-05 0.0 Total kidney 1.237E-07 7.311E-08 4.611E-08 0.0 Total lung 0.0 0.0 0.0 0.0 Total GI 1.904E-08 1.068E-08 6.465E-09 0.0 Gamma air 4.298E-01 4.298E-01 4.298E-01 4.298E-01 Beta air 9.700E-03 9.700E-03 9.700E-03 9.700E-03 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-29 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR SE SECTOR AT DISTANCE 3.7 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) DESIGN BASIS CASE Dose Adult Teen Child Infan t Ingestion bone 2.200E-09 3.569E-09 8.620E-09 0.0 Ingestion liver 3.152E-09 5.043E-09 8.832E-09 0.0 Ingestion whole body 1.803E-09 3.005E-09 6.664E-09 0.0 Ingestion thyroid 1.031E-06 1.454E-06 2.872E-06 0.0 Ingestion kidney 5.394E-09 6.535E-09 5.394E-09 0.0 Ingestion lung 0.0 0.0 0.0 0.0 Ingestion GI 8.303E-10 9.547E-10 7.563E-10 0.0 Total bone 2.200E-09 3.569E-09 8.620E-09 0.0 Total liver 3.152E-09 5.043E-09 8.832E-09 0.0 Total whole body 1.803E-09 3.005E-09 6.664E-09 0.0 Total thyroid 1.031E-06 1.454E-06 2.872E-06 0.0 Total kidney 5.394E-09 6.535E-09 5.394E-09 0.0 Total lung 0.0 0.0 0.0 0.0 Total GI 8.303E-10 9.547E-10 7.563E-10 0.0 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-30 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR NW SECTOR AT DISTANCE 0.5 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) NORMAL OPERATION CASE Dose Adult Teen Child Infant Ingestion bone 6.470E-02 5.040E-02 8.703E-02 0.0 Ingestion liver 1.065E-01 8.164E-02 1.019E-01 0.0 Ingestion whole body 8.301E-02 4.040E-02 3.180E-02 0.0 Ingestion thyroid 2.400E-01 1.651E-01 2.491E-01 0.0 Ingestion kidney 3.293E-02 1.946E-02 1.227E-02 0.0 Ingestion lung 1.215E-02 1.041E-02 1.146E-02 0.0 Ingestion GI 1.136E-01 5.889E-02 3.272E-02 0.0 Total bone 6.470E-02 5.040E-02 8.703E-02 0.0 Total liver 1.065E-01 8.164E-02 1.019E-01 0.0 Total whole body 8.301E-02 4.040E-02 3.180E-02 0.0 Total thyroid 2.400E-01 1.651E-01 2.491E-01 0.0 Total kidney 3.293E-02 1.946E-02 1.227E-02 0.0 Total lung 1.215E-02 1.041E-02 1.146E-02 0.0 Total GI 1.136E-01 5.889E-02 3.272E-02 0.0 Gamma air 7.859E-01 7.859E-01 7.859E-01 7.859E-01 Beta air 2.496E-01 2.496E-01 2.496E-01 2.496E-01 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-31 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR NW SECTOR AT DISTANCE 3.6 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) NORMAL OPERATION CASE Dose Adult Teen Child Infant Ingestion bone 3.185E-02 5.189E-02 1.052E-01 0.0 Ingestion liver 3.432E-02 5.455E-02 9.099E-02 0.0 Ingestion whole body 2.788E-02 2.742E-02 2.522E-02 0.0 Ingestion thyroid 3.039E-03 4.286E-03 8.463E-03 0.0 Ingestion kidney 1.127E-02 1.365E-02 1.127E-02 0.0 Ingestion lung 3.721E-02 1.365E-02 1.127E-02 0.0 Ingestion GI 1.088E-02 1.168E-02 8.568E-03 0.0 Total bone 3.185E-02 5.189E-02 1.052E-01 0.0 Total liver 3.432E-02 5.455E-02 9.099E-02 0.0 Total whole body 2.788E-02 2.742E-02 2.522E-02 0.0 Total thyroid 3.039E-03 4.286E-03 8.463E-03 0.0 Total kidney 1.127E-02 1.365E-02 1.127E-02 0.0 Total lung 3.721E-03 6.847E-03 1.030E-02 0.0 Total GI 1.088E-02 1.168E-02 8.568E-03 0.0 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-32 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR NNW SECTOR AT DISTANCE 0.5 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) NORMAL OPERATION CASE Dose Adult Teen Child Infant Ingestion bone 3.396E-02 2.645E-02 4.568E-02 0.0 Ingestion liver 5.590E-02 4.285E-02 5.346E-02 0.0 Ingestion whole body 4.357E-02 2.120E-02 1.669E-02 0.0 Ingestion thyroid 1.260E-01 8.665E-02 1.307E-01 0.0 Ingestion kidney 1.728E-02 1.021E-02 6.442E-03 0.0 Ingestion lung 6.378E-03 5.463E-03 6.017E-03 0.0 Ingestion GI 5.964E-02 3.091E-02 1.717E-02 0.0 Total bone 3.396E-02 2.645E-02 4.568E-02 0.0 Total liver 5.590E-02 4.285E-02 5.346E-02 0.0 Total whole body 4.357E-02 2.120E-02 1.669E-02 0.0 Total thyroid 1.260E-01 8.665E-02 1.307E-01 0.0 Total kidney 1.728E-02 1.021E-02 6.442E-03 0.0 Total lung 6.378E-03 5.463E-03 6.017E-03 0.0 Total GI 5.964E-02 3.091E-02 1.717E-02 0.0 Gamma air 4.312E-01 4.312E-01 4.312E-01 4.312E-01 Beta air 1.370E-01 1.370E-01 1.370E-01 1.370E-01 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-33 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR NNW SECTOR AT DISTANCE 1.5 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) NORMAL OPERATION CASE Dose Adult Teen Child Infant Inhalation bone 1.892E-03 1.164E-03 1.037E-03 1.221E-03 Inhalation liver 6.453E-03 4.281E-03 3.282E-03 3.440E-03 Inhalation whole body 3.269E-03 1.809E-03 1.646E-03 1.686E-03 Inhalation thyroid 1.736E-02 1.368E-02 1.719E-02 2.823E-02 Inhalation kidney 3.043E-03 2.126E-03 1.125E-03 7.919E-04 Inhalation lung 7.415E-03 6.104E-03 5.589E-03 7.540E-03 Inhalation GI 3.073E-03 1.751E-03 1.594E-03 1.630E-03 External whole body 2.043E-01 2.043E-01 2.043E-01 2.043E-01 External skin 2.950E-01 2.950E-01 2.950E-01 2.950E-01 Total bone 2.062E-01 2.054E-01 2.053E-01 2.055E-01 Total liver 2.107E-01 2.085E-01 2.075E-01 2.077E-01 Total whole body 2.075E-01 2.061E-01 2.059E-01 2.059E-01 Total thyroid 2.216E-01 2.179E-01 2.214E-01 2.325E-01 Total kidney 2.073E-01 2.064E-01 2.054E-01 2.050E-01 Total lung 2.117E-01 2.104E-01 2.098E-01 2.118E-01 Total GI 2.073E-01 2.060E-01 2.059E-01 2.059E-01 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-34 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR NNW SECTOR AT DISTANCE 3.6 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) NORMAL OPERATION CASE Dose Adult Teen Child Infant Ingestion bone 1.672E-02 2.723E-02 5.521E-02 0.0 Ingestion liver 1.801E-02 2.863E-02 4.776E-02 0.0 Ingestion whole body 1.463E-02 1.439E-02 1.324E-02 0.0 Ingestion thyroid 1.594E-03 2.248E-03 4.439E-03 0.0 Ingestion kidney 5.915E-03 7.166E-03 5.915E-03 0.0 Ingestion lung 1.953E-03 3.594E-03 5.404E-03 0.0 Ingestion GI 5.711E-03 6.131E-03 4.497E-03 0.0 Total bone 1.672E-02 2.723E-02 5.521E-02 0.0 Total liver 1.801E-02 2.863E-02 4.776E-02 0.0 Total whole body 1.463E-02 1.439E-02 1.324E-02 0.0 Total thyroid 1.594E-03 2.248E-03 4.439E-03 0.0 Total kidney 5.915E-03 7.166E-03 5.915E-03 0.0 Total lung 1.953E-03 3.594E-03 5.404E-03 0.0 Total GI 5.711E-03 6.131E-03 4.497E-03 0.0 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-35 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR N SECTOR AT DISTANCE 0.5 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) NORMAL OPERATION CASE Dose Adult Teen Child Infant Ingestion bone 1.403E-02 1.093E-02 1.887E-02 0.0 Ingestion liver 2.309E-02 1.770E-02 2.209E-02 0.0 Ingestion whole body 1.800E-02 8.761E-03 6.895E-03 0.0 Ingestion thyroid 5.204E-02 3.580E-02 5.401E-02 0.0 Ingestion kidney 7.140E-03 4.219E-03 2.661E-03 0.0 Ingestion lung 2.635E-03 2.257E-03 2.486E-03 0.0 Ingestion GI 2.464E-02 1.277E-02 7.096E-03 0.0 Total bone 1.403E-02 1.093E-02 1.887E-02 0.0 Total liver 2.309E-02 1.770E-02 2.209E-02 0.0 Total whole body 1.800E-02 8.761E-03 6.895E-03 0.0 Total thyroid 5.204E-02 3.580E-02 5.401E-02 0.0 Total kidney 7.140E-03 4.219E-03 2.661E-03 0.0 Total lung 2.635E-03 2.257E-03 2.486E-03 0.0 Total GI 2.464E-02 1.277E-02 7.096E-03 0.0 Gamma air 2.452E-01 2.452E-01 2.452E-01 2.452E-01 Beta air 7.789E-02 7.789E-02 7.789E-02 7.789E-02 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-36 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR NNE SECTOR AT DISTANCE 0.5 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) NORMAL OPERATION CASE Dose Adult Teen Child Infant Ingestion bone 8.356E-03 6.509E-03 1.124E-02 0.0 Ingestion liver 1.375E-02 1.054E-02 1.315E-02 0.0 Ingestion whole body 1.072E-02 5.217E-03 4.106E-03 0.0 Ingestion thyroid 3.099E-02 2.131E-02 3.216E-02 0.0 Ingestion kidney 4.252E-03 2.513E-03 1.585E-03 0.0 Ingestion lung 1.569E-03 1.344E-03 1.481E-03 0.0 Ingestion GI 1.468E-02 7.605E-03 4.226E-03 0.0 Total bone 8.356E-03 6.509E-03 1.124E-02 0.0 Total liver 1.375E-02 1.054E-02 1.315E-02 0.0 Total whole body 1.072E-02 5.217E-03 4.106E-03 0.0 Total thyroid 3.099E-02 2.131E-02 3.216E-02 0.0 Total kidney 4.252E-03 2.513E-03 1.585E-03 0.0 Total lung 1.569E-03 1.344E-03 1.481E-03 0.0 Total GI 1.468E-02 7.605E-03 4.226E-03 0.0 Gamma air 1.214E-01 1.214E-01 1.214E-01 1.214E-01 Beta air 3.855E-02 3.855E-02 3.855E-02 3.855E-02 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-37 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR NE SECTOR AT DISTANCE 0.5 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) NORMAL OPERATION CASE Dose Adult Teen Child Infant Ingestion bone 1.055E-02 8.215E-03 1.418E-02 0.0 Ingestion liver 1.736E-02 1.331E-02 1.660E-02 0.0 Ingestion whole body 1.353E-02 6.585E-03 5.182E-03 0.0 Ingestion thyroid 3.911E-02 2.690E-02 4.059E-02 0.0 Ingestion kidney 5.367E-03 3.171E-03 2.000E-03 0.0 Ingestion lung 1.980E-03 1.696E-03 1.868E-03 0.0 Ingestion GI 1.852E-02 9.598E-03 5.333E-03 0.0 Total bone 1.055E-02 8.215E-03 1.418E-02 0.0 Total liver 1.736E-02 1.331E-02 1.660E-02 0.0 Total whole body 1.353E-02 6.585E-03 5.182E-03 0.0 Total thyroid 3.911E-02 2.690E-02 4.059E-02 0.0 Total kidney 5.367E-03 3.171E-03 2.000E-03 0.0 Total lung 1.980E-03 1.696E-03 1.868E-03 0.0 Total GI 1.852E-02 9.598E-03 5.333E-03 0.0 Gamma air 8.058E-02 8.058E-02 8.058E-02 8.058E-02 Beta air 2.559E-02 2.559E-02 2.559E-02 2.559E-02 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-38 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR ENE SECTOR AT DISTANCE 0.7 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) NORMAL OPERATION CASE Dose Adult Teen Child Infant Ingestion bone 4.508E-03 3.512E-03 6.064E-03 0.0 Ingestion liver 7.420E-03 5.688E-03 7.097E-03 0.0 Ingestion whole body 5.784E-03 2.815E-03 2.215E-03 0.0 Ingestion thyroid 1.671E-02 1.150E-02 1.735E-02 0.0 Ingestion kidney 2.294E-03 1.356E-03 8.551E-04 0.0 Ingestion lung 8.467E-04 7.252E-04 7.988E-04 0.0 Ingestion GI 7.918E-03 4.103E-03 2.280E-03 0.0 Total bone 4.508E-03 3.512E-03 6.064E-03 0.0 Total liver 7.420E-03 5.688E-03 7.097E-03 0.0 Total whole body 5.784E-03 2.815E-03 2.215E-03 0.0 Total thyroid 1.671E-02 1.150E-02 1.735E-02 0.0 Total kidney 2.294E-03 1.356E-03 8.551E-04 0.0 Total lung 8.467E-04 7.252E-04 7.988E-04 0.0 Total GI 7.918E-03 4.103E-03 2.280E-03 0.0 Gamma air 4.566E-02 4.566E-02 4.566E-02 4.566E-02 Beta air 1.450E-02 1.450E-02 1.450E-02 1.450E-02 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-39 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR ENE SECTOR AT DISTANCE 4.5 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) NORMAL OPERATION CASE Dose Adult Teen Child Infant Inhalation bone 6.802E-05 4.186E-05 3.728E-05 4.388E-05 Inhalation liver 2.320E-04 1.539E-04 1.180E-04 1.236E-04 Inhalation whole body 1.175E-04 6.502E-05 5.915E-05 6.060E-05 Inhalation thyroid 6.241E-04 4.919E-04 6.180E-04 1.015E-03 Inhalation kidney 1.094E-04 7.641E-05 4.045E-05 2.847E-05 Inhalation lung 2.665E-04 2.194E-04 2.009E-04 2.710E-04 Inhalation GI 1.105E-04 6.296E-05 5.729E-05 5.859E-05 External whole body 6.780E-03 6.780E-03 6.780E-03 6.780E-03 External skin 1.031E-02 1.031E-02 1.031E-02 1.031E-02 Total bone 6.848E-03 6.822E-03 6.817E-03 6.824E-03 Total liver 7.012E-03 6.934E-03 6.898E-03 6.903E-03 Total whole body 6.897E-03 6.845E-03 6.839E-03 6.840E-03 Total thyroid 7.404E-03 7.272E-03 7.398E-03 7.795E-03 Total kidney 6.889E-03 6.856E-03 6.820E-03 6.808E-03 Total lung 7.046E-03 6.999E-03 6.981E-03 7.051E-03 Total GI 6.890E-03 6.843E-03 6.837E-03 6.838E-03 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-40 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR E SECTOR AT DISTANCE 1 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) NORMAL OPERATION CASE Dose Adult Teen Child Infant Ingestion bone 4.214E-03 3.283E-03 5.668E-03 0.0 Ingestion liver 6.936E-03 5.317E-03 6.634E-03 0.0 Ingestion whole body 5.406E-03 2.631E-03 2.071E-03 0.0 Ingestion thyroid 1.562E-02 1.074E-02 1.621E-02 0.0 Ingestion kidney 2.144E-03 1.267E-03 7.993E-04 0.0 Ingestion lung 7.914E-04 6.779E-04 7.467E-04 0.0 Ingestion GI 7.401E-03 3.835E-03 2.131E-03 0.0 Total bone 4.214E-03 3.283E-03 5.668E-03 0.0 Total liver 6.936E-03 5.317E-03 6.634E-03 0.0 Total whole body 5.406E-03 2.631E-03 2.071E-03 0.0 Total thyroid 1.562E-02 1.074E-02 1.621E-02 0.0 Total kidney 2.144E-03 1.267E-03 7.993E-04 0.0 Total lung 7.914E-04 6.779E-04 7.467E-04 0.0 Total GI 7.401E-03 3.835E-03 2.131E-03 0.0 Gamma air 5.322E-02 5.322E-02 5.322E-02 5.322E-02 Beta air 1.690E-02 1.690E-02 1.690E-02 1.690E-02 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-41 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR ESE SECTOR AT DISTANCE 1 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) NORMAL OPERATION CASE Dose Adult Teen Child Infant Ingestion bone 2.163E-02 1.685E-02 2.909E-02 0.0 Ingestion liver 3.560E-02 2.729E-02 3.404E-02 0.0 Ingestion whole body 2.775E-02 1.350E-02 1.063E-02 0.0 Ingestion thyroid 8.020E-02 5.517E-02 8.324E-02 0.0 Ingestion kidney 1.101E-02 6.503E-03 4.102E-03 0.0 Ingestion lung 4.061E-03 3.479E-03 3.832E-03 0.0 Ingestion GI 3.798E-02 1.968E-02 1.094E-02 0.0 Total bone 2.163E-02 1.685E-02 2.909E-02 0.0 Total liver 3.560E-02 2.729E-02 3.404E-02 0.0 Total whole body 2.775E-02 1.350E-02 1.063E-02 0.0 Total thyroid 8.020E-02 5.517E-02 8.324E-02 0.0 Total kidney 1.101E-02 6.503E-03 4.102E-03 0.0 Total lung 4.061E-03 3.479E-03 3.832E-03 0.0 Total GI 3.798E-02 1.968E-02 1.094E-02 0.0 Gamma air 2.586E-01 2.586E-01 2.586E-01 2.586E-01 Beta air 8.215E-02 8.215E-02 8.215E-02 8.215E-02 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-42 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR ESE SECTOR AT DISTANCE 3.7 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) NORMAL OPERATION CASE Dose Adult Teen Child Infant Ingestion bone 3.101E-02 5.052E-02 1.024E-01 0.0 Ingestion liver 3.341E-02 5.311E-02 8.859E-02 0.0 Ingestion whole body 2.714E-02 2.669E-02 2.456E-02 0.0 Ingestion thyroid 2.959E-03 4.173E-03 8.239E-03 0.0 Ingestion kidney 1.097E-02 1.329E-02 1.097E-02 0.0 Ingestion lung 3.623E-03 6.667E-03 1.003E-02 0.0 Ingestion GI 1.059E-02 1.137E-02 8.341E-03 0.0 Total bone 3.101E-02 5.052E-02 1.024E-01 0.0 Total liver 3.341E-02 5.311E-02 8.859E-02 0.0 Total whole body 2.714E-02 2.669E-02 2.456E-02 0.0 Total thyroid 2.959E-03 4.173E-03 8.239E-03 0.0 Total kidney 1.097E-02 1.329E-02 1.097E-02 0.0 Total lung 3.623E-03 6.667E-03 1.003E-02 0.0 Total GI 1.059E-02 1.137E-02 8.341E-03 0.0 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-43 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR SE SECTOR AT DISTANCE 1.1 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) NORMAL OPERATION CASE Dose Adult Teen Child Infant Ingestion bone 7.411E-02 5.773E-02 9.968E-02 0.0 Ingestion liver 1.220E-01 9.351E-02 1.167E-01 0.0 Ingestion whole body 9.508E-02 4.627E-02 3.642E-02 0.0 Ingestion thyroid 2.749E-01 1.891E-01 2.853E-01 0.0 Ingestion kidney 3.771E-02 2.229E-02 1.406E-02 0.0 Ingestion lung 1.392E-02 1.192E-02 1.313E-02 0.0 Ingestion GI 1.302E-01 6.745E-02 3.748E-02 0.0 Total bone 7.411E-02 5.773E-02 9.968E-02 0.0 Total liver 1.220E-01 9.351E-02 1.167E-01 0.0 Total whole body 9.508E-02 4.627E-02 3.642E-02 0.0 Total thyroid 2.749E-01 1.891E-01 2.853E-01 0.0 Total kidney 3.771E-02 2.229E-02 1.406E-02 0.0 Total lung 1.392E-02 1.192E-02 1.313E-02 0.0 Total GI 1.302E-01 6.745E-02 3.748E-02 0.0 Gamma air 6.566E-01 6.566E-01 6.566E-01 6.566E-01 Beta air 2.085E-01 2.085E-01 2.085E-01 2.085E-01 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.3-44 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED OFFSITE DOSES FOR SE SECTOR AT DISTANCE 3.7 MI (MREM/YEAR - AIR DOSES IN MRAD/YEAR) NORMAL OPERATION CASE Dose Adult Teen Child Infant Ingestion bone 1.255E-01 2.044E-01 4.144E-01 0.0 Ingestion liver 1.352E-01 2.149E-01 3.584E-01 0.0 Ingestion whole body 1.098E-01 1.080E-01 9.935E-02 0.0 Ingestion thyroid 1.199E-02 1.690E-02 3.338E-02 0.0 Ingestion kidney 4.439E-02 5.378E-02 4.439E-02 0.0 Ingestion lung 1.466E-02 2.697E-02 4.056E-02 0.0 Ingestion GI 4.286E-02 4.602E-02 3.375E-02 0.0 Total bone 1.255E-01 2.044E-01 4.144E-01 0.0 Total liver 1.352E-01 2.149E-01 3.584E-01 0.0 Total whole body 1.098E-01 1.080E-01 9.935E-02 0.0 Total thyroid 1.199E-02 1.690E-02 3.338E-02 0.0 Total kidney 4.439E-02 5.378E-02 4.439E-02 0.0 Total lung 1.466E-02 2.697E-02 4.056E-02 0.0 Total GI 4.286E-02 4.602E-02 3.375E-02 0.0 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE Table 11.4-1 Sheet 1 of 6 RADIATION MONITORS AND READOUTS RMS RMS Detector Unit(e) Readout(s) Channel Description Elevation Range Type Location No. Indication Location Recorder Location R -1 Control room area 140 ft 0.1 to E4 G-M NE area rad rack behind main 0 RI-1 Local RR-1 RNRMA monitor mR/hr control panel 0 RM-1 RNRMA R -2 Containment area 140 ft 0.1 to E4 G-M West containment wall at 1 RI-2 Local RR-2 RNRMA monitor mR/hr personnel hatch 1 RM-2 RNRMA 2 RI-2 Local RR-2 RNRMC 2 RM-2 RNRMC R -3 Oily water separator 85 ft 10 to E6 Gamma Oily water separator room 0 RM-3 RNRMA RR-3 RNRMA effluent monitor cpm Scint. R -4 Centrifugal charging 73 ft 0.1 to E4 G-M SE corner at charging pump 1 RI-4 Local RR-4 RNRMA pump CCP3 (room #2) mR/hr 1-3, pump outside of SE corner 1 RM-4 RNRMA monitor (N-wall at charging pump 2-3) 2 RI-4 Local RR-4 RNRMC 2 RM-4 RNRMC R -6 NSSS sampling room 100 ft 0.1 to E4 G-M SW corner on U-1 1 RI-6 Local RR-6 RNRMA area monitor mR/hr (NW corner on U-2) 1 RM-6 RNRMA 2 RI-6 Local RR-6 RNRMC 2 RM-6 RNRMC R -7 Incore seal table area 115 ft 0.1 to E4 G-M South hatch at upper internals 1 RI-7 Local RR-7 RNRMA monitor mR/hr laydown area on U-1 1 RM-7 RNRMA (North on U-2) 2 RI-7 Local RR-7 RNRMC 2 RM-7 RNRMC R -10 Auxiliary building 85 ft 0.1 to E4 G-M South wall 0 RI-10 Local RR-10 RNRMA control board digital mR/hr 0 RM-10 RNRMA system area monitor R -11 Containment air 100 ft 10 to E6 Gamma Area GE 1 RM -11 RNRMB RR-11 RNRMA particulate monitor cpm Scint. 2 RM -11 RNRMD RR-11 RNRMC R -12 Containment air 100 ft 10 to E6 G-M Area GE 1 RM -12 RNRMB RR-12 RNRMA radioactive gas monitor cpm 2 RM -12 RNRMD RR-12 RNRMC R -13 RHR exhaust duct air 100 ft 10 to E6 Gamma Wall in north corridor at RHR ht 1 RM -13 RNRMB RR-13 RNRMA particulate monitor cpm Scint exchg.area - area K for U-1 2 RM -13 RNRMD RR-13 RNRMC (South for U-2) Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE Table 11.4-1 Sheet 2 of 6 RMS RMS Detector Unit(e) Readout(s) Channel Description Elevation Range Type Location No. Indication Location Recorder Location R -14 Plant vent radioactive 85ft 10 to 5E6 Beta Scint Plant vent at NE wall/ 1 RM-14 RNRMS3 EARS TSC and gas monitors cpm penetration room - area L for 1 RM-14R RNRMS4 EARS TSC R-14R U-1 (SE for U-2) 2 RM-14 RNRMS3 EARS TSC 2 RM-14R RNRMS4 EARS TSC R -15 Steam jet air ejector 104ft 10 to 5E6 Beta Scint. Turb. Bldg., Area C on wall at 1 RM-15 RNRMS4 EARS TSC and radioactive gas dischg. cpm Col. line 10 between Col. lines 1 RM-15R RNRMS4 EARS TSC R-15R monitors C&D for U-1 (line 26 for U-2) 2 RM-15 RNRMS4 EARS TSC 2 RM-15R RNRMS4 EARS TSC R -17A CCW discharge header 73 ft 10 to E6 Gamma Outside door on wall at 1 RM -17A RNRMB RR-17A RNRME and effluent monitors cpm Scint. component cooling pump room 1 RM -17B RNRMB RR-17B RNRME R -17B 2 RM -17A RNRMD RR-17A RNRMC 2 RM -17B RNRMD RR-17B RNRMC R -18 Liquid radwaste 55 ft 10 to E6 Gamma Pipe tunnel from radwaste 0 RM -18 RNRME RR-18 RNRME discharge line effluent cpm Scint. (north corridor) monitor R -19 Steam generator 100 ft 10 to E6 Gamma SE side of containment 1 RI-19 SGSP RR-19 RNRME blowdown sample cpm Scint. structure/penetration room - 1 RM -19A RNRMB effluent monitor area GE for U-1 (NE for U-2) 2 RI-19 SGSP RR-19 RNRMC 2 RM -19A RNRMD R -22 Gas decay tank 55 ft 10 to E6 G-M Pipe tunnel 1 RM-22 RNRME RR-22 RNRME radioactive gas cpm 2 RM-22 RNRMC RR-22 RNRMC discharge monitor R -23 Steam generator 100 ft 1 to E6 cpm Gamma Area GE 1 RI-23 PM205 RR-23 PM205 blowdown tank effluent Scint. 1 RR-23A RNRME to out-fall monitor 2 RI-23 PM205 RR-23 PM205 2 RR-23A RNRMC R -24 Plant vent iodine 85ft 10 to 5E6 Gamma Plant vent at area L 1 RM-24 RNRMS3 EARS TSC and monitor cpm Scint. 1 RM-24R RNRMS4 EARS TSC R-24R 2 RM-24 RNRMS3 EARS TSC 2 RM-24R RNRMS4 EARS TSC R -25 Main control room air 160 ft 0.01 to E3 Gamma Auxiliary bldg. control 1 RI-25 RCRM - - intake monitor mR/hr Scint. room air intake 2 RI-25 RCRM R -26 Main control room air 160 ft 0.01 to E3 Gamma Auxiliary bldg. control room air 1 RI-26 RCRM - - intake monitor mR/hr Scint. intake 2 RI-26 RCRM Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE Table 11.4-1 Sheet 3 of 6 RMS RMS Detector Unit(e) Readout(s) Channel Description Elevation Range Type Location No. Indication Location Recorder Location R -28 Plant vent air 85ft 10 to 5E6 Beta Scint. Plant vent at NE wall to 1 RM-28 RNRMS3 EARS TSC and particulate monitors cpm containment 1 RM-28R RNRMS4 EARS TSC R-28R structure/penetration room - 2 RM-28 RNRMS3 EARS TSC area L, for U-1 2 RM-28R RNRMS4 EARS TSC (SE for U-2) R -29 Plant vent high 155 ft 0.1 to E7 ION Plant vent on platform on NW 1 RI-29 PAM-2 RR-29 PAM-2 radiation gross gamma mR/hr side of duct for U-1 (SW for monitor U-2) 2 RI-29 PAM-2 RR-29 PAM-2 R -30 Containment high 140 ft 1 to E7 R/hr ION East stairway 1 RI-30 PAM-2 RR-30 PAM-1 range area radiation 2 RI-30 PAM-2 RR-30 PAM-1 monitor - 1 R -31 Containment high 140 ft 1 to E7 R/hr ION West stairway 1 RI-31 PAM-2 RR-31 PAM-1 range area radiation 2 RI-31 PAM-2 RR-31 PAM-1 monitor - 2 R -34 Area Monitor for Plant 85ft 0.1 to E7 ION Plant vent at NE wall 1 RC-22 Local RR-34 PAM-2 Vent Monitoring Skid mR/hr penetration room - Area L for 1 RI-34 Local U-1 2 RC-22 Local RR-34 PAM-2 (SE for U-2) 2 RI-34 Local R-41 Gas decay tank cubicle 64 ft 1 to E4 ION Gas decay tank 1 1 RI-41 ABCBDS - - radiation monitor mR/hr 2 RI-41 ABCBDS (1-1, 2-1)
- 2 R-42 Gas decay tank cubicle 64 ft 1 to E4 ION Gas decay tank 2 1 RI-42 ABCBDS - -
radiation monitor mR/hr 2 RI-42 ABCBDS (1-2, 2-2) R-43 Gas decay tank cubicle 64 ft 1 to E4 ION Gas decay tank 3 1 RI-43 ABCBDS - - radiation monitor mR/hr 2 RI-43 ABCBDS (1-3, 2-3) R-44A Containment Purge 100 ft 10 to 5E6 Beta Scint. Area L 1 RM-44A RNRMS1 EARS TSC and Exhaust cpm 1 RM-44B RNRMS2 EARS TSC R-44B 2 RM-44A RNRMS1 EARS TSC 2 RM-44B RNRMS2 EARS TSC R -48 HRSS (Sentry) post- 85 ft 0.1 to E7 ION HRSS 1 RI-48 POPLSI - - accident sampling mR/hr 2 RI-48 POPLSI room Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE Table 11.4-1 Sheet 4 of 6 RMS RMS Detector Unit(e) Readout(s) Channel Description Elevation Range Type Location No. Indication Location Recorder Location R -51 Control room 140ft 0.01 to E4 G-M NW corner of turbine building 1 RI-51 RCRM - - pressurization system mR/hr ventilation intake air monitor R -52 Control room 140 ft 0.01 to E4 G-M NW corner of turbine building 1 RI-52 RCRM - - pressurization system mR/hr ventilation intake air monitor R -53 Control room 140ft 0.01 to E4 G-M SW corner of turbine building 2 RI-53 RCRM - - pressurization system mR/hr ventilation intake air monitor R -54 Control room 140 ft 0.01 to E4 G-M SW corner of turbine building 2 RI-54 RCRM - - pressurization system mR/hr ventilation intake air monitor R -58 Spent Fuel Pool Area 140 ft 0.1 to E4 G-M West wall column line 111 and 1 RI-58 Local RR-58 RNRMA Monitor mR/hr 251 1 RI-58A PAM-2 2 RI-58 Local RR-58 RNRMC 2 RI-58A PAM-2 R -59 New Fuel Storage Area 140 ft 0.1 to E4 G-M West wall column line 153 and 1 RI-59 Local RR-59 RNRMA Monitor mR/hr 203 1 RI-59A PAM-2 2 RI-59 Local RR-59 RNRMC 2 RI-59A PAM-2 R -60 TSC-Office Area 104 ft 0.1 to E4 G-M East wall at exit door 0 RI-60 Local RR-60 Local Radiation Monitor mR/hr R -61 TSC-Operations/RMS 104 ft 0.1 to E4 G-M South wall at doorway 0 RI-61 Local RR-61 Local Area Monitor mR/hr R -62 TSC-Computations 104 ft 0.1 to E4 G-M South wall at doorway 0 RI-62 Local RR-62 Local Center Area Monitor mR/hr R -63 TSC-NRC Office Area 104 ft 0.1 to E4 G-M North wall at doorway 0 RI-63 Local RR-63 Local Monitor mR/hr R -64 TSC-HVAC Equipment 104 ft 0.1 to E4 G-M East wall at midpoint 0 RI-64 Local RR-64 Local room area monitor mR/hr R -65 TSC-laboratory area 104 ft 0.1 to E4 G-M North wall at midpoint 0 RI-65 Local RR-65 Local monitor mR/hr Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE Table 11.4-1 Sheet 5 of 6 RMS RMS Detector Unit(e) Readout(s) Channel Description Elevation Range Type Location No. Indication Location Recorder Location R -66 TSC-air particulate 104 ft 10 to E6 Beta Scint. TSC-HVAC equipment room 0 RI-66 Local RR-66 Local monitor cpm R -67 TSC-noble gas monitor 104 ft 10 to E6 Beta Scint. TSC-HVAC equipment room 0 RI-67 Local RR-67 Local cpm R -68 TSC-laboratory air 104 ft 10 to E6 Beta Scint. TSC-HVAC equipment room 0 RI-68 Local RR-68 Local particulate monitor cpm R -69 TSC-laboratory noble 104 ft 10 to E6 Beta Scint. TSC-HVAC equipment room 0 RI-69 Local RR-69 Local gas monitor cpm R -71 Main steam line noble 130 ft 10 to E6 G-M Outside NW side of 1 RI-71 RNGFFD RR-71 RNRME gas radiation monitor cpm containment - area FW for U-1 2 RI-71 RNGFFD RR-71 RNRMC (lead 1) (SW for U-2) R-72 Main steam line noble 130 ft 10 to E6 G-M Outside NW side of 1 RI-72 RNGFFD RR-72 RNRME gas radiation monitor cpm containment - area FW for U-1 2 RI-72 RNGFFD RR-72 RNRMC (lead 2) (SW for U-2) R-73 Main steam line noble 130 ft 10 to E6 G-M Outside containment area GW 1 RI-73 RNGFFD RR-73 RNRME gas radiation monitor cpm 2 RI-73 RNGFFD RR-73 RNRMC (lead 3) R-74 Main steam line noble 130 ft 10 to E6 G-M Outside containment area GW 1 RI-74 RNGFFD RR-74 RNRME gas radiation monitor cpm 2 RI-74 RNGFFD RR-74 RNRMC (lead 4) R-82 TSC-iodine monitor 104 ft 10 to E6 Gamma TSC-HVAC equipment room 0 RI-82 Local RR-82 RMPTSC cpm Scint. R-83 TSC-laboratory iodine 104 ft 10 to E6 Gamma TSC-HVAC equipment room 0 RI-83 Local RR-83 RMPTSC monitor cpm Scint. R-84 Contact inspection 115 ft 0.1 to E3 ION Solid radwaste storage area 0 RM-84 Local - - station radiation R/hr monitor R-85 One-meter inspection 115 ft 0.1 to E3 ION Solid radwaste storage area 0 RM-85 Local - - station radiation R/hr monitor R-87 Plant vent extended 85 ft 1E-12 to Beta Scint. Plant vent at NE 1 RM-14 RNRMS3 EARS TSC range radioactive gas 1E-4 amps walls/penetration room - area L 2 RM-14 RNRMS3 EARS TSC monitors for U-1 (SE for U-2) Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE Table 11.4-1 Sheet 6 of 6 RMS RMS Detector Unit(e) Readout(s) Channel Description Elevation Range Type Location No. Indication Location Recorder Location R-90 R/W Storage 115 ft 0.1 to E4 G-M R/W Storage Truck Bay 0 RI-90A Local - - mR/hr RI-90B Mechanical RM Control RM R-92 Laundry facility 132 ft 10 to E5 G-M Laundry room 0 - - RR-92 Local cpm RF-87A Particulate and iodine 85 ft Plant vent at NE 1 - - - - and grab sampling wall/penetration room - area L 2 - - - - RF-87B assembly (post- for U-1 accident) (SE for U-2) RX-55 Laundry and Radwaste 115 ft Ventilation room 0 Facility Exhaust Solid Radwaste Storage Sampler Facility RX-56 Radwaste Storage 142 ft Mezzanine Area 0 Building and Laundry Radwaste Storage Building Facility Exhaust Sampler (a) Deleted in Revision 4 (b) Deleted in Revision 9 (c) Deleted in Revision 11 (d) Post-LOCA sampling control panel (e) Units designation: 0 = 1 monitor common to both Units 1 = Unit 1 monitor 2 = Unit 2 monitor Symbol Location 1 ABCBDS Auxiliary building control board digital system, elevation 85 ft 11 RNRMA Radiation monitor rack A, control room 2 ABRV Auxiliary building roof vent access area 12 RNRMB Radiation monitor rack B, control room 3 RNGFFD Main steam line radiation monitor rack, control room 13 RNRMC Radiation Monitor rack C, control room 4 HRSS Sentry high radiation sampling room 14 RNRMD Radiation Monitor rack D, control room 5 PAM-1 Post accident monitor panel 1, control room 15 RNRME Radiation monitor rack E, control room 6 PAM-2 Post accident monitor panel 2, control room 16 SGSP Steam generation sample panel 7 POPLSI Post-LOCA sampling control panel 17 RMPTSC Radiation monitor panel, TSC 8 PM197 Panel 197, access area of auxiliary building 9 PM205 Mechanical panel 205, auxiliary building 10 RCRM Rack chlorine and radiation monitor - Unit 2 side control room Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.4-3 RADIATION MONITOR - VALVE CONTROL OPERATIONS Radiation Element Affected Valves Effect RE-17A and RE-17B SV225, RCV16 Close CCW surge tank vent. RE-18 SV233, RCV18, FCV477 Liquid radwaste control valves: Close liquid radwaste over-board and open liquid radwaste equipment drain receiver dump. RE-23 SV237, FCV160, SV238, FCV157, SV239, FCV154, (a) High radiation (1) close steam generators 1 to 4 SV240, FCV151, SV242, FCV498, SV241, FCV499 blowdown tank inlet and sample, (2) close steam generator blowdown tank outlet, (3) close blowdown tank outlet to discharge tunnel and open blowdown tank outlet to equipment drain receiver; (b) Power loss to RE (1) close steam generators 1 to 4 blowdown tank inlet and sample, (2) close steam generator blowdown tank outlet, (3) close blowdown tank outlet to discharge tunnel and open blowdown tank outlet to equipment drain receiver. RE-19 SV237, FCV160, SV238, FCV157, SV239, FCV154, High radiation (1) close steam generators 1 to 4 SV240, FCV151, SV242, FCV498, SV241, FCV499 blowdown tank inlet and sample, (2) close steam generator blowdown tank outlet, (3) close blowdown tank outlet to discharge tunnel and open blowdown tank outlet to equipment drain receiver. RE-51, RE-52, RE-53, and - Initiate control room pressurization system. RE-54 RE-44A and RE-44B Trains A and B containment vent isolation valves Closure. RE-22 SV218, RCV17 Gaseous radwaste vent closure. Revision 11 November 1996
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.5-1 SOLID RADWASTE SYSTEM INPUT VOLUMES Source Design Basis Volume Normal Operation Bases Volume Boric acid waste 11,700 gal/yr 23,190 gal/yr EPRI NP-3370 Concentrates(a) Spent ion exchange resin 1,600 ft3/yr 1,600 ft3/yr EPRI NP-3370 Expended filtration/ion 400 ft3/yr 400 ft3/yr 12 beds/yr exchange media Spent filter cartridges 240/yr 240/yr EPRI NP-3370 (requiring encapsulation) Dry active waste 210 boxes 180 boxes EPRI NP-3370, EPRI NP-2900 (a) Boric acid evaporator has been abandoned in place Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.5-2 Sheet 1 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ACTIVITY IN RADWASTE SYSTEM DEMINERALIZERS NORMAL OPERATION CASE (CURIES/YEAR) Primary Primary Letdown Letdown Letdown Spent Fuel Nuclide Mxd Bed Demin Cation Demin Mxd Bed Demin Cation Demin Anion Demin Pool Demin H-3 0.0 0.0 0.0 0.0 0.0 0.0 Cr-51 0.281E 02 0.309E-00 0.179E-02 0.179E-04 0.0 0.119E-04 Mn-54 0.282E 02 0.310E-00 0.289E-02 0.289E-04 0.0 0.693E-04 Fe-55 0.227E 01 0.249E-01 0.111E-05 0.111E-07 0.0 0.856E-13 Co-58 0.587E 03 0.646E 01 0.515E-01 0.515E-03 0.0 0.863E-03 Fe-59 0.238E 02 0.262E 00 0.185E-02 0.185E-04 0.0 0.225E-04 Co-60 0.250E 03 0.275E 01 0.267E-01 0.267E-03 0.0 0.983E-04 Sr-89 0.740E 01 0.814E-01 0.600E-03 0.600E-05 0.0 0.190E-04 Sr-90 0.175E 01 0.192E-01 0.188E-03 0.188E-05 0.0 0.113E-04 Y-90 0.173E 01 0.190E-01 0.186E-03 0.186E-05 0.0 0.112E-04 Sr-91 0.384E-01 0.422E-03 0.102E-20 0.102E-22 0.0 0.976E-10 Y-91 0.377E-01 0.415E-03 0.101E-20 0.101E-22 0.0 0.960E-10 Sr-92 0.436E-02 0.479E-04 0.310E-62 0.310E-64 0.0 0.934E-12 Y-92 0.436E-02 0.479E-04 0.310E-62 0.310E-64 0.0 0.934E-12 Zr-95 0.231E 01 0.255E-01 0.199E-03 0.199E-05 0.0 0.248E-04 Nb-95 0.349E 01 0.384E-01 0.323E-03 0.323E-05 0.0 0.391E-04 Mo-99 0.0 0.0 0.0 0.0 0.0 0.0 I-131 0.138E 04 0.0 0.265E 00 0.0 0.265E-04 0.216E-02 Te-132 0.586E 02 0.0 0.776E-03 0.0 0.776E-08 0.472E-04 I-132 0.648E 02 0.0 0.801E-03 0.0 0.102E-07 0.487E-04 I-133 0.205E-03 0.0 0.144E-07 0.0 0.144E-11 0.165E-05 Cs-134 0.146E 04 0.131E 03 0.401E 01 0.361E 01 0.0 0.817E-02 I-134 0.113E 01 0.0 0.0 0.0 0.0 0.730E-10 I-135 0.361E 02 0.0 0.971E-24 0.0 0.971E-28 0.225E-07 Cs-136 0.237E 02 0.213E 01 0.217E-01 0.195E-01 0.0 0.119E-04 Cs-137 0.301E 04 0.271E 03 0.843E 01 0.759E 01 0.0 0.255E-01 Ba-140 0.315E 01 0.346E-01 0.109E-03 0.109E-05 0.0 0.422E-05 La-140 0.330E 01 0.362E-01 0.125E-03 0.125E-05 0.0 0.476E-05 Ce-144 0.357E 01 0.393E-01 0.365E-03 0.365E-05 0.0 0.159E-04 Pr-144 0.357E-01 0.393E-01 0.365E-03 0.365E-05 0.0 0.159E-04 Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.5-2 Sheet 2 of 2 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ACTIVITY IN RADWASTE SYSTEM DEMINERALIZERS NORMAL OPERATION CASE (CURIES/YEAR) Primary Primary Letdown Letdown Letdown Spent Fuel Nuclide Mxd Bed Demin Cation Demin Mxd Bed Demin Cation Demin Anion Demin Pool Demin H-3 0.0 0.0 0.0 0.0 0.0 0.0 Cr-51 0.281E 02 0.309E-00 0.179E-02 0.179E-04 0.0 0.119E-04 Mn-54 0.282E 02 0.310E-00 0.289E-02 0.289E-04 0.0 0.693E-04 Fe-55 0.227E 01 0.249E-01 0.111E-05 0.111E-07 0.0 0.856E-13 Co-58 0.588E 03 0.646E 01 0.516E-01 0.516E-03 0.0 0.863E-03 Fe-59 0.238E 02 0.262E 00 0.185E-02 0.185E-04 0.0 0.225E-04 Co-60 0.250E 03 0.275E 01 0.267E-01 0.267E-03 0.0 0.983E-04 Sr-89 0.617E 02 6.679E 00 0.500E-02 0.500E-04 0.0 0.566E-04 Sr-90 0.146E 02 0.160E 00 0.156E-02 0.156E-04 0.0 0.684E-04 Y-90 0.144E 02 0.158E 00 0.155E-02 0.155E-04 0.0 0.677E-04 Sr-91 0.320E 00 0.352E-02 0.853E-20 0.853E-22 0.0 0.980E-10 Y-91 0.315E 00 0.346E-02 0.840E-20 0.840E-22 0.0 0.964E-10 Sr-92 0.363E-01 0.399E-03 0.258E-61 0.258E-63 0.0 0.934E-12 Y-92 0.363E-01 0.399E-03 0.258E-61 0.258E-63 0.0 0.934E-12 Zr-95 0.193E 02 0.212E 00 0.166E-02 0.166E-04 0.0 0.401E-04 Nb-95 0.291E 02 0.320E 00 0.270E-02 0.270E-04 0.0 0.672E-04 Mo-99 0.0 0.0 0.0 0.0 0.0 0.0 I-131 0.115E 05 0.0 0.221E 01 0.0 0.221E-03 0.280E-02 Te-132 0.489E 03 0.0 0.647E-02 0.0 0.647E-07 0.525E-04 I-132 0.540E 03 0.0 0.668E-02 0.0 0.853E-07 0.543E-04 I-133 0.171E 04 0.0 0.120E-06 0.0 0.120E-10 0.194E-05 Cs-134 0.120E 05 0.108E 04 0.331E 02 0.298E 02 0.0 0.345E-01 I-134 0.941E 01 0.0 0.0 0.0 0.0 0.730E-10 I-135 0.301E 03 0.0 0.809E-23 0.0 0.809E-27 0.225E-07 Cs-136 0.197E 03 0.178E 02 0.181E 00 0.163E 00 0.0 0.336E-04 Cs-137 0.251E 05 0.226E 04 0.702E 02 0.632E 02 0.0 0.131E 00 Ba-140 0.263E 02 0.289E 00 0.906E-03 0.906E-05 0.0 0.713E-05 La-140 0.275E 02 0.302E 00 0.104E-02 0.104E-04 0.0 0.792E-05 Ce-144 0.298E 02 0.327E 00 0.305E-02 0.305E-04 0.0 0.868E-04 Pr-144 0.289E 02 0.327E 00 0.305E-02 0.305E-04 0.0 0.869E-04 Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.5-4 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ACTIVITY COLLECTED IN RADWASTE FILTER CARTRIDGES AT TIME OF REPLACEMENT(a) (CURIES) DESIGN BASIS CASE RCC Letdown(b) Radwaste(c) Nuclide Primary Loop Ion Exchge Concentrates Condensate 0-1 0-2 0-3 H-3 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Cr-51 0.148E 00 0.429E-07 0.214E-07 0.214E-10 0.418E-05 0.731E-03 0.212E-05 Mn-54 0.669E-01 0.312E-07 0.156E-07 0.156E-10 0.268E-05 0.475E-03 0.136E-05 Fe-55 0.125E-01 0.280E-10 0.140E-10 0.140E-13 0.165E-07 0.153E-05 0.833E-08 Co-58 0.232E 01 0.925E-06 0.462E-06 0.462E-09 0.827E-04 0.146E-01 0.421E-04 Fe-59 0.112E 00 0.396E-07 0.198E-07 0.198E-10 0.365E-05 0.643E-03 0.186E-05 Co-60 0.481E 00 0.234E-06 0.117E-06 0.117E-09 0.198E-04 0.352E-02 0.101E-04 NORMAL OPERATION CASE RCC Letdown(b) Radwaste(c) Nuclide Primary Loop Ion Exchge Concentrates Condensate 0-1 0-2 0-3 H-3 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Cr-51 0.148E 00 0.428E-07 0.214E-07 0.214E-10 0.372E-06 0.446E-02 0.441E-06 Mn-54 0.669E-01 0.312E-07 0.156E-07 0.156E-10 0.184E-06 0.356E-02 0.301E-06 Fe-55 0.125E-01 0.280E-10 0.140E-10 0.140E-13 0.125E-07 0.313E-05 0.637E-08 Co-58 0.232E 01 0.924E-06 0.462E-06 0.462E-09 0.620E-05 0.102E 00 0.904E-05 Fe-59 0.112E 00 0.396E-07 0.198E-07 0.198E-10 0.292E-06 0.427E-02 0.393E-06 Co-60 0.481E 00 0.234E-06 0.117E-06 0.117E-09 0.134E-05 0.269E-01 0.225E-05 (a) Three cartridge replacements per cycle. (b) As Designated on Figure 11.5-6. (c) As Designated on Figure 11.5-8. Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.5-5
SUMMARY
OF RADWASTE MATERIALS SHIPMENT Quantity Quantity Possible per Shipments Shipment Material per Year Shipment per Year Method Packaged wastes Solidified liquid concentrates 0 ft3 100 ft3 0 truck Class B/C Resins 160 ft3 80 ft3 2 truck Class A Resin 400 ft3 200 ft3 2 truck Spent filter cartridges 240 100 2 truck Filtration/ion exchange media 60 ft3 200 ft3 1/2 truck Dry active wastes 50 boxes 10 boxes 5 truck Revision 19 May 2010
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.6-4 Sheet 1 of 5 HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED ENVIRONMENTAL RADIOLOGICAL MONITORING PROGRAM
SUMMARY
(PRE-OPERATIONAL RESULTS) Medium or Pathway Type and Total Lower Limit All Control Sampled Number of of Location Locations Number of (Unit of Analyses Detection(a) Name, Distance(d) Mean(1)(b) Mean(1)(b) Reportable Measurement) Performed (LLD) and Direction Range(b) Range(b) Occurrences Seawater, Tritium (12) - - None detected - 0 (pCi. L-1) Gamma Isotopic (36) - - None detected - 0 54Mn None detected 59Fe 3.79x102(c) None detected 58Co None detected 60Co None detected 65Zn None detected 95Zr None detected 95Nb None detected 131I None detected 134Cs None detected 137Cs None detected 140Ba 2.14x103(c) None detected 140La 6.09x102(c) None detected Surface water Tritium (12) - - None detected - 0 (pCi. L-1) Gross Beta (12) - Sta. 5S2, 2.94(12/12) - 0 0.6 mi, 65° 2.24-3.58 Gamma Isotopic - 0 54Mn None detected 59Fe None detected 58Co None detected 60Co None detected 65Zn None detected Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.6-4 Sheet 2 of 5 Medium or Pathway Type and Total Lower Limit All Control Sampled Number of of Location Locations Number of (Unit of Analyses Detection(a) Name, Distance(d) Mean(1)(b) Mean(1)(b) Reportable Measurement) Performed (LLD) and Direction Range(b) Range(b) Occurrences 95Zr None detected 95Nb None detected 131I None detected 134Cs None detected 137Cs None detected 140Ba 1.35x102(c) None detected 140La 3.48x101(c) None detected Drinking water Tritium (12) - None detected - 0 (pCi. L-1) Gross Beta (12) Sta. D W1, 2.4 (8/12) - 0 0.0 mi, 2.24-4.09 in plant 131-Iodine (12) None detected - 0 Gamma Isotopic (12) - 0 54Mn None detected 59Fe None detected 58Co None detected 60Co None detected 65Zn None detected 95Zr None detected 95Nb None detected 131I None detected 134Cs None detected 137Cs None detected 140Ba None detected 140La None detected Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.6-4 Sheet 3 of 5 Medium or Pathway Type and Total Lower Limit All Control Sampled Number of of Location Locations Number of (Unit of Analyses Detection(a) Name, Distance(d) Mean(1)(b) Mean(1)(b) Reportable Measurement) Performed (LLD) and Direction Range(b) Range(b) Occurrences Outfall Tritium (18) - None detected - 0 (pCi. L-1) Gamma Isotopic (18) None detected 54Mn None detected 59Fe None detected 58Co None detected 60Co None detected 65Zn None detected 95Zr None detected 95Nb None detected 131I None detected 134Cs None detected 137Cs None detected 140Ba 2.74x103(c) None detected 140La 8.35x102(c) None detected Airborne 131I (507) - None detected 0.108 (3/211) 0 (pCi. m-3) 0.0137-0.159 Gross Beta - 0.012(296/296) 0.010(211/211) (507) 0.004-0.033 0.005-0.033 Gamma Isotopic (507) - - 0 134Cs None detected None detected 137Cs None detected None detected Fish and Gamma Isotopic - - 0 seafood (79) (pCi. kg-1) 54Mn 1.46x102 None detected None detected Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.6-4 Sheet 4 of 5 Medium or Pathway Type and Total Lower Limit All Control Sampled Number of of Location Locations Number of (Unit of Analyses Detection(a) Name, Distance(d) Mean(1)(b) Mean(1)(b) Reportable Measurement) Performed (LLD) and Direction Range(b) Range(b) Occurrences 59Fe 5.19x102 None detected None detected 58Co 1.74x102 None detected None detected 60Co 2.02x102 None detected None detected 65Zn - None detected None detected 134Cs 1.50x102 None detected None detected 137Cs 1.46x102 None detected 16.4 (5/57) Milk 131I (30) - None detected None detected 0 (pCi. L-1) Gamma Isotopic (30) - 0 134Cs None detected None detected 137Cs None detected None detected 140Ba None detected None detected 140La None detected None detected Food Gamma - - 0 products Isotopic (36) (pCi. kg-1) 131I 6.65x101 None detected None detected 134Cs None detected None detected 137Cs None detected None detected Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.6-4 Sheet 5 of 5 Medium or Pathway Type and Total Lower Limit All Control Sampled Number of of Location Locations Number of (Unit of Analyses Detection(a) Name, Distance(d) Mean(1)(b) Mean(1)(b) Reportable Measurement) Performed (LLD) and Direction Range(b) Range(b) Occurrences Direct TLD Packets 1 mR/mo(e) Sta. 3S1(f) 77.2(313/313) Sta. 2F2 and 0 radiation (335) 0.4 mi, 23 49.6-106.9 4D1 (mR) mR/yr 62.0(22/22) 67.8-66.2 mR/yr 8.9 (11/11(f) 7.9-10.1 mR/mo (106.9 mR/yr) (a) Unless specified, all required LLDs were met. (b) Mean and range based upon detectable measurements only. Fraction of detectable measurements at specified locations is indicated in parentheses (1); e.g., (10/12) means 10 samples out of 12 collected showed activity. (c) A priori LLD not met due to elapse time between collection and count dates, short half-life of nuclide involved, and equipment failure. Value listed is worst case. (d) Only one station location for this sample type; therefore, no control or indicator stations are listed. (e) Sensitivity of TLD system. (f) Indicator location with Highest Annual Mean. Note: The information presented in the above table was developed in support of the original license and is considered historical. Revision 23 December 2016
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.6-11 Sheet 1 of 2 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD) Airborne Food Particulate Fish Products Sediment Water or Gases (pCi/kg, Milk (pCi/kg, (pCi/kg, Analysis (pCi/l) (pCi/m3) wet) (pCi/l) wet) dry) Gross beta 4 0.01 H-3 2000* Mn-54 15 130 Fe-59 30 260 Co-58, 60 15 130 Zn-65 30 260 Zr- Nb -95 15 I-131 1** 0.07 1 60 Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 Ba- La -140 15 15
- For surface water samples, a value of 3000 pCi/l may be used.
- If no drinking water pathway exists, a value of 15 pCi/l may be used.
Table Notation The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real" signal. Revision 12 September 1998
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.6-11 Sheet 2 of 2 For a particular measurement system (which may include radiochemical separation): 4.66 S b LLD EV2.2 2Yexp( t) where: LLD is "a priori" the lower limit of detection as defined above (as pCi per unit mass or volume) Sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute) E is the counting efficiency (as counts per transformation) V is the sample size (in units of mass or volume) 2.22 is the number of transformations per minute per picocurie Y is the fractional radiochemical yield (when applicable) is the radioactive decay constant for the particular radionuclide t is the elapsed time between sample collection (or end of the sample collection period) and time of counting The value of Sb used in the calculation of the LLD for a detection system will be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma ray spectrometry, the background will include the typical contributions of other radionuclides normally present in the samples (e.g., potassium-40 in milk samples). Analyses will be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally, background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Environmental Radiological Operating Report. Revision 12 September 1998
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.6-13 ESTIMATED RELATIVE CONCENTRATIONS ( / Q ) (a) 22-1/2° Radial Sectors Nearest NW NNW N NNE NE ENE E ESE SE Milk cow None None None None None None None None None Meat animal N/A N/A N/A N/A N/A N/A N/A N/A N/A Milk goat None None None None None None None None None Residence 1.53X10-7 4.16X10-7 None 3.89X10-8 2.00X10-8 3.64X10-8 5.89X10-8 7.07X10-7(b) None Vegetable None None None None None None None None None garden Site boundary 3.44X10-6 2.70X10-6 1.51X10-6 8.25X10-7 1.62X10-7 9.18X10-8 1.07X10-7 5.20X10-7 1.32X10-6 (a) In units of seconds per cubic meter. (b) Vegetable farm has workers with residence occupancy factor of 1/2 for inhalation and group plane pathway. Revision 11 November 1996
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 11.6-14 ESTIMATED DEPOSITIONS (/Q)(a) 22-1/2° Radial Sectors Nearest NW NNW N NNE NE ENE E ESE SE Milk cow None 1.21X10-10 6.52X10-11 4.09X10-11 4.48X10-11 6.13X10-11 1.13X10-10 3.79X10-10 Meat animal 1.50X10-8 6.71X10-9 3.47X10-9 2.18X10-9 1.43X10-9 1.64X10-9 1.67X10-9 5.53X10-9 3.55X10-8 Milk goat None None None None None None None None None Residence 3.83X10-10 8.15X10-10 None 8.75X10-11 4.00X10-11 7.85X10-11 1.41X10-10 4.77X10-9 None Vegetable None None None None None None 1.41X10-10 4.77X10-9 None garden Site boundary 1.50X10-8 6.81X10-9 3.47X10-9 2.18X10-9 1.43X10-9 1.64X10-9 1.67X10-9 5.53X10-9 3.55X10-8 (a) In units of meters-2, includes sector width and frequency of winds in each sector. Revision 11 November 1996
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FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 11.2-2 LIQUID WASTE PROCESS FLOW DIAGRAM DESIGN BASIS CASE (HISTORICAL) Revision 20 November 2011 Revision 23 December 2016
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 11.2-3 LIQUID WASTE PROCESS FLOW DIAGRAM NORMAL OPERATIONS CASE (HISTORICAL) Revision 20 November 2011 Revision 23 December 2016
L E G END PROCESS FLOW PATH
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- 2. SEE TABLE 11.2-16 FOR TOTAL ANNUAL VOLUMETRIC & ACTIVITY FLOW PAST EACH POINT "6 "
SLOWDOWN FROM STEAM GEN TORS (TYPICAL OF 4) STEAM r--"1 GENERATOR SAMPLES FROM STEAM GENERATOR SLOWDOWN TANK FROM AUX. 51EAM RECYCLE DRAIN RSCl!l'le.R PVMP!> MODE ( FIG 11.2-5)
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TURBINE- GENERATOR TO CIRCULATING \I/ATER BUILDING SUMP DISCHARGE STRUCTURE LEGEND PROCESS FLOW PATH DISCHARGE PATH * * *
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- 1. NORMAL OPERATION CASE ASSUMES BLOWDOWN IS PROCESSED VIA RECYCLE PATH
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- 3. STREAM FIS EQUIVALENT TO STREAM 60AAND60 BON FIGURE 11.2-3.
FSAR UPDATE UN!TS 1 AND 2 DIABLO CANYON SITE FIGURE 11.2-5 BLOWDOWN SYSTEM FLOW DIAGRAM RECYCLE PATH Revision 11 November l 996
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- TRITIUM AIRBORNE CONCENTRATION IN CONTAINMENT VERSUS Til!E (HISTORICAL) 10-:u.~-l..'-'-'-,L..l...'--'--'--'--'-.J..LJ-,LJ....U...Ll..LL~,---'--'-'-W...~...:...C...JL.....C...L...,f,,.~..CJ.~*~--*"""""=---~l......-,:~
20 30 35 TIME (YEARS) Revision 23 December 2016 LBVP UFSAR Change Request Radioactive Waste Management Revision 11 November 1996
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FS*AR *UPDATE UNITS 1 AND 2 PACIFIC tI ~
OCEAN DIABLO CANYON SITE FIGURE11.2-9 I SITE PLOT PLAN RADWASTE DISCHARGE Revision 11 November 1996
LAUNDRY FACILITY SOLID RADWASTE STORAGE VENTILATION EXHAUST BU ILO ING VENT ILATl ON EXHAUST UNIT UNIT 2 o o o or--1---, 00 I I I NO T E S FUEL HANDLING I FUEL HANDLING I RADWASTE FILTERS 1. FUEL HANDLING BUILDING VENTILATION, AUXILIARY I BUILDING VENTILATION, CONTAINMENT VEHTILATJON, MAIN C0ff0£HSER AIR EJECTOR, ANO GLAND STEAM
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I 3. Condenser vacuum pump (shared between CONTAINMENT STRUCTURE AUXILIARY BUILDING I CONTAINMENT STRUCTURE units) exhausts to the atmosphere. I STEAM GENERATOR N0,1-2 I -------aJM/ STEAM ..GENERATOR NO. 2 - 2
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I-"' 2*4 lll:'--\-STEAM GENERATOR NO. 2-3 MAIN STEAM SAFETY VALVES ST!AM GENER.ATOR HO.. I - I n (': 1 .. 3 U3 MAIN STEAM SAFETY VALVES 10,. POWER RELIEF YM..VE 10 % POWER RELIEF VALVE lt----ls-srEAM GENERATOR NO. 2 - 4 STEAM GENERATOR NO, z- I . I I I GAS DECAY TANKS ( i%~W*ER-.-,-,,-;.-=-,s~r----~*1,~-PO-WER--R£,-,,.~.*-lYE-S=~~------~------------- J -.o v, E~ ~,T REHEAT SAFTY VALVES EAST REHEAT SAFETY VAL\IES TURBINE -GENERATOR BUILDING TURBINE-GENERATOR BUILDING IZZZZZZZZZZZZZZZZZ1111,*111111175zz1,1111. /,' l ,.*- .:_*:..117.1.Z.1.Z.1.Z.1.Zul.1.1_1L'17~,L'LILZLZ.Lt-1L'1Z~tr......1.t~,:..*',....._:.L/~tulLZ.1.Z.1./:.,(:...LZLI.Lt~Zc...,L'Lt~'..,l:.,l:.,luZ.:.*1Z.1.Z:.,Zc.1..1.1.,.J.l..,1.1.t:..L.ZLi.1./.1.(.1.*.1.i.J.I TUUIHE GENERATOl BU~LDING OORI'll<<JOUS VENT I TDRBINE GENDtATOil BUIUlllC CONTINUOUS VDI _ / (note 2) (note 2)
~ E S T REHEAT SAFETY VALVES Condenser Vacuum Pump (note 3) UNIT UNIT 2 FSAR UPDATE UNITS I AND 2 DIABLO CANYON SITE FIGURE 11,3-4 l GASEOUS WASTE SYSTEMS' RELEASE POINTS Revision 22 May 2015 Revision 11 November 1996
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 11.4-1 Sheet 1 of 2 RADIATION MONITORING SYSTEM Revision 18 October 2008 Revision 23 December 2016
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 11.4-1 Sheet 2 of 2 RADIATION MONITORING SYSTEM Revision 18 October 2008 Revision 23 December 2016
WASTE MOBILE
!NPUTfROM B.!.C!WMU . $0LIDIFIED FIGURE 115* ,--+ Si:QNCENTRA TES PROCESSING SYSTEM PROCESSING LINERS SYSTEM MOBILE 1;!01.J:C!I[lf; p
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EBQ!;;tS:S:IHQ S::l'.SltM SYSTEM LUU:RSlHI cs INPUT FROM I SPENT RESIN MOBILE 6ADWASTE SOLIDIFIED QB FIGURE 11.5*3---+i PROCESSING SYSTEM PR~ESSl~G DEWA.TERED I SYSTEM LINERSlHICS ( INPUT FROM SPENT FIL TEA MQJill.f RAOWMTE . SOLIDIFIED OR ABSORBED FIGURE 11.5-6 --+ CARTRIDGE PAOCE§SING DRUMS OR 11.S.7 AND 11.5
** PROCESSING SYSTEM LINERS I ORY ACTIVE I MISC. ORY ACTIVE WASTE l'JAlli PROCESSING ,_-----------------------+
t CQMPACTEO OR ORUMMED. SYSTEM ~ FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 11.5 -1 SOLID RADWASTE SYSTEM Revision 11 November 1996
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 11.5-3 SPENT RESIN FLOW DIAGRAM Revision 20 November 2011
( X RADWASTE STORAGE BUILDING UNIT NO. 1 UNIT NO. 2 CONTAINMENT CONTAINMENT D D D D
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ADMIN. D BLDG. UNIT NO. 1 UNIT NO. 2 TURBINE GENERATOR TURBINE GENERATOR D ( FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 11.5-4 LOCATION OF ONSITE STORAGE FACILITY ( Revision 16 June 2005
REACTOR COOLANT LETDOWN FILTER FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 11.5-6 CHEMICAL AND VOLUME CONTROL SYSTEM DISPLAYING FILTERS Revision Revision 20 November 23 December 2016 2011 LBVP UFSAR Change Request Radioactive Waste Management
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SPENT FUEL POOL SKIMMER FILTER i=-ILTER 1-1 SPENT FUE;L HEAT SPENT FUEL f_OOL EXCHANGER SPENT FUEL POOL OEMINERAUZER REAJEUNG WATER REFUELING STORAGE TANK
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SPENT FUEL POOL RESIN TRAP FILTER FSAR UPDATE UNITS I AND 2 DIABLO CANYON SITE FIGURE 11.5-7 SPENT FUEL POOL DISPLAYING FILTERS Revision 11 November 1996
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LAUNDRY I DISTILLATE TANK EQUIPM[NT QAAIN R{CEl~tR$ WASTE FILTEIP ( 0-2) WASTE MEDIA DEMINERALIZER WASTE FIi.TU ( 0-5 ) I E!J.li!l I
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f!EC~IVEA~ CIRCUI.ATIIIG WATER OVEl!BOARD ( 0-3) WASTE FILTER ( 0-4) FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 11_5-8 LIQUID RADWASTE SYSTEM DISPLAYING FILTERS Revision 11 November 1996
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- 10 COVER OF FILTER VESSEL Revision 11 November 1996
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