ML16180A128: Difference between revisions
StriderTol (talk | contribs) (Created page by program invented by StriderTol) |
StriderTol (talk | contribs) (Created page by program invented by StriderTol) |
||
| (One intermediate revision by the same user not shown) | |||
| Line 2: | Line 2: | ||
| number = ML16180A128 | | number = ML16180A128 | ||
| issue date = 05/05/2016 | | issue date = 05/05/2016 | ||
| title = | | title = Revision 26 to the Updated Final Safety Analysis Report - Table of Contents | ||
| author name = | | author name = | ||
| author affiliation = Exelon Generation Co, LLC | | author affiliation = Exelon Generation Co, LLC | ||
| Line 16: | Line 16: | ||
=Text= | =Text= | ||
{{#Wiki_filter: | {{#Wiki_filter:GINNA/UFSAR 1 INTRODUCTION AND GENERAL DESCRIPTION OF THE 1 PLANT | ||
==1.1 INTRODUCTION== | |||
2 REFERENCES FOR SECTION 1.1 4 1.2 GENERAL PLANT DESCRIPTION 5 1.2.1 SITE AND ENVIRONMENT 5 1.2.2 | |||
==SUMMARY== | |||
PLANT DESCRIPTION 5 1.2.3 STRUCTURES 6 1.2.3.1 General 6 1.2.3.2 Containment 6 1.2.3.3 Auxiliary Building 7 1.2.3.4 Intermediate Building (See Drawings 33013-2101, 33013-2102, 9 33013-2105, 33013-2106, 33013-2107, 33013-2113, 33013-2114, 33013-2115, and 33013-2121) 1.2.3.5 Turbine Building 10 1.2.3.6 Control Building 11 1.2.3.7 All-Volatile-Treatment Building 12 1.2.3.8 Standby Auxiliary Feedwater Pump Building 12 1.2.3.9 Screen House 12 1.2.3.10 Service Building 12 1.2.3.11 Diesel Generator Building 13 1.2.3.12 Old Steam Generator Storage Facility 13 1.2.3.13 Canister Preparation Building 13 1.2.3.14 ISFSI Transfer Path and Storage Pad 14 1.2.3.15 Administration Building 14 1.2.4 NUCLEAR STEAM SUPPLY SYSTEM 14 1.2.5 REACTOR AND PLANT CONTROL 15 1.2.6 WASTE DISPOSAL SYSTEM 15 1.2.7 FUEL HANDLING SYSTEM 15 1.2.8 TURBINE AND AUXILIARIES 16 1.2.9 ELECTRICAL SYSTEM 16 1.2.10 ENGINEERED SAFETY FEATURES PROTECTION SYSTEMS 16 1.2.11 DESIGN HIGHLIGHTS 17 1.2.11.1 Power Level 17 Page 1 of 9 Revision 26 5 /2016 | |||
GINNA/UFSAR 1.2.11.2 Reactor Coolant Loops 17 1.2.11.3 Peak Specific Power 17 1.2.11.4 Fuel Clad 17 1.2.11.5 Fuel Assembly Design 17 1.2.11.6 Engineered Safety Features 18 1.2.11.7 Emergency Power 18 1.2.12 STATION WATER USE 18 1.2.13 FACILITY SAFETY CONCLUSIONS 19 1.3 COMPARISON TABLES 20 1.3.1 COMPARISONS WITH SIMILAR FACILITY DESIGNS 20 1.3.2 COMPARISON OF FINAL AND PRELIMINARY SAFETY 20 ANALYSIS REPORT INFORMATION (Historical) 1.3.2.1 Partial Length Rod Cluster Control Assemblies 20 1.3.2.2 Burnable Shim Rods 20 1.3.2.3 Safety Injection System Trip Signal 20 1.3.2.4 Containment Spray System Signal 20 1.3.2.5 Safety Injection System Accumulators 21 1.3.2.6 Spray Additive 21 1.3.2.7 Rod Stop and Reactor Trip on Startup 21 1.3.2.8 Miniature Neutron Flux Detectors 21 1.3.2.9 Core Thermocouples 21 1.3.2.10 Initial Leak Rate Test Method 21 1.3.2.11 Auxiliary Building Ventilation Filters 21 1.3.2.12 Control Center Buses 21 1.3.2.13 Condenser Circulating Water Flow 21 1.3.2.14 Ramp Loading Range 21 1.3.2.15 Condensate Storage Tanks Capacity 22 1.3.2.16 Fuel Transfer System Drive 22 1.3.2.17 Steam Line Flow Nozzles 22 1.3.3 Comparison of Uprate Parameters 22 Table 1.3-1 COMPARISON OF DESIGN PARAMETERS WITH POINT 23 BEACH Page 2 of 9 Revision 26 5 /2016 | |||
GINNA/UFSAR Table 1.3-2 COMPARISON OF DESIGN PARAMETERS WITH SAN 33 ONOFRE AND CONNECTICUT YANKEE Table 1.3-3 COMPARISON OF GINNA AND KEWAUNEE UPRATE NSSS 35 DESIGN PARAMETERS 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 36 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 38 1. | |||
==5.1 INTRODUCTION== | |||
38 1.5.2 DEVELOPMENT OF THE FINAL CORE DESIGN AND FINAL 38 THERMAL-HYDRAULIC AND PHYSICS PARAMETERS 1.5.3 CORE STABILITY 39 1.5.3.1 Core Power Distribution 39 1.5.3.2 Out-of-Core Ion Chambers 39 1.5.3.3 In-Core Control Equipment 40 1.5.3.4 Startup Test Program 40 1.5.4 DEVELOPMENT OF LONG ION CHAMBERS 41 1.5.5 CONTROL ROD EJECTION AND DROPPED CONTROL ROD 42 ACCIDENT ANALYSES 1.5.6 CHARCOAL FILTERS 43 1.5.7 REACTOR COOLANT PUMP CONTROLLED LEAKAGE 43 SEALS 1.5.8 SAFETY INJECTION SYSTEM 44 1.5.8.l Development of Safety Injection System Design 44 1.5.8.2 Development of Core Cooling Analysis 44 1.5.9 DEVELOPMENT OF DESIGN, INSPECTION, AND ACCEP- 44 TANCE CRITERIA FOR PRESTRESSED REINFORCED-CON-CRETE PRESSURE VESSELS 1.5.9.1 Rock Anchors 45 1.5.9.1.1 Design Criteria and Assumptions 45 1.5.9.1.2 Test Verification and Results 45 1.5.9.2 Rock Anchor Grout 45 1.5.9.3 Tendon Inspection and Acceptance Criteria 46 1.5.9.4 Wall Tendons 46 1.5.9.4.1 Corrosion Protection 46 1.5.9.4.2 Inspection and Acceptance 47 Page 3 of 9 Revision 26 5 /2016 | |||
GINNA/UFSAR 1.5.10 DEVELOPMENT OF CONTAINMENT HYDROGEN RECOM- 47 BINER REFERENCES FOR SECTION 1.5 48 1.6 MATERIAL INCORPORATED BY REFERENCE 49 1.7 DRAWINGS AND OTHER DETAILED INFORMATION 59 1.7.1 ELECTRICAL, INSTRUMENTATION, AND CONTROL DRAW- 59 INGS 1.7.2 PIPING AND INSTRUMENTATION DIAGRAMS (P&ID) 59 1.7.3 OTHER DETAILED INFORMATION 59 Table 1.7-1 ELECTRICAL, INSTRUMENTATION, AND CONTROL DRAW- 60 INGS Table 1.7-2 PIPING AND INSTRUMENTATION DIAGRAMS (P &ID 61 1.8 CONFORMANCE TO NRC REGULATORY GUIDES 68 1.8.1 CONFORMANCE TO AEC SAFETY GUIDES 68 1.8.1.1 Safety Guide 1 -Net Positive Suction Head for Emergency Core 68 Cooling and Containment Heat Removal System Pumps 1.8.1.2 Safety Guide 2 -Thermal Shock to Reactor Pressure Vessels 68 1.8.1.3 Safety Guide 3 -Assumptions Used for Evaluating the Potential 69 Radiological Consequences of a Loss-of-Coolant Accident for Boil-ing Water Reactors 1.8.1.4 Safety Guide 4 -Assumptions Used for Evaluating the Potential 69 Radiological Consequences of a Loss-of-Coolant Accident for Pres-surized Water Reactors 1.8.1.5 Safety Guide 5 - Assumptions Used for Evaluating the Potential 69 Radiological Consequences of a Steam Line Break Accident for Boil-ing Water Reactors 1.8.1.6 Safety Guide 6 -Independence Between Redundant Standby (Onsite) 69 Power Sources and Between Their Distribution Systems 1.8.1.7 Safety Guide 7 -Control of Combustible Gas Concentrations in Con- 70 tainment Following a Loss-of-Coolant Accident 1.8.1.8 Safety Guide 8 -Personnel Selection and Training 70 1.8.1.9 Safety Guide 9 -Selection of Diesel-Generator Set Capacity for 70 Standby Power Supplies 1.8.1.10 Safety Guide 10 -Mechanical (Cadweld) Splices in Reinforcing Bars 71 of Concrete Containments 1.8.1.11 Safety Guide 11 -Instrument Lines Penetrating Primary Reactor Con- 72 tainment 1.8.1.12 Safety Guide 12 -Instrumentation for Earthquakes 72 Page 4 of 9 Revision 26 5 /2016 | |||
GINNA/UFSAR 1.8.1.13 Safety Guide 13 -Fuel Storage Facility Design Basis 72 1.8.1.14 Safety Guide 14 -Reactor Coolant Pump Flywheel Integrity 73 1.8.1.15 Safety Guide 15 -Testing of Reinforcing Bars for Concrete Structures 74 1.8.1.16 Safety Guide 16 -Reporting of Operating Information 75 1.8.1.17 Safety Guide 17 -Protection Against Industrial Sabotage 75 1.8.1.18 Safety Guide 18 - Structural Acceptance Test for Concrete Primary 76 Reactor Containments 1.8.1.18.1 Structural Integrity Test 76 1.8.1.18.2 Instrumentation 76 1.8.1.18.3 Displacement Measurements 77 1.8.1.18.4 Strain Measurements 78 1.8.1.18.5 Test Results 78 1.8.1.19 Safety Guide 19 -Nondestructive Examination of Primary Contain- 79 ment Liners 1.8.1.19.1 Test Provisions 79 1.8.1.19.2 Examination of Welds 79 1.8.1.19.3 Pressure Tests 80 1.8.1.19.4 Quality Control Provisions 80 1.8.1.20 Safety Guide 20 -Vibration Measurements on Reactor Internals 81 1.8.1.21 Safety Guide 21 -Measuring and Reporting Effluents From Nuclear 81 Power Plants 1.8.1.22 Safety Guide 22 -Periodic Testing of Protection System Actuation 82 Functions 1.8.1.23 Safety Guide 23 -Onsite Meteorological Programs 82 1.8.1.24 Safety Guide 24 -Assumptions Used for Evaluating the Potential 83 Radiological Consequences of a Pressurized Water Reactor Radioac-tive Gas Storage Tank Failure 1.8.1.25 Safety Guide 25 - Assumptions Used for Evaluating the Potential 83 Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors 1.8.1.26 Safety Guide 26 -Quality Group Classification and Standards 83 1.8.1.27 Safety Guide 27 -Ultimate Heat Sink 83 1.8.1.28 Safety Guide 28 -Quality Assurance Program Requirements 84 1.8.1.29 Safety Guide 29 -Seismic Design Classification 84 1.8.2 CONFORMANCE TO DIVISION I REGULATORY GUIDES 85 Page 5 of 9 Revision 26 5 /2016 | |||
GINNA/UFSAR 1.8.2.1 Regulatory Guide 1.4 -Assumptions Used for Evaluating the Poten- 85 tial Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors 1.8.2.2 Regulatory Guide 1.10 -Mechanical (Cadweld) Splices in Reinforc- 85 ing Bars of Category I Concrete Structures 1.8.2.3 Regulatory Guide 1.15 -Testing of Reinforcing Bars for Category I 85 Concrete Structures 1.8.2.4 Regulatory Guide 1.16 -Reporting of Operating Information 85 1.8.2.5 Regulatory Guide 1.17 -Protection of Nuclear Plants Against Indus- 85 trial Sabotage 1.8.2.6 Regulatory Guide 1.18 -Structural Acceptance Test for Concrete Pri- 85 mary Reactor Containments 1.8.2.7 Regulatory Guide 1.19 -Nondestructive Examination of Primary 85 Containment Liner Welds 1.8.2.8 Regulatory Guide 1.26, Revision 3 - Quality Group Classifications & 86 Standards for Water, Steam, and Radioactive -Waste Containing Components of Nuclear Power Plants 1.8.2.9 Regulatory Guide 1.29, Revision 3 -Seismic Design Classification 86 1.8.2.10 Regulatory Guide 1.30 - Quality Assurance Requirements for the 86 Installation, Inspection, and Testing of Instrumentation and Electrical Equipment 1.8.2.11 Regulatory Guide 1.31 -Control of Stainless Steel Welding 87 1.8.2.12 Regulatory Guide 1.32 -Use of IEEE Standard 308-1971, Criteria for 88 Class IEE Electric Systems for Nuclear Power Generating Stations 1.8.2.13 Regulatory Guide 1.33 -Quality Assurance Program Requirements 89 (Operation) 1.8.2.14 Regulatory Guide 1.34 -Control of Electroslag Weld Properties 89 1.8.2.15 Regulatory Guide 1.35 -Inservice Surveillance of Ungrouted Ten- 89 dons in Prestressed Concrete Containment Structures 1.8.2.16 Regulatory Guide 1.36, Revision 0 -Nonmetallic Thermal Insulation 90 for Austenitic Stainless Steel 1.8.2.17 Regulatory Guide 1.37, Revision 0 -Quality Assurance for Cleaning 90 of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants 1.8.2.18 Regulatory Guide 1.38, Revision 2 - Quality Assurance Require- 91 ments for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants 1.8.2.19 Regulatory Guide 1.39 -Housekeeping Requirements for Water- 92 Cooled Nuclear Power Plants Page 6 of 9 Revision 26 5 /2016 | |||
GINNA/UFSAR 1.8.2.20 Regulatory Guide 1.40 -Qualification Tests of Continuous-Duty 92 Motors Installed Inside the Containment of Water-Cooled Nuclear Power Plants 1.8.2.21 Regulatory Guide 1.41 - Preoperational Testing of Redundant Onsite 92 Electric Power Systems to Verify Proper Load Group Assignments 1.8.2.22 Regulatory Guide 1.42 -Interim Licensing Policy on As Low As 93 Practicable for Gaseous Radioiodine Releases from Light-Water-Cooled Nuclear Power Reactors 1.8.2.23 Regulatory Guide 1.43 -Control of Stainless Steel Weld Cladding of 93 Low-Alloy Steel Components 1.8.2.24 Regulatory Guide 1.44 - Control of the Use of Sensitized Stainless 94 Steel 1.8.2.25 Regulatory Guide 1.45 -Reactor Coolant Pressure Boundary Leakage 95 Detection System 1.8.2.26 Regulatory Guide 1.46 -Protection Against Pipe Whip Inside Con- 96 tainment 1.8.2.27 Regulatory Guide 1.47 -Bypassed and Inoperable Status Indication 96 for Nuclear Power Plant Safety Systems 1.8.2.28 Regulatory Guide 1.48 -Design Limits and Loading Combinations 96 for Seismic Category I Fluid System Components 1.8.2.29 Regulatory Guide 1.49 -Power Levels of Water-Cooled Nuclear 97 Power Plants 1.8.2.30 Regulatory Guide 1.50 -Control of Preheat Temperature for Welding 97 of Low-Alloy Steel 1.8.2.31 Regulatory Guide 1.51 -Inservice Inspection of ASME Code Class 2 98 and 3 Nuclear Power Plant Components 1.8.2.32 Regulatory Guide 1.52 - Design, Testing, and Maintenance Criteria 98 for Atmospheric Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants 1.8.2.33 Regulatory Guide 1.53 -Application of the Single-Failure Criterion 99 To Nuclear Power Plant Protection Systems 1.8.2.34 Regulatory Guide 1.54, Revision 0 -Quality Assurance Requirements 99 for Protective Coatings Applied to Water-Cooled Nuclear Power Plants 1.8.2.35 Regulatory Guide 1.55 -Concrete Placement in Seismic Category I 100 Structures 1.8.2.36 Regulatory Guide 1.57 -Design Limits and Loading Combinations 100 for Metal Primary Reactor Containment System Components 1.8.2.37 Regulatory Guide 1.59 -Design-Basis Floods for Nuclear Power 100 Plants Page 7 of 9 Revision 26 5 /2016 | |||
GINNA/UFSAR 1.8.2.38 Regulatory Guide 1.94, Revision 1 - Quality Assurance Installation, 101 Inspections, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants 1.8.2.39 Regulatory Guide 1.143, Revision 1 - Design Guidance for Radioac- 101 tive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants 1.8.3 CONFORMANCE TO IEEE CRITERIA 101 1.8.3.1 Criteria for Protection Systems for Nuclear Power Generating Sta- 101 tions (IEEE 279-1971) 1.8.3.2 Class 1E Electric Systems for Nuclear Power Generating Stations 101 (IEEE 308-1971) 1.8.3.2.1 Principal Design Criteria 101 1.8.3.2.2 Alternating Current Power Systems 102 1.8.3.2.2.1 General 102 1.8.3.2.2.2 Distribution Systems 103 1.8.3.2.2.3 Preferred Power Supply 103 1.8.3.2.2.4 Standby Power Supply 103 1.8.3.2.3 Direct Current Power Systems 104 1.8.3.2.3.1 General 104 1.8.3.2.3.2 Distribution System 104 1.8.3.2.3.3 Battery Supply 104 1.8.3.2.3.4 Battery Charger Supply 104 1.8.3.2.3.5 Protective Devices 105 1.8.3.2.3.6 Performance Discharge Test Provisions 105 1.8.3.2.4 Vital Instrumentation and Control Power Systems 105 1.8.3.2.5 Surveillance Requirements 105 1.8.3.3 Electrical Penetration Assemblies in Containment Structures for 106 Nuclear Fueled Power Generating Stations (IEEE 317 - April 1971) 1.8.3.4 Qualifying Class I Electric Equipment for Nuclear Power Generating 106 Stations (IEEE 323-April 1971) 1.8.3.5 Type Tests of Continuous Duty Class I Motors Installed Inside the 107 Containment of Nuclear Power Generating Stations (IEEE 334-1971) 1.8.3.6 Installation, Inspection, and Testing Requirements for Instrumenta- 107 tion and Electric Equipment During the Construction of Nuclear Power Generating Stations (IEEE 336-1971) 1.8.3.7 Trial Use Criteria for the Periodic Testing of Nuclear Power Generat- 108 ing Station Protection Systems (IEEE 338-1971) | |||
Page 8 of 9 Revision 26 5 /2016 | |||
GINNA/UFSAR 1.8.3.8 Seismic Qualification of Class I Electrical Equipment for Nuclear 108 Power Generating Stations (IEEE 344-1971) | |||
REFERENCES FOR SECTION 1.8 109 FIGURES Figure 1.2-1 Ginna Station Plot Plan Page 9 of 9 Revision 26 5 /2016 | |||
GINNA/UFSAR 2 SITE CHARACTERISTICS 1 2.1 GEOGRAPHY AND DEMOGRAPHY 2 2.1.1 SITE LOCATION AND DESCRIPTION 2 2.1.2 EXCLUSION AREA AUTHORITY AND CONTROL 2 2.1.3 POPULATION DISTRIBUTION 3 REFERENCES FOR SECTION 2.1 6 2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES 7 2.2.1 LOCATIONS AND ROUTES 7 2. | |||
==2.2 DESCRIPTION== | |||
7 Railroads 7 Pipelines 7 Waterways 7 Airports 8 Military Facilities 8 Toxic Chemicals 9 Onsite Toxic Chemicals 9 Offsite Toxic Chemicals 10 REFERENCES FOR SECTION 2.2 11 Table 2.2-1 TYPICAL INDUSTRIES IN WAYNE COUNTY (CIRCA 1969) 12 Table 2.2-2 TYPICAL INDUSTRIES IN THE ROCHESTER AREA OF MON- ROE COUNTY (CIRCA 1969) (LOCATED 18 MILES WEST OF THE SITE) 15 2.3 METEOROLOGY 16 2.3.1 REGIONAL CLIMATOLOGY 16 2.3.2 LOCAL METEOROLOGY 16 2.3.3 ONSITE METEOROLOGICAL MEASUREMENTS PROGRAM 17 2.3.4 DIFFUSION ESTIMATES 19 Long-Term Diffusion Characteristics 19 Meteorological Data 19 Airflow Trajectory and Terrain Influences 19 Atmospheric Diffusion Model 20 Source Configuration Considerations 21 Unobstructed Release Point 21 Obstructed Release Point 22 Removal Mechanisms 23 Summary of Plant Discharges 23 Input Assumptions 23 Results 24 Page 1 of 10 Revision 26 5/2016 | |||
GINNA/UFSAR Accident Analysis Diffusion Characteristics 24 Nuclear Regulatory Commission Evaluation (Historical) 24 Rochester Gas and Electric Corporation Evaluation (Historical) 26 Current Approved Evaluation 26 Conclusions 27 REFERENCES FOR SECTION 2.3 28 Table 2.3-1 WIND VELOCITY | |||
==SUMMARY== | |||
GINNA SITE TOWER, 50 FT. | |||
TOWER (FEBRUARY 1965 - JANUARY 1967, INCLUSIVE) 30 Table 2.3-2 WIND VELOCITY | |||
==SUMMARY== | |||
GINNA SITE TOWER, 150 FT. | |||
TOWER (FEBRUARY 1965 - JANUARY 1967, INCLUSIVE) 31 Table 2.3-3 WIND VELOCITY | |||
==SUMMARY== | |||
GINNA SITE TOWER, 250 FT. | |||
TOWER (FEBRUARY 1965 - JANUARY 1967, INCLUSIVE) 32 Table 2.3-4 WIND VELOCITY | |||
==SUMMARY== | |||
(HOURS) ROCHESTER AIRPORT FIVE YEARS 33 Table 2.3-5 WIND VELOCITY | |||
==SUMMARY== | |||
(HOURS) DURING PRECIPITA TION ROCHESTER AIRPORT 34 Table 2.3-6 WIND VELOCITY | |||
==SUMMARY== | |||
(HOURS) ROCHESTER COAST GUARD STATION (1951 - 1955) 35 Table 2.3-7 | |||
==SUMMARY== | |||
OF METEOROLOGICAL DATA GINNA SITE 36 Table 2.3-8a JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 37 Table 2.3-8b JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 38 Table 2.3-8c JOINT FREQUENCY TABLES OF WIND SPEED AND DIREC TION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 39 Table 2.3-8d JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 40 Table 2.3-8e JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 41 Table 2.3-8f JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 42 Page 2 of 10 Revision 26 5/2016 | |||
GINNA/UFSAR Table 2.3-8g JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 43 Table 2.3-8h JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 44 Table 2.3-8i JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 45 Table 2.3-8j JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 46 Table 2.3-8k JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 47 Table 2.3-8l JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 48 Table 2.3-8m JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION F ROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 49 Table 2.3-8n JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 50 Table 2.3-9a JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 51 Table 2.3-9b JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEM PERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 52 Table 2.3-9c JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 53 Table 2.3-9d JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 54 Page 3 of 10 Revision 26 5/2016 | |||
GINNA/UFSAR Table 2.3-9e JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 55 Table 2.3-9f JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 56 Table 2.3-9g JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 57 Table 2.3-9h JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 58 Table 2.3-9i JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 59 Table 2.3-9j JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 60 Table 2.3-9k JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 61 Table 2.3-9l JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 62 Table 2.3-9m JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 63 Table 2.3-9n JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 64 Page 4 of 10 Revision 26 5/2016 | |||
GINNA/UFSAR Table 2.3-10 GASEOUS DISCHARGE POINTS AT THE GINNA SITE 65 Table 2.3-11 VENT DESIGN INFORMATION FOR GINNA 66 Table 2.3-12 TABULATION OF INPUT ASSUMPTIONS FOR CALCULATIONS 67 Table 2.3-13 TOPOGRAPHIC ELEVATIONS FEET (MSL) FOR GINNA SITE PLANT GRADE IS 270 FEET 68 Table 2.3-14 ANNUAL DIFFUSION AND DEPOSITION ESTIMATES FOR ALL RECEPTOR LOCATIONS, RELEASE POINT: PLANT VENTS, WAKE-SPLIT 69 Table 2.3-15 GRAZING SEASON DIFFUSION AND DEPOSITION ESTIMATES FOR LIVESTOCK RECEPTOR LOCATIONS, RELEASE POINT: | |||
PLANT VENTS, WAKE-SPLIT 70 Table 2.3-16 GRAZING SEASON DIFFUSION AND DEPOSITION ESTIMATES FOR ALL RECEPTOR LOCATIONS, RELEASE POINT: PLANT VENTS, WAKE-SPLIT 71 Table 2.3-17 ANNUAL DIFFUSION AND DEPOSITION ESTIMATES FOR ALL RECEPTOR LOCATIONS, RELEASE POINT: GROUND RELEASE IN BUILDING WAKE 72 Table 2.3-18 GRAZING SEASON DIFFUSION AND DEPOSITION ESTIMATES FOR LIVESTOCK RECEPTOR LOCATIONS, RELEASE POINT: | |||
ASSUMED GROUND RELEASE IN BUILDING WAKE 73 Table 2.3-19 GRAZING SEASON DIFFUSION AND DEPOSITION ESTIMATES FOR ALL RECEPTOR LOCATIONS, RELEASE POINT: ASSUMED GROUND RELEASE IN BUILDING WAKE 74 Table 2.3-20 EXCLUSION AREA BOUNDARY DISTANCES 75 Table 2.3-21 KRPavan DIRECTION-DEPENDENT EAB x/Q | |||
==SUMMARY== | |||
76 Table 2.3-22 KRPavan DIRECTION-DEPENDENT LPZ X/Q | |||
==SUMMARY== | |||
78 2.4 HYDROLOGIC ENGINEERING 80 2.4.1 HYDROLOGIC DESCRIPTION 80 2.4.2 FLOODS 80 Flood Design Considerations 80 Effects of Local Intense Precipitation 81 2.4.3 PROBABLE MAXIMUM FLOOD ON STREAMS AND RIVERS 81 Flood Evaluation Summary 81 Derivation of Probable Maximum Flood 82 Deleted 82 2.4.4 LAKE ONTARIO SURGE FLOODING 82 2.4.5 ICE EFFECTS 82 2.4.6 COOLING WATER CANALS AND RESERVOIRS 83 2.4.7 FLOODING PROTECTION REQUIREMENTS 83 2.4.8 LOW WATER CONSIDERATIONS 83 2.4.9 DISPERSION, DILUTION, AND TRAVEL TIMES OF RELEASES OF LIQUID EFFLUENTS IN SURFACE WATERS 84 Page 5 of 10 Revision 26 5/2016 | |||
GINNA/UFSAR Near-Shore Lake Currents 84 Dispersion of Regulated Radioactive Liquid Releases 84 Regulated Radioactive Liquid Releases 84 Liquid Dispersion 85 Effect of Local Recirculation 86 Concentration of Nearest Public Water Supply Intake 87 Environmental Monitoring Program 87 Dispersion of Accidental Radioactive Liquid Releases 87 Accidental Releases to the Lake 87 Accidental Spills on the Ground 88 2.4.10 DISPERSION, DILUTION, AND TRAVEL TIMES OF RELEASES OF LIQUID EFFLUENTS IN SURFACE WATERS 89 Design-Basis Groundwater Level 89 Groundwater Protection Program 89 Water Use 90 REFERENCES FOR SECTION 2.4 92 Table 2.4-1 DEER CREEK OVERFLOW | |||
==SUMMARY== | |||
TABLE 94 Table 2.4-2 INDUSTRIAL AND MUNICIPAL WATER SUPPLIES 95 2.5 GEOLOGY, SEISMOLOGY, AND GEOTECHNICAL ENGINEER ING 97 2.5.1 BASIC GEOLOGIC AND SEISMIC INFORMATION 97 Regional Geology 97 Site Geology 98 2.5.2 VIBRATORY GROUND MOTION 99 Seismicity 99 Maximum Earthquake Potential 99 Surface Faulting 100 Nearby Regional Faulting 100 Ginna Site Vicinity Faulting 101 Ginna Excavation 102 2.5.3 STABILITY OF SLOPES 102 General 102 Onsite Slopes 103 Stability Analyses 103 Failure Evaluation 103 REFERENCES FOR SECTION 2.5 104 Table 2.5-1 EARTHQUAKE ACTIVITY NEAR ATTICA, NEW YORK 106 Table 2.5-2 MATERIAL PROPERTIES USED IN THE NRC STAFF ANALYSIS 107 OF SLOPE STABILITY Page 6 of 10 Revision 26 5/2016 | |||
GINNA/UFSAR FIGURES Figure 2.1-1 Location of the R. E. Ginna Nuclear Power Plant Figure 2.1-2 R. E. Ginna Site Figure 2.1-3 Figure Deleted Figure 2.1-4 Projection of Population Distribution 0-5 Miles Figure 2.1-5 1980 Population Estimates 0-5 Miles Figure 2.1-5a 1992 Population Estimates 0-10 Miles Figure 2.1-6 Projection of Population Distribution 0-40 Miles Figure 2.1-7 Location of Ginna Site Figure 2.1-8 Population Centers Over 2000 Figure 2.3-1 Climate of the Ginna Site Region Figure 2.3-2 Wind Direction Patterns Long Period Averages Figure 2.3-3 Sensor Placements, Primary Meteorological Tower Figure 2.3-4 Sensor Placements, Backup (Substation 13A) Meteorological Tower Figure 2.3-5 Ginna 1966, 50-Ft Wind Rose Figure 2.3-6 Ginna 1967, 50-Ft Wind Rose Figure 2.3-7 Ginna 1973-1974, 50-Ft Wind Rose Figure 2.3-8 Ginna 1975, 33-Ft Wind Rose Figure 2.3-9 Site Plan - Activity Release Points and Elevation Figure 2.4-1 Lake Ontario Levels Figure 2.4-2 General North-South Cross Section Ginna Site Figure 2.4-3 FIGURE DELETED Figure 2.4 Ginna Site Layout and Topography Figure 2.5-1 Plot Plan Showing Boring Locations Figure 2.5-2 Epicentral Location Map Figure 2.5-3 NRC Systematic Evaluation Program, Site Specific Spectrum, Ginna Site (5% Damping) | |||
Page 7 of 10 Revision 26 5/2016 | |||
GINNA/UFSAR APPENDIX 2A PROBABLE MAXIMUM FLOOD AND LOW WATER CONDITIONS 1 2A.1 ESTIMATE OF WAVE RUNUP ON VERTICAL PLANT WALL ROBERT EMMETT GINNA NUCLEAR POWER PLANT ROCHESTER, NEW YORK 3 2A.1.1 GENERAL 3 2A.1.2 DISCUSSION OF FACTORS 3 2A.1.3 ANALYSIS 4 2A.1.4 DISCUSSION 4 2A. | |||
==1.5 CONCLUSION== | |||
S 4 REFERENCES 6 2A.2 MAXIMUM PROBABLE WATER LEVELS IN LAKE ONTARIO AT THE ROBERT EMMETT GINNA NUCLEAR POWER PLANT SITE 7 2A. | |||
==2.1 INTRODUCTION== | |||
7 2A.3 ANALYSES OF ALTERNATIVE HYDROLOGIC AND METEOROLOGIC CRITERIA 8 A. LAKE ONTARIO REGULATION 8 B. RAINFALL 10 C. EXTRATROPICAL CYCLONES 11 D. TROPICAL CYCLONES 12 2A.4 DESIGN STORM ANALYSIS FOR MAXIMUM PROBABLE WATER LEVEL 13 2A.4.1 EXTRATROPICAL STORM ANALYSIS 13 2A.4.2 TROPICAL STORM ANALYSIS 14 2A.4.3 EXTREME LOW WATER LEVEL 16 CONCLUSIONS 16 BIBLIOGRAPHY REFERENCES 18 APPENDIX 2A FIGURES Figure 1 Lake Ontario Bottom Profile Figure Exhibit 1 A typical northeaster with a single center and a single frontal system, occurring 0100 EST, April 2, 1958, four hours before a peak surge of 3.5 ft. at Boston, Mass. | |||
Figure Exhibit 2 A complex northeaster with two low centers and a double frontal system, occurring 1900 EST, March 1, 1914, three hours before a peak surge of 4.1 ft. at Portland, Maine. | |||
Figure Exhibit 3 Lake Ontario Stage Data Page 8 of 10 Revision 26 5/2016 | |||
GINNA/UFSAR APPENDIX 2B DRIFT AND DISPERSION CHARACTERISTICS OF LAKE ONTARIO NEARSHORE WATERS ROCHESTER, NEW YORK TO SODUS BAY, NEW YORK 20 2B.1 DRIFT AND DISPERSION CHARACTERISTICS OF LAKE ONTARIO NEARSHORE WATERS ROCHESTER, NEW YORK TO SODUS BAY, NEW YORK 21 2B.1.1 Summary 21 2B.1.2 Introduction 22 2B.1.3 General comment on effect of discharges on Lake Ontario 23 2B.1.4 Currents in Lake Ontario 23 2B.1.5 Tracer releases 25 2B.1.6 Discussion 30 2B.1.7 Point source 31 2B.1.8 Line source 31 2B.1.9 Diffuser source 31 2B.1.10 Jet source 32 Table 1 Temperature (C) at a station 6000 feet offshore of Brookwood 34 APPENDIX A to Observed Tracer Distributions (Parts per Billion) 35 APPENDIX 2B APPENDIX B to Wind Speed and Direction Observations Meteorological Tower on APPENDIX 2B Brookwood Site Anemometer - Elevation 150 Feet 36 Table 1 37 APPENDIX 2B THE EFFECT ON LAKE DILUTION OF MOMENT MIXING 47 ATTACHMENT I Table 1 52 Table 2 53 APPENDIX 2B FIGURES Figure 1 Upper - Lake Current, Lower - Wind Speed Figure 2 Figure 3 Figure 4 Maximum concentration versus distance, May 1965 East Distribution Figure 5 Maximum concentration versus distance, July 1965 East Distribution Figure 6 Maximum concentration versus distance, October 1965 East Distribution Figure 7 Maximum concentration versus distance, May 1965 West Distribution Figure 8 Maximum concentration versus distance, October 1965 West Distribution Figure 9 Figure 10 Probable temperature elevation (degrees F) in upper six feet of Lake Ontario for horizontal canal discharge at the Brookwood site. | |||
Page 9 of 10 Revision 26 5/2016 | |||
GINNA/UFSAR APPENDIX A to APPENDIX 2B FIGURES Figure 1 Figure 2 Figure 3 APPENDIX 2B ATTACHMENT I Figure 3 APPENDIX 2C REPORT, SUPPLEMENTARY FOUNDATION STUDIES, 56 PROPOSED BROOKWOOD NUCLEAR POWER PLANT (R. E. | |||
GINNA NUCLEAR POWER PLANT), ONTARIO, NEW YORK 2C.1 INTRODUCTION 58 2C.1.1 GENERAL 58 2C.1.2 PURPOSE 58 2C.1.3 SCOPE OF WORK 58 2C.1.4 SITE CONDITIONS 58 2C.2 DISCUSSION AND RECOMMENDATIONS 60 2C.2.1 GENERAL 60 2C.2.2 FOUNDATION INSTALLATION PROCEDURES 60 2C.2.3 FOUNDATION DESIGN CRITERIA 61 2C.2.4 TURBINE-GENERATOR FOUNDATION 61 2C.3 Report Appendix - Field Explorations and Laboratory Tests 63 2C.3.1 FIELD EXPLORATIONS 63 2C.3.2 LABORATORY TESTS 63 Table 1 | |||
==SUMMARY== | |||
OF SOIL STRENGTH TEST DATA 65 APPENDIX 2C FIGURES Figure 1 Plate 1 - Plot Plan Figure 2 Plate 2 - Foundation Design Data Figure 3 Soil Sampler Type U Figure 4 Methods of Performing Unconfined Compression and Triaxial Compression Tests Figure 5 Plate A-1A - Log of Borings (Borings 201 through 202) | |||
Figure 6 Plate A-1B - Log of Borings (Borings 203 through 207) | |||
Figure 7 Plate A Unified Soil Classification System and Key to Test Data Page 10 of 10 Revision 26 5/2016 | |||
GINNA/UFSAR 3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA 2 3.1.1 ATOMIC INDUSTRIAL FORUM DESIGN CRITERIA 2 3.1.1.1 Overall Plant Requirements 2 3.1.1.1.1 Quality Standards 2 3.1.1.1.2 Performance Standards 4 3.1.1.1.3 Fire Protection 5 3.1.1.1.4 Sharing of Systems 5 3.1.1.1.5 Records Requirements 5 3.1.1.2 Protection by Multiple Fission Product Barriers 6 3.1.1.2.1 Reactor Core Design 6 3.1.1.2.2 Suppression of Power Oscillations 7 3.1.1.2.3 Overall Power Coefficient 7 3.1.1.2.4 Reactor Coolant Pressure Boundary 7 3.1.1.2.5 Reactor Containment 8 3.1.1.3 Nuclear and Radiation Controls 9 3.1.1.3.1 Control Room 9 3.1.1.3.2 Instrumentation and Controls Systems 9 3.1.1.3.3 Fission Process Monitors and Controls 10 3.1.1.3.4 Core Protection Systems 11 3.1.1.3.5 Engineered Safety Features Protection Systems 11 3.1.1.3.6 Monitoring Reactor Coolant Leakage 12 3.1.1.3.7 Monitoring Radioactivity Releases 13 3.1.1.3.8 Monitoring Fuel and Waste Storage 13 3.1.1.4 Reliability and Testability of Protection Systems 14 3.1.1.4.1 Protection Systems Reliability 14 3.1.1.4.2 Protection Systems Redundancy and Independence 15 3.1.1.4.2.1 Reactor Trip Circuits 15 3.1.1.4.2.2 Engineered Safety Features Initiation Circuits 15 3.1.1.4.3 Single-Failure Definition (Category B) 16 3.1.1.4.4 Separation of Protection and Control Instrumentation Systems 16 3.1.1.4.5 Protection Against Multiple Disability for Protection Systems 16 3.1.1.4.6 Emergency Power for Protection Systems 16 3.1.1.4.7 Demonstration of Functional Operability of Protection Systems 17 3.1.1.4.8 Protection Systems Failure Analysis Design 17 Page 1 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.1.1.5 Reactivity Control 18 3.1.1.5.1 Redundancy of Reactivity Control 18 3.1.1.5.2 Reactivity Hot Shutdown Capability 18 3.1.1.5.3 Reactivity Shutdown Capability 18 3.1.1.5.4 Reactivity Hold-Down Capability 19 3.1.1.5.5 Reactivity Control Systems Malfunction 19 3.1.1.5.6 Maximum Reactivity Worth of Control Rods 20 3.1.1.6 Reactor Coolant Pressure Boundary 20 3.1.1.6.1 Reactor Coolant Pressure Boundary Capability 20 3.1.1.6.2 Reactor Coolant Pressure Boundary Rapid Propagation Failure Preven- 21 tion 3.1.1.6.3 Reactor Coolant Pressure Boundary Brittle Fracture Prevention 22 3.1.1.6.4 Reactor Coolant Pressure Boundary Surveillance 22 3.1.1.7 Engineered Safety Features 23 3.1.1.7.1 Engineered Safety Features Basis for Design 23 3.1.1.7.2 Reliability and Testability of Engineered Safety Features 24 3.1.1.7.3 Emergency Power 24 3.1.1.7.4 Missile Protection 25 3.1.1.7.5 Engineered Safety Features Performance Capability 26 3.1.1.7.6 Engineered Safety Features Components Capability 26 3.1.1.7.7 Accident Aggravation Prevention 27 3.1.1.7.8 Emergency Core Cooling System (ECCS) Capability 27 3.1.1.7.9 Inspection of Emergency Core Cooling System (ECCS) 28 3.1.1.7.10 Testing of Emergency Core Cooling System (ECCS) Components 28 3.1.1.7.11 Testing of Emergency Core Cooling System (ECCS) 28 3.1.1.7.12 Testing of Operational Sequence of Emergency Core Cooling System 28 (ECCS) 3.1.1.7.13 Containment Design Basis 29 3.1.1.7.14 Nil Ductility Transition Temperature Requirement for Containment 29 Material 3.1.1.7.15 Reactor Coolant Pressure Boundary Outside Containment 30 3.1.1.7.16 Containment Heat Removal Systems 30 3.1.1.7.17 Containment Isolation Valves 30 3.1.1.7.18 Initial Leakage Rate Testing of Containment 30 3.1.1.7.19 Periodic Containment Leakage Rate Testing 31 3.1.1.7.20 Provisions for Testing of Penetrations 31 3.1.1.7.21 Provisions for Testing of Isolation Valves 31 Page 2 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.1.1.7.22 Inspection of Containment Pressure-Reducing Systems 32 3.1.1.7.23 Testing of Containment Pressure-Reducing Systems Components 32 3.1.1.7.24 Testing of Containment Spray Systems 32 3.1.1.7.25 Testing of Operational Sequence of Containment Pressure-Reducing 32 Systems 3.1.1.7.26 Inspection of Air Cleanup Systems 33 3.1.1.7.27 Testing of Air Cleanup Systems Components 33 3.1.1.7.28 Testing Air Cleanup System 33 3.1.1.7.29 Testing of Operational Sequence of Air Cleanup Systems 33 3.1.1.8 Fuel and Waste Storage Systems 34 3.1.1.8.1 Prevention of Fuel Storage Criticality 34 3.1.1.8.2 Fuel and Waste Storage Decay Heat 34 3.1.1.8.3 Fuel and Waste Storage Radiation Shielding 35 3.1.1.8.4 Protection Against Radioactivity Release From Spent Fuel and Waste 35 Storage 3.1.1.9 Control of Releases of Radioactivity to the Environment 35 3.1.2 GENERAL DESIGN CRITERIA 36 3.1.2.1 Overall Requirements 36 3.1.2.1.1 General Design Criterion 1 - Quality Standards and Records 37 3.1.2.1.2 General Design Criterion 2 - Design Bases for Protection Against Nat- 38 ural Phenomena 3.1.2.1.3 General Design Criterion 3 - Fire Protection 38 3.1.2.1.4 General Design Criterion 4 - Environmental and Missile Design Bases 39 3.1.2.1.5 General Design Criterion 5 - Sharing of Structures, Systems, and Com- 39 ponents 3.1.2.2 Protection by Multiple Fission Product Barriers 39 3.1.2.2.1 General Design Criterion 10 - Reactor Design 39 3.1.2.2.2 General Design Criterion 11 - Reactor Inherent Protection 40 3.1.2.2.3 General Design Criterion 12 - Suppression of Reactor Power Oscilla- 40 tions 3.1.2.2.4 General Design Criterion 13 - Instrumentation and Control 40 3.1.2.2.5 General Design Criterion 14 - Reactor Coolant Pressure Boundary 41 3.1.2.2.6 General Design Criterion 15 - Reactor Coolant System Design 41 3.1.2.2.7 General Design Criterion 16 - Containment Design 42 3.1.2.2.8 General Design Criterion 17 - Electrical Power Systems 42 3.1.2.2.9 General Design Criterion 18 - Inspection and Testing of Electrical 44 Power Systems 3.1.2.2.10 General Design Criterion 19 - Control Room 44 Page 3 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.1.2.3 Protection and Reactivity Control Systems 45 3.1.2.3.1 General Design Criterion 20 - Protection Systems Functions 45 3.1.2.3.2 General Design Criterion 21 - Protection System Reliability and Test- 45 ability 3.1.2.3.3 General Design Criterion 22 - Protection System Independence 46 3.1.2.3.4 General Design Criterion 23 - Protection System Failure Modes 46 3.1.2.3.5 General Design Criterion 24 - Separation of Protection and Control 47 Systems 3.1.2.3.6 General Design Criterion 25 - Protection System Requirements for 47 Reactivity Control Malfunctions 3.1.2.3.7 General Design Criterion 26 - Reactivity Control System Redundancy 48 and Capability 3.1.2.3.8 General Design Criterion 27 - Combined Reactivity Control System 48 Capability 3.1.2.3.9 General Design Criterion 28 - Reactivity Limits 49 3.1.2.3.10 General Design Criterion 29 - Protection Against Anticipated Opera- 49 tional Occurrences 3.1.2.4 Fluid Systems 49 3.1.2.4.1 General Design Criterion 30 - Quality of Reactor Coolant Pressure 49 Boundary 3.1.2.4.2 General Design Criterion 31 - Fracture Prevention of Reactor Coolant 50 Pressure Boundary 3.1.2.4.3 General Design Criterion 32 - Inspection of Reactor Coolant Pressure 51 Boundary 3.1.2.4.4 General Design Criterion 33 - Reactor Coolant Makeup 51 3.1.2.4.5 General Design Criterion 34 - Residual Heat Removal 52 3.1.2.4.6 General Design Criterion 35 - Emergency Core Cooling 52 3.1.2.4.7 General Design Criterion 36 - Inspection of Emergency Core Cooling 53 System (ECCS) 3.1.2.4.8 General Design Criterion 37 - Testing of Emergency Core Cooling 53 Systems (ECCS) 3.1.2.4.9 General Design Criterion 38 - Containment Heat Removal 54 3.1.2.4.10 General Design Criterion 39 - Inspection of Containment Heat 54 Removal System 3.1.2.4.11 General Design Criterion 40 - Testing of Containment Heat Removal 54 System 3.1.2.4.12 General Design Criterion 41 - Containment Atmosphere Cleanup 55 3.1.2.4.13 General Design Criterion 42 - Inspection of Containment Atmosphere 56 Cleanup Systems 3.1.2.4.14 General Design Criterion 43 - Testing of Containment Atmosphere 56 Cleanup Systems 3.1.2.4.15 General Design Criterion 44 - Cooling Water 57 Page 4 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.1.2.4.16 General Design Criterion 45 - Inspection of Cooling Water System 57 3.1.2.4.17 General Design Criterion 46 - Testing of Cooling Water System 58 3.1.2.5 Reactor Containment 58 3.1.2.5.1 General Design Criterion 50 - Containment Design Basis 58 3.1.2.5.2 General Design Criterion 51 - Fracture Prevention of Containment 59 Pressure Boundary 3.1.2.5.3 General Design Criterion 52 - Capability for Containment Leakage 59 Rate Testing 3.1.2.5.4 General Design Criterion 53 - Provisions for Containment Testing and 59 Inspection 3.1.2.5.5 General Design Criterion 54 - Piping Systems Penetrating Containment 60 3.1.2.5.6 General Design Criterion 55 - Reactor Coolant Pressure Boundary Pen- 60 etrating Containment 3.1.2.5.7 General Design Criterion 56 - Primary Containment Isolation 61 3.1.2.5.8 General Design Criterion 57 - Closed System Isolation Valves 62 3.1.2.6 Fuel and Radioactivity Control 62 3.1.2.6.1 General Design Criterion 60 - Control of Releases of Radioactive 62 Materials to the Environment 3.1.2.6.2 General Design Criterion 61 - Fuel Storage and Handling and Radioac- 62 tivity Control 3.1.2.6.3 General Design Criterion 62 - Prevention of Criticality in Fuel Storage 63 and Handling 3.1.2.6.4 General Design Criterion 63 - Monitoring Fuel and Waste Storage 63 3.1.2.6.5 General Design Criterion 64 - Monitoring Radioactivity Releases 64 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYS- 66 TEMS 3. | |||
==2.1 INTRODUCTION== | |||
66 3.2.2 SYSTEMATIC EVALUATION PROGRAM EVALUATION 66 3.2.2.1 Fracture Toughness 67 3.2.2.1.1 Pressurizer 67 3.2.2.1.2 Accumulators 68 3.2.2.1.3 Component Cooling Water (CCW) Pumps 68 3.2.2.1.4 Service Water Pumps 68 3.2.2.1.5 Main Steam Piping and Valves 69 3.2.2.1.6 Feedwater Piping and Valves 69 3.2.2.2 Radiography Requirements 69 3.2.2.2.1 Class 2 Pressure Vessels 69 3.2.2.2.2 Class 1 and 2 Welded Joints 70 3.2.2.2.3 Main Steam and Feedwater Piping 70 Page 5 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.2.2.3 Valve Design 71 3.2.2.4 Pump Design 71 3.2.2.5 Storage Tank Design 72 Table 3.2-1 CLASSIFICATION OF STRUCTURES, SYSTEMS, AND 74 COMPONENTS 3.3 WIND AND TORNADO LOADINGS 84 3. | |||
==3.1 INTRODUCTION== | |||
84 3.3.2 STRUCTURAL UPGRADE PROGRAM EVALUATION 84 3.3.2.1 Structural Evaluation Approach 84 3.3.2.1.1 Requirements 84 3.3.2.1.2 Structural Evaluation Process 84 3.3.2.1.3 Structural Evaluation Computer Program 85 3.3.2.1.4 Input Load Criteria 85 3.3.2.1.5 General Assumptions 86 3.3.2.1.6 Load Combinations and Acceptance Criteria 87 3.3.2.2 Structural Evaluation 88 3.3.2.2.1 Primary Member Evaluation 88 3.3.2.2.2 Secondary Member Evaluation 89 3.3.2.2.3 Connections and Anchorages Evaluation 89 3.3.2.2.4 Exterior Shell Evaluation 90 3.3.2.2.4.1 Siding 90 3.3.2.2.4.2 Concrete Masonry Block Walls 90 3.3.2.2.4.3 Architectural Items 91 3.3.2.3 Results of the Structural Evaluation 91 3.3.2.3.1 Primary Members 91 3.3.2.3.1.1 General 91 3.3.2.3.1.2 Severe Environmental Conditions 91 3.3.2.3.1.3 Extreme Snow Load Condition 92 3.3.2.3.1.4 132-mph Tornado 92 3.3.2.3.1.5 188-mph Tornado 92 3.3.2.3.1.6 250-mph Tornado 92 3.3.2.3.2 Secondary Members 93 3.3.2.3.3 Connections and Anchorages 93 3.3.2.3.4 Exterior Shell 94 3.3.2.3.4.1 Metal Siding 94 3.3.2.3.4.2 Roof Decking 94 Page 6 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.3.2.3.4.3 Block Walls 94 3.3.3 TORNADO MISSILES AND SAFE SHUTDOWN APPROACH 94 3.3.3.1 Background 94 3.3.3.2 Shutdown Methodology 95 3.3.3.2.1 Assumptions 95 3.3.3.2.2 Shutdown Details 95 3.3.3.3 Required Components 96 3.3.3.3.1 Refueling Water Storage Tank (RWST) 96 3.3.3.3.2 Electrical Buses 14, 17, and 18 96 3.3.3.3.3 Main Steam Lines A and B, and Main Feedwater Lines A and B 97 3.3.3.3.3.1 Results - Steel Rod 97 3.3.3.3.3.2 Results - Utility Pole 97 3.3.3.3.3.3 Failure of Block Walls 97 3.3.3.3.4 Surface of the Spent Fuel Pool 98 3.3.3.3.5 Diesel Generators and Their Fuel Supply 98 3.3.3.3.6 Relay Room 98 3.3.3.3.7 Service Water System 99 3.3.3.3.8 Standby Auxiliary Feedwater System 99 3.3.3.3.9 Instrumentation 99 3.3.3.3.10 Cable Tunnel 100 3.3.4 DESIGN TORNADO 100 3.3.4.1 Introduction 100 3.3.4.2 Safety Assessment 100 3.3.4.3 Reserve Plant Capacity 101 3.3.4.4 System Reserve Capacity 102 3.3.5 STRUCTURAL UPGRADE PROGRAM 103 3.3.5.1 Introduction 103 3.3.5.2 Criteria Changes 103 3.3.5.2.1 First Stage Review 103 3.3.5.2.2 Second Stage Review 104 3.3.5.3 Stability Evaluation 105 3.3.5.3.1 Primary Members 105 3.3.5.3.2 Connections and Anchorages 105 3.3.5.4 NRC Technical Evaluation Report (SEP Topic III-2) Open Items 106 3.3.5.4.1 Effective Tornado Loadings 106 3.3.5.4.2 Structural Loadings 107 Page 7 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.3.5.4.3 Structural Acceptance Criteria 107 3.3.5.4.4 Structural Systems 107 3.3.5.5 SEP Topic III-7.B, Loads, Load Combinations, and Design Criteria 108 3.3.5.6 Diesel Generator Component Operability 109 3.3.5.7 Conclusions 109 3.3.6 INTERMEDIATE BUILDING BLOCK WALL REINFORCEMENT 110 Table 3.3-1 PRIMARY MEMBER FAILURES PER LOADING COMBINATION 113 3.4 WATER LEVEL (FLOOD) DESIGN 114 3.4.1 FLOOD PROTECTION 114 3.4.1.1 Flood Protection Measures for Seismic Category I Structures 114 3.4.1.1.1 Introduction 114 3.4.1.1.2 Lake Ontario Flood Protection 114 3.4.1.1.3 Deer Creek Flood Protection 114 3.4.1.2 Permanent Dewatering System 115 3.4.2 FLOODING DUE TO FAILURE OF TANKS 115 3.4.3 ROOF DRAINAGE 116 3.5 MISSILE PROTECTION 119 3.5.1 INTERNALLY GENERATED MISSILES 119 3.5.1.1 Introduction 119 3.5.1.1.1 Design Criteria 119 3.5.1.1.2 Systematic Evaluation Program 119 3.5.1.2 Turbine Missiles 120 3.5.1.2.1 Introduction 120 3.5.1.2.2 Turbine Inspection Program 121 3.5.1.2.3 Systematic Evaluation Program Topic III-4 121 3.5.1.3 Effects of Internally Generated Missiles on Systems and Equipment 122 3.5.1.3.1 Systems Needed to Perform Safety Functions 122 3.5.1.3.1.1 Reactor Coolant System 122 3.5.1.3.1.2 Emergency Core Cooling System (ECCS) 123 3.5.1.3.1.3 Containment Heat Removal and Atmosphere Cleanup Systems 124 3.5.1.3.1.4 Chemical and Volume Control System 125 3.5.1.3.1.5 Residual Heat Removal System 126 3.5.1.3.1.6 Component Cooling Water System 126 3.5.1.3.1.7 Service Water System 126 3.5.1.3.1.8 Diesel-Generator Auxiliary Systems 127 Page 8 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.5.1.3.1.9 Main Steam System 127 3.5.1.3.1.10 Feedwater and Condensate Systems 128 3.5.1.3.1.11 Preferred Auxiliary Feedwater System 128 3.5.1.3.1.12 Standby Auxiliary Feedwater System (SAFW) 128 3.5.1.3.1.13 Ventilation Systems for Vital Areas 129 3.5.1.3.1.14 Combustible Gas Control System 129 3.5.1.3.2 Systems Whose Failure May Result in Activity Release 129 3.5.1.3.2.1 Spent Fuel Pool Cooling System 129 3.5.1.3.2.2 Sampling System 130 3.5.1.3.2.3 Waste Disposal System 130 3.5.1.3.2.4 Containment Shutdown Purge System 130 3.5.1.3.2.5 Instrument and Service Air Systems 130 3.5.1.3.3 Electrical Systems 131 3.5.1.3.3.1 Diesel Generators 131 3.5.1.3.3.2 Station Batteries 131 3.5.1.3.3.3 480-Volt Switchgear 131 3.5.1.3.3.4 Control Room 131 3.5.1.3.3.5 Cable Spreading/Relay Room 131 3.5.2 EXTERNALLY GENERATED MISSILES 132 3.5.2.1 Tornado Missiles 132 3.5.2.2 Site Proximity Missiles 132 3.5.2.2.1 Design Criteria 132 3.5.2.2.2 Nearby Hazardous Activities 132 3.5.2.2.3 Aircraft Hazards 133 3.6 PROTECTION AGAINST THE DYNAMIC EFFECTS ASSOCI- 135 ATED WITH THE POSTULATED RUPTURE OF PIPING 3.6.1 POSTULATED PIPING FAILURES IN FLUID SYSTEMS INSIDE 135 CONTAINMENT 3.6.1.1 Evaluation Procedure 135 3.6.1.1.1 Pipe Selection 135 3.6.1.1.2 Effects-Oriented Evaluation 136 3.6.1.1.3 Mechanistic Evaluation 136 3.6.1.2 Required Equipment 137 3.6.1.3 Safety Analysis 137 3.6.1.3.1 Single-Failure Considerations 137 3.6.1.3.1.1 Introduction 137 3.6.1.3.1.2 Containment Fan Coolers 138 Page 9 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.6.1.3.1.3 Low-Pressure Safety Injection Isolation Valves 138 3.6.1.3.2 High-Energy Line Break Effects 138 3.6.1.3.2.1 Introduction 138 3.6.1.3.2.2 Alternate Charging 139 3.6.1.3.2.3 Residual Heat Removal Pump Suction 139 3.6.1.3.2.4 Reactor Coolant Pump Seal-Water to Seals 140 3.6.1.3.2.5 Letdown Line 140 3.6.1.3.2.6 Charging Line 141 3.6.1.3.2.7 Steam Generator Blowdown Lines 142 3.6.1.3.2.8 Main Steam and Feedwater Lines 142 3.6.1.3.2.9 Residual Heat Removal Pump Discharge Line 145 3.6.1.3.2.10 Standby Auxiliary Feedwater Lines 145 3.6.1.3.2.11 Accumulator Lines and Branch Lines 145 3.6.1.3.2.12 Auxiliary Spray Line 148 3.6.1.3.2.13 Reactor Coolant System 149 3.6.1.3.2.14 Pressurizer Surge Line 149 3.6.1.3.2.15 Pressurizer Spray Lines 152 3.6.1.3.2.16 Pressurizer Safety and Relief Lines 152 3.6.2 POSTULATED PIPING FAILURES IN FLUID SYSTEMS OUTSIDE 153 CONTAINMENT 3.6.2.1 Introduction and Summary 153 3.6.2.1.1 Initial Evaluation 153 3.6.2.1.2 Systematic Evaluation Program Reevaluation 154 3.6.2.2 Evaluation Procedure 155 3.6.2.2.1 Initial Evaluation 155 3.6.2.2.2 Systematic Evaluation Program Reevaluation 156 3.6.2.3 Analysis Criteria 157 3.6.2.3.1 December 18, 1972, AEC Letter Evaluation Criteria 157 3.6.2.3.2 Systematic Evaluation Program Criteria 157 3.6.2.3.2.1 High-Energy Fluid Systems Piping 157 3.6.2.3.2.2 Moderate-Energy Fluid System Piping 159 3.6.2.3.2.3 Type of Breaks and Leakage Cracks in Fluid System Piping 160 3.4.1.1.1.1 Assumptions 161 3.4.1.1.1.2 Effects of Piping Failure 162 3.4.1.2 Analysis in Response to December 18, 1972, AEC Letter 162 Page 10 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.4.1.2.1 Rupture Load Analysis 162 3.4.1.2.2 Main Steam System Load Analysis 163 3.4.1.2.3 Feedwater System Load Analysis 163 3.4.1.2.4 Jet Impingement Load Analysis 163 3.4.1.2.5 Pipe Whip Analysis for Main Steam and Feedwater Piping 164 3.4.1.2.5.1 Analytical Methods 164 3.4.1.2.5.2 Results of Analysis 164 3.4.1.2.6 Blowdown Analysis 165 3.4.1.2.6.1 Main Steam Blowdown Analysis 165 3.4.1.2.6.2 Feedwater Blowdown Analysis 165 3.4.1.2.7 Compartment Pressurization Analysis 166 3.4.1.2.7.1 Main Steam Line Ruptures 166 3.4.1.2.7.2 Building Pressurization for a Branch Line Rupture 166 3.4.1.2.8 Flooding Analysis 166 3.4.1.2.8.1 Intermediate Building Flooding 166 3.4.1.2.8.2 Screen House and Turbine Building Flooding 167 3.4.1.3 Systematic Evaluation Program Analysis 167 3.4.1.3.1 Zone Reevaluation Performed as Part of the Systematic Evaluation 167 Program Review 3.4.1.3.1.1 Screen House 167 3.4.1.3.1.2 Intermediate Building 168 3.4.1.3.1.3 Turbine Building Main Steam and Main Feedwater Line Breaks 169 3.4.1.3.1.4 Structural Analysis of the Turbine Building for Pressurization 170 3.4.1.3.1.5 Battery Room/Mechanical Equipment Room Flooding 172 3.4.1.3.1.6 Auxiliary Feedwater Line Breaks on the 253-Ft Elevation of the Inter- 172 mediate Building 3.4.1.3.1.7 Relay Room and Air Handling Room 172 3.4.1.3.1.8 Auxiliary Building 173 3.4.1.3.2 Main Steam Safety and Relief Valves 174 3.4.1.3.2.1 Pipe Failures in the Intermediate Building 174 3.6.2.5.2.2 Pipe Failures in the Turbine Building 175 3.6.2.5.2.3 Decay Heat Removal Following Blowdown from Both Steam Genera- 176 tors 3.6.2.5.2.4 Conclusions 177 Table 3.6-1 LINES PENETRATING CONTAINMENT WHICH NORMALLY OR 181 OCCASIONALLY EXPERIENCE HIGH-ENERGY SERVICE CON-DITIONS Page 11 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR Table 3.6-2 LINES INSIDE CONTAINMENT BUT NOT PENETRATING CON- 183 TAINMENT WHICH NORMALLY OR OCCASIONALLY EXPERI-ENCE HIGH-ENERGY SERVICE CONDITIONS Table 3.6-3 CONTAINMENT PIPE DATA 184 3.7 SEISMIC DESIGN 186 3.7.1 SEISMIC INPUT 186 3.7.1.1 Introduction 186 3.7.1.1.1 Original Seismic Classification 186 3.7.1.1.2 Seismic Reevaluation 187 3.7.1.1.2.1 Scope of Reevaluation 187 3.7.1.1.2.2 Reevaluation Criteria 187 3.7.1.2 Design Response Spectra 188 3.7.1.3 Design Time-History 188 3.7.1.4 Critical Damping Values 189 3.7.1.5 Supporting Media for Seismic Category I Structures 189 3.7.2 SEISMIC SYSTEM ANALYSIS 190 3.7.2.1 Seismic Analysis Methods 190 3.7.2.1.1 Original Seismic Analysis 190 3.7.2.1.2 Seismic Reevaluation 191 3.7.2.2 Natural Frequencies and Response Loads 192 3.7.2.3 Procedure Used for Mathematical Modeling 192 3.7.2.4 Soil-Structure Interaction 192 3.7.2.5 Development of Floor Response Spectra 192 3.7.2.6 Combination of Earthquake Directional Components 193 3.7.2.7 Combination of Modal Responses 193 3.7.2.8 Interaction of Nonseismic Structures with Seismic Category I Struc- 193 tures 3.7.2.9 Use of Constant Vertical Static Factors 194 3.7.3 SEISMIC SUBSYSTEM ANALYSIS 194 3.7.3.1 Seismic Analysis Methods 194 3.7.3.1.1 Original Design 194 3.7.3.1.1.1 Piping and Tanks 194 3.7.3.1.1.2 Steam Generator 195 3.7.3.1.1.3 Control Rod Drive Mechanisms 195 3.7.3.1.1.4 Reactor Internals 195 3.7.3.1.1.5 Reactor Vessel 196 Page 12 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.7.3.1.1.6 Pressurizer 196 3.7.3.1.2 Seismic Reevaluation 197 3.7.3.2 Basis for Selection of Frequencies 198 3.7.3.3 Use of Equivalent Static Analysis 198 3.7.3.4 Three Components of Earthquake Motion 198 3.7.3.5 Combination of Modal Responses 199 3.7.3.6 Analytical Procedures for Piping 199 3.7.3.6.1 Residual Heat Removal System Line from Reactor Coolant System 199 Loop A to Containment 3.7.3.6.2 Steam Line from Steam Generator B to Containment 200 3.7.3.6.3 Pressurizer Safety and Relief Lines 200 3.7.3.6.3.1 Analytical Methods 200 3.7.3.6.3.2 Transfer Matrix Method 201 3.7.3.6.3.3 Stiffness Matrix Formulation 202 3.7.3.7 Seismic Piping Upgrade Program 203 3.7.3.7.1 Program Scope 203 3.7.3.7.2 Piping Selection Criteria 203 3.7.3.7.3 Selected Lines 204 3.7.3.7.3.1 Reactor Coolant System 204 3.7.3.7.3.2 Main Steam 204 3.7.3.7.3.3 Main Feedwater 204 3.7.3.7.3.4 Auxiliary Feedwater 204 3.7.3.7.3.5 Safety Injection 205 3.7.3.7.3.6 Residual Heat Removal 205 3.7.3.7.3.7 Containment Spray 205 3.7.3.7.3.8 Chemical and Volume Control System 206 3.7.3.7.3.9 Steam Generator Blowdown 206 3.7.3.7.3.10 Service Water System 206 3.7.3.7.3.11 Component Cooling Water 207 3.7.3.7.3.12 Standby Auxiliary Feedwater 208 3.7.3.7.4 Codes and Standards 208 3.7.3.7.5 Analytical Procedures 208 3.7.3.7.5.1 General 208 3.7.3.7.5.2 Damping Values 208 3.7.3.7.5.3 Combination of Modal Responses 209 Page 13 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.7.3.7.5.4 Safe Shutdown Earthquake Stresses 211 3.7.3.7.5.5 Small Piping Analysis 212 3.7.3.7.5.6 Branch Line Analysis 212 3.7.3.7.5.7 Piping Beyond Scope of Upgrade Program 212 3.7.3.7.6 Piping System Models 213 3.7.3.7.7 Valve Model 214 3.7.3.7.8 Equipment Model 214 3.7.3.7.9 Interaction Effects 214 3.7.3.7.10 Support Model 214 3.7.3.7.10.1 Deviations 214 3.7.3.7.10.2 Support-Welded Attachments 215 3.7.4 SEISMIC INSTRUMENTATION 216 Table 3.7-1 ORIGINAL AND CURRENT RECOMMENDED DAMPING 218 VALUES Table 3.7-2 MODAL FREQUENCIES OF THE INTERCONNECTED 219 BUILDING MODEL Table 3.7-3 EQUIPMENT AND LOCATIONS WHERE IN-STRUCTURE 221 SPECTRA WERE GENERATED FOR THE SYSTEMATIC EVALUATION PROGRAM 3.8 DESIGN OF SEISMIC CATEGORY I STRUCTURES 222 3.8.1 CONTAINMENT 222 3.8.1.1 General Description 222 3.8.1.1.1 Containment Structure 222 3.8.1.1.2 Waterproofing 223 3.8.1.1.3 Rock Anchors 223 3.8.1.1.4 Construction Sequence 223 3.8.1.1.5 Steel Reinforcement 225 3.8.1.2 Mechanical Design Bases 226 3.8.1.2.1 General 226 3.8.1.2.2 Design Loads 226 3.8.1.2.3 Design Stress Criteria 227 3.8.1.2.3.1 Limiting Loads 227 3.8.1.2.3.2 Load Factors 228 3.8.1.2.3.3 Maximum Thermal Load 228 3.8.1.2.4 Load Capacity 229 Page 14 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.8.1.2.4.1 Reinforced Concrete 229 3.8.1.2.4.2 Prestressed Concrete 230 3.8.1.2.4.3 Liner 232 3.8.1.2.4.4 Rock 233 3.8.1.2.5 Codes and Standards 233 3.8.1.2.5 Codes and Standards Steam Generator Replacement (Dome 236 Opening Repairs 3.8.1.3 Seismic Design 238 3.8.1.3.1 Initial Seismic Design 238 3.8.1.3.2 Seismic Reanalysis 239 3.8.1.4 Containment Detailed Design 239 3.8.1.4.1 Stress Analysis 239 3.8.1.4.1.1 Analysis Methods 239 3.8.1.4.1.2 Analysis Results 240 3.8.1.4.1.3 Analysis for Steam Generator Replacement Dome Openings 241 3.8.1.4.2 Rock Anchors 241 3.8.1.4.2.1 Rock Anchor Design 241 3.8.1.4.2.2 Preinstallation Grouting Test 242 3.8.1.4.2.3 Previous Applications 243 3.8.1.4.2.4 Rock Hold-Down Capacity 243 3.8.1.4.2.5 Hold-Down Factor of Safety 245 3.8.1.4.2.6 Installation 245 3.8.1.4.3 Tendons 246 3.8.1.4.3.1 General Design 246 3.8.1.4.3.2 Seismic Considerations 248 3.8.1.4.3.3 Stressing Procedure 250 3.8.1.4.3.4 Corrosion Protection 251 3.8.1.4.4 Hinge Design 253 3.8.1.4.4.1 Tension Bars 253 3.8.1.4.4.2 Liner Knuckle 255 3.8.1.4.4.3 Elastomer Bearing Pads 256 3.8.1.4.5 Concrete 258 3.8.1.4.5.1 Radial Shear 258 3.8.1.4.5.2 Longitudinal Shears 258 3.8.1.4.5.3 Horizontal Shear 259 3.8.1.4.5.4 Anchorage Stresses 260 Page 15 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.8.1.4.5.5 Shell Stress Analytical Procedures 261 3.8.1.4.6 Insulation 266 3.8.1.4.7 Liner 267 3.8.1.4.7.1 Vibrations 267 3.8.1.4.7.2 Anchorage Fatigue Analysis 267 3.8.1.4.7.3 Base Slab Liner 267 3.8.1.4.7.4 Liner Stresses 268 3.8.1.4.7.5 Liner Buckling 269 3.8.1.4.7.6 Liner Corrosion Allowance 273 3.8.1.5 Penetrations 273 3.8.1.5.1 General 273 3.8.1.5.2 Electrical Penetrations 274 3.8.1.5.3 Piping Penetrations 275 3.8.1.5.4 Access Hatch and Personnel Locks 275 3.8.1.5.5 Fuel Transfer Penetration 276 3.8.1.5.6 Typical Penetration Analysis 277 3.8.1.5.6.1 Loss-of-Coolant Accident 277 3.8.1.5.6.2 Loss-of-Coolant Accident Plus Earthquake 279 3.8.1.5.7 Penetration Reinforcement Analyzed for Pipe Rupture 280 3.8.1.6 Quality Control and Material Specifications 281 3.8.1.6.1 Concrete 281 3.8.1.6.1.1 Ultimate Compressive Strength 281 3.8.1.6.1.2 Quality Control Measures 281 3.8.1.6.1.3 Concrete Suppliers 282 3.8.1.6.1.4 Concrete Specifications 283 3.8.1.6.1.5 Admixtures 285 3.8.1.6.2 Mild Steel Reinforcement 287 3.8.1.6.3 Cadwell Splices 288 3.8.1.6.4 Radial Tension Bars 289 3.8.1.6.5 Containment Liner 289 3.8.1.6.5.1 Fabrication and Workmanship 289 3.8.1.6.5.2 Penetrations 290 3.8.1.6.5.3 Welding 290 3.8.1.6.5.4 Erection Tolerances 291 3.8.1.6.5.5 Painting 291 3.8.1.6.6 Elastomer Pads 292 Page 16 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.8.1.6.7 Tendons 292 3.8.1.6.7.1 Materials 292 3.8.1.6.7.2 Tests and Inspection 293 3.8.1.6.8 Liner Insulation 293 3.8.1.7 Testing and Inservice Inspection Requirements 294 3.8.1.7.1 Construction Phase Testing 294 3.8.1.7.1.1 Liner 294 3.8.1.7.1.2 Prestressing Tendons 295 3.8.1.7.1.3 Concrete Reinforcement 295 3.8.1.7.1.4 Concrete 296 3.8.1.7.1.5 Elastomer Bearing Pads 297 3.8.1.7.1.6 Rock Anchor Tests 298 3.8.1.7.1.7 Large Opening Reinforcements 299 3.8.1.7.1.8 Liner Insulation 299 3.8.1.7.2 General Description of the Structural Integrity Test 299 3.8.1.7.2.1 Pressurization 299 3.8.1.7.2.2 Measurements 300 3.8.1.7.2.3 Test Pressure Justification 302 3.8.1.7.2.4 Test Results 302 3.8.1.7.2.5 Containment Return to Service Testing Post 1996 Steam Generator 302 Replacement 3.8.1.7.3 Postoperational Surveillance 303 3.8.1.7.3.1 Leakage Monitoring 303 3.8.1.7.3.2 Initial Tendon Surveillance Program 303 3.8.1.7.3.3 Current Tendon Surveillance Program 304 3.8.1.7.3.4 Current Tendon Surveillance Program Results 305 3.8.1.7.3.5 Test on Rock Anchors 306 3.8.1.7.3.6 Inservice Inspection 306 3.8.2 STRUCTURAL REANALYSIS PROGRAM 307 3.8.2.1 Design Codes, Criteria, and Load Combinations - SEP Topic III-7.B 307 3.8.2.1.1 Introduction 307 3.8.2.1.1.1 Seismic Category I Structures 307 3.8.2.1.1.2 Structural Codes 308 3.8.2.1.1.3 Code Comparison 310 3.8.2.1.2 Assessment of Design Codes and Load Changes for Concrete 310 Structures 3.8.2.1.2.1 Columns With Spliced Reinforcing 310 Page 17 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.8.2.1.2.2 Brackets and Corbels (Not on the Containment Shell) 311 3.8.2.1.2.3 Elements Loaded in Shear With No Diagonal Tension (Shear Friction) 312 3.8.2.1.2.4 Structural Walls - Primary Load Carrying 313 3.8.2.1.2.5 Elements Subject to Temperature Variations 314 3.8.2.1.2.6 Areas of Containment Shell Subject to Peripheral Shear 315 3.8.2.1.2.7 Areas of Containment Shell Subject to Torsion 316 3.8.2.1.2.8 Brackets and Corbels (On the Containment Shell) 316 3.8.2.1.2.9 Areas of Containment Shell Subject to Biaxial Tension 316 3.8.2.1.2.10 Steel Embedments Transmitting Loads to Concrete 317 3.8.2.1.3 Assessment of Design Codes and Load Changes for Steel Structures 317 3.8.2.1.3.1 Shear Connectors in Composite Beams 318 3.8.2.1.3.2 Composite Beams With Steel Deck 318 3.8.2.1.3.3 Hybrid Girders 318 3.8.2.1.3.4 Compression Elements 319 3.8.2.1.3.5 Tension Members 319 3.8.2.1.3.6 Coped Beams 319 3.8.2.1.3.7 Moment Connections 320 3.8.2.1.3.8 Lateral Bracing 320 3.8.2.1.3.9 Steel Embedments 320 3.8.2.1.4 Summary 322 3.8.2.2 Structural Reevaluation of Containment 322 3.8.2.2.1 Introduction 322 3.8.2.2.2 Containment Temperature 323 3.8.2.1.1 Containment Pressure 323 3.8.2.1.2 Seismic Loads 323 3.8.2.1.3 Design and Analysis Procedures 324 3.8.2.1.3.1 Containment Model 324 3.8.2.1.3.2 Seismic and Loss-of-Coolant Accident Loads 324 3.8.2.1.3.3 Pressure, Seismic, and Operating Temperature Loads 325 3.8.2.1.4 Structural Acceptance Criteria 326 3.8.2.1.5 Structural Evaluation of Containment 326 3.8.2.1.5.1 Seismic Analysis 326 3.8.2.1.5.2 Load Combinations 327 3.8.2.1.6 Structural Evaluation of Large Openings 328 3.8.2.1.7 Structural Evaluation of Tension Rods 328 Page 18 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.8.2.2 Dome Liner Reevaluation 328 3.8.2.2.1 Dome Liner Studs 328 3.8.2.2.2 Loads 328 3.8.2.2.2.1 Loss-of-Coolant Accident 328 3.8.2.2.2.2 Steam Line Break 329 3.8.2.2.3 Model Definition 329 3.8.2.2.3.1 General Dome Model 329 3.8.2.2.3.2 Insulation Termination Region Model 329 3.8.2.2.4 Analysis 330 3.8.2.2.4.1 Controlling Loads 330 3.8.2.2.4.2 Liner-Stud Interaction 330 3.8.2.2.4.3 Effect of Internal Pressure on Liner Buckling 332 3.8.2.2.5 Results and Conclusions 333 3.8.2.2.5.1 Insulation Termination Region 333 3.8.2.2.5.2 General Dome 334 3.8.2.2.5.3 Effect of Internal Pressure on Liner Buckling and Stud Integrity 335 3.8.2.2.6 Overall Conclusions 337 3.8.3 CONTAINMENT INTERNAL STRUCTURES 337 3.8.3.1 Description of the Internal Structures 337 3.8.3.2 Applicable Codes, Standards, and Specifications 338 3.8.3.3 Loads and Load Combinations 338 3.8.3.3.1 Load Combinations Considered 338 3.8.3.3.2 Applicable Load Combinations 338 3.8.3.4 Design and Analysis Procedures 339 3.8.3.4.1 Original Design 339 3.8.3.4.2 Systematic Evaluation Program Reevaluation 340 3.8.3.5 Method of Analysis 340 3.8.3.6 Structural Acceptance Criteria 341 3.8.3.7 Structural Evaluation 341 3.8.4 OTHER SEISMIC CATEGORY I STRUCTURES 341 3.8.4.1 Description of the Structures 341 3.8.4.1.1 Auxiliary Building 342 3.8.4.1.2 Control Building 342 3.8.4.1.3 Diesel Generator Building 343 3.8.4.1.4 Intermediate Building 343 Page 19 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.8.4.1.5 Standby Auxiliary Feedwater Building 344 3.8.4.1.6 Screen House 344 3.8.4.1.7 Turbine Building 345 3.8.4.1.8 Service Building 345 3.8.4.1.9 Interconnected Building Complex 346 3.8.4.1.10 Canister Preparation Building (CPB) 346 3.8.4.2 Applicable Codes, Standards, and Specifications 347 3.8.4.3 Loads and Load Combinations 347 3.8.4.4 Design and Analysis Procedures 347 3.8.4.4.1 Original Design and Analysis Procedures 347 3.8.4.4.2 SEP Reevaluation Design and Analysis Procedures 348 3.8.4.4.2.1 Mathematical Model 348 3.8.4.4.2.2 Method of Analysis 350 3.8.4.4.2.3 Structural Evaluation 351 3.8.4.5 Masonry Walls 352 3.8.4.5.1 Applicable Walls 352 3.8.4.5.2 Loads and Load Combinations 352 3.8.4.5.3 Stress Analysis 354 3.8.4.5.3.1 Computer Program 354 3.8.4.5.3.2 Seismic Analysis 354 3.8.4.5.4 Interstory Drift 355 3.8.4.5.5 Multi-Wythe Walls 355 3.8.4.5.6 Block Pullout 355 3.8.4.5.7 Structural Acceptance Criteria - Allowable Stresses 355 3.8.4.5.7.1 Normal Operating Conditions 355 3.8.4.5.7.2 Safe Shutdown Earthquake 356 3.8.4.5.8 Evaluation Results 356 3.8.4.5.8.1 General 356 3.8.4.5.8.2 Inelastic Analysis 357 3.8.4.5.8.3 Wall Modifications 357 3.8.4.5.9 Materials, Quality Control, and Special Construction 358 Techniques 3.8.5 FOUNDATIONS 359 Page 20 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR Table 3.8-1a COMPUTER PROGRAM SAND INPUT FOR CONTAINMENT 364 SEISMIC ANALYSIS - DIMENSIONS AND FORMULA Table 3.8-1b COMPUTER PROGRAM SAND INPUT FOR CONTAINMENT 365 SEISMIC ANALYSIS - DIMENSION CALCULATIONS Table 3.8-1c COMPUTER PROGRAM SAND INPUT FOR CONTAINMENT 366 SEISMIC ANALYSIS - NATURAL FREQUENCIES AND | |||
===RESPONSE=== | |||
Table 3.8-2 MAJOR STRUCTURES FOR WHICH PRESTRESSED ROCK 367 ANCHORS WERE USED Table 3.8-3 PROPERTIES AND TESTS FOR CONTAINMENT ANCHOR AND 369 TENDON CORROSION INHIBITOR Table 3.8-4 ALLOWABLE STRESSES 370 Table 3.8-5a CONTAINMENT STRUCTURE STRESSES - LOADING #1 DEAD 371 LOAD Table 3.8-5b CONTAINMENT STRUCTURE STRESSES - LOADING #2 FINAL 372 PRESTRESS - 636 K/TENDON Table 3.8-5c CONTAINMENT STRUCTURE STRESSES - LOADING #3 374 OPERATING TEMPERATURE - WINTER Table 3.8-5d CONTAINMENT STRUCTURE STRESSES - LOADING #4 376 OPERATING TEMPERATURE - SUMMER Table 3.8-5e CONTAINMENT STRUCTURE STRESSES - LOADING #5 377 INTERNAL PRESSURE Table 3.8-5f CONTAINMENT STRUCTURE STRESSES - LOADING #6 378 ACCIDENT TEMPERATURE - P = 60 PSIG, T = 286F Table 3.8-5g CONTAINMENT STRUCTURE STRESSES - LOADING #7 379 ACCIDENT TEMPERATURE - P = 90 PSIG, T = 312F Table 3.8-5h CONTAINMENT STRUCTURE STRESSES - LOADING #8 0.10G 381 EARTHQUAKE - HORIZONTAL + VERTICAL COMPONENT Table 3.8-6a CONTAINMENT STRUCTURE LOADING COMBINATIONS - 382 LOAD NUMBERS 1 THROUGH 48 Table 3.8-6b CONTAINMENT STRUCTURE LOADING COMBINATIONS - 384 KEY TO SYMBOLS Table 3.8-7 CONCRETE COVER REQUIRED FOR REINFORCING STEEL 385 Table 3.8-8 ELASTOMER PADS PROPERTIES 386 Table 3.8-9 ROCK ANCHOR A - UPLIFT TEST WITH JACKING FRAME, 387 MAY 19, 1966 Table 3.8-10 DESIGN CODE COMPARISON 388 Table 3.8-11 ACI 318-63 VERSUS ACI 349-76 CODE COMPARISONS 390 Table 3.8-12 ACI 301-63 VERSUS ACI 301-72 (REVISED 1975) COMPARISON 392 Table 3.8-13 ACI 318-63 VERSUS ASME B&PV CODE, SECTION III, 393 DIVISION 2, 1980 CODE COMPARISON Table 3.8-14 ASME B&PV CODE, SECTION III, DIVISION 2, 1980 (ACI 359-80) 394 VERSUS ACI 318-63 CODE COMPARISION Page 21 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR Table 3.8-15 LIST OF STRUCTURAL ELEMENTS TO BE EXAMINED 395 Table 3.8-16 MASSES, MOMENT OF INERTIA (I), FLEXURAL AREA (A), 397 AND SHEAR AREA (As) FOR THE LLNL MODEL Table 3.8-17 MODAL FREQUENCIES FOR THE LAWRENCE LIVERMORE 398 NATIONAL LABORATORY CONTAINMENT SHELL MODEL Table 3.8-18 RESPONSE VALUES FOR REGULATORY GUIDE 1.60 399 HORIZONTAL (0.17g) AND VERTICAL (0.11g) SPECTRA INPUT Table 3.8-19 PEAK HARMONIC AMPLITUDES OF THE SEISMIC LOAD ON 400 CYLINDER AND DOME OF THE CONTAINMENT SHELL Table 3.8-20 MATERIAL PROPERTIES FOR STEEL, CONCRETE, AND FOAM 401 INSULATION Table 3.8-21 MAXIMUM DISPLACEMENTS OF 5/8-INCH S6L STUDS IN THE 402 INSULATION TERMINATION REGION Table 3.8-22 MAXIMUM DISPLACEMENT OF STUDS IN GENERAL DOME 403 Table 3.8-23 LOAD DEFINITIONS 404 3.9 MECHANICAL SYSTEMS AND COMPONENTS 406 3.9.1 SPECIAL TOPICS FOR MECHANICAL COMPONENTS 406 3.9.1.1 Design Transients 406 3.9.1.1.1 Load Combinations 406 3.9.1.1.2 Cyclic Loads 406 3.9.1.1.2.1 Thermal and Pressure Cyclic Loads 406 3.9.1.1.2.2 Pressurizer Surge Line 406 3.9.1.1.2.3 Unisolable Connections to the Reactor Coolant System 407 3.9.1.1.3 Transient Hydraulic Loads 408 3.9.1.1.4 Operating-Basis Earthquake 408 3.9.1.1.5 Safe Shutdown Earthquake 408 3.9.1.1.6 Secondary System Fluid Flow Instability (Water Hammer) 408 3.9.1.1.7 Loss-of-Coolant Accident 408 3.9.1.2 Computer Programs Used in Analysis 409 3.9.1.3 Experimental Stress Analysis 410 3.9.1.3.1 Plastic Model Analysis 410 3.9.1.3.2 Plastic Model Details 410 3.9.2 DYNAMIC TESTING AND ANALYSIS 412 3.9.2.1 Piping Systems 412 3.9.2.1.1 General 412 3.9.2.1.2 Seismic Category I Piping, 2-1/2 Inch Nominal Size and Larger 413 3.9.2.1.2.1 Static Analysis 413 3.9.2.1.2.2 Dynamic Analysis 413 Page 22 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.9.2.1.2.3 Residual Heat Removal System Line From Reactor Coolant System 414 Loop A to Containment 3.9.2.1.2.4 Steam Line From Steam Generator B to Containment 415 3.9.2.1.2.5 Charging Line 416 3.9.2.1.3 Seismic Category I Piping, 2-Inch Nominal Size and Under, Original 416 Design 3.9.2.1.4 Pressurizer Safety and Relief Valve Discharge Piping 416 3.9.2.1.4.1 1972 Analysis 416 3.9.2.1.4.2 NUREG 0737, Item II.D.1 Analysis 417 3.9.2.1.5 Main Steam Header Dynamic Load Factor Analysis 418 3.9.2.1.5.1 Extended Power Uprate Considerations 419 3.9.2.1.6 Secondary System Water Hammer 419 3.9.2.1.6.1 Analysis 419 3.9.2.1.6.2 Evaluation Results 420 3.9.2.1.6.3 Corrective Actions 420 3.9.2.1.6.4 Extended Power Uprate Considerations 421 3.9.2.1.7 Velan Swing Check Valves 421 3.9.2.1.8 Seismic Piping Upgrade Program 421 3.9.2.2 Safety-Related Mechanical Equipment 422 3.9.2.2.1 Original Seismic Input and Behavior Criteria 422 3.9.2.2.2 Current Seismic Input 423 3.9.2.2.3 Systematic Evaluation Program 423 3.9.2.2.4 Systematic Evaluation Program Reevaluation of Selected Mechanical 424 Components for Design Adequacy 3.9.2.2.4.1 Essential Service Water (SW) Pumps 424 3.9.2.2.4.2 Component Cooling Heat Exchanger 425 3.9.2.2.4.3 Component Cooling Surge Tank 425 3.9.2.2.4.4 Diesel-Generator Air Tanks 425 3.9.2.2.4.5 Boric Acid Storage Tank 426 3.9.2.2.4.6 Refueling Water Storage Tank (RWST) 426 3.9.2.2.4.7 Motor-Operated Valves 427 3.9.2.2.4.8 Steam Generators 427 3.9.2.2.4.9 Reactor Coolant Pumps 428 3.9.2.2.4.10 Pressurizer 428 3.9.2.2.4.11 Control Rod Drive Mechanism 429 3.9.2.3 Dynamic Response Analysis of Reactor Internals Under Operational 429 Flow Transients and Steady-State Conditions Page 23 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.9.2.3.1 Design Criteria 430 3.9.2.3.1.1 General 430 3.9.2.3.1.2 Critical Internals 430 3.9.2.3.1.3 Allowable Stress Criteria 431 3.9.2.3.2 Blowdown and Force Analysis 431 3.9.2.3.2.1 Computer Program 431 3.9.2.3.2.2 Blowdown Model 432 3.9.2.3.2.3 LATFORC MODEL 433 3.9.2.3.2.4 FORCE2 MODEL 433 3.9.2.3.3 Fuel Assembly Thimbles 434 3.9.2.3.4 Dynamic System Analysis of Reactor Internals Under Loss-of-Coolant 434 Accident (LOCA) 3.9.2.3.4.1 Mathematical Model of the Reactor Pressure Vessel (RPV) System 434 3.9.2.3.4.2 Analytical Methods 436 3.9.2.3.4.3 RPV Internal Hydraulic Loads 436 3.9.2.3.4.4 Reactor Coolant Loop Mechanical Loads 438 3.9.2.3.4.5 Results of the Analysis 438 3.9.2.3.5 Transverse Guide Tube Excitation by Blowdown Forces 438 3.9.2.3.5.1 General 438 3.9.2.3.6.1 Reactor Pressure Vessel System Thermal-Hydraulic Analysis 440 3.9.2.3.6.2 Bypass Flow Analysis 440 3.9.2.3.6.3 Thermal Analysis of the Baffle/Barrel Region 441 3.9.2.3.6.4 Pressure Drop Across the Baffle Plate Analyses 441 3.9.2.3.6.5 Flow Induced Vibration 441 3.9.2.3.6.6 Reactor Internals Structural Integrity 441 3.9.2.3.6.7 Control Rod Performance 441 3.9.2.3.6.8 Vessel/Internals/Fuel/Control Rod Response During Loca Conditions 442 3.9.2.3.6.9 Summary of Conclusions 442 3.9.2.4 Asymmetric Loss-of-Coolant Accident Loading Analysis 442 3.9.2.5 Seismic Evaluation of Reactor Vessel Internals 442 3.9.2.5.1 Analysis Procedure 442 3.9.2.5.2 Analysis Results 443 3.9.3 COMPONENT SUPPORTS AND CORE SUPPORT STRUCTURES 444 3.9.3.1 Loading Combinations, Design Transients, and Stress Limits 444 3.9.3.2 Component Supports 444 3.9.3.2.1 Reactor Vessel 444 Page 24 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.9.3.2.2 Steam Generators 445 3.9.3.2.3 Reactor Coolant Pumps 445 3.9.3.2.4 Pressurizer 446 3.9.3.2.5 Reactor Coolant Piping 446 3.9.3.3 Pipe Supports 446 3.9.3.3.1 Original Analysis 446 3.9.3.3.2 IE Bulletin Reanalysis 446 3.9.3.3.3 Seismic Piping Upgrade Program 447 3.9.3.3.3.1 Applicable Supports 447 3.9.3.3.3.2 Load Combinations and Stress Limits 447 3.9.3.3.3.3 Structural Requirements 447 3.9.3.3.4 Base Plate Flexibility 449 3.9.3.3.5 Snubbers 449 3.9.3.3.5.1 Design Loads 449 3.9.3.3.5.2 Surveillance Program 450 3.9.4 CONTROL ROD DRIVE SYSTEMS 450 3.9.4.1 Description 450 3.9.4.1.1 General 450 3.9.4.1.2 Latch Assembly 451 3.9.4.1.3 Pressure Vessel 452 3.9.4.1.4 Operating Coil Stack 452 3.9.4.1.5 Drive Shaft Assembly 452 3.9.4.1.6 Position Indicator Coil Stack 452 3.9.4.2 Design Loads, Stress Limits, and Allowable Deformation 452 3.9.4.3 Control Rod Drive Mechanism Housing Mechanical Failure Evaluation 453 3.9.4.3.1 Housing Description 453 3.9.4.3.2 Effects of Rod Travel Housing Longitudinal Failures 453 3.9.4.3.3 Effect of Rod Travel Housing Circumferential Failures 453 3.9.4.3.4 Summary 454 3.9.5 REACTOR PRESSURE VESSEL INTERNALS 454 3.9.5.1 Design Arrangements 454 3.9.5.1.1 Lower Core Support Structure 454 3.9.5.1.1.1 Support Structure Assembly 454 3.9.5.1.1.2 Lower Core Plate 454 3.9.5.1.1.3 Thermal Shield 455 3.9.5.1.1.4 Coolant Flow Passages 456 Page 25 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.9.5.1.1.5 Support and Alignment Arrangements 456 3.9.5.1.2 Upper Core Support Assembly 456 3.9.5.1.3 In-Core Instrumentation Support Structures 457 3.9.5.2 Loading Conditions 458 3.9.5.3 Design Bases 458 3.9.6 INSERVICE INSPECTION OF PUMPS AND VALVES 459 3.9.6.1 General 459 3.9.6.2 Inservice Testing of Pumps 459 3.9.6.3 Inservice Testing of Valves 460 3.9.7 Extended Power Uprate (EPU) 460 Table 3.9-1 ORIGINAL DESIGN LOADING COMBINATIONS AND STRESS 464 LIMITS Table 3.9-2 RESIDUAL HEAT REMOVAL LOOP A STRESS | |||
==SUMMARY== | |||
465 Table 3.9-3 MAIN STEAM LINE-LOOP B STRESS | |||
==SUMMARY== | |||
466 Table 3.9-4 CHARGING LINE STRESS | |||
==SUMMARY== | |||
467 Table 3.9-5 LOAD COMBINATIONS AND ACCEPTANCE CRITERIA FOR 468 PRESSURIZER SAFETY AND RELIEF VALVE PIPING AND SUPPORTS - UPSTREAM OF VALVES Table 3.9-6 LOAD COMBINATIONS AND ACCEPTANCE CRITERIA FOR 469 PRESSURIZER SAFETY AND RELIEF VALVE PIPING AND SUPPORTS - SEISMICALLY DESIGNED DOWNSTREAM PORTION Table 3.9-7 DEFINITIONS OF LOAD ABBREVIATIONS 470 Table 3.9-8 LOADING COMBINATIONS AND STRESS LIMITS FOR PIPING 471 FOR SEISMIC UPGRADE PROGRAMS Table 3.9-9 ALLOWABLE STEAM GENERATOR NOZZLE LOADS 472 Table 3.9-10 REACTOR COOLANT PUMP AUXILIARY NOZZLE UMBRELLA 473 LOADS Table 3.9-11 SYSTEMATIC EVALUATION PROGRAM STRUCTURAL 476 BEHAVIOR CRITERIA FOR DETERMINING SEISMIC DESIGN ADEQUACY Table 3.9-12 MECHANICAL COMPONENTS SELECTED FOR SEP SEISMIC 477 REVIEW Table 3.9-13 MAXIMUM STRESS HOT-LEG BREAK (ORIGINAL ANALYSIS) 478 Table 3.9-14 MAXIMUM STRESS COLD-LEG BREAK (ORIGINAL 479 ANALYSIS) | |||
Table 3.9-15 MAXIMUM CORE BARREL STRESS AND DEFLECTION UNDER 480 HOT-LEG BLOWDOWN (ORIGINAL ANALYSIS) | |||
Table 3.9-16a MAXIMUM STRESS INTENSITIES AND DEFLECTION COLD- 481 LEG BLOWDOWN (ORIGINAL ANALYSIS) - IN THE UPPER BARREL Page 26 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR Table 3.9-16b MAXIMUM STRESS INTENSITIES AND DEFLECTION COLD- 482 LEG BLOWDOWN (ORIGINAL ANALYSIS) - AT THE UPPER BARREL ENDS Table 3.9-17 CORE BARREL STRESSES (ORIGINAL ANALYSIS) 483 Table 3.9-18 CORE BARREL STRESSES (ORIGINAL ANALYSIS) 484 Table 3.9-19 CORE BARREL STRESSES (ORIGINAL ANALYSIS) 485 Table 3.9-20 CORE BARREL STRESSES (ORIGINAL ANALYSIS) 486 Table 3.9-21 CORE BARREL STRESSES (ORIGINAL ANALYSIS) 487 Table 3.9-22 CORE BARREL STRESSES (ORIGINAL ANALYSIS) 489 Table 3.9-23a LOAD COMBINATIONS AND ALLOWABLE STRESS LIMITS 490 FOR PRIMARY EQUIPMENT SUPPORTS EVALUATION - FOR PLANT EVENTS Table 3.9-23b LOAD COMBINATIONS AND ALLOWABLE STRESS LIMITS 491 FOR PRIMARY EQUIPMENT SUPPORTS EVALUATION - | |||
DEFINITION OF LOADING CONDITIONS FOR PRIMARY EQUIPMENT SUPPORTS EVALUATION IN TABLE 3.9-23a Table 3.9-24 RESIDUAL HEAT REMOVAL LOOP A SUPPORT LOADS1 492 CALCULATED FOR IE BULLETIN 79-07 Table 3.9-25a MAIN STEAM LINE LOOP B SUPPORT LOADS2 CALCULATED 495 FOR IE BULLETIN 79 SEISMIC SUPPORT Table 3.9-25b MAIN STEAM LINE LOOP B NOZZLE LOADS CALCULATED 496 FOR IE BULLETIN 79 NOZZLE LOADS Table 3.9-26 CHARGING LINE SUPPORT LOADSa CALCULATED FOR IE 497 BULLETIN 79-07 Table 3.9-27 LOADING COMBINATIONS AND STRESS LIMITS FOR 502 SUPPORTS ON PIPING SYSTEMS Table 3.9-28 ANALYSIS OF TYPICAL PIPE SUPPORT BASE PLATES 503 CALCULATED FOR IE BULLETIN 79-02 Table 3.9-29 INTERNALS DEFLECTIONS UNDER ABNORMAL OPERATION 504 3.10 SEISMIC QUALIFICATION OF SEISMIC CATEGORY I INSTRU- 505 MENTATION AND ELECTRICAL EQUIPMENT 3.10.1 SEISMIC QUALIFICATION CRITERIA 505 3.10.1.1 Original Criteria 505 3.10.1.2 Current Criteria 505 3.10.2 SEISMIC QUALIFICATION OF ELECTRICAL EQUIPMENT AND 506 INSTRUMENTATION 3.10.2.1 Introduction 506 3.10.2.2 Battery Racks 507 3.10.2.3 Motor Control Centers 1L and 1M 507 3.10.2.4 Switchgear 508 Page 27 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.10.2.5 Control Room Electrical Panels 508 3.10.2.6 Electrical Cable Raceways 509 3.10.2.7 Constant Voltage Transformers 509 3.10.3 SEISMIC QUALIFICATION OF SUPPORTS OF ELECTRICAL 509 EQUIPMENT AND INSTRUMENTATION 3.10.3.1 Equipment Addressed 510 3.10.3.2 Raceway Anchorages 510 3.10.3.2.1 Test Program 510 3.10.3.2.2 Test Loads 511 3.10.3.2.3 Expansion Anchor Test Results 512 3.10.3.2.4 Frictional Anchor Test Results 512 3.10.3.2.5 Embedded Anchor Test Results 513 3.10.3.3 Class 1E Equipment Anchorage Qualification Program 513 3.10.3.4 Conclusions 514 3.10.4 FUNCTIONAL CAPABILITY OF COMPONENTS 514 3.10.5 SEISMIC CATEGORY I TUBING 514 3.10.5.1 Codes and Standards 514 3.10.5.1.1 Tubing Design Requirements 515 3.10.5.1.2 Tubing Supports Design Requirements 515 3.10.5.2 Load Conditions 516 3.10.5.2.1 Tubing 516 3.10.5.2.2 Tubing Supports 516 3.10.5.3 Routing Requirements 517 Table 3.10-1 MAJOR CLASS 1E COMPONENTS AND THE BASIS FOR 520 SEISMIC QUALIFICATION Table 3.10-2 ELECTRICAL COMPONENTS SELECTED FOR SEISMIC 522 REVIEW Table 3.10-3 SHELL ANCHOR TEST | |||
==SUMMARY== | |||
523 Table 3.10-4 FRICTION BOLT TEST RESULT | |||
==SUMMARY== | |||
524 Table 3.10-5 CATEGORY 3 ANCHORS TEST | |||
==SUMMARY== | |||
525 Table 3.10-6 STRESS LIMITS FOR TUBING 526 3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRI- 527 CAL EQUIPMENT 3. | |||
==11.1 BACKGROUND== | |||
527 3.11.1.1 Initial Design Considerations 527 Page 28 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 3.11.1.2 Review of Environmental Qualification of Safety-Related Electrical 527 Equipment 3.11.2 Equipment Identification 528 3.11.3 IDENTIFICATION OF LIMITING ENVIRONMENTAL CONDI- 528 TIONS 3.11.3.1 Inside Containment 528 3.11.3.1.1 Post Loss-of-Coolant Accident Environment 528 3.11.3.1.2 Post Main Steam Line Break Environment 530 3.11.3.2 Auxiliary Building 530 3.11.3.2.1 Heating, Ventilation, and Air Conditioning 530 3.11.3.2.2 Loss of Ventilation 531 3.11.3.2.3 Radiation Levels 532 3.11.3.2.4 Flooding 532 3.11.3.3 Intermediate Building 532 3.11.3.4 Cable Tunnel 533 3.11.3.5 Control Building 533 3.11.3.6 Diesel Generator Rooms 534 3.11.3.7 Turbine Building 534 3.11.3.8 Auxiliary Building Annex 535 3.11.3.9 Screen House 535 3.11.4 EQUIPMENT QUALIFICATION INFORMATION 535 3.11.5 ENVIRONMENTAL QUALIFICATION PROGRAM 535 Table 3.11-1 ENVIRONMENTAL SERVICE CONDITIONS FOR EQUIPMENT 540 DESIGNED TO MITIGATE DESIGN-BASIS EVENTS Table 3.11-2 ESTIMATES FOR TOTAL AIRBORNE GAMMA DOSE 549 CONTRIBUTORS IN CONTAINMENT TO A POINT IN THE CONTAINMENT CENTER - GINNA STATION Table 3.11-3 ESTIMATES FOR TOTAL AIRBORNE BETA DOSE 551 CONTRIBUTORS IN CONTAINMENT TO A POINT IN THE CONTAINMENT CENTER - GINNA STATION Table 3.11-4 ESTIMATES FOR TOTAL AIRBORNE GAMMA DOSE 553 CONTRIBUTORS IN CONTAINMENT TO A POINT IN THE CONTAINMENT CENTER, REGULATORY GUIDE 1.89, REVISION 1 Table 3.11-5 ESTIMATES FOR TOTAL AIRBORNE BETA DOSE 555 CONTRIBUTORS IN CONTAINMENT TO A POINT IN THE CONTAINMENT CENTER, REGULATORY GUIDE 1.89, REVISION 1 Table 3.11-6 GINNA STATION/REGULATORY GUIDE 1.89, APPENDIX D, 557 COMPARISON OF POSTACCIDENT RADIATION ENVIRONMENT ASSUMPTIONS Page 29 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR FIGURES Figure 3.7-1 Seismic Response Spectra, 8%g Housner Model Figure 3.7-2 Seismic Response Spectra, 20%g Housner Model Figure 3.7-3 NRC Systematic Evaluation Program Site Specific Spectrum, Ginna Site (5% Damping) | |||
Figure 3.7-4 Comparison of the Housner Response Spectrum for 2% of Critical Damping with the 7% Regulatory Guide 1.60 Spectrum Figure 3.7-5 In-Structure Response Spectra for Interconnected Building, Half-Area and Full-Area Models Figure 3.7-6 Containment Building and Complex of Interconnected Seismic Cate- gory I and Nonseismic Structures, Plan View Figure 3.7-7 Horizontal Response Spectra - SEP Systematic Evaluation Program Figure 3.7-8 Steam Generator Mathematical Model Figure 3.7-9 Mathematical Model of Reactor Vessel Figure 3.7-10 Seismic Average Acceleration Spectrum Design Earthquake, 1% Damping Figure 3.7-11 Locations Where In-Structure Response Spectra Were Generated in Interconnected Building Complex Figure 3.7-12 SEP Response Spectra for Pressurizer PR-1 (Containment Building Elevation 253 ft) for 3%, 5%, and 7% Damping Figure 3.7-13 SEP Response Spectra for Control Rod Drive (Containment Building Elevation 253 ft) for 3%, 5%, 7% Damping Figure 3.7-14 SEP Response Spectra for Control Rod Drive (Containment Building Elevation 278 ft) for 3%, 5%, and 7% Damping Figure 3.7-15 SEP Response Spectra for Steam Generator SG-1A (Containment Building Elevation 250 ft) for 3%, 5%, and 7% Damping Figure 3.7-16 SEP Response Spectra for Steam Generator SG-1A (Containment Building Elevation 278 ft) for 3%, 5%, and 7% Damping Figure 3.7-17 SEP Response Spectra for Steam Generator SG-1B (Containment Building Elevation 250 ft) for 3%, 5%, and 7% Damping Figure 3.7-18 SEP Response Spectra for Steam Generator SG-1B (Containment Building Elevation 278 ft) for 3%, 5%, and 7% Damping Figure 3.7-19 SEP Response Spectra for Reactor Coolant Pump Rp-1A (Containment Building Elevation 247 ft) for 3%, 5%, and 7% Damping Figure 3.7-20 SEP Response Spectra for Reactor Coolant Pump RP-1B (Containment Building Elevation 247 ft) for 3%, 5%, and 7% Damping Figure 3.7-21 SEP Equipment Response Spectra for 3%, 5%, and 7% Damping at Auxiliary Building Platform (Elevation 281 ft 6 in) | |||
Figure 3.7-22 SEP Equipment Response Spectra for 3%, 5%, and 7% Damping at Auxiliary Building Heat Exchanger 35 (Elevation 281 ft 6 in) | |||
Page 30 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR Figure 3.7-23 SEP Equipment Response Spectra for 3%, 5%, and 7% Damping at Auxiliary Building Surge Tank 34 Figure 3.7-24 SEP Equipment Response Spectra for 3%, 5%, and 7% Damping at Auxiliary Building Boric Acid Storage Tank 34 Figure 3.7-25 SEP Equipment Response Spectra for 3%, 5%, and 7% Damping at Auxiliary Building Operating Floor (Elevation 271 ft 6 in) | |||
Figure 3.7-26 SEP Equipment Response Spectra for 3%, 5%, and 7% Damping at Control Building Basement Floor (Elevation 250 ft 0 in) | |||
Figure 3.7-27 SEP Equipment Response Spectra for 3%, 5%, and 7% Damping at Control Building Relay Room Floor (Elevation 269 ft 9 in) | |||
Figure 3.7-28 SEP Equipment Response Spectra for 3%, 5%, and 7% Damping at Control Room Floor (Elevation 289 ft 9 in) | |||
Figure 3.7-29 Residual Heat Removal Line Inside Containment Figure 3.7-30 Lumped Mass Model - Steam Line B Figure 3.7-31 Structural Model, Pressurizer Safety and Relief Line Figure 3.8-1 Containment Cross Section and Details Figure 3.8-2 Containment Mat Foundation and Ring Girder Figure 3.8-3 Containment Mat Foundation, Reinforcement and Details Figure 3.8-4 Containment Wall Reinforcement and Details Figure 3.8-5 Containment Dome Reinforcement and Details Figure 3.8-6 Containment Miscellaneous Embedded Back-Up Steel Figure 3.8-7 Tendon Vent Cans and Grease Fill Connections Figure 3.8-8 Temperature Gradients - Operating Conditions Figure 3.8-9 Earthquake Meridional Forces Figure 3.8-10 Containment Dynamic Analysis Model Figure 3.8-11 Ginna Containment Mode Shapes Figure 3.8-12 Ginna Containment - Earthquake Response Figure 3.8-13 Moments, Shears, Deflection, Tensile Force, and Hoop Tension Dia-grams Load Combination A Figure 3.8-14 Moments, Shears, Deflection, Tensile Force, and Hoop Tension Dia-grams Load Combination B Figure 3.8-15 Moments, Shears, Deflection, Tensile Force, and Hoop Tension Dia-grams Load Combination C Figure 3.8-16 Tendon to Rock Coupling Figure 3.8-17 Containment - Top Tendon Access Figure 3.8-18 Containment Miscellaneous Steel Tendon Conduit - Hinge Detail Figure 3.8-19 Liner Knuckle Dimensions Figure 3.8-20 Containment Base to Cylinder Model Page 31 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR Figure 3.8-21 Containment Dome to Cylinder Discontinuity Model Figure 3.8-22 Cracked Wall Shear Modulus Analysis Figure 3.8-23 Liner Shear Stress Analysis Figure 3.8-24 Windgirder, Shear Channels, and Shear Studs Figure 3.8-25 Cylinder Liner Plate Support Model Figure 3.8-26 Containment Penetration Details Figure 3.8-27 Containment Penetration Details (Typical) | |||
Figure 3.8-28 Composite Drawing Electrical Penetration Figure 3.8-29 Containment Penetrations Section and Details Figure 3.8-30 Containment Equipment Hatch Figure 3.8-31 Containment Personnel Hatch Figure 3.8-32 Containment - Fuel Transfer Tube Penetration Figure 3.8-33 Containment Penetrations Arrangements and Location Figure 3.8-34 Test Coupon - Containment Concrete Shell Figure 3.8-35 Cadweld Splice Test Results Figure 3.8-36 Quality Control Chart for 5000 PSI Concrete Figure 3.8-37 Neoprene Base Hinge Load Deformation Specimen 1 Figure 3.8-38 Neoprene Base Hinge Load Deformation Specimen 2 Figure 3.8-39 Rock Anchor Test A-1 Figure 3.8-40 Containment - Rock Anchor A Test Figure 3.8-41 Containment - Rock Anchor B Test Figure 3.8-42 Containment - Rock Anchor C Test Figure 3.8-43 Accident Temperature Transient Inside the Containment Used for Liner Analysis Figure 3.8-44 Accident Pressure Transient Inside the Containment Used for Liner Analysis Figure 3.8-45 Plan View of the Facade Structure and Containment Figure 3.8-46 Accident Temperature Gradient Through the Uninsulated Containment Shell After 94 Seconds Figure 3.8-47 Accident Temperature Gradient Through the Uninsulated Containment Shell After 380 Seconds Figure 3.8-48 Ginna Containment Structure Figure 3.8-49 Liner Stud Interaction Models Figure 3.8-50 Accident Temperature Distribution in the Steel Liner Figure 3.8-51 Force Displacement Curve for 3/4 in. Headed Studs Figure 3.8-52 Force Displacement Curve for 5/8 in. S6L Studs Page 32 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR Figure 3.8-53 Strut Buckling Under P and Delta T Figure 3.8-54 Pressure Effect on Liner Buckling Comparison With LOCA Figure 3.8-55 Reactor Containment Internal Structures Figure 3.8-56 Containment Interior Structures Model for STARDYNE Figure 3.8-57 Schematic Plan View of Major Ginna Structures Figure 3.8-58 Three-Dimensional View of Interconnected Building Complex Figure 3.8-59 Flow Chart of the Analysis of the Interconnected Building Complex Figure 3.8-60 Masonry Wall Reevaluation, Wall Location Plan, Lower Levels Figure 3.8-61 Masonry Wall Reevaluation, Wall Location Plan, Intermediate Levels Figure 3.8-62 Masonry Wall Reevaluation, Wall Location Plan, Operating Levels Figure 3.9-1 Steam-Generator Water Hammer Preliminary Forcing Function Figure 3.9-2 Plastic Model of Reactor Coolant System - Plan View Figure 3.9-3 Lumped Mass Dynamic Model of PCV 434 Figure 3.9-4 Lumped Mass Dynamic Model of PCV 435 Figure 3.9-5 Comparison of WHAM Results With LOFT Semi-Scale Blowdown Experiments, Test No. 519 Figure 3.9-6 Comparison of WHAM Results With LOFT Semi-Scale Blowdown Experiments, Test No. 560 Figure 3.9-6a Steam Generator Upper Support Systems Figure 3.9-7 Control Rod Drive Mechanism Assembly Figure 3.9-8 Control Rod Drive Mechanism Schematic Figure 3.9-9 Reactor Vessel Internals Figure 3.9-10 Detailed View of Reactor Vessel Internals Figure 3.10-1 Q-Deck Detail Figure 3.10-2 Unistrut Detail Figure 3.10-3 Threaded Insert Detail Poured in Place Anchor Figure 3.10-4 Tray Support Types for Friction Bolt Testing Figure 3.11-1 Containment Volume and Reactor Power LOCA Dose Corrections Page 33 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR Appendix 3A INITIAL EVALUATION OF CAPABILITY TO WITHSTAND TOR- 558 NADOES 3A.1 INTRODUCTION AND CONCLUSIONS 559 3A.2 IDENTIFICATION OF CRITICAL SYSTEMS AND STRUCTURES 561 3A.3 TORNADO EFFECTS ON STRUCTURES 562 3A.3.1 GENERAL 562 3A.3.2 REACTOR CONTAINMENT 562 3A.3.3 AUXILIARY BUILDING 562 3A.3.4 INTERMEDIATE BUILDING 563 3A.3.5 DIESEL-GENERATOR ANNEX 563 3A.3.6 SCREEN HOUSE 563 3A.3.7 CONTROL ROOM 564 3A.3.8 SERVICE BUILDING 564 3A.3.9 CABLE TUNNELS 564 3A.4 TORNADO EFFECTS ON THE SYSTEMS REQUIRED FOR HOT 565 SHUTDOWN 3A.4.1 DECAY HEAT REMOVAL 565 3A.4.1.1 Steam Relief System 565 3A.4.1.2 Auxiliary Feedwater System 565 3A.4.1.3 Service Water System 566 3A.4.2 REACTIVITY CONTROL 567 3A.4.2.1 Boration System 567 3A.4.2.2 Boration Using Refueling Water 567 3A.4.3 CONTAINMENT VENTILATION SYSTEM 568 3A.4.4 EMERGENCY POWER SUPPLY SYSTEM 569 3A.4.5 CONTROL SYSTEM 569 3A.4.5.1 Control Room 569 3A.4.5.2 Systems of Batteries 569 3A.4.5.3 Steam-Generator Level and Pressure Indicators, Pressurizer Pressure 569 and Level Control 3A.5 TORNADO EFFECT ON SPENT FUEL POOL 571 Appendix 3A Figures Figure 1 Boration System Figure 2 Site Plot Plan Page 34 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR Figure 3 Diesel Generator Annex - Elevation 253 ft 6 in. | |||
Figure 4 Screen House Layout Figure 5 Steam Relief Valves Figure 6 Auxiliary Feedwater Pumps Figure 7 Component Cooling System Figure 8 Spent Fuel Storage Pool, Plan View Figure 9 Spent Fuel Storage Pool, Section View Appendix 3B DESIGN OF LARGE OPENING REINFORCEMENTS FOR CON- 572 TAINMENT VESSEL Table of Contents 573 Summary 576 I. Design Bases 576 II. GENERAL DESCRIPTION 576 III. STRESS DISTRIBUTION AROUND A CIRCULAR HOLE IN A 576 CIRCULAR CYLINDRICAL SHELL IV. ANALYSIS OF STRESSES AROUND LARGE OPENINGS 576 V. VERIFICATION OF REINFORCEMENT ADEQUACY 577 | |||
: 1. DESIGN BASES 579 1.1 General 579 1.2 Design Loads 579 1.3 Load Combinations 579 1.4 Material Stress/Strain Criteria 580 1.5 Test Condition 582 1.6 Operating Condition 582 | |||
: 2. GENERAL DESCRIPTION OF OPENING REINFORCEMENT 583 2.1 Introduction 583 2.2 Rebar for Discontinuity Stresses 583 2.3 Normal Shear at Edge of Opening 583 2.4 Prestressing 583 | |||
: 3. STRESS DISTRIBUTION AROUND A CIRCULAR HOLE IN A 584 CIRCULAR CYLINDRICAL SHELL 3.1 Introduction 584 3.2 Finite Element Method 585 3.3 Applications of Three-Dimensional Photoelasticity 586 | |||
: 4. ANALYSIS OF THE STRESSES AROUND LARGE OPENINGS IN 588 THE R. E. GINNA SECONDARY CONTAINMENT VESSEL Page 35 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 4.1 Verification of Finite-element Method of Analysis 588 4.2 General Considerations Concerning Methods of Analysis of Reinforced 589 Concrete Structures in the Cracked Condition 4.3 Stress Analysis in Cracked and Uncracked Conditions Under Operating 590 and Accident Loads 4.3.2 Basic Loading Conditions 592 4.3.3 Effect of Concrete Cracking 595 4.3.4 Effect of Creep and Shrinkage 597 | |||
: 5. Verification of Design Criteria 598 5.1 Basis For Verification of Shell Loading Capacity Due to Primary Loads 598 (Principal Stress-resultants and Principal Stress-couples) 5.2 Interaction Diagram 599 5.3 Reinforcing Steel 600 5.4 Maximum Liner Stresses 600 5.5 Penetration Barrel 600 5.6 Normal Shear 601 5.7 Rebar Anchorage 602 5.8 Tendon Losses 603 5.9 Summary of Design and Conclusions 604 Table 4-1 Load Combinations 608 Table 4-2 Stress Around Equipment Hatch-Loading (Uncracked Shell) 609 Table 4-3 Stress Around Equipment Hatch-Loading (Cracked Shell) 611 Table 5-1 Maximum Liner Stresses Stress tangent to the edge in Ksi 619 Appendix A to EFFECT OF CONCRETE CREEP AND THE SUSTAINED OPER- 620 APPENDIX 3B ATING STRESSES ON STRESS DISTRIBUTION AROUND OPEN-INGS IN A RAPIDLY PRESSURIZED REINFORCED CONCRETE VESSEL 3B.A EFFECT OF CONCRETE CREEP AND THE SUSTAINED OPER- 621 ATING STRESSES ON STRESS DISTRIBUTION AROUND OPEN-INGS IN A RAPIDLY PRESSURIZED REINFORCED CONCRETE VESSEL Appendix B TO EARTHQUAKE ANALYSIS 628 APPENDIX 3B 3B.B Earthquake Analysis 629 ADDENDUM TO ADDENDUM TO THE REPORT ON: DESIGN OF LARGE OPEN- 630 APPENDIX 3B ING REINFORCEMENTS FOR CONTAINMENT VESSEL 3B.C Introduction 631 1 Design 632 1.1 Concrete Shear 632 1.2 Interaction Diagrams 632 Page 36 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 1.3 Earthquake Design 632 1.4 Thermal Gradients 632 1.5 Penetration Material 633 1.6 Working Strength Design 633 1.7 Anchorage Plate Bearing Stress 633 1.8 Insulated Liner Temperature Increase 633 1.9 High Strength Rebar 633 1.10 Proof Test Instrumentation 633 1.11 Operating Conditions 634 1.12 Shear - Diagonal Tension 634 1.13 Normal Shears 635 1.14 Radial Shear at the Periphery of the Opening 635 1.15 Accident Temperature Effects 635 1.16 Analytical Model for Different Load Combinations 635 1.17 Shear Reinforcement 635 1.18 Equation (5.11) 636 1.19 Rebar Located Away from the Barrel 636 1.20 Verification of Analysis 637 1.21 Test Problem 638 1.22 Accident Temperature 638 2 Construction 639 2.1 Construction Schedule 639 2.2 Concrete Removal 639 2.3 Concrete Work 639 2.4 Retensioning Tendons 640 2.5 Rebar Splices 640 2.6 Tendon Conduit 640 Table I STRESS AROUND EQUIPMENT HATCH LOADING CONDITION 641 NO. 4 - Accident Temperature Appendix 3B Figures Figures Appendix 3B Figures Figure 1 Figure 2 Figure 3 Stress Distribution Around Openings in Cylindrical Shells Figure 4 Grid for Finite Element Analysis of the Stresses Around Openings Page 37 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR Figure 5 Membrane Stress Around Opening Edge (Vessel Subject to Internal Pressure) | |||
Figure 6 Surface Stresses Around Opening Edge (Vessel Subject to Internal Pressure) | |||
Figure 7 Hoop Stresses Along Longitudinal Axis (Vessel Subject to Internal Pressure) | |||
Figure 8 Axial Stresses Along Transverse Axis (vessel Subject to Inernal Pres-sure) | |||
Figure 9 Hoop Stress-Resultant No Along Symmetry Axes (Test Problem) | |||
Figure 10 Layer Thickness And Destination Figure 11 Nodal Forces Due to Curvature of Tendons in the Neighborhood of Opening Figure 12 Stress Distribution Around Openings (Thermal Gradient Near Equip-ment Opening) | |||
Figure 13 Steady State Temperature Distributions - Winter Gradient Figure 14 Stress Distribution Around Openings (Effect of Bond Failure Along Terminated Rebars) | |||
Figure 15 Hoop Stress-Resultant Along Horizontal And Vertical Symmetry Axes (Internal Pressure = 69 PSI) | |||
Figure 16 Shell Displacements (Final Vertical Prestress) | |||
Figure 17 Shell Displacements (69 PSI Internal Pressure) | |||
Figure 18 Interaction Diagram for Axial Compression/Tension and Bending Figure 19 Interaction Diagram Ring Steel Direction Elements No. 73 & 74 Figure 20 Interaction Diagram Elements No. 97, 100, & 101 Figure 21 Interaction Diagram Elements No. 97, 100, & 101 Figure 22 Interaction Diagram Elements No. 33, 55, 66, & 77 Figure 23 Interaction Diagram Element No. 77 Figure 24 Interaction Diagram Element No. 55 Drawings Figure Drawing 1 Reactor Containment Vessel - Equipment/Personnel Access Reinforce-ment - Enlarged Sections Figure Drawing 2 Reactor Containment Vessel - Equipment Access Opening Reinforce-ment - Stretch-out & Sections Figure I Comparison of H.H. & GAI Results Hoop Stress Resultants Along Horizontal and Vertical Symmetry Axes (Internal Pressure = 69 PSI) | |||
Figure Drawing 1 Reactor Containment Vessel - Equipment/Personnel Access Reinforcement - Enlarged Sections Page 38 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR Figure Drawing 2 Reactor Containment Vessel - Equipment Access Opening Reinforcement - Stretch-out & Sections Figure Drawing 3 Large Openings - Pour Schedule Appendix 3C CONTAINMENT SHELL STRESS CALCULATION RESULTS 642 Table 3C-1 CONTAINMENT SHELL STRESS CALCULATION RESULTS 643 Appendix 3D CONTAINMENT TENDON ANCHORAGE HARDWARE CAPAC- 668 ITY TESTS Compressive Load Tests of 90 Wire Tendon Base Plate - Test on Con- 669 crete Stand Compressive Load Tests of 90 Wire Tendon Base Plate - Test on Con- 673 crete Stand Compression Tests of 90-Wire Anchor Head Assembly 681 Compression Tests of 90-Wire Anchor Head Assembly 683 Load Tests of Coupler and Adaptor 90-11 690 Load Tests of Coupler and Adaptor 90-11 692 90 Wire Tendon Test 696 90 Wire Tendon Test 697 90 Wire Tendon Test 698 Load Tests of 90-X7 Coupler 702 Appendix 3E CONTAINMENT LINER INSULATION PREOPERATIONAL 704 TESTS BM Containment Insulation SP-5290 Ginna Plant 705 Report No. E455-T-268, VINYLCEL (4 pcf) - Water Vapor Permeabil- 707 ity and Humid Aging Tests Report No. E455-T-266, VINYLCEL (4 pcf) - Effect of Heat and Pres- 711 sure Report No. E455-T-258, VINYLCEL - Resistance to Flame Exposure 718 Appendix 3F | |||
==SUMMARY== | |||
OF STRUCTURAL DESIGN CODE COMPARISON 740 Table of Contents 741 3F.1 INTRODUCTION 742 Table 3F.2-1 AISC 1963 VERSUS AISC 1980 | |||
==SUMMARY== | |||
OF CODE 743 COMPARISON Table 3F.3-1 ACI 318-63 VERSUS ACI 349-76 | |||
==SUMMARY== | |||
OF CODE 747 COMPARISON Table 3F.4-1 ACI 301-63 VERSUS ACI 301-72 (REVISED 1975) | |||
==SUMMARY== | |||
OF 756 CODE COMPARISON Table 3F.5-1 ACI 318-63 VERSUS ASME B&PV CODE, SECTION III, 762 DIVISION 2, 1980, | |||
==SUMMARY== | |||
OF CODE COMPARISON | |||
: 1. Support load combination is seismic plus deadweight. | |||
: 2. Support load combination is seismic plus deadweight. | |||
Page 39 of 39 Revision 26 5/2016 | |||
GINNA/UFSAR 4 REACTOR 1 4.1 | |||
==SUMMARY== | |||
DESCRIPTION 2 4.1.1 REACTOR CORE 2 4.1.2 WESTINGHOUSE OPTIMIZED FUEL ASSEMBLIES/422 VAN- 2 TAGE + FUEL ASSEMBLIES 4.1.3 RECONSTITUTED FUEL ASSEMBLIES 4 4.1.4 STARTUP REPORT 5 | |||
==4.1 REFERENCES== | |||
FOR SECTION 4.1 6 4.2 FUEL SYSTEM DESIGN 7 4.2.1 DESIGN BASES 7 4.2.1.1 Performance Objectives 7 4.2.1.2 Principal Design Criteria 7 4.2.1.2.1 Reactor Core Design 7 4.2.1.2.2 Suppression of Power Oscillations 9 4.2.1.2.3 Redundancy of Reactivity Control 9 4.2.1.2.4 Reactivity MODE 3 (Hot Shutdown) Capability 9 4.2.1.2.5 Reactivity Shutdown Capability 10 4.2.1.2.6 Reactivity Holddown Capability 10 4.2.1.2.7 Reactivity Control Systems Malfunction 11 4.2.1.2.8 Maximum Reactivity Worth of Control Rods 12 4.2.1.2.9 Conformance With 1972 General Design Criteria 12 4.2.1.3 Safety Limits 13 4.2.1.3.1 Nuclear Limits 13 4.2.1.3.2 Reactivity Control Limits 13 4.2.1.3.3 Thermal and Hydraulic Limits 13 4.2.1.3.4 Mechanical Limits 14 4.2.1.3.4.1 Reactor Internals 14 4.2.1.3.4.2 Fuel Assemblies 15 4.2.1.3.4.3 Control Rods 16 4.2.1.3.4.4 Control Rod Drive Assembly 16 4.2.2 FUEL SYSTEM DESIGN DESCRIPTION 17 Page 1 of 5 Revision 26 5/2016 | |||
GINNA/UFSAR 4 REACTOR 1 4.2.3 CORE COMPONENTS DESIGN DESCRIPTION 17 4.2.3.1 Fuel Assembly 17 4.2.3.1.1 Top Nozzle, Springs, and Clamps 18 4.2.3.1.2 Bottom Nozzle 18 4.2.3.1.3 Guide Thimbles 19 4.2.3.1.4 Instrumentation Tube 19 4.2.3.1.5 Grid Assemblies 19 4.2.3.1.6 Fuel Rods 20 4.2.3.1.7 Fuel Assembly Joints and Connections 20 4.2.3.1.8 Fuel Assembly Identification 21 4.2.3.2 Control Rods 21 4.2.3.3 Neutron Source Assemblies 22 4.2.3.4 Plugging Devices 22 4.2.3.5 Fuel Pellet and Cladding Design Considerations 23 4.2.3.6 Reload Fuel Design 24 4.2.3.6.1 Reload Fuel Design - Westinghouse Optimized Fuel 24 4.2.3.6.2 Reload Fuel Design - Westinghouse OFA/VANTAGE + Fuel 24 4.2.3.6.3 Reload Fuel Design - Westinghouse 422V+ Fuel 24 4.2.3.7 Fuel Assembly and Rod Cluster Control Assembly Tests 24 4.2.3.7.1 Reactor Evaluation Center Tests 24 4.2.3.7.2 Loading and Handling Tests 24 4.2.3.7.3 Axial and Lateral Bending Tests 25 4.2.4 DESIGN EVALUATION 25 4.2.4.1 Fuel and Cladding Evaluation - Original Core 25 4.2.4.2 Design Evaluation - Reload Optimized Fuel Assembly, OFA/VAN- 26 TAGE+ Fuel Assembly, and 422 VANTAGE+ Fuel Assembly Designs 4.2.4.2.1 Introduction 26 4.2.4.2.2 Fuel Design 26 4.2.4.2.3 Design for Seismic and Loss-of-Coolant Accident Forces 26 4.2.4.2.4 Emergency Core Cooling System (ECCS) Calculation Loss-of-Coolant 26 Accident Cladding Models 4.2.4.2.5 Initial Fuel Conditions for Transient Analysis 27 Page 2 of 5 Revision 26 5/2016 | |||
GINNA/UFSAR 4 REACTOR 1 4.2.4.2.6 Predicted Clad Collapse Time 27 4.2.4.2.7 Nuclear Design 27 4.2.4.2.8 Fuel Assembly Hydraulic Lift-Off 28 4.2.4.2.9 Thermal-Hydraulic Analysis 28 4.2.4.2.9.1 Sensitivity Factors 28 4.2.4.2.9.2 WRB-1 Correlation 28 4.2.4.2.9.3 Rod Bow Penalties 29 4.2.4.2.9.4 DNBR Design Limits 30 4.2.4.3 Design Evaluation of Reconstituted Fuel Assemblies 30 4.2.5 CORE COMPONENTS TESTS AND INSPECTIONS 30 | |||
==4.2 REFERENCES== | |||
FOR SECTION 4.2 31 Table 4.2-1 NUCLEAR DESIGN DATA 33 Table 4.2-2 CORE MECHANICAL DESIGN PARAMETERS 35 Table 4.2-3 FUEL DESIGN 38 Table 4.2-4 KINETIC PARAMETERS USED IN TRANSIENT ANALYSIS 39 (WESTINGHOUSE OFA/VANTAGE+ AND 422V+ GINNA FUEL ASSEMBLY 14 x 14 FUEL) 4.3 RELOAD CORE NUCLEAR DESIGN 40 4.3.1 PRELIMINARY DESIGN PHASE 40 4.3.2 DETERMINATION OF NUCLEAR-RELATED KEY SAFETY 41 PARAMETERS 4.3.2.1 Reactivity Control Aspects 41 4.3.2.1.1 Insertion Limits 42 4.3.2.1.2 Total Rod Worth 43 4.3.2.1.3 Trip Reactivity 43 4.3.2.1.4 Differential Rod Worths 43 4.3.2.1.5 Summary 44 4.3.2.2 Core Reactivity Parameters and Coefficients 44 4.3.2.2.1 Moderator Temperature Coefficient 44 4.3.2.2.2 Fuel Temperature Coefficient 45 4.3.2.2.3 Boron Worth 45 4.3.2.2.4 Delayed Neutrons 45 4.3.2.2.5 Prompt Neutron Lifetime 45 Page 3 of 5 Revision 26 5/2016 | |||
GINNA/UFSAR 4.3.2.2.6 Summary 46 4.3.2.3 Reactor Core Power Distribution 46 4.3.3 EVALUATION OF RELOADS WITH OFA/VANTAGE+ AND 422V+ 46 FUEL ASSEMBLIES 4.3.4 TESTS FOR REACTIVITY ANOMALIES 47 | |||
==4.3 REFERENCES== | |||
FOR SECTION 4.3 48 4.4 THERMAL AND HYDRAULIC DESIGN 49 4.4.1 DESIGN BASIS 49 4. | |||
==4.2 DESCRIPTION== | |||
AND EVALUATION OF THE THERMAL-HYDRAU- 49 LIC DESIGN AND ANALYSIS OF RELOAD CORES 4.4.2.1 Hydraulic Evaluation 49 4.4.2.2 Thermal and Hydraulic Key Safety Parameters 49 4.4.2.2.1 Engineering Hot-Channel Factors 50 4.4.2.2.2 Axial Fuel Stack Shrinkage 50 4.4.2.2.3 Fuel Temperatures 50 4.4.2.2.4 Rod Internal Pressure 50 4.4.2.2.5 Core Thermal Limits 51 4.4.2.2.6 Key Safety Parameters for Specific Events 52 4.4.2.3 VIPRE Code 52 4.4.2.3.1 Steady-State Analysis 53 4.4.2.3.2 Transient Analysis 53 4.4.3 THERMAL-HYDRAULIC METHODOLOGY FOR OFA/VANTAGE+ 53 and 422V+ FUEL ASSEMBLY DESIGN EVALUATION 4.4.3.1 General 53 4.4.3.2 Rod Bow 55 4.4.4 THERMAL AND HYDRAULIC TESTS AND INSPECTIONS 55 4.4.5 REACTOR COOLANT FLOW MEASUREMENT 55 4.4.5.1 Pump Power 56 4.4.5.2 Secondary Heat Balance 56 4.4.5.3 Elbow Tap Differential Pressure 56 4.4.5.4 Core Exit Thermocouple 56 4.4.5.5 Pump Power-Differential Pressure 57 4.4.5.6 Experience 58 4.4.5.7 Low Flow Trip Setpoint 59 4.4.5.8 Precision Calorimetric Measurement for Reactor Coolant System Flow 59 | |||
==4.4 REFERENCES== | |||
FOR SECTION 4.4 63 Page 4 of 5 Revision 26 5/2016 | |||
GINNA/UFSAR Table 4.4-1 THERMAL AND HYDRAULIC DESIGN PARAMETERS 65 4.5 REACTOR MATERIALS 67 4.5.1 CONTROL ROD DRIVE SYSTEM STRUCTURAL MATERIALS 67 4.5.2 REACTOR INTERNALS MATERIALS 67 4.6 FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEM 68 FIGURES Figure 4.2-1 Typical Rod Cluster Control Assembly Figure 4.2-2 Fuel Assembly and Control Cluster Cross Section Figure 4.2-3 14 x 14 OFA and 422V+ Fuel Assemblies Figure 4.2-4 OFA and 422V+ Top Nozzle Assemblies Figure 4.2-5 Debris Filter Bottom Nozzle Figure 4.2-6 Optimized Guide Thimble Assembly Figure 4.2-7 Optimized Instrumentation Tube Figure 4.2-8 Mid-Grid Connection Figure 4.2-9 Removable Top Nozzle and Top Grid Connection Figure 4.3-1 Control Rod Cluster Groups Figure 4.4-1 Typical Pump Power Versus Flow Curves Page 5 of 5 Revision 26 5/2016 | |||
GINNA/UFSAR 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 1 5.1 | |||
==SUMMARY== | |||
DESCRIPTION 2 5.1.1 GENERAL 2 5.1.2 PERFORMANCE OBJECTIVES 2 5.1.3 DESIGN CRITERIA 2 5.1.3.1 Quality Standards 3 5.1.3.2 Performance Standards 3 5.1.3.3 Records Requirements 4 5.1.3.4 Missile Protection 4 5.1.3.5 Reactor Coolant Pressure Boundary 4 5.1.3.6 Monitoring Reactor Coolant Leakage 5 5.1.3.7 Reactor Coolant Pressure Boundary Capability 6 5.1.3.8 Reactor Coolant Pressure Boundary Rapid Propagation Failure Prevention 6 5.1.3.9 Reactor Coolant Pressure Boundary Surveillance 7 5.1.3.10 Adequacy of Reactor Coolant System Design Relative to 1972 10 CFR 50, 8 Appendix A, Criteria 5.1.4 DESIGN CHARACTERISTICS 8 5.1.4.1 Design Pressure 8 5.1.4.2 Design Temperature 9 5.1.5 CYCLIC LOADS 9 5.1.6 SERVICE LIFE 9 5.1.7 RELIANCE ON INTERCONNECTED SYSTEMS 10 5.1.8 SYSTEM INCIDENT POTENTIAL 10 Table 5.1-1 REACTOR COOLANT SYSTEM PRESSURE SETTINGS 12 Table 5.1-2 REACTOR COOLANT PIPING DESIGN DATA 13 Table 5.1-3 REACTOR COOLANT SYSTEM DESIGN PRESSURE DROP 14 Table 5.1-4 THERMAL AND LOADING CYCLES 15 5.2 INTEGRITY OF THE REACTOR COOLANT PRESSURE BOUNDARY 16 5.2.1 COMPLIANCE WITH CODES 16 5.2.1.1 System Integrity 16 5.2.1.2 Codes and Classifications 17 5.2.1.2.1 Code Requirements 17 5.2.1.2.2 Quality Control 17 5.2.1.2.3 Field Erection Procedures 18 5.2.1.3 Seismic Loads 18 Page 1 of 8 Revision 26 5/2016 | |||
GINNA/UFSAR 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.2.2 OVERPRESSURIZATION PROTECTION 19 5.2.2.1 Normal Operation 19 5.2.2.2 Low Temperature Overpressure Protection (LTOP) System 19 5.2.2.2.1 Design Bases 20 5.2.2.2.2 System Description 20 5.2.2.2.3 System Evaluation 21 5.2.2.2.3.1 General 21 5.2.2.2.3.2 Mass Addition Case 22 5.2.2.2.3.3 Heat Addition at 60F 22 5.2.2.2.3.4 Heat Addition at 320F 23 5.2.2.2.3.5 Administrative Controls 23 5.2.2.2.4 Tests and Inspections 24 5.2.3 REACTOR COOLANT PRESSURE BOUNDARY MATERIALS 24 5.2.3.1 Material Specifications 24 5.2.3.1.1 Nondestructive Examination of Materials and Components Prior to Operation 24 5.2.3.1.1.1 Quality Assurance Program 24 5.2.3.1.1.2 Welding and Heat Treatment 25 5.2.3.1.2 Quality Assurance for Electroslag Welds 26 5.2.3.1.2.1 Piping Elbows 26 5.2.3.1.2.2 Reactor Coolant Pump Casings 26 5.2.3.1.2.3 Reactor Coolant Pump Field Erection and Welding 28 5.2.3.2 Compatibility With Reactor Coolant 28 5.2.4 INSERVICE INSPECTION AND TESTING OF THE REACTOR COOLANT SYS- 29 TEM PRESSURE BOUNDARY 5.2.4.1 Inservice Inspection Program 29 5.2.4.2 Inspection Areas and Components 29 5.2.4.2.1 Accessible Components and Areas 29 5.2.4.2.2 Accessible Areas During Refueling 31 5.2.4.3 Accessibility 31 5.2.4.4 Examination Methods 32 5.2.4.5 Evaluation of Examination Results 33 5.2.4.6 Repair Requirements 33 5.2.4.7 Pressure Testing 33 5.2.4.8 Exemptions 33 Page 2 of 8 Revision 26 5/2016 | |||
GINNA/UFSAR 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.2.5 DETECTION OF LEAKAGE THROUGH REACTOR COOLANT PRESSURE 34 BOUNDARY 5.2.5.1 Leakage Detection Methods 34 5.2.5.2 Leakage Limitations 35 5.2.5.3 Locating Leaks 36 5.2.5.4 Leakage Detection System Descriptions 36 5.2.5.4.1 Containment Air Particulate and Radiogas Monitor 36 5.2.5.4.1.1 Air Particulate Monitor 36 5.2.5.4.1.2 Sensitivity Assumptions 36 5.2.5.4.1.3 Leakage Detection Threshold 38 5.2.5.4.1.4 Radiogas Monitor 39 5.2.5.4.2 Humidity Detector 39 5.2.5.4.3 Condensate Measuring System 39 5.2.5.4.4 Liquid Inventory in Process Systems and Containment Sumps 40 5.2.5.5 Leakage Detection System Evaluation 40 Table 5.2-1 REACTOR COOLANT SYSTEM CODE REQUIREMENTS 44 Table 5.2-2 MATERIALS OF CONSTRUCTION OF THE REACTOR COOLANT SYSTEM 45 COMPONENTS Table 5.2-3 REACTOR COOLANT SYSTEM QUALITY ASSURANCE PROGRAM 46 Table 5.2-4 Table DELETED 49 Table 5.2-5 REACTOR COOLANT PRESSURE BOUNDARY TO CONTAINMENT LEAK- 50 AGE DETECTION SYSTEMS Table 5.2-6 REACTOR COOLANT PRESSURE BOUNDARY INTERSYSTEM LEAKAGE 51 DETECTION SYSTEMS Table 5.2-7 SEQUENCE OF EVENTS - MASS ADDITION CASE 52 Table 5.2-8 HEAT ADDITION AT 60F - SEQUENCE OF EVENTS 53 Table 5.2-9 HEAT ADDITION AT 320F - SEQUENCE OF EVENTS 54 5.3 REACTOR VESSEL 55 5.3.1 REACTOR VESSEL MATERIALS 55 5.3.1.1 Reactor Vessel Description 55 5.3.1.2 Material Specifications 56 5.3.1.3 Testing and Surveillance 57 5.3.2 PRESSURE-TEMPERATURE LIMITS 57 Page 3 of 8 Revision 26 5/2016 | |||
GINNA/UFSAR 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.3.2.1 Thermal and Pressure Loadings 57 5.3.2.2 Pressure-Temperature Limits 58 5.3.2.3 Pressure-Temperature Limit Calculation 59 5.3.2.4 Irradiation Effect on Pressure-Temperature Limit 59 5.3.2.5 Heatup and Cooldown Rates 60 5.3.3 REACTOR VESSEL INTEGRITY 60 5.3.3.1 Safety Factors 60 5.3.3.2 Material Surveillance Program 61 5.3.3.3 Surveillance Program Analysis 62 5.3.3.3.1 Results Summary 63 5.3.3.3.2 Charpy V-Notch Impact Test Results 65 5.3.3.3.3 Tension Test Results 66 5.3.3.3.4 Radiation Analysis and Neutron Dosimetry 66 5.3.3.4 Analysis of Effects of Loss of Coolant and Safety Injection on the Reactor Vessel 66 5.3.3.4.1 Reactor Vessel 66 5.3.3.4.2 Safety Injection Nozzles 68 5.3.3.4.3 Fuel Assembly Grid Springs 68 5.3.3.4.4 Core Barrel and Thermal Shield 68 5.3.3.4.5 Subsequent Analyses of Reactor Vessel 68 5.3.3.5 Pressurized Thermal Shock 69 Table 5.3-1 REACTOR VESSEL SPECIFICATIONS 74 Table 5.3-2 REACTOR VESSEL DESIGN DATA 75 Table 5.3-3 REACTOR VESSEL MATERIALS 76 Table 5.3-4 IDENTIFICATION OF BELTLINE MATERIALS 77 Table 5.3-5 BELTLINE MATERIAL CHEMICAL COMPOSITION (WEIGHT PERCENT) 78 Table 5.3-6a MECHANICAL PROPERTIES OF BELTLINE MATERIALS - FORGINGS 79 Table 5.3-6b MECHANICAL PROPERTIES OF BELTLINE MATERIALS 80 Table 5.3-7 | |||
==SUMMARY== | |||
OF PRIMARY-PLUS-SECONDARY STRESS INTENSITY FOR 81 COMPONENTS OF THE REACTOR VESSEL Table 5.3-8 | |||
==SUMMARY== | |||
OF CUMULATIVE FATIGUE USAGE FACTORS FOR COMPO- 83 NENTS OF THE REACTOR VESSEL Table 5.3-9 | |||
==SUMMARY== | |||
OF SURVEILLANCE CAPSULE RESULTS 84 Table 5.3-10 COMPARISON OF SURVEILLANCE MATERIAL 30 FT-LB TRANSITION 85 TEMPERATURE SHIFTS AND UPPER SHELF ENERGY DECREASES WITH REGULATORY GUIDE 1.99, REVISION 2, PREDICTIONS Page 4 of 8 Revision 26 5/2016 | |||
GINNA/UFSAR 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.4 COMPONENT AND SUBSYSTEM DESIGN 87 5.4.1 REACTOR COOLANT PUMPS 87 5.4.1.1 General Description 87 5.4.1.1.1 Centrifugal Pump 87 5.4.1.1.2 Controlled Leakage Shaft Seal 87 5.4.1.1.3 Pump Motor 88 5.4.1.1.4 Vibration Measurement 88 5.4.1.1.5 Lube Oil Leakage Collection System 89 5.4.1.2 Pump Flywheel Integrity 89 5.4.1.2.1 Pump Overspeed 89 5.4.1.2.2 Pump Flywheel Design and Fabrication 89 5.4.1.2.3 Flywheel Design Evaluation 90 5.4.1.2.4 Pump Seismic Design 90 5.4.1.2.5 Inservice Inspection Program 91 5.4.1.2.6 Conclusion 91 5.4.2 STEAM GENERATORS 91 5.4.2.1 Replacement Steam Generator Materials 92 5.4.2.2 Steam Generator Inservice Inspection 92 5.4.2.3 Replacement Steam Generator Design Evaluation 92 5.4.2.4 High Cycle Fatigue Failure of Original Steam Generator Tubes 93 5.4.3 REACTOR COOLANT PIPING 93 5.4.3.1 General 93 5.4.3.1.1 General Description 93 5.4.3.1.2 Pressure Isolation of Low-Pressure Systems 94 5.4.3.2 Reactor Coolant System Vents 94 5.4.3.2.1 General 94 5.4.3.2.2 Reactor Head Vent System Description 95 5.4.4 MAIN STEAM LINE ISOLATION SYSTEM 96 5.4.5 RESIDUAL HEAT REMOVAL (RHR) SYSTEM 97 5.4.5.1 Design Bases 97 5.4.5.2 System Design 98 5.4.5.2.1 Codes and Classifications 99 5.4.5.2.2 Components 99 Page 5 of 8 Revision 26 5/2016 | |||
GINNA/UFSAR 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.4.5.2.2.1 Heat Exchangers 99 5.4.5.2.2.2 Pumps 99 5.4.5.2.2.3 Valves 99 5.4.5.2.2.4 Piping 100 5.4.5.3 Performance Evaluation 100 5.4.5.3.1 Isolation Requirement 100 5.4.5.3.1.1 Isolation Valve Description 100 5.4.5.3.1.2 Deviations From Branch Technical Position RSB 5-1 101 5.4.5.3.2 Residual Heat Removal Overpressure Protection 102 5.4.5.3.2.1 Design Basis 102 5.4.5.3.2.2 Analysis 102 5.4.5.3.2.3 Effect of Stuck Open Relief Valve 103 5.4.5.3.3 Residual Heat Removal Pump Protection 104 5.4.5.3.4 Single-Failure Considerations 105 5.4.5.3.5 Leakage Provisions 106 5.4.5.3.6 Boron Concentration 107 5.4.5.4 Residual Heat Removal at Reduced Coolant Inventory 107 5.4.5.4.1 Generic Letter 88-17 Requirements 107 5.4.5.4.2 Containment Closure 108 5.4.5.4.3 Instrumentation for Reduced Inventory Operation 109 5.4.5.4.4 Available Equipment to Mitigate Loss of Residual Heat Removal Cooling 110 5.4.5.4.5 Reduced Inventory Procedures 110 5.4.5.4.6 Analyses 111 5.4.5.5 Tests and Inspections 112 5.4.6 MAIN STEAM AND FEEDWATER PIPING 112 5.4.7 PRESSURIZER 113 5.4.7.1 System Description 113 5.4.7.2 Seismic Evaluation 114 5.4.8 PRESSURIZER RELIEF DISCHARGE SYSTEM 115 5.4.8.1 System Description 115 5.4.8.2 System Analysis 116 5.4.9 VALVES 116 5.4.9.1 Original Valve Design 116 5.4.9.2 Valve Wall Thickness 117 5.4.9.3 Motor-Operated Valve Program 117 Page 6 of 8 Revision 26 5/2016 | |||
GINNA/UFSAR 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.4.10 SAFETY AND PRESSURIZER POWER OPERATED RELIEF VALVES (PORVs) 119 5.4.10.1 System Description 119 5.4.10.2 Performance Testing and Evaluation 120 5.4.11 COMPONENT SUPPORTS 121 5.4.11.1 Design Criteria 121 5.4.11.1.1 General 121 5.4.11.1.2 Asymmetric Loss-of-Coolant Accident Loading 121 5.4.11.1.3 Lamellar Tearing 122 5.4.11.2 Support Structures 122 5.4.11.2.1 Reactor Vessel Supports 122 5.4.11.2.2 Steam Generator Supports 123 5.4.11.2.3 Reactor Coolant Pump Supports 123 5.4.11.2.4 Pressurizer Supports 123 5.4.11.2.5 Reactor Coolant Piping Supports 123 5.4.11.2.6 Inspection and Testing 123 Table 5.4-1 REACTOR COOLANT PUMP DESIGN DATA 129 Table 5.4-2 REPLACEMENT STEAM GENERATOR DESIGN DATA 130 Table 5.4-3 REACTOR COOLANT PUMP COMPOSITE HOT PERFORMANCE CURVE 131 DATA Table 5.4-4 REACTOR COOLANT PUMPS COLD PERFORMANCE CURVE DATA FOR 133 INDIVIDUAL IMPELLERS Table 5.4-5 REACTOR VESSEL HEAD VENT EQUIPMENT PARAMETERS 134 Table 5.4-6 RESIDUAL HEAT REMOVAL SYSTEM COMPONENT DESIGN DATA 136 Table 5.4-7 PRESSURIZER DESIGN DATA 138 Table 5.4-8 PRESSURIZER RELIEF TANK DESIGN DATA 139 Table 5.4-9 VALVE AND PIPING INFORMATION 140 Page 7 of 8 Revision 26 5/2016 | |||
GINNA/UFSAR 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS FIGURES Figure 5.2-1 Figure DELETED Figure 5.2-2 Figure DELETED Figure 5.2-3 Reactor Coolant Leak Detection Sensitivity Figure 5.3-1 Reactor Vessel Schematic Figure 5.3-2 Identification and Location of Beltline Region Material Figure 5.3-3 Arrangement of Surveillance Capsules in the Reactor Vessel Figure 5.4-1 Reactor Coolant Pump Figure 5.4-2 Reactor Coolant Pump Estimated Performance Characteristics Figure 5.4-2a Reactor Coolant Pump Composite Curve, Calculated Hot Performance, Total Head and Hydraulic Efficiency Versus Flow Figure 5.4-2b Reactor Coolant Pump Composite Curve, Calculated Hot Performance, Brake Horse- power Versus Flow Figure 5.4-2c Reactor Coolant Pump Composite Curve, Calculated Cold Performance, Total Head and Hydraulic Efficiency Versus Flow Figure 5.4-2d Reactor Coolant Pump Composite Curve, Calculated Cold Performance, Brake Horsepower Versus Flow Figure 5.4-3 Reactor Coolant Pressure Shaft Seal Arrangement Figure 5.4-4 Reactor Coolant Pump Flywheel Figure 5.4-5 Reactor Coolant Pump Flywheel Primary Stress at Operating Speed Figure 5.4-6 Replacement Steam Generator Figure 5.4-7 Figure DELETED Figure 5.4-8 Pressurizer Figure 5.4-9 Pressurizer Relief Tank Page 8 of 8 Revision 26 5/2016 | |||
GINNA/UFSAR 6 ENGINEERED SAFETY FEATURES 1 6.1 ENGINEERED SAFETY FEATURES INTRODUCTION AND MATERIALS 2 6. | |||
==1.1 INTRODUCTION== | |||
2 6.1.2 ENGINEERED SAFETY FEATURES MATERIALS 2 6.1.2.1 Postaccident Environmental Conditions 2 6.1.2.1.1 General 2 6.1.2.1.2 Design-Basis Accident Temperature-Pressure Cycle 3 6.1.2.1.3 Design-Basis Accident Radiation Environment 4 6.1.2.1.4 Design Chemical Composition of the Emergency Core Cooling Solution 4 6.1.2.1.5 Trace Composition of Emergency Core Cooling Solution 5 6.1.2.2 Materials of Construction in the Containment 5 6.1.2.3 Corrosion of Metals of Construction in Design-Basis Emergency Core Cooling 6 Solu- tion 6.1.2.3.1 Corrosion Resistance 6 6.1.2.3.2 Caustic Stress Cracking Resistance 7 6.1.2.4 Corrosion of Metals of Construction by Trace Contaminants in Emergency Core 7 Cool- ing Solution 6.1.2.4.1 Low Temperature of Emergency Core Cooling Solution 8 6.1.2.4.2 Low Chloride Concentration of Emergency Core Cooling Solution 8 6.1.2.4.3 Alkaline Nature of the Emergency Core Cooling Solution 9 6.1.2.4.4 Summary 9 6.1.2.5 Corrosion of Aluminum Alloys 9 6.1.2.6 Compatibility of Protective Coatings With the Postaccident Environment 10 6.1.2.7 Evaluation of the Compatibility of Concrete-Emergency Core Cooling Solution in 10 the Postaccident Environment 6.1.2.8 Miscellaneous Materials of Construction 12 6.1.2.8.1 Sealants 12 6.1.2.8.2 Containment Recirculation Fan Cooler (CRFC) Materials 12 6.1.2.8.3 Polyvinyl Chloride Protective Coating 13 6.1.2.8.4 Vinylcel Insulation 14 6.1.2.9 Organic Materials Evaluation 15 Table 6.1-1 REVIEW OF SOURCES OF VARIOUS ELEMENTS IN CONTAINMENT 18 AND THEIR EFFECTS ON MATERIALS OF CONSTRUCTION Table 6.1-2 MATERIALS OF CONSTRUCTION IN GINNA STATION CONTAINMENT 20 Table 6.1-3 INVENTORIES OF ALUMINUM INSIDE CONTAINMENT BUILDING 21 Table 6.1-4 CORROSION OF MATERIALS IN SODIUM BORATE SOLUTION 22 Page 1 of 14 Revision 26 5/2016 | |||
GINNA/UFSAR 6 ENGINEERED SAFETY FEATURES 1 Table 6.1-5 GINNA Post-LOCA CONTAINMENT TEMPERATURES 23 Table 6.1-6 CONCRETE SPECIMEN TEST DATA 24 Table 6.1-7 EVALUATION OF SEALANT MATERIALS FOR USE IN THE CONTAINMENT 25 6.2 CONTAINMENT SYSTEMS 26 6.2.1 CONTAINMENT SYSTEM STRUCTURE 26 6.2.1.1 Design Basis 26 6.2.1.1.1 Principal Design Criteria 27 6.2.1.1.1.1 General 27 6.2.1.1.1.2 Quality Standards 27 6.2.1.1.1.3 Performance Standards 27 6.2.1.1.1.4 Fire Protection 28 6.2.1.1.1.5 Records Requirement 28 6.2.1.1.1.6 Reactor Containment 29 6.2.1.1.1.7 Reactor Containment Design Basis 30 6.2.1.1.1.8 Nil Ductility Transition Requirement for Containment Material 30 6.2.1.1.2 Supplementary Accident Criteria 30 6.2.1.1.3 Energy and Material Release 31 6.2.1.2 Containment Integrity Evaluation 31 6.2.1.2.1 Systematic Evaluation Program (SEP) Evaluation 31 6.2.1.2.1.1 Introduction 31 6.2.1.2.1.2 NRC Analyses 32 6.2.1.2.1.3 Summary 32 6.2.1.2.2 Mass and Energy Release Safety Analysis 32 6.2.1.2.2.1 Loss-of-Coolant (LOCA) Mass and Energy Releases 32 6.2.1.2.2.2 Input Parameters and Assumptions 34 6.2.1.2.2.3 Desription of Analyses 38 6.2.1.2.2.4 Mass and Energy Release Data 41 6.2.1.2.2.5 Long-Term Mass and Energy Releases 44 6.2.1.2.2.6 Long-Term LOCA Containment Response 45 6.2.1.2.2.7 Description of the LOCA GOTHIC Containment Model 49 6.2.1.2.2.8 Results 53 6.2.1.2.3 Secondary System Pipe Break Analysis 55 6.2.1.2.3.1 Event Analysis 55 6.2.1.2.3.2 Protective Features 55 6.2.1.2.3.3 Single Failures Assumed 56 6.2.1.2.3.4 Operator Actions Assumed 56 6.2.1.2.3.5 Chronological Description of Event 56 6.2.1.2.3.6 Impact on Fission Product Barriers 57 Page 2 of 14 Revision 26 5/2016 | |||
GINNA/UFSAR 6.2.1.2.3.7 Reactor Core and Plant System Evaluation 57 6.2.1.2.3.8 Input Parameters and Initial Conditions 58 6.2.1.2.3.9 Methodology 58 6.2.1.2.3.10 Acceptance Criteria 58 6.2.1.2.3.11 Results 58 6.2.1.2.3.12 Radiological Consequences 58 6.2.1.2.3.13 Conclusion 59 6.2.1.3 Evaluation of Containment Internal Structures 59 6.2.1.3.1 Introduction 59 6.2.1.3.2 Reactor Coolant Loop Compartment Pressure 59 6.2.1.3.3 Thermal Gradients 60 6.2.1.3.4 Reactor Vessel and Steam Generator Annulus Pressure 60 6.2.1.3.5 Seismic Evaluation 61 6.2.1.3.6 Technical Evaluation for Extended Power Uprate (EPU) Conditions 61 6.2.1.3.6.1 Introduction 61 6.2.1.3.6.2 Input Parameters and Assumptions 61 6.2.1.3.6.3 Acceptance Criteria 62 6.2.1.3.6.4 Description of Analysis 63 6.2.1.3.6.5 Short-Term LOCA M&E Releases Results 63 6.2.1.4 Minimum Operating Conditions 64 6.2.1.5 Instrumentation Requirements 64 6.2.1.5.1 Pressure 64 6.2.1.5.2 Sump Level 64 6.2.1.5.3 Radiation 65 6.2.1.5.4 Containment Temperature and Dewpoint 65 6.2.2 CONTAINMENT HEAT REMOVAL SYSTEMS 65 6.2.2.1 Containment Recirculation Fan Cooler (CRFC) System 66 6.2.2.1.1 Design Bases 66 6.2.2.1.1.1 Capacity 66 6.2.2.1.1.2 Design Objectives 67 6.2.2.1.1.3 Special Features 68 6.2.2.1.2 System Design 68 6.2.2.1.2.1 System Description 68 6.2.2.1.2.2 Design Analysis 69 6.2.2.1.2.3 Redundancy Provisions 70 6.2.2.1.2.4 Actuation Provisions 71 6.2.2.1.2.5 Environmental Protection 71 6.2.2.1.3 Design Evaluation 72 6.2.2.1.4 Tests and Inspections 73 6.2.2.1.5 Instrumentation 73 Page 3 of 14 Revision 26 5/2016 | |||
GINNA/UFSAR 6.2.2.2 Containment Spray System 74 6.2.2.2.1 Design Bases 74 6.2.2.2.1.1 Design Criteria 74 6.2.2.2.1.2 Performance Objectives 75 6.2.2.2.1.3 Service Life 76 6.2.2.2.1.4 Codes and Classifications 76 6.2.2.2.2 System Design 76 6.2.2.2.2.1 Operational Requirements 76 6.2.2.2.2.2 Refueling Water Storage Tank (RWST) 77 6.2.2.2.2.3 Containment Spray Pumps 77 6.2.2.2.2.4 Liquid Jet Eductor 77 6.2.2.2.2.5 Spray Ring Headers 77 6.2.2.2.2.6 Spray Nozzles 78 6.2.2.2.2.7 Environmental Qualification 78 6.2.2.2.2.8 System Tests 78 6.2.2.2.3 Design Evaluation 78 6.2.2.2.3.1 Design Basis 78 6.2.2.2.3.2 Heat Transfer Calculations 79 6.2.2.2.3.3 Reliance on Interconnected Systems 81 6.2.2.2.3.4 Reliability Considerations 81 6.2.2.2.3.5 Containment Spray Pump Net Positive Suction Head Requirements 81 6.2.2.2.3.6 Equipment Protection 81 6.2.2.2.4 Minimum Operating Conditions 82 6.2.2.2.5 Tests and Inspections 82 6.2.2.2.6 Instrumentation 82 6.2.2.2.6.1 Interlock and Control Features 82 6.2.2.2.6.2 Control Room and Local Indication 82 6.2.3 SECONDARY CONTAINMENT 83 6.2.4 CONTAINMENT ISOLATION SYSTEM 83 6.2.4.1 Design Criteria 83 6.2.4.2 Design Basis 84 6.2.4.2.1 Functional Requirements 84 6.2.4.2.2 Seismic Design 85 6.2.4.3 System Design 85 6.2.4.3.1 Isolation Valve Parameters Tabulation 86 6.2.4.3.2 Isolation Valve Operability 86 6.2.4.4 Design Evaluation 87 6.2.4.4.1 Current Safety Criteria 87 6.2.4.4.2 Class 1 Penetrations (Outgoing Lines, Reactor Coolant System) 88 6.2.4.4.2.1 Applicable Lines 88 Page 4 of 14 Revision 26 5/2016 | |||
GINNA/UFSAR 6.2.4.4.2.2 Class 1 Penetration Evaluation 88 6.2.4.4.3 Class 2 (Outgoing Lines) 89 6.2.4.4.3.1 Applicable Lines 89 6.2.4.4.3.2 Class 2 Evaluation 89 6.2.4.4.4 Class 3 (Incoming Lines) 90 6.2.4.4.4.1 Class 3A Penetrations 91 6.2.4.4.4.2 Class 3B Penetrations 91 6.2.4.4.5 Class 4 Penetrations (Closed System, Missile Protected) 92 6.2.4.4.5.1 Applicable Lines 92 6.2.4.4.5.2 Class 4 Evaluation 93 6.2.4.4.6 Class 5 Penetrations (Special Service) 94 6.2.4.4.6.1 Applicable Lines 94 6.2.4.4.6.2 Class 5 Evaluation 94 6.2.4.4.7 Special Cases 95 6.2.4.4.8 Instrumentation and Controls Evaluation 95 6.2.4.4.9 Containment Purging During Normal Plant Operation 95 6.2.5 COMBUSTIBLE GAS CONTROL IN THE CONTAINMENT 96 6.2.5.1 Introduction 97 6.2.5.2 Hydrogen Recombiner System 98 6.2.5.2.1 Description 98 6.2.5.2.2 Containment Venting 99 6.2.5.3 Design Evaluation 99 6.2.5.3.1 Hydrogen Production and Accumulation 99 6.2.5.3.1.1 Zirconium-Water Reaction 100 6.2.5.3.1.2 Radiolytic Hydrogen Generation 100 6.2.5.3.1.3 Corrosion of Materials 101 6.2.5.3.1.4 Initial Inventory in the RCS and Pressurizer 102 6.2.5.3.2 Effect of Recombiners 102 6.2.6 CONTAINMENT LEAKAGE TESTING 103 6.2.6.1 Containment Design Leakage 103 6.2.6.2 Tests and Inspections 103 6.2.6.2.1 Design Criteria 103 6.2.6.2.2 Initial Containment Leakage Rate Testing 103 6.2.6.2.3 Periodic Containment Leakage Rate Testing (Type A Tests) 104 6.2.6.2.4 Provisions for Testing of Type B Penetrations 104 6.2.6.2.5 Provisions for Testing of Isolation Valves (Type C) 105 6.2.6.3 Leakage Test Compliance with 10 CFR 50, Appendix J 105 Table 6.2-1 SYSTEM PARAMETERS INITIAL CONDITIONS 111 Table 6.2-2 SAFETY INJECTION FLOW - MINIMUM SAFEGUARDS 112 Table 6.2-3 SAFETY INJECTION FLOW - MAXIMUM SAFEGUARDS 113 Page 5 of 14 Revision 26 5/2016 | |||
GINNA/UFSAR Table 6.2-4 LOCA M&E RELEASE ANALYSIS - CORE DECAY HEAT FRACTION 114 Table 6.2-5 DOUBLE-ENDED HOT LEG BREAK BLOWDOWN M&E RELEASE 116 Table 6.2-6 DOUBLE-ENDED HOT LEG BREAK - MASS BALANCE 120 Table 6.2-7 DOUBLE-ENDED HOT LEG BREAK - ENERGY BALANCE 121 Table 6.2-8 DOUBLE-ENDED PUMP SUCTION BREAK MIN SI BLOWDOWN M&E 122 RELEASE Table 6.2-9 DOUBLE-ENDED PUMP SUCTION BREAK MIN SI 126 Reflood M&E Release Table 6.2-10 DOUBLE-ENDED PUMP SUCTION BREAK 132 Min SI Principle Parameters During Reflood Table 6.2-11 DOUBLE-ENDED PUMP SUCTION BREAK 134 Post Reflood M&E Release-Minimum Safeguards Table 6.2-12 DOUBLE-ENDED PUMP SUCTION BREAK MASS BALANCE - MIN SI 141 Table 6.2-13 DOUBLE-ENDED PUMP SUCTION BREAK ENERGY BALANCE - 142 MINIMUM SAFEGUARDS Table 6.2-14 Table DELETED 143 Table 6.2-15 Table DELETED 144 Table 6.2-16 CONTAINMENT RESPONSE ANALYSIS PARAMETERS 145 Table 6.2-17 CONTAINMENT RECIRCULATION FAN COOLER HEAT REMOVAL 146 CAPABIL- ITY AS A FUNCTION OF CONTAINMENT STEAM SATURATION TEMPERA- TURE Table 6.2-18 LOCA CONTAINMENT RESPONSE ANALYSIS RECIRCULATION 147 SYSTEM ALIGNMENT PARAMETERS Table 6.2-19 CONTAINMENT STRUCTURAL HEAT SINK INPUT 148 Table 6.2-20 MATERIAL PROPERTIES FOR CONTAINMENT STRUCTURAL HEAT SINKS 151 Table 6.2-21 DOUBLE-ENDED HOT LEG BREAK SEQUENCE OF EVENTS 152 Table 6.2-22 DOUBLE-ENDED PUMP SUCTION BREAK SEQUENCE OF EVENTS 153 (Minimum Safeguards) | |||
Table 6.2-23 LOCA CONTAINMENT RESPONSE RESULTS 155 Table 6.2-24 INITIAL CONDITIONS AND MAJOR ASSUMPTIONS FOR THE STEAMLINE 156 BREAK MASS AND ENERGY RELEASE MODEL (LIMITING CONTAINMENT PRESSURE CASE) | |||
Table 6.2-25 MAJOR CONTAINMENT ANALYSIS ASSUMPTIONS 157 Table 6.2-26 SEQUENCE OF EVENTS 158 STEAMLINE BREAK, VITAL BUS FAILURE Table 6.2-27 CONTAINMENT SPRAY PUMP DESIGN PARAMETERS 159 Table 6.2-28 SINGLE FAILURE ANALYSIS - CONTAINMENT SPRAY SYSTEM 160 Table 6.2-29 CONTAINMENT PIPING PENETRATIONS AND ISOLATION BOUNDARIES 161 Table 6.2-30 CONTAINMENT PIPING PENETRATIONS AND ISOLATION BOUNDARIES 178 | |||
- NOTES FOR TABLE 6.2-29 Table 6.2-31 CONTAINMENT PIPING PENETRATIONS AND ISOLATIONBOUNDARIES - 181 LEGEND FOR Table 6.2-29 Table 6.2-32 EFFECT OF LOSS OF AIR OR POWER SUPPLY TO AIR-OPERATED VALVES 182 Page 6 of 14 Revision 26 5/2016 | |||
GINNA/UFSAR Table 6.2-33 ESSENTIAL AND NONESSENTIAL SYSTEM CONTAINMENT PENETRATIONS 184 Table 6.2-34 PARAMETERS AND ASSUMPTIONS USED TO DETERMINE HYDROGEN GENERATION (HISTORICAL) 188 Table 6.2-35 FISSION PRODUCT DECAY ENERGY IN SUMP SOLUTION (HISTORICAL) 189 Table 6.2-36 FISSION PRODUCT DECAY ENERGY IN THE CORE (HISTORICAL) 190 6.3 EMERGENCY CORE COOLING SYSTEM (ECCS) 191 6.3.1 DESIGN CRITERIA 191 6.3.1.1 Emergency Core Cooling System (ECCS) Capability 191 6.3.1.2 Inspection of Emergency Core Cooling System (ECCS) 192 6.3.1.3 Testing of Emergency Core Cooling System (ECCS) and Components 192 6.3.1.4 Testing of Operational Sequence of Emergency Core Cooling System (ECCS) 192 6.3.1.5 Service Life 193 6.3.1.6 Codes and Classifications 193 6.3.2 SYSTEM DESIGN AND OPERATION 193 6.3.2.1 System Description 193 6.3.2.1.1 General 193 6.3.2.1.2 Injection Phase 196 6.3.2.1.3 Recirculation Phase 197 6.3.2.2 Component Description 197 6.3.2.2.1 Accumulators 197 6.3.2.2.2 Safety Injection Pumps 198 6.3.2.2.2.1 Operation 198 6.3.2.2.2.2 Pump Design and Fabrication 199 6.3.2.2.3 Refueling Water Storage Tank (RWST) 199 6.3.2.2.4 Heat Exchangers 200 6.3.2.2.5 Boric Acid Storage Tanks 200 6.3.2.2.6 Containment Sump B 201 6.3.2.2.7 Valves 201 6.3.2.2.7.1 General 201 6.3.2.2.7.2 Motor-Operated Valves 202 6.3.2.2.7.3 Manual Valves 203 6.3.2.2.7.4 Accumulator Check Valves 203 6.3.2.2.7.5 Leakage Limitations 204 6.3.2.2.8 Piping 205 6.3.2.2.8.1 General 205 6.3.2.2.8.2 Design Criteria 205 6.3.2.2.8.3 Design Review 205 6.3.2.2.8.4 Materials 206 6.3.2.2.8.5 Welding and Fabrication 206 Page 7 of 14 Revision 26 5/2016 | |||
GINNA/UFSAR 6.3.2.2.8.6 Packaging 207 6.3.2.2.9 Motors 207 6.3.2.3 System Operation 207 6.3.2.3.1 Separation 207 6.3.2.3.2 System Actuation 207 6.3.2.3.3 Injection Phase 208 6.3.2.3.4 Recirculation Phase 208 6.3.2.3.5 Steam Line Break Protection 209 6.3.2.3.6 Safety Injection System Leakage Outside Containment 210 6.3.3 DESIGN EVALUATION 210 6.3.3.1 Range of Core Protection 210 6.3.3.1.1 Safety Injection Requirements Versus Break Size 210 6.3.3.1.2 Makeup System Capacity 211 6.3.3.1.3 System Evaluation 211 6.3.3.2 System Response 212 6.3.3.3 Safety Injection System Switchover From Injection to Recirculation 212 6.3.3.4 Boron Precipitation During Long-Term Cooling 214 6.3.3.5 Single Failure Analysis 214 6.3.3.6 Passive Systems 215 6.3.3.7 Emergency Flow to the Core 215 6.3.3.8 Recirculation Loop Leakage 216 6.3.3.9 Safety Injection Pump Net Positive Suction Head Requirements 216 6.3.3.10 Seismic Analysis 218 6.3.3.11 MODE 4 (Hot Standby) LOCA Evaluation 218 6.3.3.12 Alternate RCS Injection (BDB) 219 6.3.4 MINIMUM OPERATING CONDITIONS 220 6.3.5 TESTS AND INSPECTIONS 220 6.3.5.1 Inspection 220 6.3.5.2 System Testing 220 6.3.5.3 Components Testing 221 6.3.5.4 Operational Sequence Testing 222 6.3.5.5 Gas Intrusion Management Program 223 6.3.6 INSTRUMENTATION 223 6.3.6.1 Containment Sump Level 223 6.3.6.2 Refueling Water Storage Tank (RWST) Level 224 6.3.6.3 Accumulator Pressure and Level 224 6.3.6.4 Boric Acid Storage Tank Level 224 6.3.6.5 Residual Heat Exchanger Flow and Temperature 224 Page 8 of 14 Revision 26 5/2016 | |||
GINNA/UFSAR 6.3.6.6 Safety Injection Line Flow 224 6.3.6.7 Safety Injection Pumps Discharge Pressure 224 6.3.6.8 Pump Energization 225 6.3.6.9 Valve Position 225 Table 6.3-1 QUALITY STANDARDS OF SAFETY INJECTION SYSTEM COMPONENTS 228 Table 6.3-2 ACCUMULATOR DESIGN PARAMETERS 232 Table 6.3-3 SAFETY INJECTION SYSTEM PUMPS DESIGN PARAMETERS 233 Table 6.3-4 REFUELING WATER STORAGE TANK (RWST) DESIGN PARAMETERS 234 Table 6.3-5 RESIDUAL HEAT REMOVAL HEAT EXCHANGERS DESIGN PARAMETERS 235 Table 6.3-6 RECIRCULATION LOOP LEAKAGE INFORMATION USED IN 236 ORIGINAL ANALYSIS Table 6.3-7 INSTRUMENTATION READOUTS ON THE CONTROL BOARD FOR 237 OPERA- TOR MONITORING DURING RECIRCULATION Table 6.3-8 SAFETY INJECTION VALVE OPERATION AND INTERLOCKS 239 Table 6.3-9 SINGLE FAILURE ANALYSIS - SAFETY INJECTION SYSTEM 241 6.4 HABITABILITY SYSTEMS 243 6.4.1 DESIGN CRITERION 243 6.4.2 SYSTEM DESIGN 244 6.4.2.1 Definition of Control Room Envelope (CRE) 244 6.4.2.2 Ventilation System Design 244 6.4.2.2.1 Normal HVAC System - NORMAL and PURGE Modes of Operation 244 6.4.2.2.2 CREATS System - EMERGENCY Mode of Operation 244 6.4.2.3 Leak Tightness 245 6.4.2.4 Interaction with Other Zones and Pressure-containing Equipment. 245 6.4.2.4.1 Interaction with the Turbine Building 246 6.4.2.4.2 Interaction with the Relay Room 246 6.4.2.5 Shielding Design 246 6.4.2.6 System Operational Procedures 246 6.4.3 DESIGN EVALUATIONS 247 6.4.3.1 Radiological Analysis 247 6.4.3.2 Protection from Toxins 248 6.4.3.2.1 Chlorine 249 6.4.3.2.2 Ammonia 249 6.4.3.2.3 Halon 249 6.4.3.2.4 Refrigerant 249 6.4.3.2.5 Sodium Hypochlorite 249 6.4.3.2.6 Carbon Dioxide 250 6.4.3.3 Protection from Smoke and Fire 250 Page 9 of 14 Revision 26 5/2016 | |||
GINNA/UFSAR 6 ENGINEERED SAFETY FEATURES 1 6.4.3.3.1 Internal Sources of Smoke and Fire 250 6.4.3.3.2 External Sources of Smoke and Fire 250 6.4.3.4 Protection from Temperature Extremes 251 6.4.4 TESTS AND INSPECTIONS 251 6.4.5 INSTRUMENTATION REQUIREMENT 252 Table 6.4-1 Control Room Habitability Radiological Evaluation - Assumptions and Results 255 Table 6.4-2 CORE ACTIVITIES 257 6.5 FISSION PRODUCT REMOVAL SYSTEMS 258 6.5.1 ENGINEERED SAFETY FEATURE FILTER SYSTEMS 258 6.5.1.1 Introduction 258 6.5.1.2 Containment Air Filtration System 258 6.5.1.2.1 Design Basis 258 6.5.1.2.2 System Design 259 6.5.1.2.2.1 General Description 259 6.5.1.2.2.2 Charcoal Filters 259 6.5.1.2.2.3 HEPA Filters 260 6.5.1.2.2.4 Protection From Sodium Hydroxide Attack 260 6.5.1.2.2.5 Fire Protection 261 6.5.1.2.3 Design Evaluation 262 6.5.1.2.3.1 Decay Heat Generation in the Charcoal Filters 262 6.5.1.2.3.2 Decay Heat Dissipation With Normal Air Flow 262 6.5.1.2.3.3 Decay Heat Dissipation With Loss of Air Flow 262 6.5.1.2.4 Tests and Inspections 263 6.5.1.2.4.1 HEPA Filter Tests 263 6.5.1.2.4.2 Charcoal Filter Tests 264 6.5.1.2.4.3 System Tests 264 6.5.1.2.5 Instrumentation Requirements 264 6.5.1.3 Generic Letter 96-06 Requirements 265 6.5.1.4 Generic Letter 99-02 Requirements 266 6.5.2 CONTAINMENT SPRAY AND NaOH SYSTEMS 266 6.5.2.1 System Design and Operation 266 6.5.2.1.1 Spray Additive Tank 266 6.5.2.1.2 Effect of Sodium Hydroxide and Boric Acid Mixing 267 6.5.2.1.3 Iodine Retention 269 6.5.2.2 Iodine Effectiveness Evaluation of the Containment Spray and NaOH Systems 270 Page 10 of 14 Revision 26 5/2016 | |||
GINNA/UFSAR 6 ENGINEERED SAFETY FEATURES 1 6.5.2.2.1 Purpose of Chemical Modification 270 6.5.2.2.1.1 Thermal Capacity 270 6.5.2.2.1.2 Absorption of Iodine in Refueling Water Spray 270 6.5.2.2.1.3 Iodine Absorption with Sodium Hydroxide Addition 271 6.5.2.2.1.4 Spray Absorption Process for Iodine Removal 272 6.5.2.2.2 Technical Basis for Iodine Removal Factor 272 6.5.2.2.2.1 Analytical Model and Assumptions 272 6.5.2.2.2.2 Removal of Elemental Iodine 275 6.5.2.2.2.3 Removal of Other Airborne Forms of Iodine 276 6.5.2.2.2.4 Experimental Verification 276 Table 6.5-1 DATA FOR CHARCOAL FILTER EVALUATION 280 6.6 INSERVICE INSPECTION OF CLASS 2 AND 3 COMPONENTS 281 6. | |||
==6.1 INTRODUCTION== | |||
281 6.6.2 INSERVICE INSPECTION PROGRAM | |||
==SUMMARY== | |||
281 6.6.2.1 Scope 281 6.6.2.2 Inspection Intervals 281 6.6.2.3 Extent and Frequency 281 6.6.2.4 Examination Methods 282 6.6.2.5 Evaluation of Examination Results 282 6.6.2.6 System Pressure Testing 282 6.6.2.7 Records and Reports 282 6.6.2.8 Exemptions 282 FIGURES Figure 6.1-1 Design-Basis Accident, Containment Temperature Profile Figure 6.1-2 Design-Basis Accident, Containment Pressure Profile Figure 6.1-3 Postaccident Core Materials Design Conditions Figure 6.1-4 Containment Atmosphere Total Gamma Dose Figure 6.1-5 Containment Atmosphere Total Beta Dose Figure 6.1-6 pH of Unadjusted Boric Acid Solutions Figure 6.1-7 Titration Curve for Boric Acid With Sodium Hydroxide Figure 6.1-8 Temperature-Concentration Relation For Caustic Corrosion of Austenitic Stainless Steel Figure 6.1-9 Aluminum Corrosion Rates in LOCA Environment Figure 6.1-10 Post LOCA Containment Hydrogen Production Rate Figure 6.1-11 Boron Loss From Boron-Concrete Reaction Following a Design-Basis Accident Figure 6.1-12 Post LOCA Containment Hydrogen Production Figure 6.2-1 Containment Atmosphere Pressure, Double-Ended Hot Leg Break Figure 6.2-2 Containment Atmosphere Temperature, Double-Ended Hot Leg Break Figure 6.2-3 Containment Sump Temperature, Double-Ended Hot Leg Break Page 11 of 14 Revision 26 5/2016 | |||
GINNA/UFSAR Figure 6.2-4 Containment Atmosphere Pressure, Double-Ended Pump Suction Break - | |||
Minimum Safeguards Figure 6.2-5 Containment Atmosphere Temperature, Double-Ended Pump Suction Break - | |||
Mini- mum Safeguards Figure 6.2-6 Containment Sump Temperature, Double-Ended Pump Suction Break - | |||
Minimum Safeguards Figure 6.2-7 1.1-Ft2 Break Case 25B With 102% Power and Diesel Failure Assumed, Containment Pressure Versus Time Figure 6.2-8 1.4-Ft2 Break Case 13A With 102% Power and Vital Bus Failure Assumed, Contain- ment Steam Temperature Versus Time Figure 6.2-9 1.1-Ft2 Break Case 25B With 102% Power and Diesel Failure Assumed, Containment Steam Temperature Versus Time Figure 6.2-10 Reactor Containment Fan Cooler, Accident Versus Normal Air Flow Figure 6.2-11 Figure DELETED Figure 6.2-12 Figure Deleted Figure 6.2-13 Steam Generator Inspection and Maintenance Cabling Access Penetration 2 Figure 6.2-13a Fuel Transfer Tube Penetration 29 Figure 6.2-14 Reactor Coolant System Charging Line Penetration 100 Figure 6.2-15 Safety Injection System Pentrations 101, 110b, and 113 Figure 6.2-16 Alternate Charging Line Penetration 102 Figure 6.2-17 Construction Fire Service Water Penetration 103 Figure 6.2-18 Containment Spray Header A Penetration 105 Figure 6.2-19 Reactor Coolant Pump A Seal Water Line Penetration 106 Figure 6.2-20 Sump A Discharge Penetration 107 Figure 6.2-21 Reactor Coolant Pump Seal Water Return and Excess Letdown Penetration 108 Figure 6.2-22 Containment Spray Header B Penetration 109 Figure 6.2-23 Reactor Coolant Pump B Seal Water Line Penetration 110a Figure 6.2-24 Residual Heat Removal to Loop B Cold Leg Penetration 111 Figure 6.2-25 Letdown Line from Reactor Coolant System Penetration 112 Figure 6.2-26 Standby Auxiliary Feedwater to Steam Generators A and B Penetrations 119 and 123b Figure 6.2-27 Nitrogen to Accumulators Penetration 120a Figure 6.2-28 Pressurizer Relief Tank Gas Analyzer Penetration 120b Figure 6.2-29 Pressurizer Relief Tank Makeup Water Penetration 121a Figure 6.2-30 Pressurizer Relief Tank N2 Penetration 121b Figure 6.2-31 Containment Pressure Transmitters PT-945 and PT-946 Penetration 121c Figure 6.2-32 Reactor Coolant Drain Tank to Gas Analyzer Penetration 123a Figure 6.2-33 Component Cooling Water to and from Excess Letdown Heat Exchanger Penetrations 124a and 124c Figure 6.2-34 Containment Postaccident Air Sample (C Fan) Penetrations 124b and 124d Figure 6.2-35 Component Cooling Water from Reactor Coolant Pump 1B Penetration 125 Figure 6.2-36 Component Cooling Water from Reactor Coolant Pump 1A Penetration 126 Page 12 of 14 Revision 26 5/2016 | |||
GINNA/UFSAR Figure 6.2-37 Component Cooling Water to Reactor Coolant Pump 1A Penetration 127 Figure 6.2-38 Component Cooling Water to Reactor Coolant Pump B Penetration 128 Figure 6.2-39 Reactor Coolant Drain Tank Gas Header Penetration 129 Figure 6.2-40 Component Cooling Water from and to Reactor Support Coolers Penetrations 130 and 131 Figure 6.2-41 Mini-Purge Exhaust Penetration 132 Figure 6.2-42 Residual Heat Removal from Loop A Hot Leg Penetration 140 Figure 6.2-43 Sump B to Reactor Coolant Drain Tank Pump A Penetration 141 Figure 6.2-44 Sump B to Reactor Coolant Drain Tank Pump B Penetration 142 Figure 6.2-45 Reactor Coolant Drain Tank Discharge Penetration 143 Figure 6.2-46 Reactor Compartment Cooling Unit A Supply and Return Penetrations 201a and 209b Figure 6.2-47 Reactor Compartment Cooling Unit B Supply and Return Penetrations 201b and 209a Figure 6.2-48 Hydrogen Recombiner B (Main and Pilot) Penetrations 202a and 202b Figure 6.2-49 Containment Pressure Transmitters PT-947 and PT-948 Penetration 203a Figure 6.2-50 Figure DELETED Figure 6.2-51 Purge Supply Penetration 204 Figure 6.2-52 Reactor Coolant System Loop B Hot Leg Sample Penetration 205 Figure 6.2-53 Pressurizer Liquid Sample Penetration 206a Figure 6.2-54 Steam Generator A Sample Penetration 206b Figure 6.2-55 Pressurizer Steam Sample Penetration 207a Figure 6.2-56 Steam Generator B Sample Penetration 207b Figure 6.2-57 Hydrogen Recombiner A and B Oxygen Makeup Penetration 210 Figure 6.2-58 Purge Exhaust Penetration 300 Figure 6.2-59 Figure Deleted Figure 6.2-60 Hydrogen Recombiner A (Main and Pilot) Penetrations 304a and 304b Figure 6.2-61 Containment Postaccident Air Sample Penetrations 305a, 305c, and 305d Figure 6.2-62 Containment Air Sample (Return) Penetration 305b Figure 6.2-63 Containment Air Sample Outlet Penetration 305e Figure 6.2-64 Fire Service Water Penetration 307 Figure 6.2-65 Service Water for Containment Fan Coolers, Penetrations 308, 311, 312, 315, 316, 319, 320, and 323 Figure 6.2-66 Mini-Purge Supply Penetration 309 Figure 6.2-67 Instrument Air Penetration 310a Figure 6.2-68 Service Air Penetration 310b Figure 6.2-69 Leakage Test Depressurization Penetration 313 Figure 6.2-70 Leakage Test Supply Penetration 317 Figure 6.2-71 Steam Generator A Blowdown Penetration 321 Figure 6.2-72 Steam Generator B Blowdown Penetration 322 Page 13 of 14 Revision 26 5/2016 | |||
GINNA/UFSAR Figure 6.2-73 Demineralized Water Penetration 324 Figure 6.2-74 Containment H2 Monitors Penetrations 332a, 332b, and 332d Figure 6.2-75 Containment Pressure Transmitters PT-944, PT-949, and PT-950 Penetration 332c Figure 6.2-76 Main Steam from Steam Generator A Penetration 401 Figure 6.2-77 Main Steam from Steam Generator B Penetration 402 Figure 6.2-78 Main and Auxiliary Feedwater to Steam Generators A and B Penetrations 403 and 404 Figure 6.2-79 Figure DELETED Sheet 1 - | |||
Figure 6.2-79 Figure DELETED Sheet 2 - | |||
Figure 6.2-80 Containment Hydrogen Production With and Without Recombiner Figure 6.3-1 Sheet Figure DELETED 1-Figure 6.3-1 Sheet Figure DELETED 2-Figure 6.3-2 Safety Injection Pump Performance Characteristics Figure 6.3-3 Residual Heat Removal Pump Reactor Injection Capability Figure 6.3-4 Range of Core Protection Provided by Various Components of the Safety Injection Sys- tem Figure 6.5-1 Carbon Cell Banking Arrangement Figure 6.5-2 Filters - Containment Unit Figure 6.5-3 Iodine Partition Coefficient and pH in the Containment Versus Time Figure 6.5-4 Pressure Dependence of the Ratio vG/t Page 14 of 14 Revision 26 5/2016 | |||
GINNA/UFSAR 7 INSTRUMENTATION AND CONTROLS 1 | |||
==7.1 INTRODUCTION== | |||
2 7.1.1 IDENTIFICATION OF SAFETY-RELATED SYSTEMS 2 7.1.2 IDENTIFICATION OF SAFETY CRITERIA 3 7.1.2.1 General Design Criteria 3 7.1.2.2 Compliance with IEEE 279-1971 3 7.1.2.2.1 Design Basis 3 7.1.2.2.2 Requirements 4 7.1.2.2.2.1 Operability 4 7.1.2.2.2.2 Testability 5 7.1.2.2.2.3 Control of Protective Actions 5 7.2 REACTOR TRIP SYSTEM (RTS) 8 7.2.1 DESIGN BASES 8 7.2.1.1 Design Criteria 8 7.2.1.1.1 Fuel Damage Limits 8 7.2.1.1.2 Reliability and Testability 9 7.2.1.1.3 Redundancy and Independence 10 7.2.1.1.4 Effects of Adverse Conditions 10 7.2.1.1.5 Testing While In Operation 11 7.2.1.1.6 Fail Safe Design 11 7.2.1.1.7 Single Failure Criterion 11 7.2.1.2 Seismic Design 12 7.2.1.3 Operating Environment 12 7. | |||
==2.2 DESCRIPTION== | |||
13 7.2.2.1 Logic Train 14 7.2.2.1.1 Sensors 14 7.2.2.1.2 Process and Nuclear Instrumentation 15 7.2.2.1.3 Protection Cabinets 15 7.2.2.1.4 Logic Relay Cabinets 15 7.2.2.1.5 Trip Breakers 16 7.2.2.2 Reactor Trips 17 7.2.2.2.1 General 17 7.2.2.2.2 Manual Trip 17 7.2.2.2.3 High-Nuclear-Flux (Power Range) Trip 17 7.2.2.2.4 High-Nuclear-Flux (Intermediate Range) Trip 18 Page 1 of 9 Revision 26 5/2016 | |||
GINNA/UFSAR 7 INSTRUMENTATION AND CONTROLS 1 7.2.2.2.5 High-Nuclear-Flux (Source Range) Trip 18 7.2.2.2.6 Overtemperature Delta T Trip 18 7.2.2.2.7 Overpower Delta T Trip 18 7.2.2.2.8 Low Pressurizer Pressure Trip 18 7.2.2.2.9 High Pressurizer Pressure Trip 19 7.2.2.2.10 High Pressurizer Water Level Trip 19 7.2.2.2.11 Low Reactor Coolant Flow Trip 19 7.2.2.2.12 Safety Injection System Actuation Trip 20 7.2.2.2.13 Turbine Trip/Reactor Trip 20 7.2.2.2.14 Low-Low Steam-Generator Water Level Trip 20 7.2.2.3 Interlocks 20 7.2.2.4 Permissive Circuits 21 7.2.2.4.1 P-1 Permissive 21 7.2.2.4.2 P-2 Permissive 21 7.2.2.4.3 P-3 Permissive 21 7.2.2.4.4 P-4 Permissive 21 7.2.2.4.5 P-6 Permissive 21 7.2.2.4.6 P-7 Permissive 22 7.2.2.4.7 P-8 Permissive 22 7.2.2.4.8 P-9 Permissive 22 7.2.2.4.9 P-10 Permissive 22 7.2.2.5 Alarms 22 7.2.2.6 Design Features 23 7.2.2.6.1 Isolation of Redundant Protection Channels 23 7.2.2.6.1.1 Channelized Design 23 7.2.2.6.1.2 Separation 24 7.2.2.6.2 Channel Bypass or Removal from Operation 24 7.2.2.6.3 Capability for Test and Calibration 24 7.2.2.6.4 Information Readout and Indication of Bypass 25 7.2.2.6.5 Physical Isolation 25 7.2.2.6.6 Sensor Line Separation 25 7.2.2.6.7 Instrument Line Identification 26 7.2.3 ANALYSIS 26 7.2.3.1 Reactor Trip System (RTS) and Departure From Nucleate Boiling 26 7.2.3.2 Core Protection System 26 Page 2 of 9 Revision 26 5/2016 | |||
GINNA/UFSAR 7 INSTRUMENTATION AND CONTROLS 1 7.2.3.2.1 Overpower Protection 26 7.2.3.2.2 Overtemperature Protection 27 7.2.4 REACTOR TRIP SIGNAL TESTING 27 7.2.4.1 Analog Channel Testing 27 7.2.4.2 Logic Channel Testing 28 7.2.4.2.1 Planned Tests 28 7.2.4.2.2 Test Procedure 29 7.2.4.2.3 Logic Channel Test Panels 29 7.2.4.3 Trip Breaker Testing and Preventive Maintenance 30 7.2.5 INTERACTION OF CONTROL AND PROTECTION SYSTEMS 30 7.2.5.1 Introduction 30 7.2.5.2 Specific Control and Protection Interactions 30 7.2.5.2.1 Nuclear Flux 30 7.2.5.2.2 Coolant Temperature 31 7.2.5.2.3 Pressurizer Pressure 31 7.2.5.2.4 Pressurizer Level 32 7.2.6 ANTICIPATED-TRANSIENT-WITHOUT-SCRAM MITIGATION 33 SYSTEM ACTUATION CIRCUITRY Table 7.2-1 Table DELETED 36 Table 7.2-2 PERMISSIVE CIRCUITS 37 Table 7.2-3 REACTOR TRIP FUNCTION SETPOINTS 38 7.3 ENGINEERED SAFETY FEATURES SYSTEMS 40 7.3.1 DESIGN CRITERIA 40 7.3.1.1 Protection Systems 40 7.3.1.2 Redundancy and Independence 41 7.3.1.3 Testing While In Operation 42 7.3.1.4 Fail Safe Design 42 7.3.2 SYSTEM DESCRIPTION 43 7.3.2.1 Initiating Circuitry 43 7.3.2.2 System Functions 44 7.3.2.2.1 Steam Line Isolation 44 7.3.2.2.2 Feedwater Line Isolation 44 7.3.2.3 Sensing and Display Instrumentation 45 7.3.2.3.1 Reactor Vessel Level Indication System 45 7.3.2.3.2 Containment Pressure 45 7.3.2.3.3 Containment Sump Level 46 Page 3 of 9 Revision 26 5/2016 | |||
GINNA/UFSAR 7.3.2.3.4 Accumulator Level and Pressure 46 7.3.2.3.5 Refueling Water Storage Tank Level (RWST) 46 7.3.2.3.6 Sodium Hydroxide Tank Level and Flow 46 7.3.2.3.7 Safety Injection Pumps Discharge Pressure and Flow 46 7.3.2.3.8 Residual Heat Removal (Low-Head Safety Injection) Flow 46 7.3.2.3.9 Pump Energization 46 7.3.2.3.10 Valve Position 46 7.3.2.3.11 Residual Heat Exchangers 47 7.3.2.3.12 Alarms 47 7.3.2.3.13 Air Coolers 47 7.3.2.3.14 Local Instrumentation 47 7.3.2.4 Engineered Safety Features Reset Controls 47 7.3.3 DESIGN EVALUATION 48 7.3.3.1 Engineered Safety Features Systems Isolation 48 7.3.3.2 Loss of Voltage or Degraded Voltage on Engineered Safety Features Bus 48 7.3.4 TESTING 48 7.3.4.1 Analog Channel Testing 48 7.3.4.2 Logic Channel Testing 49 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 51 7. | |||
==4.1 DESCRIPTION== | |||
51 7.4.1.1 Reactor Trip System (RTS) 51 7.4.1.2 Auxiliary Feedwater Systems 52 7.4.1.3 Main Steam System 52 7.4.1.4 Service Water System 53 7.4.1.5 Chemical and Volume Control System 53 7.4.1.6 Component Cooling Water System (CCW) 54 7.4.1.7 Residual Heat Removal System 54 7.4.1.8 Electrical Instrumentation and Power Systems 55 7.4.2 EVALUATION 55 7.4.3 EMERGENCY SHUTDOWN CONTROL 55 7.4.3.1 General 55 7.4.3.2 Residual Heat Removal 56 7.4.3.3 Reactivity Control 57 7.4.3.4 Pressurizer Pressure and Level Control 57 7.4.3.5 Electrical Systems 57 7.4.3.6 Startup of Other Equipment 58 7.4.3.7 Indication and Controls Provided Outside the Control Room 58 7.4.3.7.1 Local Panel Indication 58 7.4.3.7.2 Local Motor Controls 59 Page 4 of 9 Revision 26 5/2016 | |||
GINNA/UFSAR 7.4.3.7.3 Valve Control 59 7.4.3.7.4 Pressurizer Heater Control 60 7.4.3.7.5 Lighting 60 7.4.3.7.6 Communications 60 7.4.3.7.7 Electrical Systems 60 7.4.4 ALTERNATIVE SHUTDOWN SYSTEM 60 7.4.4.1 System Description 60 7.4.4.2 Alternative Shutdown Stations 61 7.4.4.2.1 Charging Pump Room (Primary Station) (see Section 7.4.3.7.1 F) 61 7.4.4.2.2 Intermediate Building North (Primary Station) (see Section 7.4.3.7.1 E) 62 7.4.4.2.3 Emergency Diesel Generator Area (Support Station) (see Section 7.4.3.7.7) 62 7.4.4.2.4 480-Volt Alternating Current Bus 14 (Support Station) 62 7.4.4.2.5 Battery Rooms 1A and 1B (Support Station) 62 7.4.4.2.6 Motor Control Centers 1C and 1D (Support Station) 62 7.4.4.2.7 480-Volt Alternating Current Bus 18 (Support Station) 63 7.4.4.2.8 Selected Safe Shutdown Systems 63 Table 7.4-1 FUNCTIONS FOR SHUTDOWN AND COOLDOWN 65 Table 7.4-2 SAFE SHUTDOWN INSTRUMENTS 66 Table 7.4-3 SAFE SHUTDOWN SYSTEMS POWER SOURCE AND LOCATION 69 Table 7.4-4 APPENDIX R ALTERNATIVE SHUTDOWN METHODS AND 71 CONTROL LOCATIONS 7.5 SAFETY-RELATED DISPLAY INSTRUMENTATION 74 7.5.1 CONTROL ROOM 74 7.5.1.1 Description 74 7.5.1.1.1 General 74 7.5.1.1.2 Main Control Board 74 7.5.1.1.3 Other Control Room Displays 75 7.5.1.2 Design Review 76 7.5.2 SAFETY PARAMETER DISPLAY 76 Table 7.5-1 COMPARISON OF GINNA STATION POSTACCIDENT 78 INSTRUMENTATION TO REGULATORY GUIDE 1.97, REVISION 3, CRITERIA 7.6 OTHER INSTRUMENTATION SYSTEMS REQUIRED FOR SAFETY 94 7.6.1 OVERPRESSURE PROTECTION DURING LOW POWER OPERATION 94 7.6.2 AUXILIARY FEEDWATER SYSTEM AUTOMATIC INITIATION AND 95 FLOW INDICATION 7.6.3 SUBCOOLING METER 95 7.6.4 DIRECT CURRENT POWER SYSTEM BUS VOLTAGE MONITORING 95 AND ANNUNCIATION Page 5 of 9 Revision 26 5/2016 | |||
GINNA/UFSAR 7 INSTRUMENTATION AND CONTROLS 1 7.6.5 REACTOR VESSEL LEVEL INDICATION SYSTEM 96 7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY 99 7. | |||
==7.1 DESCRIPTION== | |||
99 7.7.1.1 General 99 7.7.1.1.1 Reactor Control System 99 7.7.1.1.2 Steam Dump Control System 99 7.7.1.1.3 Reactivity Control 100 7.7.1.1.4 Reactor Control System Operation 100 7.7.1.1.5 Pressurizer Pressure and Water Level Control System 101 7.7.1.1.6 Steam Dump System 101 7.7.1.2 Rod Control System 102 7.7.1.2.1 Control Group Control 102 7.7.1.2.1.1 General 102 7.7.1.2.1.2 Rod Control Input Signals 102 7.7.1.2.1.3 Rod Control Program 103 7.7.1.2.2 Shutdown Group Control 104 7.7.1.2.3 Control Rod Drive Performance 104 7.7.1.2.4 Control Rod Power Supply System 104 7.7.1.2.4.1 General 104 7.7.1.2.4.2 Control Rod Power Supply Connections 105 7.7.1.2.5 Control Rod Power Supply Evaluation 106 7.7.1.2.5.1 Alternating Current Power Connections 106 7.7.1.2.5.2 Direct Current Power Connections 107 7.7.1.2.5.3 Evaluation Summary 108 7.7.1.2.6 Rod Position Indication System 109 7.7.1.2.6.1 Microprocessor System 109 7.7.1.2.6.2 Digital System 110 7.7.1.2.6.3 Actual Position Indication 110 7.7.1.2.6.4 Demand Position Indication 110 7.7.1.2.6.5 Rod Deviation Alarm 110 7.7.1.2.7 Pulse-to-Analog Converter 111 7.7.1.2.8 Interlocks and Rod Stops 111 7.7.1.2.9 Rod Insertion Limit Circuit 112 7.7.1.2.10 Rod Drop Protection 112 7.7.1.2.11 Asymmetric Rod Cluster Control Assembly Withdrawal 113 Page 6 of 9 Revision 26 5/2016 | |||
GINNA/UFSAR 7 INSTRUMENTATION AND CONTROLS 1 7.7.1.2.12 Rod Control Cabinet Cooling 114 7.7.1.3 Pressurizer Pressure and Level Control 114 7.7.1.3.1 Pressure Control 114 7.7.1.3.2 Level Control 115 7.7.1.4 Turbine Bypass 115 7.7.1.5 Steam Generator Level Control 116 7.7.1.6 Steam Generator Overfill Protection 118 7.7.2 CONTROL SYSTEM EVALUATION 118 7.7.2.1 Plant Stability 118 7.7.2.2 Step Load Changes Without Turbine Bypass 119 7.7.2.3 Loading and Unloading 119 7.7.2.4 Loss of Load With Turbine Bypass 120 7.7.2.5 Turbine Trip With Reactor Trip 120 7.7.2.6 Control Rod Misalignment 121 7.7.2.6.1 General 121 7.7.2.6.2 Consequences of Rod Misalignment 121 7.7.2.6.3 Analysis of Control Rod Misalignment 122 7.7.2.6.4 Redundant Checks for Control Rod Malfunction 122 7.7.2.6.4.1 Operator Checks 122 7.7.2.6.4.2 Additional Periodic Tests 123 7.7.2.6.4.3 Details of Instrumentation System 123 7.7.2.6.4.4 Power Range Nuclear Instrumentation 123 7.7.2.6.4.5 Thermocouples 124 7.7.2.6.4.6 In-Core Movable Detectors 124 7.7.2.6.4.7 Summary 125 7.7.2.6.5 Expected Instrument Response to Control Rod Misalignment Ginna Station 125 7.7.2.6.6 Plant Startup Tests 126 7.7.3 NUCLEAR INSTRUMENTATION SYSTEM 126 7.7.3.1 Design Basis 126 7.7.3.2 System Design 127 7.7.3.2.1 Source Range Description 128 7.7.3.2.2 Intermediate Range Description 129 7.7.3.2.3 Power Range Description 130 7.7.3.2.4 Dropped Rod Protection 131 7.7.3.2.5 Audio Count Rate Channel 131 Page 7 of 9 Revision 26 5/2016 | |||
GINNA/UFSAR 7 INSTRUMENTATION AND CONTROLS 1 7.7.3.2.6 Recorders 131 7.7.3.2.7 Power Supply 132 7.7.3.2.8 Equipment Locations 132 7.7.3.3 System Evaluation 132 7.7.4 IN-CORE INSTRUMENTATION 132 7.7.4.1 Design Basis 132 7.7.4.2 System Design 132 7.7.4.2.1 General 132 7.7.4.2.2 Thermocouples 133 7.7.4.2.3 Movable Miniature Neutron Flux Detectors 133 7.7.4.2.4 Control and Readout System 134 7.7.5 REACTOR COOLANT TEMPERATURE INDICATION 135 7.7.6 PLANT PROCESS COMPUTER SYSTEM AND SAFETY ASSESSMENT SYS- 136 TEM 7.7.6.1 General 136 7.7.6.2 Plant Process Computer System 137 7.7.6.3 Safety Parameter Display System 138 Table 7.7-1 OUT-OF-PHASE CURRENTS (AMPS) 140 Table 7.7-2 ROD STOPS 141 Table 7.7-3 EXPECTED MAXIMUM VARIATIONS BETWEEN 142 SYMMETRICALLY LOCATED DETECTORS FIGURES Figure 7.2-1 Reactor Protection Systems Figure 7.2-2 Figure DELETED Figure 7.2-3 Figure DELETED Figure 7.2-4 Figure DELETED Figure 7.2-5 Figure DELETED Figure 7.2-6 Figure DELETED Figure 7.2-7 Figure DELETED Figure 7.2-8 Figure DELETED Figure 7.2-9 Figure DELETED Figure 7.2-10 Figure DELETED Figure 7.2-11 Figure DELETED Figure 7.2-12 Design Philosophy to Achieve Isolation Between Channels Figure 7.2-13 Figure DELETED Figure 7.2-14 Channel Configuration (Channel 1 Typical) Tavg / T Control and Protection System Page 8 of 9 Revision 26 5/2016 | |||
GINNA/UFSAR Figure 7.2-15 Analog System Symbols Figure 7.2-16 Analog Channel Testing Arrangement Figure 7.2-17 Trip Logic Channels Figure 7.2-18 Analog Channels Figure 7.2-19 Logic Channel Test Panels Figure 7.2-20 Electrical Diagram - Undervoltage Coil and Shunt Trip Assembly Figure 7.3-1 Sheet 1 - Figure DELETED Figure 7.3-1 Sheet 2 - Figure DELETED Figure 7.3-2 Figure DELETED Figure 7.3-3 Figure DELETED Figure 7.3-4 Actuation Circuits of Engineered Safety Features Figure 7.3-5 Figure DELETED Figure 7.3-6 Analog and Logic Channel Testing Figure 7.5-1 Control Room Layout Figure 7.7-1 Reactor Control System Figure 7.7-2 Simplified Block Diagram of Reactor Control Systems Figure 7.7-3 Control Group - Rod Drive System Figure 7.7-4 Power Supply to Rod Control Equipment and Control Rod Drive Mechanisms Figure 7.7-4a Illustration of MRPI Indication Figure 7.7-5 Figure DELETED Figure 7.7-6 Nuclear Protection System Figure 7.7-7 Power Range Nuclear Detector Locations Figure 7.7-8 Location of Control Rods and Instrumentation Figure 7.7-9 Power Range Nuclear Instrumentation System Figure 7.7-10 Neutron Detectors and Range of Operation Figure 7.7-11 In-Core Instrumentation - Assembly Figure 7.7-12 Sheet 1 - In-Core Instrumentation, Details Figure 7.7-12 Sheet 2 - In-Core Instrumentation, Details Figure 7.7-13 Typical Arrangement of Moveable Miniature Neutron Flux Detector System Figure 7.7-14 Advanced Digital Feedwater Control System Input Signal Validation Figure 7.7-15 Advanced Digital Feedwater Control System Flow Controller and Cv Demand Figure 7.7-16 Advanced Digital Feedwater Control System Valve Sequence and Tracking Logic Page 9 of 9 Revision 26 5/2016 | |||
GINNA/UFSAR 8 ELECTRIC POWER 1 | |||
==8.1 INTRODUCTION== | |||
2 8.1.1 GENERAL 2 8.1.2 OFFSITE POWER DESCRIPTION 2 8.1.3 ONSITE POWER DESCRIPTION 3 8.1.4 PRINCIPAL DESIGN CRITERIA 3 8.1.4.1 Performance Standards 3 8.1.4.2 Emergency Power 4 8.1.4.3 Adequacy of Electrical Design Relative to 1972 Criteria 5 8.1.4.4 Potential Risk of Station Blackout 6 8.1.4.5 Station Blackout Program 7 8.1.4.5.1 Assumptions 7 8.1.4.5.2 Ventilation 8 8.1.4.5.3 Plant Classification 8 8.1.4.5.4 Diesel Generator Reliability 8 8.1.4.5.5 Diesel Generator Cold Starts 9 8.1.4.6 Fukushima - Diverse and Flexible Coping Strategies (FLEX) 9 8.1.4.6.1 Regulatory Requirements 9 8.1.4.6.2 Site Response to NRC Orders 9 8.1.4.6.3 Summary of FLEX Strategies 10 8.1.4.6.3.1 Reactor Core Cooling and Heat Removal 10 8.1.4.6.3.2 RCS Inventory and Reactivity Control 10 8.1.4.6.3.3 Containment Integrity 11 8.1.4.6.3.4 Spent Fuel Pool Cooling 12 8.1.4.6.3.5 Electric Power 12 8.2 OFFSITE POWER SYSTEM 15 8. | |||
==2.1 DESCRIPTION== | |||
15 8.2.1.1 Transmission System 15 8.2.1.1.1 Step-up Transformers 15 8.2.1.1.2 Transmission Lines 15 8.2.1.1.3 Circuit Breakers 16 Page 1 of 4 Revision 26 5/2016 | |||
GINNA/UFSAR 8.2.1.1.4 Protective Relay Circuits 16 8.2.1.2 Station Auxiliary (Startup) Transformers 12A and 12B 17 8.2.2 ANALYSIS 19 8.2.2.1 Transmission System 19 8.2.2.1.1 Loss of Ginna Station Output 19 8.2.2.1.2 Switchyard Direct Current Power System 19 8.2.2.1.3 Transmission Network Protective Features 19 8.2.2.1.4 Northeast Power Coordinating Council Load-Shedding Practice 20 8.2.2.2 Station Auxiliary (Startup) Transformers 12A and 12B 21 8.2.2.2.1 Original Ginna Station Design 21 8.2.2.2.2 Transformer Failure Rates 21 8.2.2.2.3 Backup Auxiliary Transformers 22 8.2.2.3 Radiation Exposure During Restoration of Power 23 Table 8.2-1 DIRECT RADIATION DOSE RATES1 (REM/HR) 25 Table 8.2-2 INHALATION DOSE RATES2 (REM/SEC) 26 8.3 ONSITE POWER SYSTEM 27 8.3.1 ALTERNATING CURRENT POWER SYSTEM 27 8.3.1.1 Description 27 8.3.1.1.1 Single-Line Diagrams 27 8.3.1.1.2 Station Unit Transformer 27 8.3.1.1.3 The 4160-Volt System 28 8.3.1.1.4 The 480-Volt System 28 8.3.1.1.4.1 480-Volt Buses 28 8.3.1.1.4.2 Class 1E Trains 29 8.3.1.1.5 The 120-Volt Alternating Current System 30 8.3.1.1.5.1 Instrument Bus 1A 30 8.3.1.1.5.2 Instrument Bus 1B 31 8.3.1.1.5.3 Instrument Bus 1C 31 8.3.1.1.5.4 Instrument Bus 1D 31 8.3.1.1.6 Emergency Power 32 8.3.1.1.6.1 Emergency Power Sources 32 Page 2 of 4 Revision 26 5/2016 | |||
GINNA/UFSAR 8.3.1.1.6.2 Diesel-Generator Rapid Startup and Loading 32 8.3.1.1.6.3 Diesel-Generator Protective Trips 33 8.3.1.1.6.4 Fuel Oil Supply 33 8.3.1.1.6.5 Diesel-Generator Startup Logic 34 8.3.1.1.6.6 Emergency Power Supply 34 8.3.1.1.6.7 Alternative Shutdown Provisions 36 8.3.1.1.6.8 Regulatory Review of Diesel-Generator Capability 36 8.3.1.2 Analysis 36 8.3.1.2.1 Evaluation of Layout and Load Distribution 36 8.3.1.2.2 Diesel Generators 37 8.3.1.2.3 Normal Power Sources 37 8.3.1.2.4 Reliability Assurance 37 8.3.1.2.4.1 Redundancy 37 8.3.1.2.4.2 Sequencing Circuits 37 8.3.1.2.4.3 Sequencing Relays 38 8.3.1.2.4.4 Engineered Safety Features Actuation 39 8.3.1.2.4.5 Separation 39 8.3.1.2.4.6 Fuse Coordination 40 8.3.1.2.4.7 Overload and Short Circuit Protection 40 8.3.1.2.5 Instrument Bus Evaluation 40 8.3.1.2.6 Loss of Offsite Power Under Accident Conditions 41 8.3.1.2.6.1 Operator Actions 41 8.3.1.2.6.2 Reliability Assurance 42 8.3.1.2.7 Degraded Grid Voltage 43 8.3.1.2.7.1 Susceptibility to Degraded Grid Voltage Conditions 43 8.3.1.2.7.2 Adequacy of Onsite Power System Voltages 44 8.3.1.3 Containment Electrical Penetrations 44 8.3.1.4 Independence of Redundant Systems 45 8.3.1.4.1 Criteria Relating to Cable-Tray Loading and Separation 45 8.3.1.4.2 Separation of Redundant Circuits 46 8.3.1.4.3 Quality Assurance 47 Page 3 of 4 Revision 26 5/2016 | |||
GINNA/UFSAR 8.3.2 DIRECT CURRENT POWER SYSTEMS 47 8.3.2.1 Description 47 8.3.2.1.1 Direct Current System 47 8.3.2.1.2 Battery Room 48 8.3.2.1.3 Battery Chargers 48 8.3.2.1.4 Technical Support Center Battery 49 8.3.2.2 Analysis 49 8.3.2.3 Direct Current Fuse Coordination 50 8.3.3 FIRE PROTECTION FOR CABLE SYSTEMS 52 Table 8.3-1a ENGINEERED SAFETY FEATURES ACTUATION (ESFAS) 56 SEQUENCE ACTION (TRAIN A) | |||
Table 8.3-1b ENGINEERED SAFETY FEATURES ACTUATION (ESFAS) 57 SEQUENCE ACTION - DIESEL GENERATOR LOADING Table 8.3-2a DIESEL GENERATOR LOADING (TRAIN A) 58 Table 8.3-2b DIESEL GENERATOR LOADING (TRAIN B) 59 Table 8.3-3 CONTAINMENT ELECTRICAL PENETRATIONS 60 Table 8.3-4 MAJOR BATTERY LOADS 62 | |||
: 1. At listed locations following a loss-of-coolant accident. | |||
: 2. Within 200 ft of containment following a loss-of-coolant accident. | |||
FIGURES Figure 8.1-1 Electrical Distribution System Figure 8.2-1 Transmission Connections Page 4 of 4 Revision 26 5/2016 | |||
GINNA/UFSAR 9 AUXILIARY SYSTEMS 1 9.1 FUEL STORAGE AND HANDLING 2 9.1.1 NEW FUEL STORAGE 2 9.1.2 SPENT FUEL STORAGE 2 9.1.2.1 Design Criteria 3 9.1.2.1.1 General 3 9.1.2.1.2 Effective Multiplication Factor 4 9.1.2.1.3 Protection Against Damage 4 9.1.2.1.4 Storage Capacity 4 9.1.2.1.5 Fuel Pool Cooling System Instrumentation 5 9.1.2.1.6 Seismic Design 5 9.1.2.1.7 Fuel Handling System 6 9.1.2.1.8 Minimum Center-to-Center Spacing 6 9.1.2.1.9 Stability of Fuel Storage Racks 6 9.1.2.1.10 Fuel Pool Leakage Prevention 7 9.1.2.1.11 Depth of Water Over Fuel 7 9.1.2.1.12 Fixed Neutron Poisons 7 9.1.2.1.13 Bearing Loads on Pool Liner 8 9.1.2.2 Description 8 9.1.2.2.1 Spent Fuel Pool (SFP) 8 9.1.2.2.2 Spent Fuel Storage Racks 9 9.1.2.3 Design Evaluation 10 9.1.2.4 Nuclear Analysis 11 9.1.2.4.1 Methods of Analysis 11 9.1.2.4.1.1 Criticality Methodology 14 9.1.2.4.1.2 Criticality Analysis of Consolidated Rod Storage Canisters in Spent 16 Fuel Racks 9.1.2.4.1.3 Summary of Criticality Results 17 9.1.2.4.2 Accident Analysis 19 9.1.2.4.2.1 Fresh Fuel Storage Racks 19 Page 1 of 15 Revision 26 4/2016 | |||
GINNA/UFSAR 9.1.2.4.2.2 Spent Fuel Storage Racks 19 9.1.2.5 Thermal-Hydraulic Analysis 20 9.1.2.6 Radiological Evaluation 20 9.1.2.7 Radiological Consequences of Tornado Missile Accident (TMA) 21 9.1.2.8 Radiological Consequences of a Dropped Consolidated Canister 21 9.1.3 SPENT FUEL POOL COOLING 22 9.1.3.1 Design Bases 22 9.1.3.2 System Design and Operation 24 9.1.3.2.1 System Design 24 9.1.3.2.2 System Operation 24 9.1.3.2.3 Suction Lineup Using Spent Fuel Pool (SFP) Pump A 25 9.1.3.3 Spent Fuel Pool (SFP) Cooling System Components 25 9.1.3.3.1 Spent Fuel Pool (SFP) Heat Exchangers 26 9.1.3.3.2 Spent Fuel Pool (SFP) Pumps 26 9.1.3.3.3 Spent Fuel Pool (SFP) Filter 26 9.1.3.3.4 Spent Fuel Pool (SFP) Strainer 26 9.1.3.3.5 Spent Fuel Pool (SFP) Demineralizer 26 9.1.3.3.6 Spent Fuel Pool (SFP) Skimmer 26 9.1.3.3.7 Spent Fuel Pool (SFP) Valves 26 9.1.3.3.8 Spent Fuel Pool (SFP) Piping 27 9.1.3.4 System Evaluation 27 9.1.3.4.1 Thermal-Hydraulic Analysis 27 9.1.3.4.1.1 Heat Removal Requirements 27 9.1.3.4.1.2 Service Water Temperature 27 9.1.3.4.1.3 Analysis of Heat Removal System 28 9.1.3.4.1.4 Cooling Water Flow in Fuel Pool 28 9.1.3.4.1.5 Cooling Analysis of Individual Fuel Assemblies 28 9.1.3.4.1.6 Cooling Analysis of Consolidated Fuel Canisters 30 9.1.3.4.1.7 Normal 1/3 Core Off-Load Cooling Capability 31 9.1.3.4.1.8 Full Core Off-Load Cooling Capability 32 9.1.3.4.2 Leakage Provisions 33 Page 2 of 15 Revision 26 4/2016 | |||
GINNA/UFSAR 9.1.3.4.3 Interruption of Spent Fuel Pool (SFP) Cooling 34 9.1.3.5 Minimum Operating Conditions 36 9.1.4 FUEL HANDLING SYSTEMS 36 9.1.4.1 Reactor Cavity 36 9.1.4.2 Refueling Canal 37 9.1.4.3 Fuel Handling Equipment 37 9.1.4.3.1 Auxiliary Building Crane 37 9.1.4.3.2 New Fuel Elevator 38 9.1.4.3.3 Spent Fuel Pool (SFP) Bridge 38 9.1.4.3.4 Fuel Transfer System 38 9.1.4.3.5 Manipulator Crane 39 9.1.4.3.6 Reactor Vessel Head Lifting Device 40 9.1.4.3.7 Reactor Internals Lifting Device 41 9.1.4.3.8 Rod Cluster Control Assembly Changing Fixture 41 9.1.4.3.9 Upper Internals Storage Stand 42 9.1.4.3.10 125 Ton Cask Handling Crane 42 9.1.4.4 Fuel Handling/Refueling Tools 42 9.1.4.4.1 New Fuel Assembly Handling Tool 42 9.1.4.4.2 Spent Fuel Handling Tool 42 9.1.4.4.2.1 Spent Fuel Cask Loading Tool (SFCLT) 43 9.1.4.4.3 Burnable Poison Rod Assembly Handling Tool 43 9.1.4.4.4 Control Rod Drive Shaft Tool 44 9.1.4.4.5 Thimble Plug Handling Tool 44 9.1.4.4.6 Irradiation Sample Handling Tool 44 9.1.4.4.7 Stud Tensioners 45 9.1.4.4.8 Portable Rod Cluster Control Assemble (RCCA) Tool 45 9.1.4.5 Fuel Handling System Operation During MODE 6 (Refueling) 45 9.1.4.5.1 Introduction 45 9.1.4.5.2 Preparation Phase 45 9.1.4.5.3 MODE 6 (Refueling) Phase 46 9.1.4.5.4 Reactor Reassembly 47 Page 3 of 15 Revision 26 4/2016 | |||
GINNA/UFSAR 9.1.4.6 Fuel Handling System Evaluation 47 9.1.4.6.1 Incident Protection 48 9.1.4.6.2 Malfunction Analysis 48 9.1.4.7 Minimum Operating Conditions 48 9.1.4.8 Tests and Inspections 48 9.1.5 CONTROL OF HEAVY LOADS 48 9.1.5.1 Conduct of Heavy Loads Movements 49 Table 9.1-1 FUEL PARAMETERS EMPLOYED IN THE CRITICALITY 57 ANALYSIS Table 9.1-2 TORNADO MISSILE ACCIDENT DOSE ANALYSIS 58 ASSUMPTIONS Table 9.1-2 TORNADO MISSILE ACCIDENT DOSE ANALYSIS 59 ASSUMPTIONS (OFFSITE X/Q) | |||
Table 9.1-2 TORNADO MISSILE ACCIDENT DOSE ANALYSIS 60 ASSUMPTIONS (OFFSITE BREATHING RATES) | |||
Table 9.1-3 SPENT FUEL POOL (SFP) COOLING SYSTEM RATING 61 Table 9.1-4 SPENT FUEL POOL (SFP) COOLING SYSTEM COMPONENT 62 DATA Table 9.1-5 OFFSITE AND CONTROL ROOM DOSES FOR THE SPENT FUEL 64 POOL TORNADO MISSILE ACCIDENT Table 9.1-6 HEAT-UP TIMES ASSOCIATED WITH LOSS OF SPENT FUEL 65 POOL COOLING 9.2 WATER SYSTEMS 66 9.2.1 SERVICE WATER (SW) SYSTEM 66 9.2.1.1 Design Bases 66 9.2.1.2 Description 66 9.2.1.2.1 General Description 66 9.2.1.2.2 Service Water System Design 68 9.2.1.2.3 Service Water System Initiation on Loss of Offsite Power 68 9.2.1.2.4 Containment Cooling Coils 69 9.2.1.2.5 Radiation Monitors 70 9.2.1.2.6 Service Water Fouling 70 9.2.1.3 Design Evaluation 71 9.2.1.4 Postaccident Conditions 72 9.2.1.4.1 Recirculation Phase 72 9.2.1.4.2 Limiting Steam Line Break Events 73 9.2.1.4.3 Accident Considerations With Offsite Power Available 73 9.2.1.4.4 Postulated Service Water Pump Discharge Check Valve Failure 74 Page 4 of 15 Revision 26 4/2016 | |||
GINNA/UFSAR 9.2.1.5 Tests and Inspections 74 9.2.2 COMPONENT COOLING WATER (CCW) SYSTEM 75 9.2.2.1 Design Bases 75 9.2.2.2 System Design and Operation 75 9.2.2.3 Component Description 77 9.2.2.4 System Evaluation 78 9.2.2.4.1 Availability and Reliability 78 9.2.2.4.1.1 Accessibility 78 9.2.2.4.1.2 Seismic Design 78 9.2.2.4.1.3 Loss of Component Cooling Water System 78 9.2.2.4.1.4 Component Cooling Water Surge Tank 79 9.2.2.4.1.5 Safety-Related Functions 79 9.2.2.4.1.6 Flow-Induced Vibration 80 9.2.2.4.2 Leakage Provisions 81 9.2.2.4.2.1 Introduction 81 9.2.2.4.2.2 Leakage Detection 81 9.2.2.4.2.3 Relief Valves 82 9.2.2.4.3 Incident Control 82 9.2.2.4.4 Malfunction Analysis 83 9.2.2.5 Instrumentation Requirements 84 9.2.2.6 Minimum Operating Conditions 84 9.2.2.7 Tests and Inspections 84 9.2.3 DEMINERALIZED WATER MAKEUP SYSTEM 84 9.2.4 CONDENSATE STORAGE FACILITIES 85 Table 9.2-1 LOADS SUPPLIED BY SERVICE WATER (SW) SYSTEM 88 Table 9.2-2 MAJOR SERVICE WATER SYSTEM FLOWS 90 Table 9.2-3 COMPONENT COOLING LOOP COMPONENT DATA 93 Table 9.2-4 FAILURE ANALYSIS OF PUMPS, HEAT EXCHANGERS, AND 95 VALVES Table 9.2-5 MINIMUM ALLOWED COMPONENTS FOR THE COMPONENT 96 COOLING WATER (CCW) SYSTEM 9.3 PROCESS AUXILIARIES 97 9.3.1 INSTRUMENT AND SERVICE AIR SYSTEMS 97 9.3.1.1 System Description 97 9.3.1.2 Component Description 98 Page 5 of 15 Revision 26 4/2016 | |||
GINNA/UFSAR 9.3.1.2.1 Compressors 98 9.3.1.2.2 Aftercoolers 98 9.3.1.2.3 Air Receivers 99 9.3.1.2.4 Filters and Dryers 99 9.3.2 SAMPLING SYSTEMS 100 9.3.2.1 Nuclear Sampling System 100 9.3.2.1.1 Design Bases 100 9.3.2.1.1.1 Functional Requirements 100 9.3.2.1.1.2 Operational Requirements 101 9.3.2.1.2 System Design and Operation 101 9.3.2.1.2.1 Sampling System 101 9.3.2.1.2.2 Reactor Coolant Samples 102 9.3.2.1.2.3 Chemical and Volume Control System Samples 102 9.3.2.1.2.4 Steam Generator Liquid Samples 102 9.3.2.1.2.5 Sample Sink 103 9.3.2.1.2.6 Instrumentation 103 9.3.2.1.2.7 Steam Generator Blowdown 103 9.3.2.1.3 Component Description 103 9.3.2.1.3.1 Sample Heat Exchangers 103 9.3.2.1.3.2 Delay Coil 103 9.3.2.1.3.3 Sample Pressure Vessels 103 9.3.2.1.3.4 Sample Sink 104 9.3.2.1.3.5 Piping and Fittings 104 9.3.2.1.3.6 Instrumentation 104 9.3.2.1.3.7 Valves 104 9.3.2.1.4 System Evaluation 105 9.3.2.1.4.1 Availability and Reliability 105 9.3.2.1.4.2 Leakage Provisions 105 9.3.2.1.4.3 Malfunction Analysis 105 9.3.2.1.5 Minimum Operating Conditions 105 9.3.2.1.6 Tests and Inspections 105 9.3.2.2 Nonnuclear Sampling System 105 9.3.2.2.1 Steam Generator Blowdown Sampling 105 9.3.2.2.2 Hotwell Sampling 106 Page 6 of 15 Revision 26 4/2016 | |||
GINNA/UFSAR 9.3.2.2.3 Condensate Sampling 106 9.3.2.2.4 Feedwater Sampling 106 9.3.2.2.5 Main Steam Sampling 106 9.3.2.2.6 Heater Drain Tank Sampling 106 9.3.2.2.7 Sampling Cooling 106 9.3.2.3 Postaccident Sampling System 107 9.3.2.3.1 Design Bases 107 9.3.2.3.2 System Description 108 9.3.2.3.3 Component Description and Operation 108 9.3.2.3.3.1 Liquid and Gas Sample Panel 108 9.3.2.3.3.2 Gas Sampling 109 9.3.2.3.3.3 Liquid Sampling 110 9.3.2.3.3.4 Instrument Panel 111 9.3.2.3.3.5 Electrical Control Panel 111 9.3.2.3.3.6 Postaccident Sampling System Coolers 111 9.3.2.3.3.7 Postaccident Sampling System Waste Tank 112 9.3.2.3.3.8 Postaccident Sampling System Waste Transfer Pump 112 9.3.2.3.3.9 Postaccident Sampling System Waste Tank Evacuating Compressor 112 9.3.2.3.3.10 Containment Sump A Sample Pump 113 9.3.3 EQUIPMENT AND FLOOR DRAINS SYSTEMS 113 9.3.4 CHEMICAL AND VOLUME CONTROL SYSTEM 113 9.3.4.1 Design Bases 113 9.3.4.1.1 Redundancy of Reactivity Control 113 9.3.4.1.2 Reactivity Holddown Capability 114 9.3.4.1.3 Reactivity Hot Shutdown Capability 114 9.3.4.1.4 Reactivity Shutdown Capability 115 9.3.4.1.5 Codes and Classifications 115 9.3.4.2 System Design and Operation 115 9.3.4.2.1 General 115 9.3.4.2.2 Letdown and Charging Systems 116 9.3.4.2.2.1 General 116 9.3.4.2.2.2 Charging Pump Control 117 9.3.4.2.3 Seal-Water Injection System 118 9.3.4.2.4 Reactor Makeup Control System 118 9.3.4.2.4.1 System Description 118 Page 7 of 15 Revision 26 4/2016 | |||
GINNA/UFSAR 9.3.4.2.4.2 Automatic Makeup 120 9.3.4.2.4.3 Dilution 120 9.3.4.2.4.4 Boration 120 9.3.4.2.5 Boron Recycle System 121 9.3.4.2.5.1 System Description 121 9.3.4.2.5.2 Alarm Functions 122 9.3.4.2.6 Heat Tracing System 122 9.3.4.3 Component Description 123 9.3.4.3.1 Letdown and Charging Systems 123 9.3.4.3.1.1 Regenerative Heat Exchanger 123 9.3.4.3.1.2 Letdown Orifices 123 9.3.4.3.1.3 Nonregenerative Heat Exchanger 123 9.3.4.3.1.4 Mixed-Bed Demineralizers 124 9.3.4.3.1.5 Cation Bed Demineralizer 124 9.3.4.3.1.6 Deborating Demineralizers 124 9.3.4.3.1.7 Resin Fill Tank 124 9.3.4.3.1.8 Reactor Coolant Filter 125 9.3.4.3.1.9 Volume Control Tank 125 9.3.4.3.1.10 Charging Pumps 125 9.3.4.3.1.11 Charging Pump Leakoff Tank 126 9.3.4.3.1.12 Charging Pump Dampener 126 9.3.4.3.1.13 Excess Letdown Heat Exchanger 126 9.3.4.3.2 Seal-Water Injection System 126 9.3.4.3.2.1 Seal-Water Heat Exchanger 126 9.3.4.3.2.2 Seal-Water Filter 127 9.3.4.3.2.3 Seal-Water Injection Filters 127 9.3.4.3.3 Reactor Makeup Control System 127 9.3.4.3.3.1 Boric Acid Filter 127 9.3.4.3.3.2 Boric Acid Storage Tanks 127 9.3.4.3.3.3 Batching Tank 128 9.3.4.3.3.4 Boric Acid Tank Heaters 128 9.3.4.3.3.5 Boric Acid Transfer Pumps 128 9.3.4.3.3.6 Boric Acid Blender 128 9.3.4.3.3.7 Chemical Mixing Tank 128 9.3.4.3.3.8 Heat Tracing 129 Page 8 of 15 Revision 26 4/2016 | |||
GINNA/UFSAR 9.3.4.3.3.9 Reactor Makeup Water Pumps 129 9.3.4.3.3.10 Reactor Makeup Water Tank 129 9.3.4.3.4 Boron Recycle System 129 9.3.4.3.4.1 Holdup Tanks 129 9.3.4.3.4.2 Holdup Tank Recirculation Pump 129 9.3.4.3.4.3 Gas Stripper Feed Pumps 130 9.3.4.3.4.4 Base Removal Ion Exchanger 130 9.3.4.3.4.5 Cation Ion Exchanger 130 9.3.4.3.4.6 Ion Exchanger Filter 130 9.3.4.3.4.7 Gas Stripper Equipment 130 9.3.4.3.4.8 Boric Acid Evaporator Equipment 131 9.3.4.3.4.9 Evaporator Condensate Demineralizers 131 9.3.4.3.4.10 Condensate Filter 131 9.3.4.3.4.11 Concentrates Filter 131 9.3.4.3.4.12 Concentrates Holding Tank 132 9.3.4.3.4.13 Concentrates Holding Tank Transfer Pumps 132 9.3.4.3.4.14 Monitor Tanks 132 9.3.4.3.4.15 Monitor Tank Pump 132 9.3.4.3.5 Valves 132 9.3.4.3.6 Piping 133 9.3.4.4 System Evaluation 133 9.3.4.4.1 Availability and Reliability 133 9.3.4.4.2 Seismic Analysis 134 9.3.4.4.3 Leakage Prevention 134 9.3.4.4.4 Incident Control 134 9.3.4.4.5 Malfunction Analysis 135 9.3.4.4.5.1 System Failures 135 9.3.4.4.5.2 Inadvertent Dilution 135 9.3.4.4.5.3 Alternative Methods of Boration 135 9.3.4.4.5.4 Inadvertent Dilution of Boric Acid Storage Tanks 136 9.3.4.4.5.5 Loss of Seal Injection Water 136 9.3.4.4.6 Overpressurization Protection 136 9.3.4.4.6.1 Suction Lines 136 Page 9 of 15 Revision 26 4/2016 | |||
GINNA/UFSAR 9.3.4.4.6.2 Discharge Lines 137 9.3.4.4.7 Galvanic Corrosion 137 9.3.4.4.8 Control of Tritium 137 9.3.4.4.9 Reactor Coolant Activity Concentration Calculations 138 9.3.4.4.9.1 Computation Method 138 9.3.4.4.9.2 Tritium Production 140 9.3.4.4.9.3 Radioactivity Monitoring 140 9.3.4.4.9.4 Technical Specifications Limits 140 9.3.4.4.9.5 Tritium Limit 140 9.3.4.4.9.6 R.E. Ginna Normal Operation RCS and Secondary Coolant Sources 140 9.3.4.5 Minimum Operating Conditions 141 Table 9.3-1 NUCLEAR PROCESS SAMPLING SYSTEM CODE 143 REQUIREMENTS Table 9.3-2 NUCLEAR PROCESS SAMPLING SYSTEM COMPONENTS 144 Table 9.3-3 MALFUNCTION ANALYSIS OF NUCLEAR PROCESS 146 SAMPLING SYSTEM Table 9.3-4 POSTACCIDENT SAMPLING SYSTEM FUNCTIONAL 147 REQUIREMENTS Table 9.3-5 LIQUID AND GAS SAMPLE PANEL ANALYTICAL EQUIPMENT 148 REQUIREMENTS Table 9.3-6 CHEMICAL AND VOLUME CONTROL SYSTEM 149 PERFORMANCE PARAMETERS Table 9.3-7 PRINCIPAL COMPONENT DATA | |||
==SUMMARY== | |||
151 Table 9.3-8 MALFUNCTION ANALYSIS OF CHEMICAL AND VOLUME 153 CONTROL SYSTEM Table 9.3-9 REACTOR COOLANT SYSTEM EQUILIBRIUM ACTIVITIES 154 Table 9.3-10 ARAMETERS USED IN THE 1811 MWT UPRATE CALCULATION 156 OF REACTOR COOLANT FISSION PRODUCT ACTIVITIES Table 9.3-11a PARAMETERS USED IN THE CALCULATION OF TRITIUM 157 PRODUCTION IN THE REACTOR COOLANT - BASIC ASSUMPTIONS Table 9.3-11b CALCULATION OF TRITIUM PRODUCTION IN THE REACTOR 158 COOLANT Table 9.3-12 ANSI/ANS 18.1-1999 NORMAL SOURCE INPUT PARAMETERS 159 Table 9.3-12 ANSI/ANS 18.1-1999 NORMAL SOURCE INPUT PARAMETERS 160 Page 10 of 15 Revision 26 4/2016 | |||
GINNA/UFSAR 9.4 AIR CONDITIONING, HEATING, COOLING, AND VENTILATION SYSTEMS 162 9.4.1 CONTAINMENT VENTILATION SYSTEM 162 9.4.1.1 Design Bases 162 9.4.1.1.1 Design Objectives 162 9.4.1.1.2 Design Criteria 163 9.4.1.2 System Design 164 9.4.1.2.1 Introduction 164 9.4.1.2.2 Containment Recirculation Cooling and Filtration System 164 9.4.1.2.3 Control Rod Drive Mechanism Cooling System 165 9.4.1.2.4 Reactor Compartment Cooling System 166 9.4.1.2.5 Refueling Water Surface and Purge System 166 9.4.1.2.6 Containment Auxiliary Charcoal Filter System 166 9.4.1.2.7 Containment Post-accident Charcoal Filter System 167 9.4.1.2.8 Containment Shutdown Purge System 167 9.4.1.2.9 Containment Mini-Purge System 168 9.4.1.2.10 Penetration Cooling System 168 9.4.2 AUXILIARY BUILDING VENTILATION SYSTEM 169 9.4.2.1 Design Basis 169 9.4.2.2 System Design and Operation 169 9.4.2.2.1 System Design Objective 169 9.4.2.2.2 Charcoal Filter Circuit 169 9.4.2.2.3 System Operation 170 9.4.2.3 System Components 171 9.4.2.3.1 Auxiliary Building Air Handling Unit 171 9.4.2.3.2 Auxiliary Building Exhaust Fan 1C 171 9.4.2.3.3 Auxiliary Building Exhaust Fans 1A and 1B 171 9.4.2.3.4 Auxiliary Building Exhaust Fan 1G 171 9.4.2.3.5 Auxiliary Building Charcoal Filter Fans 1A and 1B 172 9.4.2.3.6 Penetration Cooling Fans 1A and 1B 172 9.4.2.3.7 Pump Area Coolers 172 9.4.2.3.8 Intermediate Building Supply and Exhaust Fans 172 9.4.2.3.9 Steam Isolation Dampers 173 9.4.2.4 System Evaluation 173 9.4.2.4.1 Effect of Loss of Cooling on Pumps and Valves 173 9.4.2.4.2 Revised Auxiliary Building Loss of Cooling Analysis 174 Page 11 of 15 Revision 26 4/2016 | |||
GINNA/UFSAR 9.4.2.4.2.1 AUXILIARY BUILDING TEMPERATURE WITH MINIMUM 175 SERVICE WATER FLOW 9.4.2.4.3 Effect of Loss of Offsite Power on Ventilation Flow 176 9.4.3 CONTROL ROOM AREA VENTILATION SYSTEM 176 9.4.4 SPENT FUEL POOL AREA VENTILATION SYSTEM 176 9.4.5 TURBINE BUILDING VENTILATION SYSTEM 176 9.4.6 SERVICE BUILDING VENTILATION SYSTEM 177 9.4.7 ALL-VOLATILE-TREATMENT BUILDING VENTILATION 177 SYSTEM 9.4.7.1 Introduction 177 9.4.7.2 Summary Description of the System 178 9.4.7.2.1 Compressor and Booster Pump Area Ventilation System 178 9.4.7.2.2 Demineralizer Area Ventilation System 179 9.4.7.2.3 Demineralizer Area Control Room System 179 9.4.7.2.4 Heating System 179 9.4.8 TECHNICAL SUPPORT CENTER VENTILATION SYSTEM 179 9.4.8.1 System Description 179 9.4.8.2 System Operation 180 9.4.8.2.1 Cooling Systems 180 9.4.8.2.2 Heating Systems 181 9.4.9 ENGINEERED SAFETY FEATURES VENTILATION SYSTEMS 181 9.4.9.1 Engineered Safety Features Equipment Ventilation and Cooling 182 9.4.9.2 Relay Room 182 9.4.9.3 Battery Rooms 183 9.4.9.4 Essential Auxiliary Systems 183 9.4.9.5 Diesel Generators 184 9.4.9.6 Standby Auxiliary Feedwater System (SAFW) 185 9.4.9.6.1 System Operation 185 9.4.9.6.2 Controls and Instrumentation 186 9.4.9.7 Post-accident Fan Coolers and Charcoal Filters 187 9.4.10 STATION HEATING STEAM SYSTEM 187 Table 9.4-1 CONTAINMENT VENTILATION SYSTEM PRINCIPAL 189 COMPONENT DATA | |||
==SUMMARY== | |||
9.5 OTHER AUXILIARY SYSTEMS 192 Page 12 of 15 Revision 26 4/2016 | |||
GINNA/UFSAR 9.5.1 FIRE PROTECTION 192 9.5.1.1 Design Basis Summary 192 9.5.1.1.1 Defense-in-Depth 192 9.5.1.1.2 NFPA 805 Performance Criteria 193 9.5.1.1.3 Codes of Record 194 9.5.1.1.4 Required Systems 194 9.5.1.1.5 Definition of Power Block Structures 195 9.5.1.2 System Design 195 9.5.1.2.1 General 195 9.5.1.2.2 Fire Detection and Signaling Systems 196 9.5.1.2.3 Fire Suppression Systems 197 9.5.1.2.3.1 Water Supply 197 9.5.1.2.3.2 Fire Pumps 197 9.5.1.2.3.3 Piping and Valves 198 9.5.1.2.3.4 Fire Hydrants 199 9.5.1.2.3.5 Yard Loop 199 9.5.1.2.3.6 Interior Hose Stations 199 9.5.1.2.3.7 Water Suppression Systems 200 9.5.1.2.3.8 Gas Suppression Systems 201 9.5.1.2.3.9 Portable Fire Extinguishers 202 9.5.1.2.3.10 Wet Chemical Suppression System 202 9.5.1.2.4 Other Design Considerations 202 9.5.1.2.4.1 Smoke Removal 202 9.5.1.2.4.2 Breathing Equipment 202 9.5.1.2.4.3 Control Building Ventilation 202 9.5.1.2.4.4 Reactor Coolant Pump Motor Oil Collection System 203 9.5.1.2.4.5 Floor Drains and Curbs 203 9.5.1.2.4.6 Lighting Systems 203 9.5.1.2.4.7 Communications 203 9.5.1.2.4.8 Electrical Cable Insulation 203 9.5.1.2.4.9 Fire Barriers 204 9.5.1.2.4.10 Electrical Cable Penetrations 205 9.5.1.2.4.11 Piping and Duct Penetrations 205 9.5.1.2.4.12 Cable Separation 205 Page 13 of 15 Revision 26 4/2016 | |||
GINNA/UFSAR 9.5.1.2.4.13 Spray Shields 206 9.5.1.2.4.14 Construction Joints 206 9.5.1.3 Safety Evaluation 206 9.5.1.4 Fire Protection Program Documentation, Configuration Control 206 and Quality Assurance 9.5.2 COMMUNICATIONS SYSTEMS 207 9.5.2.1 Public Address System 207 9.5.2.2 Telephone Systems 207 9.5.2.3 Radio Systems 208 9.5.2.4 Offsite Communications 208 9.5.2.5 Emergency Communications With the NRC 209 9.5.3 LIGHTING SYSTEMS 209 9.5.4 DIESEL GENERATOR FUEL OIL STORAGE AND TRANSFER 210 SYSTEM 9.5.5 DIESEL GENERATOR COOLING SYSTEM 211 9.5.6 DIESEL GENERATOR STARTING SYSTEM 211 9.5.7 DIESEL GENERATOR LUBRICATION SYSTEM 212 9.5.8 DIESEL GENERATOR COMBUSTION AIR INTAKE AND 212 EXHAUST Table 9.5-1 FIRE SERVICE WATER HOSE REEL LOCATIONS 215 Table 9.5-2 Power Block Buildings 217 FIGURES Figure 9.1-1 Fuel Handling Structures Figure 9.1-2 Figure DELETED Figure 9.1-3 Arrangement of Spent Fuel Storage Racks Figure 9.1-4 Figure DELETED Figure 9.1-5 Figure DELETED Figure 9.1-6 Figure DELETED Figure 9.1-7 Spent Fuel Pool Cooling Cycle Figure 9.1-8 Figure DELETED Figure 9.1-9 Figure DELETED Figure 9.1-10 Reactor Vessel Head Lifting Device Figure 9.1-11 Fuel Handling Devices Figure 9.3-1 Maximum Tritium Activity Released to Primary Coolant Figure 9.5-1 Figure DELETED Page 14 of 15 Revision 26 4/2016 | |||
GINNA/UFSAR Figure 9.5-2 Figure DELETED Figure 9.5-3 Figure DELETED Figure 9.5-4 Figure DELETED Figure 9.5-5 Figure DELETED Figure 9.5-6 Diesel Engine Lubricating Oil Systems (Simplified) | |||
Page 15 of 15 Revision 26 4/2016 | |||
GINNA/UFSAR 10 STEAM AND POWER CONVERSION SYSTEM 1 | |||
==10.1 INTRODUCTION== | |||
2 10.1.1 | |||
==SUMMARY== | |||
DESCRIPTION 2 10.1.1.1 Functional Description 2 10.1.1.2 Radioactivity 3 10.1.1.3 Major Systems 3 10.1.2 DESIGN BASES 4 10.1.2.1 System Design 4 10.1.2.2 Codes and Classifications 5 10.1.3 SYSTEM EVALUATION 5 10.1.3.1 Variables Limits Functions 5 10.1.3.2 Transient Effects 6 10.1.3.3 Secondary-Primary Interactions 7 Table 10.1-1 STEAM AND POWER CONVERSION SYSTEM COMPONENT 8 DESIGN PARAMETERS 10.2 TURBINE GENERATOR AND CONTROLS 11 10.2.1 MAIN TURBINE 11 10.2.1.1 Description 11 10.2.1.2 Turbine Controls 11 10.2.1.2.1 Description 11 10.2.1.2.2 Automatic Load Reduction 12 10.2.1.3 Turbine Disk Integrity 12 10.2.1.4 Turbine Supervisory Instrumentation 13 10.2.2 MAIN GENERATOR 13 10.2.3 ELECTROHYDRAULIC CONTROL SYSTEM 14 10.2.3.1 Function 14 10.2.3.2 Components 14 10.2.3.3 Alarms and Controls 15 10.2.3.4 Turbine Trip Devices 16 10.2.3.4.1 Overspeed Trip Mechanism 16 Page 1 of 5 Revision 26 5/2016 | |||
GINNA/UFSAR 10.2.3.4.2 Auxiliary Governor 17 10.2.3.4.3 Protective Trip Devices 17 10.2.3.4.4 Testing and Inspection 18 10.3 MAIN STEAM SYSTEM 21 10.3.1 DESIGN BASIS 20 10.3.2 SYSTEM DESCRIPTION 20 10.3.2.1 Flow Path 20 10.3.2.2 Steam Generators 21 10.3.2.3 Steam Piping 22 10.3.2.4 Main Steam Safety Valves (MSSV) 22 10.3.2.5 Atmospheric Relief Valves (ARV) 22 10.3.2.6 Main Steam Isolation Valves 23 10.3.2.7 Main Steam Non-Return Check Valves 23 10.3.2.8 Main Steam Header 24 10.3.2.9 Main Turbine Stop Valves and Control Valves 24 10.3.2.10 Moisture Separator Reheaters 24 10.3.2.11 Reheater Stop and Intercept Valves 25 10.3.3 INSTRUMENTATION REQUIREMENTS 25 10.4 CONDENSATE AND FEEDWATER SYSTEMS 27 10. | |||
==4.1 DESCRIPTION== | |||
27 10.4.2 FLOW PATH 27 10.4.3 MAIN CONDENSERS 27 10.4.4 CONDENSATE SYSTEM 28 10.4.4.1 Condensate Pumps 28 10.4.4.2 Condensate Booster Pumps 29 10.4.4.3 Low-Pressure Heaters 29 10.4.4.4 Condensate Bypass Valve 30 10.4.5 FEEDWATER SYSTEM 30 10.4.5.1 Main Feedwater Pumps 30 10.4.5.2 High-Pressure Heaters 31 Page 2 of 5 Revision 26 5/2016 | |||
GINNA/UFSAR 10.4.5.3 Feedwater Flow Control 31 10.4.5.4 Feedwater Flow Measurement 31 10.5 AUXILIARY FEEDWATER SYSTEMS 33 10. | |||
==5.1 INTRODUCTION== | |||
33 10.5.2 DESIGN BASES 33 10.5.2.1 Functional Requirements 33 10.5.2.2 Preferred Auxiliary Feedwater System 34 10.5.2.3 Standby Auxiliary Feedwater System (SAFW) 34 10.5.3 SYSTEMS OPERATION AND DESCRIPTION 35 10.5.3.1 Preferred Auxiliary Feedwater System 35 10.5.3.1.1 Normal Lineup 35 10.5.3.1.2 Startup and Cooldown Operations 35 10.5.3.1.3 Transient Operations 36 10.5.3.1.4 System Description 36 10.5.3.2 Standby Auxiliary Feedwater System (SAFW) 38 10.5.4 DESIGN EVALUATION 38 10.5.4.1 System Evaluation 38 10.5.4.2 Alternating Current Independence of the Turbine-Driven Auxiliary 39 Feedwater Pump (TDAFW) 10.5.5 INSTRUMENTATION 40 10.5.5.1 Motor-Driven Auxiliary Feedwater Pump (MDAFW) Controls 40 10.5.5.2 Preferred Auxiliary Feedwater System Initiation 40 10.5.5.3 Auxiliary Feedwater System Alarms 41 10.5.5.4 Auxiliary Feedwater Performance Indications 42 10.5.5.5 Control From Outside the Control Room 42 10.6 CIRCULATING WATER SYSTEM 45 10.6.1 DESIGN BASES 45 10.6.2 SYSTEM DESCRIPTION 45 10.6.2.1 Intake Structure 45 10.6.2.2 Inlet Tunnel 46 10.6.2.3 Traveling Screens 46 Page 3 of 5 Revision 26 5/2016 | |||
GINNA/UFSAR 10.6.2.4 Circulating Water Pumps 46 10.6.2.5 Condenser Inlet and Outlet Valves 46 10.6.2.6 Condensate Cooler 46 10.6.2.7 Screen House 46 10.6.2.8 Piping and Discharge Canal 47 10.6.2.9 Flooding Protection 47 10.6.3 INSTRUMENTATION AND CONTROL 48 10.6.4 INSERVICE INSPECTION 48 10.7 OTHER FEATURES OF THE STEAM AND POWER CONVER- 50 SION SYSTEM 10.7.1 STEAM DUMP SYSTEM 50 10.7.2 HEATER DRAIN SYSTEM 51 10.7.3 EXTRACTION STEAM SYSTEM 51 10.7.4 CONDENSATE STORAGE SYSTEM 52 10.7.5 STEAM-GENERATOR BLOWDOWN AND BLOWDOWN 53 RECOVERY SYSTEM 10.7.5.1 Steam-Generator Blowdown System 53 10.7.5.2 Blowdown Recovery System 54 10.7.5.3 Blowdown System Operation 54 10.7.6 MAIN TURBINE AND GENERATOR AUXILIARY SYSTEMS 55 10.7.6.1 Gland Sealing Steam and Exhaust System 56 10.7.6.2 Air Ejectors 57 10.7.6.3 Vacuum Priming System 57 10.7.6.4 Exhaust Hood Spray System 57 10.7.6.5 Turbine Lube-Oil System 58 10.7.6.6 Generator Hydrogen Cooling System 59 10.7.6.7 Generator Seal-Oil System 60 Page 4 of 5 Revision 26 5/2016 | |||
GINNA/UFSAR 10.7.6.8 Generator Exciter Cooling 60 10.7.7 SECONDARY CHEMISTRY CONTROL 60 10.7.7.1 All-Volatile-Treatment Chemistry 60 10.7.7.1.1 Background 60 10.7.7.1.2 All-Volatile-Treatment Chemistry Control 60 10.7.7.2 Nonnuclear Sampling System 62 10.7.7.3 Water Chemistry Monitoring Program 62 10.7.7.4 Catalytic Oxygen Removal System 62 10.7.7.5 Condensate Polishing Demineralizer System 63 10.7.7.5.1 System Description 63 10.7.7.5.2 Resin Transfer 64 10.7.7.5.3 Regeneration (Drawing 33013-1910, Sheets 1 and 2) 64 10.7.7.5.4 Waste Disposal (Drawing 33013-1912) 65 10.7.7.6 Chemical Dispersant 65 10.7.8 EROSION/CORROSION MONITORING PROGRAM 66 Table 10.7-1 COMPUTERIZED SECONDARY WATER CHEMISTRY MONI- 67 TORING SYSTEM FIGURES Figure 10.2-1 Turbine Control and Protection System Figure 10.6-1 Figure DELETED Sheet 1 - | |||
Figure 10.6-1 Figure DELETED Sheet 2 - | |||
Figure 10.6-2 Screen House Area Plot Plan Figure 10.6-3 Circulating Water Intake Cross Section Page 5 of 5 Revision 26 5/2016 | |||
GINNA/UFSAR 11 RADIOACTIVE WASTE MANAGEMENT 1 11.1 DESIGN CRITERIA AND SOURCE TERMS 2 11.1.1 GENERAL DESIGN CRITERIA 2 11.1.1.1 AIF General Design Criterion 70 (1967) 2 11.1.1.2 Appendix A General Design Criteria (1972) 3 11.1.2 SOURCE TERMS 3 11.1.2.1 Liquid Sources 4 11.1.2.2 Gaseous Sources 4 11.1.2.3 Radioactivity Inputs 4 11.2 LIQUID WASTE MANAGEMENT SYSTEM 7 11.2.1 DESIGN BASES 7 11.2.2 SYSTEM DESCRIPTION 7 11.2.2.1 Laundry and Hot Shower Tanks 7 11.2.2.2 Chemical Drain Tank 7 11.2.2.3 Reactor Coolant Drain Tank and Pumps 7 11.2.2.4 Waste Holdup Tank 8 11.2.2.4.1 Liquid Waste Sources 8 11.2.2.4.2 Waste Holdup Tank Discharge 9 11.2.2.5 Auxiliary Building Sump Tank, Sump Tank Pumps, and Sump Pumps 9 11.2.2.6 Waste Evaporator 10 11.2.2.7 Evaporator Feed Tank 10 11.2.2.8 Evaporator Feed Tank Pumps 11 11.2.2.9 Concentrator 11 11.2.2.10 Distillate Tank 12 11.2.2.11 Waste Condensate Demineralizers 12 11.2.2.12 Waste Condensate Tanks 12 11.2.2.13 Vendor Supplied Demineralization System 12 11.2.2.14 High Conductivity Waste Tank 13 11.2.2.15 Retention Tank 13 11.2.2.16 Neutralizing Tank 13 Page 1 of 4 Revision 26 5/2016 | |||
GINNA/UFSAR 11.2.2.17 Monitor Tanks 13 11.2.2.18 Radwaste Control System 13 11.2.2.19 Dry Cleaning Unit 13 11.2.2.20 Piping and Valves 14 11.2.3 LIQUID EFFLUENT RELEASE CONCENTRATIONS AND DOSES 14 11.2.3.1 Liquid Release Control 14 11.2.3.2 Dose Calculations 14 11.2.3.3 Accidental Spill of Liquid Radwastes 16 Table 11.2-1 WASTE DISPOSAL COMPONENTS CODE REQUIREMENTS 18 Table 11.2-2a LIQUID WASTE SYSTEM COMPONENT | |||
==SUMMARY== | |||
DATA - 19 TANKS Table 11.2-2b LIQUID WASTE SYSTEM COMPONENT | |||
==SUMMARY== | |||
DATA - 20 PUMPS Table 11.2-2c LIQUID WASTE SYSTEM COMPONENT | |||
==SUMMARY== | |||
DATA - 21 COMPRESSORS Table 11.2-3 LIQUID WASTE DISPOSAL SYSTEM PERFORMANCE DATA 22 Table 11.2-4a LIQUID EFFLUENTS, 10 CFR 50, APPENDIX I CALCULATIONS, 23 ASSUMPTIONS (HISTORICAL) | |||
Table 11.2-4b LIQUID EFFLUENTS, 10 CFR 50, APPENDIX I CALCULATIONS, 24 ASSUMPTIONS1 (HISTORICAL) | |||
Table 11.2-5 LIQUID EFFLUENTS, 10 CFR 50, APPENDIX I CALCULATIONS, 25 RESULTSa. (HISTORICAL) | |||
Table 11.2-6 MAXIMUM INDIVIDUAL DOSES FROM LIQUID EFFLUENTS 28 (MREM/YEAR) (HISTORICAL) | |||
Table 11.2-7 ESTIMATED ANNUAL DOSES TO THE PUBLIC DUE TO NOR- 30 MAL OPERATION LIQUID RADWASTE EFFLUENTS - CORE POWER LEVEL 1811 MWt 11.3 GASEOUS WASTE MANAGEMENT SYSTEM 31 11.3.1 DESIGN BASES 31 11.3.2 SYSTEM DESCRIPTION 31 11.3.2.1 Operation 31 11.3.2.2 Components 31 11.3.2.2.1 Waste Gas Compressors 32 11.3.2.2.2 Gas Decay Tanks 32 Page 2 of 4 Revision 26 5/2016 | |||
GINNA/UFSAR 11.3.2.2.3 Waste Disposal Panel 33 11.3.2.2.4 Gas Analyzer 33 11.3.2.2.5 Nitrogen Manifold 34 11.3.2.2.6 Hydrogen Manifold 34 11.3.2.2.7 Valves 34 11.3.3 GASEOUS RADIOACTIVE RELEASES 34 Table 11.3-1 GASEOUS EFFLUENTS, 10 CFR 50, APPENDIX I 37 CALCULATIONS, ASSUMPTIONS2 (HISTORICAL) | |||
Table 11.3-2 GASEOUS EFFLUENTS, 10 CFR 50, APPENDIX I 39 CALCULATIONS, RESULTS3 (HISTORICAL) | |||
Table 11.3-3 MAXIMUM INDIVIDUAL DOSES FROM GASEOUS EFFLUENTS 41 (HISTORICAL) | |||
Table 11.3-4 ESTIMATED ANNUAL DOSES TO THE PUBLIC DUE TO 43 NORMAL OPERATION GASEOUS RADWASTE EFFLUENTS - | |||
CORE POWER LEVEL 1811 MWt 11.4 SOLID WASTE MANAGEMENT SYSTEM 44 11. | |||
==4.1 DESCRIPTION== | |||
44 11.4.1.1 General 44 11.4.1.1.1 Types of Solid Waste 44 11.4.1.1.2 Sludge 44 11.4.1.1.3 Oily Waste 44 11.4.1.1.4 Bead Resin 44 11.4.1.1.5 Spent Filters 45 11.4.1.1.6 Dry Active Waste 45 11.4.1.2 Spent Resin Storage Tanks 45 11.4.1.3 Storage Facilities 45 11.4.2 SOLID WASTE ESTIMATES 46 11.4.3 PROCESS CONTROL PROGRAM 46 Table 11.4-1 ANNUAL SHIPMENT OF SOLID WASTE (JULY 1990-JUNE 1991) 48 11.5 PROCESS AND EFFLUENT RADIATION MONITORING AND 49 SAMPLING SYSTEMS 11.5.1 DESIGN BASES 49 11.5.2 SYSTEM DESCRIPTION 49 Page 3 of 4 Revision 26 5/2016 | |||
GINNA/UFSAR 11.5.2.1 General 49 11.5.2.2 Process Radiation Monitoring System 50 11.5.2.2.1 General Description 50 11.5.2.2.2 Containment Iodine Monitor 50 11.5.2.2.3 Plant Vent Iodine Monitor 50 11.5.2.2.4 Containment Particulate and Noble Gas Monitors 51 11.5.2.2.5 Containment Vent High-Range Effluent Monitor 51 11.5.2.2.6 Plant Vent Particulate Monitors 51 11.5.2.2.7 Plant Vent Noble Gas and High-Range Effluent Monitor 51 11.5.2.2.8 Air Ejector and Gland Steam Exhaust Monitors 52 11.5.2.2.9 Containment Service Water Monitor 53 11.5.2.2.10 Component Cooling Water Monitor 53 11.5.2.2.11 Liquid Waste Disposal Monitor 53 11.5.2.2.12 Steam-Generator Blowdown Monitor 53 11.5.2.2.13 Spent Fuel Pool (SFP) Heat Exchanger Service Water Monitors 54 11.5.2.2.14 Retention Tank Monitor 54 11.5.2.2.15 High Conductivity Waste Tank Monitor 54 11.5.2.2.16 Control Room Radiation Monitors 55 11.5.2.3 Tritium Sampling 55 11.5.3 DESIGN EVALUATION 55 11.5.4 ENVIRONMENTAL RADIOACTIVITY MONITORING PRO- 55 GRAM Figures Figure 11.5-1 Gaseous Radwaste Treatment Systems Effluent Paths and Controls Figure 11.5-2 Liquid Radwaste Treatment Systems Effluent Paths and Controls Page 4 of 4 Revision 26 5/2016 | |||
GINNA/UFSAR 12 RADIATION PROTECTION 1 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES 2 ARE AS LOW AS IS REASONABLY ACHIEVABLE 12.1.1 ALARA PROGRAM 2 12.1.2 ORGANIZATIONAL RESPONSIBILITIES 2 12.1.3 RADIATION PROTECTION PROGRAM 2 12.2 RADIATION SOURCES 3 Table 12.2-1 SHIELDING SOURCE TERMS (T=0) (HISTORICAL) 6 Table 12.2-2 SHIELDING SOURCE ACTIVITY at T=0 hrs - Power Level 1811 10 MWt 12.3 RADIATION PROTECTION DESIGN FEATURES 16 12.3.1 DESIGN CRITERIA 16 12.3.1.1 Conformance to 1967 Design Criteria 16 12.3.1.2 Conformance to 1972 Design Criteria 16 12.3.2 SHIELDING 17 12.3.2.1 Design Basis 17 12.3.2.2 Shielding Design 19 12.3.2.2.1 Primary Shield 19 12.3.2.2.2 Secondary Shield 19 12.3.2.2.3 Containment Structure 20 12.3.2.2.4 Fuel Handling (Water and Pool) Shield 20 12.3.2.2.5 Auxiliary Shielding 21 12.3.2.2.6 Shielding Design Modifications 21 12.3.2.2.7 Containment Accessibility Procedure 22 12.3.3 VENTILATION 22 12.3.3.1 Gas Collection and Decay Tank System 23 12.3.3.2 Plant Ventilation System 23 12.3.3.3 Containment Ventilation System 24 12.3.3.4 Air Ejector and Gland Seal Exhaust System 24 12.3.4 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONI- 24 TORING INSTRUMENTATION 12.3.4.1 Introduction 24 Page 1 of 3 Revision 26 5/2016 | |||
GINNA/UFSAR 12.3.4.2 Description 25 12.3.4.3 Radiation Monitoring System Detectors 26 12.3.5 EQUIPMENT AND SYSTEM DECONTAMINATION 27 12.3.5.1 Design Basis 27 12.3.5.2 Methods of Decontamination 27 12.3.5.3 Decontamination Facilities 27 Table 12.3-1 PLANT ZONE CLASSIFICATIONS 30 Table 12.3-2a PRIMARY SHIELD NEUTRON FLUXES AND DESIGN PARAME- 31 TERS - CALCULATED NEUTRON FLUXES (HISTORICAL) | |||
Table 12.3-2b PRIMARY SHIELD NEUTRON FLUXES AND DESIGN PARAME- 32 TERS - DESIGN PARAMETERS (HISTORICAL) | |||
Table 12.3-3 SECONDARY SHIELD DESIGN PARAMETERS (HISTORICAL) 33 Table 12.3-4 CONTAINMENT STRUCTURE DESIGN PARAMETERS (HISTOR- 34 ICAL) | |||
Table 12.3-5 REFUELING CANAL AND SPENT FUEL POOL DESIGN PARAM- 35 ETERS (HISTORICAL) | |||
Table 12.3-6 PRINCIPAL AUXILIARY SHIELDING (HISTORICAL) 36 12.4 DOSE ASSESSMENT 37 12.4.1 OPERATION IN MODES 1 AND 2 37 12.4.2 FUEL HANDLING OPERATIONS 37 12.4.3 POSTACCIDENT CONDITIONS 38 12.4.3.1 Summary 38 12.4.3.2 Methodology 39 12.4.3.2.1 Calculation of Dose Rates 39 12.4.3.2.2 Doses to Personnel During Postaccident Access to Vital Areas 39 12.4.3.3 Areas That May Require Access for Postaccident Operations 40 12.4.3.3.1 Hydrogen Recombiner Control Panel (Area A) 40 12.4.3.3.2 Postaccident Containment Air Sample Penetration No. 203 (Area B) 40 12.4.3.3.3 Nuclear Sample Room (Area C) 40 12.4.3.3.4 Primary Chemistry Laboratory (Area D) 41 12.4.3.3.5 Count Room (Area E) 41 12.4.3.3.6 Postaccident Containment Sample Penetration No. 305 (Area F) 41 12.4.3.3.7 Radwaste Control Panel (Area G) 41 12.4.3.3.8 Safeguards Bus 16 (Area H) 41 Page 2 of 3 Revision 26 5/2016 | |||
GINNA/UFSAR 12.4.3.3.9 Safeguards Bus 14 (Area I) 42 12.4.3.3.10 Postaccident Containment Air Sample Penetration (Area J) 42 12.4.3.3.11 Auxiliary Building Heating, Ventilation, and Air Conditioning (Area 43 K) 12.4.3.3.12 Spent Fuel Pool (SFP) and Auxiliary Building Heating, Ventilation, 43 and Air Conditioning Filters (Area L and Area M) 12.4.3.3.13 Control Access High Efficiency Particulate Air and Charcoal Filters 43 (Area N) 12.4.3.3.14 Control Room 43 12.4.3.3.15 Control Room 43 Table 12.4-1 RADIATION MONITORING SYSTEM READINGS (1983) 45 Table 12.4-2 RADIATION SURVEY READINGS IN PLANT AREAS (1983) 46 Table 12.4-3 Table DELETED 48 Table 12.4-4 EXPOSURE RATES FOR VITAL AREAS AS A FUNCTION OF 49 TIME (R/hr) | |||
Table 12.4-5 VITAL AREA RADIATION DOSE | |||
==SUMMARY== | |||
50 12.5 RADIATION PROTECTION PROGRAM ADMINISTRATION 51 12.5.1 ORGANIZATION 51 12.5.2 EXPOSURE CONTROL PROGRAM 51 12.5.2.1 External Exposure 51 12.5.2.2 Internal Exposure 51 12.5.2.3 Respiratory Protection 52 12.5.2.4 Radioactive Sources Control 52 12.5.2.5 Medical Examinations 52 12.5.3 SURVEILLANCE PROGRAM 53 12.5.3.1 Surveys 53 12.5.3.2 Radiation Work Permits 53 12.5.3.3 Access Control, Posting, and Labeling 53 12.5.3.3.1 Restricted Areas 53 12.5.3.3.2 Access Control 53 12.5.3.3.3 Protective Apparel 54 12.5.4 RADIATION PROTECTION FACILITIES AND EQUIPMENT 54 Table 12.5-1 Table DELETED 56 FIGURES Figure 12.5-1 Table DELETED Page 3 of 3 Revision 26 5/2016 | |||
GINNA/UFSAR 13 CONDUCT OF OPERATIONS 1 13.1 ORGANIZATIONAL STRUCTURE OF GINNA NUCLEAR GENERATING 2 STATION 13.1.1 ORIGINAL CONSTRUCTION ORGANIZATION 2 13.1.1.1 Design and Construction Activities (Project Phase) 2 13.1.1.2 Preoperational Activities 2 13.1.2 CORPORATE NUCLEAR OPERATIONS ORGANIZATION 2 13.1.2.1 President and Chief Executive Officer, CENG (CEO) 3 13.1.2.2 Senior Vice President - Nuclear Operations Chief Nuclear Officer CENG 3 (COO/ CNO) 13.1.2.3 DELETED 13.1.3 GINNA OPERATING ORGANIZATION 3 13.1.3.1 Site Vice President (SVP) 3 13.1.3.2 Plant Manager 3 13.1.3.3 Director, Site Engineering 4 13.1.3.3.1 Engineering Supervision 4 13.1.3.3.1.1 Department Engineer(s) 4 13.1.3.3.1.2 Reactor Engineer(s) 4 13.1.3.4 Director, Site Work Management 4 13.1.3.5 Director, Site Training 5 13.1.3.6 Manager, Site Security 5 13.1.3.7 Director, Site Operations 5 13.1.3.8 Manager, Site Radiation Protection 6 13.1.3.9 Manager, Site Chemistry, Environment and Radwaste 7 13.1.3.10 Director, Site Maintenance 8 13.1.4 QUALIFICATIONS OF PLANT PERSONNEL 8 13.1.4.1 Qualifications of Plant Staff 8 13.1.4.2 Qualifications of Incumbent Plant Personnel 8 13.2 TRAINING PROGRAM 10 13.2.1 PLANT STAFF TRAINING PROGRAM 10 13.2.1.1 Objectives 10 13.2.1.2 Initial Training Programs (Historical) 10 13.2.1.2.1 Personnel Selection 10 13.2.1.2.2 Nuclear Theory 10 13.2.1.2.3 Plant Systems and Operations 11 13.2.1.2.4 Final Phase 11 13.2.1.3 Onsite Training Prior to Startup 12 Page 1 of 3 Revision 26 5/2016 | |||
GINNA/UFSAR 13.2.2 REPLACEMENT AND RETRAINING OF PERSONNEL 13 13.2.2.1 Licensed Operator Replacement and Requalification Training 13 13.2.2.2 Replacement and Retraining of Unlicensed Personnel 13 13.2.2.3 General Employee Training 13 REFERENCES FOR SECTION 13.2 15 Table 13.2-1 INITIAL GINNA STATION PERSONNEL TRAINING 16 13.3 EMERGENCY PLANNING 17 13.4 REVIEW AND AUDIT 18 13.4.1 ONSITE REVIEW 18 13.4.2 INDEPENDENT REVIEW 18 13.4.3 AUDIT PROGRAM 18 13.4.3.1 Nuclear Safety Review Board 18 13.4.3.2 Quality Performance & Assessment Group 18 13.5 PLANT PROCEDURES 19 13.5.1 ADMINISTRATIVE PROCEDURES 19 13.5.1.1 Conformance With Regulatory Guide 1.33 19 13.5.1.2 Preparation of Procedures 19 13.5.1.3 Description of Administrative Procedures 19 13.5.2 OPERATING AND MAINTENANCE PROCEDURES 19 13.5.2.1 Control Room Operating Procedures 19 13.5.2.2 Site Contingency Procedures 20 13.5.2.2.1 General 20 13.5.2.2.2 Adverse Weather Conditions 20 13.5.2.2.2.1 High Winds 20 13.5.2.2.2.2 Tornadoes 20 13.5.2.2.2.3 Ice Storms 20 13.5.2.2.3 High Water or Flood Emergency Plan 21 13.5.2.2.4 Earthquake Emergency Plan 21 13.5.2.2.5 Fire Emergency Plan 21 13.5.2.3 Other Procedures 22 REFERENCES FOR SECTION 13.5 24 13.6 INDUSTRIAL SECURITY 25 FIGURES Figure 13.1-1 Site Vice President Figure 13.1-2 Plant Manager Figure 13.1-3 Site Engineering Figure 13.1-4 Site Work Management Page 2 of 3 Revision 26 5/2016 | |||
GINNA/UFSAR Figure 13.1-5 Site Training Figure 13.1-6 Site Operations Figure 13.1-7 Site Radiation Protection Figure 13.1-8 Site Chemistry, Environment and Radwaste Figure 13.1-9 Site Maintenance Page 3 of 3 Revision 26 5/2016 | |||
GINNA/UFSAR 14 INITIAL TEST PROGRAM 1 14.1 | |||
==SUMMARY== | |||
OF TEST PROGRAM AND OBJECTIVES 2 14.1.1 STARTUP AND POWER TESTING AT 1300 MEGAWATTS THER- 2 MAL 14.1.1.1 Summary 2 14.1.1.2 Tests Prior to Reactor Fueling 3 14.1.1.2.1 Summary 3 14.1.1.2.2 Test Objectives 4 14.1.1.3 Final Plant Preparation 10 14.1.1.3.1 Core Loading 10 14.1.1.3.2 Postloading Tests 11 14.1.1.4 Initial Testing in the Operating Reactor 12 14.1.1.4.1 General 12 14.1.1.4.2 Initial Criticality 12 14.1.1.4.3 Zero Power Testing 12 14.1.1.4.4 Power Level Escalation 13 14.1.1.4.5 Post Startup Surveillance and Testing Requirements 13 14.1.2 POWER TEST PROGRAM TO 1520 MEGAWATTS THERMAL 14 REFERENCES FOR SECTION 14.1 15 Table 14.1-1 INITIAL TESTING | |||
==SUMMARY== | |||
- INITIAL CRITICALITY 16 THROUGH 100 - HOUR ACCEPTANCE TEST 14.2 INITIAL ORGANIZATION AND STAFFING 19 14.2.1 INITIAL STARTUP AND OPERATING STRUCTURE 19 14.2.2 ONSITE TRAINING PRIOR TO STARTUP 19 14.3 TEST PROCEDURES 21 14.3.1 PRE FUEL LOADING TESTS 21 14.3.2 POST FUEL LOADING TESTS 21 14.4 CONDUCT OF TEST PROGRAM 22 14.4.1 CONDUCT OF INITIAL TEST PROGRAM 22 14.4.2 REVIEW, EVALUATION, AND APPROVAL OF TEST RESULTS 23 14.4.3 TEST RECORDS 23 Page 1 of 6 Revision 26 5/2016 | |||
GINNA/UFSAR 14.5 TEST PROGRAM SCHEDULE 24 14.5.1 INITIAL CRITICALITY TO ACCEPTANCE 24 14.5.2 1520 MEGAWATTS THERMAL POWER TEST PROGRAM 25 14.6 INDIVIDUAL TEST DESCRIPTIONS 26 14.6.1 INITIAL STARTUP AND POWER TEST PROGRAM 26 14.6.1.1 Safety Injection Systems Preoperational Tests 26 14.6.1.1.1 Safety Injection Test 26 14.6.1.1.2 Accumulator Blowdown Test 30 14.6.1.1.3 Safety Injection Flow Test 33 14.6.1.1.4 Containment Spray System Test 34 14.6.1.1.5 Residual Heat Removal System Test 34 14.6.1.1.6 Safeguards Systems Operational Test 34 14.6.1.1.7 Emergency Diesel Generator Test 35 14.6.1.1.8 Direct Current Test 36 14.6.1.2 Preoperational Instrumentation and Control Tests 36 14.6.1.2.1 Reactor Coolant System Pressure Comparison Test 36 14.6.1.2.2 Resistance Temperature Detector Cross Calibration Test 36 14.6.1.2.3 Steam Generator Manual Control and Level Instrumentation Test 37 14.6.1.2.4 Rod Position Indication System Test 37 14.6.1.2.5 Rod Stepping Test 38 14.6.1.2.6 Rod Cluster Control Assembly Drop Time and Partial Length Rods 38 Operational Tests 14.6.1.2.7 In-Core Thermocouples Test 38 14.6.1.2.8 Movable In-Core Detector System Test 38 14.6.1.2.9 Reactor Makeup Blender and Boric Acid Transfer Pumps Operational 38 Test 14.6.1.2.10 Pressurizer Level Control Test 39 14.6.1.2.11 Pressurizer Pressure Control Test 40 14.6.1.2.12 Steam Dump Test 41 14.6.1.2.13 Radiation Monitoring System Operational Test 41 14.6.1.2.14 Reactor Coolant System Flow Measurement Test 42 14.6.1.2.15 Nuclear Instrumentation Test 42 14.6.1.3 Safety and Relief Valve Tests 42 Page 2 of 6 Revision 26 5/2016 | |||
GINNA/UFSAR 14.6.1.3.1 Pressurizer Safety Valve Test 42 14.6.1.3.2 Main Steam Safety Valve Test 42 14.6.1.4 Waste Systems Tests 43 14.6.1.4.1 Liquid Waste Concentration Demonstration Test 43 14.6.1.4.2 Waste Disposal System Gaseous Waste Test 43 14.6.1.4.3 Liquid Waste Processing Test 44 14.6.1.5 Reactor Coolant System Measurement Tests 44 14.6.1.5.1 Reactor Vessel Internals Measurement Test 44 14.6.1.5.2 Reactor Coolant System Vibration Test 46 14.6.1.5.3 Preoperational Reactor Coolant System Leakage Test 46 14.6.1.5.4 Reactor Coolant System Thermal Expansion Test 47 14.6.1.5.5 Flow Coastdown Test 47 14.6.1.5.6 Natural Circulation Test 47 14.6.1.6 Miscellaneous Safety-Related Tests 48 14.6.1.6.1 Backfeed from the 115-Kilovolt Grid Test 48 14.6.1.6.2 Blackout Test Without Safety Injection Test 49 14.6.1.6.3 Main Steam Isolation Valve Test 49 14.6.1.6.4 Fire Service Water Test 50 14.6.1.6.5 Electrical System Logic Test 50 14.6.1.6.6 Reactor Trip System (RTS) Operational Test 50 14.6.1.6.7 Reactor Coolant System Hydro Test 51 14.6.1.6.8 Ventilation Systems Test 51 14.6.1.6.9 Preoperational Containment Vessel Leak Rate Test 51 14.6.1.6.10 Structural Integrity Test 52 14.6.1.6.11 Reactor Trip System (RTS) Operation Time Response Test 53 14.6.1.7 Operational and Transient Tests 53 14.6.1.7.1 Ten Percent Load Swing Test at Thirty Percent Power 53 14.6.1.7.2 Generator Trip Test 54 14.6.1.7.3 Ten Percent Load Swing Test at Seventy-Five Percent Power Level 54 14.6.1.7.4 Fifty Percent Load Reduction from Seventy-Five Percent Power Level 54 Test 14.6.1.7.5 One Hundred Percent Power Level Transient Tests 55 Page 3 of 6 Revision 26 5/2016 | |||
GINNA/UFSAR 14.6.1.7.6 Operational Dynamic Rod Drop Test 57 14.6.1.7.7 Delta T Zero Power Alignment and Delta T Channel Span Adjustment 58 Tests 14.6.1.7.8 Nuclear Instrumentation Calibration and Reactor Coolant System Flow 58 Confirmation 14.6.1.7.9 Ex-Core In-Core Calibration Test 61 14.6.1.8 Startup Physics Testing 61 14.6.1.8.1 Introduction 61 14.6.1.8.2 Power Distribution Measurements 62 14.6.1.8.3 Zero Power Critical Boron Concentrations 62 14.6.1.8.4 Reactivity Coefficients and Shutdown Margin 62 14.6.1.8.5 Ejected and Dropped Rod Worths 62 14.6.1.8.6 Xenon Oscillation Test 62 14.6.2 POWER TEST PROGRAM TO 1520 MEGAWATTS THERMAL 63 14.6.2.1 Test Description 63 14.6.2.2 Steam Generator Moisture Carryover Tests 65 14.6.2.3 Assembly Delta T Measurements 65 14.6.2.4 Plant Radiation Surveys 65 14.6.2.5 Reactor Physics Measurements 66 14.6.2.5.1 Zero Power Measurements 66 14.6.2.5.2 At-Power Measurements 66 Table 14.6-1a ACCUMULATOR BLOWDOWN TEST RESULTS - OBSERVED 69 RESULTS Table 14.6-1b ACCUMULATOR BLOWDOWN TEST RESULTS - PREDICTION 70 OF FINAL PRESSURE Table 14.6-1c ACCUMULATOR BLOWDOWN TEST RESULTS - PIPE RESIS- 71 TANCE Table 14.6-2 BEGINNING OF CYCLE ZERO POWER CRITICAL BORON CON- 72 CENTRATIONS Table 14.6-3 REACTIVITY COEFFICIENTS AND SHUTDOWN MARGIN 73 Table 14.6-4 EJECTED AND DROPPED ROD WORTHS 74 Table 14.6-5 AVERAGE INCREASE IN CONTAINMENT RADIATION LEVELS 75 DURING UPRATING PROGRAM TO 1520 MEGAWATTS THER-MAL Table 14.6-6 | |||
==SUMMARY== | |||
OF MEASURED PARAMETERS AT HOT ZERO 76 POWER PRIOR TO UPRATING TO 1520 MEGAWATTS THERMAL Page 4 of 6 Revision 26 5/2016 | |||
GINNA/UFSAR Table 14.6-7 SELECTED DATA FOR FLUX MAPS 77 FIGURES Figure 14.2-1 Initial Plant Organization Figure 14.6-1 Original Safety Injection System Functional Test Data Sheet (Typical) | |||
Sheet 1 - | |||
Figure 14.6-1 Original Safety Injection System Functional Test Data Sheet (Typical) | |||
Sheet 2 - | |||
Figure 14.6-2 Rod Cluster Control Assembly Drop Time Test Results (Typical) | |||
Sheet 1 - | |||
Figure 14.6-2 Rod Cluster Control Assembly Drop Time Test Results (Typical) | |||
Sheet 2 - | |||
Figure 14.6-2 Rod Cluster Control Assembly Drop Time Test Results (Typical) | |||
Sheet 3 - | |||
Figure 14.6-2 Rod Cluster Control Assembly Drop Time Test Results (Typical) | |||
Sheet 4 - | |||
Figure 14.6-3 Partial-Length Rods Operation Test Results (Typical) | |||
Figure 14.6-4 Radiation Monitoring System Operational Test Data Sheet (Typical) | |||
Sheet 1 - | |||
Figure 14.6-4 Radiation Monitoring System Operational Test Data Sheet (Typical) | |||
Sheet 2 - | |||
Figure 14.6-5 Pressurizer Safety Valve Test Data Sheet Figure 14.6-6 Reactor Vessel Internals Displacement Indicators Figure 14.6-7 Reactor Vessel Accelerometers Sheet 1 - | |||
Figure 14.6-7 Reactor Vessel Accelerometers Sheet 2 - | |||
Figure 14.6-7 Reactor Vessel Accelerometers Sheet 3 - | |||
Figure 14.6-7 Reactor Vessel Accelerometers Sheet 4 - | |||
Figure 14.6-8 Reactor Coolant System Vibration Test Data Sheet Sheet 1 - | |||
Figure 14.6-8 Reactor Coolant System Vibration Test Data Sheet Sheet 2 - | |||
Figure 14.6-9 Reactor Coolant Pump Flow Coastdown Test Sheet 1 - | |||
Figure 14.6-9 Reactor Coolant Pump Flow Coastdown Test Sheet 2 - | |||
Figure 14.6-9 Reactor Coolant Pump Flow Coastdown Test Sheet 3 - | |||
Page 5 of 6 Revision 26 5/2016 | |||
GINNA/UFSAR Figure 14.6-9 Reactor Coolant Pump Flow Coastdown Test Sheet 4 - | |||
Figure 14.6-9 Reactor Coolant Pump Flow Coastdown Test Sheet 5 - | |||
Figure 14.6-10 Natural Circulation Test Data Sheet 1 - | |||
Figure 14.6-10 Natural Circulation Test Data Sheet 2 - | |||
Figure 14.6-11 Preoperational Containment Leak Rate Test Data Figure 14.6-12 Preoperational Containment Leak Rate Test, Pressurization System Figure 14.6-13 50% Load Reduction From 75% Power Level Test, Recorded Process Sheet 1 - Variables Figure 14.6-13 50% Load Reduction From 75% Power Level Test, Recorded Process Sheet 2 - Variables Figure 14.6-13 50% Load Reduction From 75% Power Level Test, Recorded Process Sheet 3 - Variables Figure 14.6-13 50% Load Reduction From 75% Power Level Test, Recorded Process Sheet 4 - Variables Figure 14.6-13 50% Load Reduction From 75% Power Level Test, Recorded Process Sheet 5 - Variables Figure 14.6-14 Operational Dynamic Rod Drop Test, Nuclear Power Channel Signals Figure 14.6-15 Ginna 1520 Megawatts Power Test Program, March 1 Through April 15, 1972 Figure 14.6-16 Ginna Uprating, Control Bank D Differential and Integral Worth, Cycle 1B 7800 MWd/MTU, Hot Zero Power Figure 14.6-17 Ginna Uprating, Control Bank C Differential and Integral Worth, Cycle 1B 7800 MWd/MTU, Hot Zero Power Figure 14.6-18 Ginna Uprating, Boron Concentration Versus Reactivity Insertion, Cycle 1B 7800 MWd/MTU, Hot Zero Power Figure 14.6-19 Ginna Uprating, Isothermal Temperature Coefficient Versus Boron Concentration, Cycle 1B 7800 MWd/MTU, Hot Zero Power Figure 14.6-20 Relative Power During Uprating: 1300, 1380, 1455 MWt Figure 14.6-21 Relative Power During Uprating: 1455 MWt Figure 14.6-22 Ginna Uprating, Power Range Output Versus Core Power, Channel NE41 Figure 14.6-23 Ginna Uprating, Axial Offset Calibration, Channel 41 Output Normal-ized to 1520 MWt Figure 14.6-24 Relative Power at 1520 MWt, April 12, 1972 Page 6 of 6 Revision 26 5/2016 | |||
15 ACCIDENT ANALYSES 1 15.0 GENERAL 2 15.0.1 INITIAL CONDITIONS 2 15.0.1.1 Assumed Values of Initial Conditions 2 15.0.2 POWER DISTRIBUTION 3 15.0.3 REACTIVITY COEFFICIENTS ASSUMED IN THE ACCIDENT 4 ANALY- SES 15.0.4 ROD CLUSTER CONTROL ASSEMBLY INSERTION 4 CHARACTERIS- TICS 15.0.5 TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN THE 4 ACCI- DENT ANALYSES 15.0.6 INSTRUMENTATION DRIFT AND CALORIMETRIC ERRORS - 5 POWER RANGE NEUTRON FLUX 15.0.7 COMPUTER CODES 5 15.0.7.1 FACTRAN 6 15.0.7.2 RETRAN 6 15.0.7.3 TWINKLE 6 15.0.7.4 VIPRE 7 15.0.7.5 ADVANCED NODAL CODE (ANC) 7 15.0.8 CLASSIFICATION OF PLANT CONDITIONS 7 15.0.8.1 Condition I - Normal Operation 8 15.0.8.2 Condition II - Faults of Moderate Frequency 8 15.0.8.3 Condition III - Infrequent Faults 8 15.0.8.4 Condition IV - Limiting Faults 8 15.0.9 UFSAR Re-write 9 15.0.9.1 General Layout 9 15.0.9.2 Interpretation of Operator Action Times 9 | |||
==15.0 REFERENCES== | |||
FOR SECTION 15.0 10 Table 15.0-1 NSSS PCWG Parameters for Ginna Station Uprate Program 11 Table 15.0-1 NSSS PCWG Parameters for Ginna Station Uprate Program 12 Table 15.0-2 Non-LOCA Analysis Limits and Analysis Results 13 Table 15.0-3 Non-LOCA Plant Initial Condition Assumptions 16 Table 15.0-4 Pressurizer and Main Steam System (MSS) Pressure Relief 17 Assump- tions Table 15.0-5 Core Kinetics Parameters and Reactivity Feedback Coefficients 21 Table 15.0-6 Summary of RPS and ESFAS Functions Actuated 22 Page 1 of 24 Revision 26 5/2016 | |||
Table 15.0-7 Overtemperature and Overpower T Setpoints 25 Table 15.0-8 DETERMINATION OF MAXIMUM OVERPOWER TRIP POINT 26 | |||
- POWER RANGE NEUTRON FLUX CHANNEL - BASED ON NOMINAL SETPOINT CONSIDERING INHERENT INSTRU-MENT ERRORS Table 15.0-9 Summary of Initial Conditions and Computer Codes Used 27 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 30 15.1.1 DECREASE IN FEEDWATER TEMPERATURE 30 15.1.1.1 Description of Event 30 15.1.1.2 Frequency of Event 31 15.1.1.3 Event Analysis 31 15.1.1.3.1 Protective Features 31 15.1.1.3.2 Single Failures Assumed 31 15.1.1.3.3 Operator Actions Assumed 31 15.1.1.3.4 Chronological Description of Event 31 15.1.1.3.5 Impact on Fission Product Barriers 31 15.1.1.4 Reactor Core and Plant System Evaluation 32 15.1.1.4.1 Input Parameters and Initial Conditions 32 15.1.1.4.2 Methodology 32 15.1.1.4.3 Acceptance Criteria 32 15.1.1.4.4 Results 32 15.1.1.5 Radiological Consequences 32 15.1.1.6 Conclusions 32 15.1.2 INCREASE IN FEEDWATER FLOW 33 15.1.2.1 Increase in Feedwater Flow at Full Power 33 15.1.2.1.1 Description of Event 33 15.1.2.1.2 Frequency of Event 33 15.1.2.1.3 Event Analysis 33 15.1.2.1.3.1 Protective Features 34 15.1.2.1.3.2 Single Failures Assumed 34 15.1.2.1.3.3 Operator Actions Assumed 35 15.1.2.1.3.4 Chronological Description of Event 35 15.1.2.1.3.5 Impact on Fission Product Barriers 35 15.1.2.1.4 Reactor Core and Plant System Evaluation 35 15.1.2.1.4.1 Input Parameters and Initial Conditions 35 15.1.2.1.4.2 Method of Analysis 36 Page 2 of 24 Revision 26 5/2016 | |||
15.1.2.1.4.3 Acceptance Criteria 36 15.1.2.1.4.4 Results 36 15.1.2.1.5 Radiological Consequences 37 15.1.2.1.6 Conclusion 37 15.1.2.2 Increase in Feedwater Flow at Zero Power 37 15.1.2.2.1 Description of Event 37 15.1.2.2.2 Frequency of Event 37 15.1.2.2.3 Event Analysis 37 15.1.2.2.3.1 Protective Features 37 15.1.2.2.3.2 Single Failures Assumed 38 15.1.2.2.3.3 Operator Actions Assumed 38 15.1.2.2.3.4 Chronological Description of Event 38 15.1.2.2.3.5 Impact on Fission Product Barriers 38 15.1.2.2.4 Reactor Core and Plant System Evaluation 38 15.1.2.2.4.1 Input Parameters and Initial Conditions 38 15.1.2.2.4.2 Methodology 39 15.1.2.2.4.3 Acceptance Criteria 39 15.1.2.2.5 Radiological Consequences 39 15.1.2.2.6 Conclusion 39 15.1.3 EXCESSIVE LOAD INCREASE INCIDENT 39 15.1.3.1 Description of Event 39 15.1.3.2 Frequency of Event 40 15.1.3.3 Event Analysis 40 15.1.3.3.1 Protective Features 40 15.1.3.3.2 Single Failures Assumed 41 15.1.3.3.3 Operator Actions Assumed 41 15.1.3.3.4 Chronological Description of Event 41 15.1.3.3.5 Impact on Fission Product Barriers 41 15.1.3.4 Reactor Core and Plant System Evaluation 41 15.1.3.4.1 Input Parameters and Initial Conditions 41 15.1.3.4.2 Methodology 42 15.1.3.4.3 Acceptance Criteria 42 15.1.3.5 Radiological Consequences 42 15.1.3.6 Conclusions 42 15.1.4 INADVERTENT OPENING OF A STEAM GENERATOR RELIEF/ 43 SAFETY VALVE 15.1.5 SPECTRUM OF STEAM SYSTEM PIPING FAILURES INSIDE AND 43 OUTSIDE OF CONTAINMENT Page 3 of 24 Revision 26 5/2016 | |||
15.1.5.1 Description of Event 43 15.1.5.2 Frequency of Event 43 15.1.5.3 Event Analysis 44 15.1.5.3.1 Protective Features 44 15.1.5.3.2 Single Failures Assumed 45 15.1.5.3.3 Operator Actions Assumed 45 15.1.5.3.4 Chronological Description of Event 45 15.1.5.3.5 Impact on Fission Product Barriers 45 15.1.5.4 Reactor Core and Plant System Evaluation 46 15.1.5.4.1 Input Parameters and Initial Conditions 46 15.1.5.4.2 Methodology 47 15.1.5.4.3 Acceptance Criteria 48 15.1.5.4.4 Results 48 15.1.5.5 Radiological Consequences 49 15.1.5.6 Conclusions 50 15.1.5.7 Supplemental Evaluations 50 15.1.5.7.1 SEV-1073 50 15.1.5.7.2 HZP 6 Inch Steamline Break 50 15.1.5.7.3 High Steam Flow Setpoint Increase Evaluation 50 15.1.5.7.4 Steamline Rupture a Full Power 51 15.1.5.8 Potential for Containment Overpressurization 51 15.1.6 COMBINED STEAM GENERATOR ATMOSPHERIC RELIEF VALVE 51 (ARV) AND MAIN FEEDWATER REGULATING VALVE (MFRV) | |||
FAIL- URES 15.1.6.1 Description of Event 51 15.1.6.2 Frequency of Event 52 15.1.6.3 Event Analysis 52 15.1.6.3.1 Protective Features 53 15.1.6.3.2 Single Failures Assumed 53 15.1.6.3.3 Operator Actions Assumed 53 15.1.6.3.4 Chronological Description of Event 54 15.1.6.3.5 Impact on Fission Product Barriers 54 15.1.6.4 Reactor Core and Plant System Evaluation 54 15.1.6.4.1 Input Parameters and Initial Conditions 54 15.1.6.4.2 Methodology 55 15.1.6.4.3 Acceptance Criteria 56 15.1.6.4.4 Results 56 15.1.6.5 Radiological Consequences 56 Page 4 of 24 Revision 26 5/2016 | |||
15.1.6.6 Conclusions 57 | |||
==15.1 REFERENCES== | |||
FOR SECTION 15.1 58 Table 15.1-1 TIME SEQUENCE OF EVENTS FOR FEEDWATER MALFUNC- 59 TION TRANSIENTS HOT FULL POWER - SINGLE LOOP - WITH ROD CONTROL Table 15.1-2 TIME SEQUENCE OF EVENTS FOR FEEDWATER MALFUNC- 60 TION TRANSIENTS HOT FULL POWER - SINGLE LOOP - | |||
WITH- OUT ROD CONTROL Table 15.1-3 TIME SEQUENCE OF EVENTS FOR FEEDWATER MALFUNC- 61 TION TRANSIENTS HOT FULL POWER - MULTI LOOP - | |||
WITH ROD CONTROL Table 15.1-4 TIME SEQUENCE OF EVENTS FOR FEEDWATER MALFUNC- 62 TION TRANSIENTS HOT FULL POWER - MULTI LOOP - | |||
WITH- OUT ROD CONTROL Table 15.1-5 Table DELETED 63 Table 15.1-6 TIME SEQUENCE OF EVENTS FOR STEAM LINE RUPTURE 64 Table 15.1-7 | |||
==SUMMARY== | |||
OF MAIN FEEDWATER REGULATING VALVES 65 (MFRV)/STEAM GENERATOR ATMOSPHERIC RELIEF VALVE (ARV) COMBINATION FAILURE CASES EVALUATED Table 15.1-8 MSLB DOSE ANALYSIS ASSUMPTIONS 66 Table 15.1-9 RESULTS FOR MAIN STEAM LINE BREAK, REM TEDE 68 Table 15.1-10 TIME SEQUENCE OF EVENTS FOR THE COMBINED FAILURE 69 OF TWO MFRV's AND TWO ARV's AT HOT FULL POWER 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 70 15.2.1 STEAM PRESSURE REGULATOR MALFUNCTION OR FAILURE 70 THAT RESULTS IN DECREASING STEAM FLOW 15.2.2 LOSS OF EXTERNAL ELECTRICAL LOAD 70 15.2.2.1 Description of Event 70 15.2.2.2 Frequency of Event 70 15.2.2.3 Event Analysis 70 15.2.2.3.1 Protective Features 71 15.2.2.3.2 Single Failures Assumed 71 15.2.2.3.3 Operator Actions Assumed 71 15.2.2.3.4 Chronological Description of Event 72 15.2.2.3.5 Impact on Fission Product Barriers 72 15.2.2.4 Reactor Core and Plant System Evaluation 72 15.2.2.4.1 Input Parameters and Initial Conditions 72 15.2.2.4.2 Method of Analysis 73 15.2.2.4.3 Acceptance Criteria 74 15.2.2.4.4 Results 74 15.2.2.5 Radiological Consequences 75 Page 5 of 24 Revision 26 5/2016 | |||
15.2.2.6 Conclusions 75 15.2.2.7 Supplemental Evaluations 75 15.2.3 TURBINE TRIP 75 15.2.4 LOSS OF CONDENSER VACUUM 75 15.2.5 LOSS OF ALL ALTERNATING CURRENT POWER TO THE STATION 76 AUXILIARIES 15.2.5.1 Description of the event 76 15.2.5.2 Frequency of Event 76 15.2.5.3 Event Analysis 77 15.2.5.3.1 Protective Features 77 15.2.5.3.2 Single Failures Assumed 78 15.2.5.3.3 Operator Actions Assumed 78 15.2.5.3.4 Chronological Description of Event 78 15.2.5.3.5 Impact on Fission Product Barriers 78 15.2.5.4 Reactor Core and Plant System Evaluation 79 15.2.5.4.1 Input Parameters and Initial Conditions 79 15.2.5.4.2 Method of Analysis 80 15.2.5.4.3 Acceptance Criteria 80 15.2.5.4.4 Results 81 15.2.5.5 Radiological Consequences 81 15.2.5.6 Conclusions 81 15.2.5.7 Supplemental Evaluations 82 15.2.6 LOSS OF NORMAL FEEDWATER FLOW 82 15.2.6.1 Description of Event 82 15.2.6.2 Frequency of Event 83 15.2.6.3 Event Analysis 83 15.2.6.3.1 Protective Features 83 15.2.6.3.2 Single Failures Assumed 84 15.2.6.3.3 Operator Actions Assumed 84 15.2.6.3.4 Chronological Description of Event 84 15.2.6.3.5 Impact on Fission Product Barriers 84 15.2.6.4 Reactor Core and Plant System Evaluation 84 15.2.6.4.1 Input Parameters and Initial Conditions 84 15.2.6.4.2 Method of Analysis 86 15.2.6.4.3 Acceptance Criteria 86 15.2.6.4.4 Results 86 15.2.6.5 Radiological Consequences 87 15.2.6.6 Conclusions 87 Page 6 of 24 Revision 26 5/2016 | |||
15.2.6.7 Supplemental Evaluations 87 15.2.7 FEEDWATER SYSTEM PIPE BREAKS 88 15.2.7.1 Description of Event 88 15.2.7.2 Frequency of Event 88 15.2.7.3 Event Analysis 88 15.2.7.3.1 Protective Features 88 15.2.7.3.2 Single Failures Assumed 89 15.2.7.3.3 Operator Actions Assumed 90 15.2.7.3.4 Chronological Description of Event 90 15.2.7.3.5 Impact on Fission Product Barriers 90 15.2.7.4 Reactor Core and Plant System Evaluation 90 15.2.7.4.1 Input Parameters and Initial Conditions 90 15.2.7.4.2 Method of Analysis 92 15.2.7.4.3 Acceptance Criteria 92 15.2.7.4.4 Results 93 15.2.7.5 Radiological Consequences 94 15.2.7.6 Conclusions 94 | |||
==15.2 REFERENCES== | |||
FOR SECTION 15.2 95 Table 15.2-1 TIME SEQUENCE OF EVENTS FOR LOSS OF EXTERNAL ELEC- 96 TRICAL LOAD Table 15.2-2 TIME SEQUENCE OF EVENTS FOR LOSS OF OFFSITE ALTER- 97 NATING CURRENT POWER TO THE STATION AUXILIARIES Table 15.2-3 Table DELETED 98 Table 15.2-4 TIME SEQUENCE OF EVENTS FOR LOSS OF NORMAL FEED- WATER 99 FLOW Table 15.2-5 TIME SEQUENCE OF EVENTS FOR THE FEEDWATER LINE PIPE 100 BREAK (0.3 FT2 BREAK AREA) 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE 101 15.3.1 FLOW COASTDOWN ACCIDENTS 101 15.3.1.1 Description of Event 101 15.3.1.2 Frequency of Event 101 15.3.1.3 Event Analysis 101 15.3.1.3.1 Protective Features 102 15.3.1.3.2 Single Failures Assumed 103 15.3.1.3.3 Operator Actions Assumed 103 15.3.1.3.4 Chronological Description of Event 103 15.3.1.3.5 Impact on Fission Product Barriers 103 15.3.1.4 Reactor Core and Plant System Evaluation 103 Page 7 of 24 Revision 26 5/2016 | |||
15.3.1.4.1 Input Parameters and Initial Conditions 103 15.3.1.4.2 Method of Analysis 104 15.3.1.4.3 Acceptance Criteria 104 15.3.1.4.4 Results 104 15.3.1.5 Radiological Consequences 105 15.3.1.6 Conclusions 105 15.3.2 LOCKED ROTOR ACCIDENT 106 15.3.2.1 Description of Event 106 15.3.2.2 Frequency of Event 106 15.3.2.3 Event Analysis 106 15.3.2.3.1 Protective Features 106 15.3.2.3.2 Single Failures Assumed 107 15.3.2.3.3 Operator Actions Assumed 107 15.3.2.3.4 Chronological Description of Event 107 15.3.2.3.5 Impact on Fission Product Barriers 107 15.3.2.4 Reactor Core and Plant System Evaluation 108 15.3.2.4.1 Input Parameters and Initial Conditions 108 15.3.2.4.2 Method of Analysis 108 15.3.2.4.3 Acceptance Criteria 109 15.3.2.4.4 Results 110 15.3.2.5 Radiological Consequences 110 15.3.2.6 Conclusions 110 | |||
==15.3 REFERENCES== | |||
FOR SECTION 15.3 111 Table 15.3-1 TIME SEQUENCE OF EVENTS FOR LOSS OF REACTOR COOL- 112 ANT FLOW Table 15.3-2 | |||
==SUMMARY== | |||
OF LIMITING RESULTS FOR LOCKED ROTOR 113 ACCIDENT Table 15.3-3 TIME SEQUENCE OF EVENTS FOR LOCKED ROTOR INCIDENT 114 Table 15.3-4 LR Dose Analysis Assumptions 115 Table 15.13-5 RESULTS FOR LOCKED ROTOR 116 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 117 15.4.1 UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY WITH- 117 DRAWAL FROM A SUBCRITICAL CONDITION 15.4.1.1 Description of Event 117 15.4.1.2 Frequency of Event 117 15.4.1.3 Event Analysis 117 15.4.1.3.1 Protective Features 117 15.4.1.3.2 Single Failures Assumed 118 15.4.1.3.3 Operator Actions Assumed 118 15.4.1.3.4 Chronological Description of Event 118 Page 8 of 24 Revision 26 5/2016 | |||
15.4.1.3.5 Impact on Fission Product Barriers 118 15.4.1.4 Reactor Core and Plant System Evaluation 118 15.4.1.4.1 Input Parameters and Initial Conditions 118 15.4.1.4.2 Methodology 120 15.4.1.4.3 Acceptance Criteria 120 15.4.1.4.4 Results 120 15.4.1.5 Radiological Evaluation 121 15.4.1.6 Conclusions 121 15.4.2 UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY WITH- 121 DRAWAL AT POWER 15.4.2.1 Description of Event 121 15.4.2.2 Frequency of Event 121 15.4.2.3 Event Analysis 121 15.4.2.3.1 Protective Features 122 15.4.2.3.2 Single Failures Assumed 122 15.4.2.3.3 Operator Actions Assumed 122 15.4.2.3.4 Chronological Description of Event 122 15.4.2.3.5 Impact on Fission Product Barriers 122 15.4.2.4 Reactor Core and Plant System Evaluation 123 15.4.2.4.1 Input Parameters and Initial Conditions 123 15.4.2.4.2 Methodology 124 15.4.2.4.3 Acceptance Criteria 124 15.4.2.4.4 Results 125 15.4.2.5 Radiological Evaluation 126 15.4.2.6 Conclusions 126 15.4.3 STARTUP OF AN INACTIVE REACTOR COOLANT LOOP 126 15.4.3.1 Description of Event 126 15.4.3.2 Frequency of Event 127 15.4.3.3 Event Analysis 127 15.4.3.3.1 Protective Features 127 15.4.3.3.2 Single Failures Assumed 127 15.4.3.3.3 Operator Actions Assumed 127 15.4.3.3.4 Chronological Description of Event 127 15.4.3.3.5 Impact on Fission Product Barriers 128 15.4.3.4 Reactor Core and Plant System Evaluation 128 15.4.3.4.1 Input Parameters and Initial Conditions 128 15.4.3.4.2 Methodology 128 15.4.3.4.3 Acceptance Criteria 129 Page 9 of 24 Revision 26 5/2016 | |||
15.4.3.4.4 Results 129 15.4.3.4.5 Effect of 18 Month Fuel Cycle Changes 130 15.4.3.5 Radiological Evaluation 130 15.4.3.6 Conclusions 130 15.4.4 CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION 130 15.4.4.1 Description of Event 130 15.4.4.2 Frequency of Event 131 15.4.4.3 Event Analysis 131 15.4.4.3.1 Protective Features and Single Failures Assumed 131 15.4.4.3.1.1 Reactor in Mode 1 or Mode 2 131 15.4.4.3.1.2 Reactor in MODES 3 to 6 132 15.4.4.3.1.3 Indication and Alarms 132 15.4.4.3.2 Operator Actions Assumed 132 15.4.4.3.3 Chronological Description of Event 133 15.4.4.3.4 Impact on Fission Product Barriers 133 15.4.4.4 Reactor Core and Plant System Evaluation 133 15.4.4.4.1 Methodology 133 15.4.4.4.2 Acceptance Criteria 133 15.4.4.4.3 Dilution During Refueling (MODE 6) 134 15.4.4.4.3.1 Input Parameters and Initial Conditions 134 15.4.4.4.3.2 Results 135 15.4.4.4.4 Dilution During Cold Shutdown (MODE 5) 135 15.4.4.4.5 Dilution at Startup (MODE 2) 135 15.4.4.4.5.1 Input Parameters and Initial Conditions 135 15.4.4.4.5.2 Results 136 15.4.4.4.6 Dilution at Power (MODE 1) 136 15.4.4.4.6.1 Input Parameters and Initial Conditions 136 15.4.4.4.6.2 Results 137 15.4.4.4.7 Dilution from a Single Failure While in Residual Heat Removal Mode - 137 Inad- vertent Draining of the Spray Additive Tank. | |||
15.4.4.4.8 Dilution from a Single Failure While in Residual Heat Removal 137 Mode (MODE 5) -Boron Dilution from the Reactor Coolant Drain Tank. | |||
15.4.4.4.8.1 Input Parameters and Initial Conditions 137 15.4.4.4.8.2 Results 138 15.4.4.4.9 Dilution from a Single Failure While in Residual Heat Removal Mode 138 (MODE 5) -Boron Dilution Due to Resin Changing in the Purification Sys- tem. | |||
15.4.4.4.9.1 Input Parameters and Initial Conditions 138 15.4.4.4.9.2 Results 139 Page 10 of 24 Revision 26 5/2016 | |||
15.4.4.4.10 Dilution from a Single Failure While in Residual Heat Removal Mode 139 (MODE 6) -Boron Dilution from Reactor Coolant Drain Tank After Refuel- ing. | |||
15.4.4.4.10.1 Input Parameters and Initial Conditions 139 15.4.4.4.10.2 Results 139 15.4.4.5 Radiological Evaluation 140 15.4.4.6 Conclusions 140 15.4.5 RUPTURE OF A CONTROL ROD DRIVE MECHANISM HOUSING 140 | |||
- ROD CLUSTER CONTROL ASSEMBLY EJECTION 15.4.5.1 Description of Event 140 15.4.5.1.1 Nuclear Design 141 15.4.5.1.2 Effects on Adjacent Housings 141 15.4.5.2 Frequency of Event 141 15.4.5.3 Event Analysis 141 15.4.5.3.1 Protective Features 141 15.4.5.3.2 Single Failures Assumed 142 15.4.5.3.3 Operator Actions Assumed 142 15.4.5.3.4 Chronological Description of Event 142 15.4.5.3.5 Impact on Fission Product Barriers 142 15.4.5.4 Reactor Core and Plant System Evaluation 143 15.4.5.4.1 Input Parameters and Initial Conditions 143 15.4.5.4.2 Methodology 143 15.4.5.4.2.1 Average Core Analysis 144 15.4.5.4.2.2 Ejected Rod Worths and Hot Channel Factors 144 15.4.5.4.2.3 Hot Spot Analysis 144 15.4.5.4.2.4 Reactivity Feedback Weighting Factors 145 15.4.5.4.2.5 System Overpressure Analysis 145 15.4.5.4.3 Acceptance Criteria 146 15.4.5.4.4 Results 146 15.4.5.4.4.1 Beginning of Life, Full Power - Case (1) 147 15.4.5.4.4.2 Beginning of Life, Zero Power - Case (2) 147 15.4.5.4.4.3 End of Life, Full Power - Case (3) 147 15.4.5.4.4.4 End of Life, Zero Power - Case (4) 147 15.4.5.4.4.5 Pressure Surge 147 15.4.5.4.4.6 Lattice Deformations 148 15.4.5.5 Radiological Evaluation 148 15.4.5.6 Conclusions 148 15.4.6 ROD CLUSTER CONTROL ASSEMBLY DROP 148 Page 11 of 24 Revision 26 5/2016 | |||
15.4.6.1 Description of Event 148 15.4.6.2 Frequency of Event 149 15.4.6.3 Event Analysis 149 15.4.6.3.1 Protective Features 149 15.4.6.3.2 Single Failures Assumed 150 15.4.6.3.3 Operator Actions Assumed 150 15.4.6.3.4 Chronological Description of Event 150 15.4.6.3.5 Impact on Fission Product Barriers 150 15.4.6.4 Reactor Core and Plant System Evaluation 150 15.4.6.4.1 Input Parameters and Initial Conditions 150 15.4.6.4.2 Methodology 151 15.4.6.4.2.1 One or More Dropped Rod Cluster Control Assemblies From the Same Group 151 15.4.6.4.2.2 Dropped Rod Cluster Control Assembly Bank 151 15.4.6.4.2.3 Statically Misaligned Rod Cluster Control Assembly 151 15.4.6.4.3 Acceptance Criteria 151 15.4.6.4.4 Results 152 15.4.6.4.4.1 One or More Dropped Rod Cluster Control Assemblies 152 15.4.6.4.4.2 Dropped Rod Cluster Control Assembly Bank 152 15.4.6.4.4.3 Statically Misaligned Rod Cluster Control Assembly 152 15.4.6.5 Radiological Evaluation 153 15.4.6.6 Conclusions 153 | |||
==15.4 REFERENCES== | |||
FOR SECTION 15.4 154 Table 15.4-1 TIME SEQUENCE OF EVENTS FOR UNCONTROLLED ROD CLUSTER 156 CONTROL ASSEMBLY WITHDRAWAL FROM A SUB- CRITICAL Table 15.4-2 TIME SEQUENCE OF EVENTS FOR UNCONTROLLED ROD CLUSTER 157 CONTROL ASSEMBLY WITHDRAWAL AT POWER Table 15.4-3 PARAMETERS USED IN THE ANALYSIS OF THE ROD CLUS- TER 158 CONTROL ASSEMBLY EJECTION ACCIDENT Table 15.4-4 TIME SEQUENCE OF EVENTS FOR ROD CLUSTER CONTROL 159 ASSEMBLY EJECTION Table 15.4-5 REA CONTAINMENT ASSUMPTIONS 160 Table 15.4-6 RESULTS FOR REA DOSE, REM TEDE 162 15.5 INCREASE IN REACTOR COOLANT INVENTORY 163 | |||
==15.5 REFERENCES== | |||
FOR SECTION 15.5 164 15.6 DECREASE IN REACTOR COOLANT INVENTORY 165 Page 12 of 24 Revision 26 5/2016 | |||
15.6.1 INADVERTENT OPENING OF A PRESSURIZER SAFETY VALVE OR 165 PRESSURIZER POWER OPERATED RELIEF VALVE (PORV) 15.6.1.1 Description of Event 165 15.6.1.2 Frequency of Event 165 15.6.1.3 Event Analysis 165 15.6.1.3.1 Protective Features 165 15.6.1.3.2 Single Failures Assumed 165 15.6.1.3.3 Operator Actions Assumed 165 15.6.1.3.4 Chronological Description of Event 165 15.6.1.3.5 Impact on Fission Product Barriers 165 15.6.1.4 Reactor Core and Plant System Evaluation 166 15.6.1.4.1 Input Parameters and Initial Conditions 166 15.6.1.4.2 Methodology 166 15.6.1.4.3 Acceptance Criteria 166 15.6.1.4.4 Results 167 15.6.1.5 Radiological Consequences 167 15.6.1.6 Conclusions 167 15.6.2 RADIOLOGICAL CONSEQUENCES OF SMALL LINES 167 CARRYING PRIMARY COOLANT OUTSIDE CONTAINMENT 15.6.3 Steam Generator Tube Rupture 168 15.6.3.1 Description of Event 168 15.6.3.2 Frequency of Event 168 15.6.3.3 Event Analysis 169 15.6.3.3.1 Protective Features 169 15.6.3.3.2 Single Failures Assumed 170 15.6.3.3.2.1 Single Failure - Margin to Overfill 170 15.6.3.3.2.2 Single Failure - Mass Release 171 15.6.3.3.3 Operator Actions Assumed 171 15.6.3.3.3.1 Operator Actions to Terminate Tube Rupture Flow 171 15.6.3.3.3.2 Operator Actions Due to Single Failures 173 15.6.3.3.3.3 Operator Actions for Cooldown to MODE 5 (Cold Shutdown) 173 15.6.3.3.4 Chronological Description of Event 174 15.6.3.3.5 Impact on Fission Product Barriers 174 15.6.3.4 Reactor Core and Plant System Evaluation 175 15.6.3.4.1 Input Parameters and Initial Conditions 175 15.6.3.4.2 Methodology 176 15.6.3.4.3 Acceptance Criteria 177 15.6.3.4.4 Results 177 Page 13 of 24 Revision 26 5/2016 | |||
15.6.3.4.4.1 SGTR Margin to Overfill Transient Analysis 177 15.6.3.4.4.2 SGTR Mass Release Transient Analysis 179 15.6.3.5 Radiological Consequences 180 15.6.3.6 Conclusions 181 15.6.4 PRIMARY SYSTEM PIPE RUPTURES 181 15.6.4.1 Loss of Reactor Coolant from Small Ruptured Pipes or From Cracks in Large 181 Pipes Which Actuates Emergency Core Cooling System (ECCS) 15.6.4.1.1 Description of Event 181 15.6.4.1.2 Frequency of Event 182 15.6.4.1.3 Event Analysis 182 15.6.4.1.3.1 Protective Features 182 15.6.4.1.3.2 Single Failures Assumed 183 15.6.4.1.3.3 Operator Actions Assumed 183 15.6.4.1.3.4 Chronological Description of Event 183 15.6.4.1.3.5 Impact on Fission Product Barriers 184 15.6.4.1.4 Reactor Core and Plant System Evaluation 184 15.6.4.1.4.1 Input Parameters and Initial Conditions 184 15.6.4.1.4.2 Methodology 185 15.6.4.1.4.3 Acceptance Criteria 186 15.6.4.1.4.4 Results 186 15.6.4.1.4.5 Effect of Emergency Core Cooling System (ECCS) Evaluation Model Modi- 187 fications 15.6.4.1.5 Radiological Evaluation 187 15.6.4.1.6 Conclusions 187 15.6.4.2 Major Reactor Coolant System Pipe Ruptures (Loss-of-Coolant Accident) 187 15.6.4.2.1 Description of Event 187 15.6.4.2.2 Frequency of Event 189 15.6.4.2.3 Event Analysis 189 15.6.4.2.3.1 Protective Features 189 15.6.4.2.3.2 Single Failures Assumed 190 15.6.4.2.3.3 Operator Actions Assumed 190 15.6.4.2.3.4 Chronological Description of Event 190 15.6.4.2.3.5 Impact on Fission Product Barriers 192 15.6.4.2.4 Reactor Core and Plant System Evaluation 192 15.6.4.2.4.1 Input Parameters and Initial Conditions 192 15.6.4.2.4.2 Methodology 195 15.6.4.2.4.3 Acceptance Criteria 200 Page 14 of 24 Revision 26 5/2016 | |||
15.6.4.2.4.4 Results 200 15.6.4.2.5 Radiological Evaluation 201 15.6.4.2.6 Conclusions 202 | |||
==15.6 REFERENCES== | |||
FOR SECTION 15.6 203 Table 15.6-1 COMPARISON OF NOMINAL AND PLANT PARAMETERS USED 207 IN STEAM GENERATOR TUBE RUPTURE (SGTR) ANALYSIS Table 15.6-2 OPERATOR ACTION TIMES 208 Table 15.6-3 SEQUENCE OF EVENTS - MARGIN TO OVERFILL ANALYSIS 209 Table 15.6-4 OPERATOR ACTION TIMES FOR DESIGN BASIS STEAM GEN- 210 ERATOR TUBE RUPTURE ANALYSIS Table 15.6-5 SEQUENCE OF EVENTS - OFFSITE RADIATION DOSE ANALY- 211 SIS Table 15.6-6 SGTR DOSE ANALYSIS ASSUMPTIONS 212 Table 15.6-7 STEAM RELEASES AND RUPTURE FLOW 214 Table 15.6-8 RESULTS FOR SGTR, REM TEDE 215 Table 15.6-9 TIME SEQUENCE OF EVENTS - ACCIDENTAL DEPRESSURIZA- 216 TION OF THE RCS Table 15.6-10 TOTAL SMALL BREAK LOSS-OF-COOLANT ACCIDENT 217 SAFETY INJECTION AND SPILL FLOW Table 15.6-11 SMALL BREAK LOSS-OF-COOLANT ACCIDENT KEY ASSUMP- 218 TIONS Table 15.6-12 SMALL BREAK LOSS-OF-COOLANT ACCIDENT MAIN STEAM 220 SAFETY VALVE (MSSV) ASSUMPTIONS Table 15.6-13 SMALL BREAK LOSS-OF-COOLANT ACCIDENT TIME 221 SEQUENCE OF EVENTS Table 15.6-14 SMALL BREAK LOSS-OF-COOLANT ACCIDENT FUEL 222 CLADDING RESULTS Table 15.6-15 LARGE BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS 223 TIME SEQUENCE OF EVENTS FOR DECLG BREAK Table 15.6-16 Key LBLOCA Parameters and Initial Transient Assumptions for R. 224 E. Ginna Analysis Table 15.6-17 LARGE BREAK LOCA ANALYSIS SAFETY INJECTION 227 FLOW VERSUS PRESSURE Table 15.6-18a PARAMETERS FOR CONTAINMENT PRESSURE - DRY 229 CON- TAINMENT DATA Table 15.6-18b STRUCTURAL HEAT SINK DATA 230 Table 15.6-19 PLANT OPERATING RANGE ALLOWED BY THE BEST- 232 ESTIMATE LARGE BREAK LOCA ANALYSIS (R. E. GINNA) | |||
Page 15 of 24 Revision 26 5/2016 | |||
Table 15.6-20 LIMITING LARGE BREAK PCT AND OXIDATION RESULTS 234 FOR R. E. GINNA Table 15.6-21 ASSUMPTIONS FOR ANALYSIS OF RADIOLOGICAL CONSE- 235 QUENCES OF THE LOSS-OF-COOLANT ACCIDENT Table 15.6-21A LBLOCA DOSE | |||
==SUMMARY== | |||
, REM TEDE 237 Table 15.6-22 Total Core Activity (Curies) at End of 525-day Fuel Cycle - including 238 Decay Table 15.6-23 Core Inventory Fraction Released into Containment 241 Table 15.6-24 TABLE DELETED 243 15.7 RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 244 15.7.1 RADIOACTIVE GAS WASTE SYSTEM FAILURE 244 15.7.1.1 Gas Decay Tank Rupture 244 15.7.1.1.1 Description of Event 244 15.7.1.1.2 Frequency of Event 244 15.7.1.1.3 Event Analysis 244 15.7.1.1.3.1 Single Failures Assumed 245 15.7.1.1.3.2 Operator Actions Assumed 245 15.7.1.1.3.3 Chronological Description of Event 245 15.7.1.1.3.4 Impact on Fission Product Barriers 245 15.7.1.1.4 Reactor Core and Plant System Evaluation 245 15.7.1.1.4.1 Input Parameters and Initial Conditions 245 15.7.1.1.4.2 Methodology 246 15.7.1.1.4.3 Acceptance Criteria 246 15.7.1.1.4.4 Results 246 15.7.1.1.5 Radiological Evaluation 246 15.7.1.1.6 Conclusions 246 15.7.1.2 Volume Control Tank Rupture 247 15.7.1.2.1 Description of Event 247 15.7.1.2.2 Frequency of Event 247 15.7.1.2.3 Event Analysis 247 15.7.1.2.3.1 Single Failures Assumed 247 15.7.1.2.3.2 Operator Actions Assumed 247 15.7.1.2.3.3 Chronological Description of Event 247 15.7.1.2.3.4 Impact on Fission Product Barriers 248 15.7.1.2.4 Reactor Core and Plant System Evaluation 248 15.7.1.2.4.1 Input Parameters and Initial Conditions 248 15.7.1.2.4.2 Methodology 248 Page 16 of 24 Revision 26 5/2016 | |||
15.7.1.2.4.3 Acceptance Criteria 249 15.7.1.2.4.4 Results 249 15.7.1.2.5 Radiological Evaluation 249 15.7.1.2.6 Conclusions 249 15.7.2 RADIOACTIVE LIQUID WASTE SYSTEM FAILURE 249 15.7.2.1 Description of Event 249 15.7.2.2 Frequency of Event 250 15.7.2.3 Event Analysis 250 15.7.2.3.1 Single Failures Assumed 251 15.7.2.3.2 Operator Actions Assumed 251 15.7.2.3.3 Chronological Description of Event 251 15.7.2.3.4 Impact on Fission Product Barriers 251 15.7.2.4 Reactor Core and Plant System Evaluation 251 15.7.2.4.1 Input Parameters and Initial Conditions 251 15.7.2.4.2 Methodology 252 15.7.2.4.3 Acceptance Criteria 252 15.7.2.4.4 Results 252 15.7.2.4.4.1 Accidental Release of Liquid Waste Assessment 252 15.7.2.4.4.2 Spent Resin Storage Tank Assessment 253 15.7.2.4.5 Effects of 18-month Fuel Cycle 254 15.7.2.5 Radiological Evaluation 254 15.7.2.6 Conclusions 254 15.7.3 FUEL HANDLING ACCIDENTS 254 15.7.3.1 Description of Event 254 15.7.3.1.1 MODE 6 (Refueling) Preparations 254 15.7.3.1.2 Fuel Handling Equipment Safety Features 255 15.7.3.1.3 Fuel Handling Operations Precautions 256 15.7.3.1.4 Consequence of Dropped Fuel Assembly 256 15.7.3.2 Frequency of Event 257 15.7.3.3 Event Analysis 257 15.7.3.3.1 Protective Features 258 15.7.3.3.2 Single Failures Assumed 258 15.7.3.3.3 Operator Actions Assumed 258 15.7.3.3.4 Chronological Description of Event 258 15.7.3.3.5 Impact on Fission Product Barriers 258 15.7.3.4 Reactor Core and Plant System Evaluation 259 15.7.3.4.1 Input Parameters and Initial Conditions 259 Page 17 of 24 Revision 26 5/2016 | |||
15.7.3.4.2 Methodology 259 15.7.3.4.3 Acceptance Criteria 259 15.7.3.4.4 Results 259 15.7.3.5 Radiological Evaluation 259 15.7.3.6 Conclusions 260 | |||
==15.7 REFERENCES== | |||
FOR SECTION 15.7 261 Table 15.7-1 FISSION PRODUCT INVENTORY AND ACTIVITY RELEASED 264 FROM POOL Table 15.7-2 FHA DOSE ANALYSIS ASSUMPTIONS 265 Table 15.7-3 FHA DOSE. REM TEDE 266 Table 15.7-4 Table DELETED 267 Table 15.7-5 Table DELETED 268 Table 15.7-6 Table DELETED 269 15.8 ANTICIPATED TRANSIENTS WITHOUT SCRAM 270 15.8.1 ANTICIPATED TRANSIENTS WITHOUT SCRAM (atws) 270 15.8.2 frequency of event 270 15.8.3 Event Analysis 270 15.8.3.1 Single Failures Assumed 270 15.8.3.2 Operator Actions Assumed 270 15.8.3.3 Chronological Description of Event 271 15.8.3.4 Impact on Fission Product Barriers 271 15.8.4 Reactor Core and Plant System Evaluation 272 15.8.4.1 Input Parameters and Initial Conditions 272 15.8.4.2 Methodology 273 15.8.4.3 Acceptance Criteria 273 15.8.4.4 Results 273 15.8.5 Radiological Evaluation 274 15.8.6 Conclusions 274 | |||
==15.8 REFERENCES== | |||
FOR SECTION 15.8 275 FIGURES Figure 15.0-1 Core Limits and Overpower-Overtemperature Delta T Setpoints (Tref = | |||
576.0F) | |||
Figure 15.0-2 Reactivity Coefficients Used in Non-LOCA Safety Analysis Figure 15.0-3 Reactivity Insertion Scram Curves Figure 15.1-1 Feedwater Flow Increase at Full Power, Nuclear Power and Loop Average Temperature Versus Time Figure 15.1-2 Feedwater Flow Increase at Full Power, Pressurizer Pressure and Steam Gen-erator Pressure Versus Time Page 18 of 24 Revision 26 5/2016 | |||
Figure 15.1-3 Feedwater Flow Increase at Full Power, Steam Generator Mass Versus Time Figure 15.1-4 Steam Line Rupture, Multiplication Factor Versus Core Average Temperature (Calculated at 1050 psia) | |||
Figure 15.1-5 Steam Line Rupture, Integrated Doppler Defect Versus Fraction of Power Figure 15.1-6 Steam Line Rupture, 1.4ft2 Break with Power, Two Loops in Service, Core Heat Flux and Nuclear Power Versus Time Figure 15.1-7 Steam Line Rupture, 1.4ft2 Break with Power, Two Loops in Service, Pressur-izer Water Volume and Pressurizer Pressure Versus Time Figure 15.1-8 Steam Line Rupture, 1.4ft2 Break with Power, Two Loops in Service, Loop TAVG and Cold Leg Loop Temperature Versus Time Figure 15.1-9 Steam Line Rupture, 1.4ft2 Break with Power, Two Loops in Service, Faulted Loop Steam Flow and Total Feedwater Flow Versus Time Figure 15.1-10 Steam Line Rupture, 1.4ft2 Break with Power, Two Loops in Service, Core Averaged Boron and Reactivity Versus Time Figure 15.1-11 Steam Line Rupture, 1.4ft2 Break Without Power, Two Loops in Service, Core Heat Flux and Nuclear Power Versus Time Figure 15.1-12 Steam Line Rupture, 1.4ft2 Break Without Power, Two Loops in Service, Pressurizer Water Volume and Pressurizer Pressure Versus Time Figure 15.1-13 Steam Line Rupture, 1.4ft2 Break Without Power, Two Loops in Service, Loop TAVG and Cold Leg Loop Temperatures Versus Time Figure 15.1-14 Steam Line Rupture, 1.4ft2 Break Without Power, Two Loops in Service, Faulted Loop Steam Flow and Total Feedwater Flow Versus Time Figure 15.1-15 Steam Line Rupture, 1.4ft2 Break without Power, Two Loops in Service, Core Averaged Boron and Reactivity Versus Time Figure 15.1-16 Steam Line Rupture, 1.4ft2 Break with Power, One Loop in Service, Core Heat Flux and Nuclear Power Versus Time Figure 15.1-17 Steam Line Rupture, 1.4ft2 Break with Power, One Loop in Service, Pressur-izer Water Volume and Pressurizer Pressure Versus Time Figure 15.1-18 Steam Line Rupture, 1.4ft2 Break with Power, One Loop in Service, Loop TAVG and Cold Leg Loop Temperatures Versus Time Figure 15.1-19 Steam Line Rupture, 1.4ft2 Break with Power, One Loop in Service, Faulted Loop Steam Flow and Total Feedwater Flow Versus Time Figure 15.1-20 Steam Line Rupture, 1.4ft2 Break with Power, One Loop in Service, Core Averaged Boron and Reactivity Versus Time Figure 15.1-21 Combined Atmospheric Relief Valve and Main Feedwater Regulating Valve Failure, Nuclear Power and Core Heat Flux Versus Time Figure 15.1-22 Combined Atmospheric Relief Valve and Main Feedwater Regulating Valve Failure, Loop Average Temperature and Pressurizer Pressure Versus Time Figure 15.1-23 Combined Atmospheric Relief Valve and Main Feedwater Regulating Valve Failure, DNBR Versus Time Figure 15.1-24 Combined Atmospheric Relief Valve and Main Feedwater Regulating Valve Failure, Steam Generator Level and Steam Generator Mass Versus Time Page 19 of 24 Revision 26 5/2016 | |||
Figure 15.2-1 Loss of Load, with Automatic Pressure Control, Nuclear Power and DNBR Versus Time Figure 15.2-2 Loss of Load, with Automatic Pressure Control, RCS Average Temperature and Pressurizer Water Volume Versus Time Figure 15.2-3 Loss of Load, with Automatic Pressure Control, Steam Generator Pressure and Pressurizer Pressure Versus Time Figure 15.2-4 Loss of Load, Without Pressure Control, Nuclear Power Versus Time Figure 15.2-5 Loss of Load, Without Pressure Control, RCS Average Temperature and Pres-surizer Water Volume Versus Time Figure 15.2-6 Loss of Load, Without Pressure Control, Steam Generator Pressure and Reac-tor Coolant System Pressures Versus Time Figure 15.2-7 Loss of Load, Peak MSS Pressure Case, Nuclear Power Versus Time Figure 15.2-8 Loss of Load, Peak MSS Pressure Case, RCS Average Temperature and Pres-surizer Water Volume Versus Time Figure 15.2-9 Loss of Load, Peak MSS Pressure Case, Steam Generator Pressure and Pres-surizer Pressure Versus Time Figure 15.2-10 Figure Deleted Figure 15.2-11 Figure Deleted Figure 15.2-12 Figure Deleted Figure 15.2-13 Loss of Offsite Alternating Current Power to the Station Auxiliaries, Nuclear Power and Pressurizer Pressure Versus Time Figure 15.2-14 Loss of Offsite Alternating Current Power to the Station Auxiliaries, Pressur-izer Water Volume and Pressurizer Steam Relief Rate Versus Time Figure 15.2-15 Loss of Offsite Alternating Current Power to the Station Auxiliaries, Reactor Coolant Flow and Core Inlet/Outlet Temperatures Versus Time Figure 15.2-16 Loss of Offsite Alternating Current Power to the Station Auxiliaries, Steam Generator Mass and Steam Generator Pressure Versus Time Figure 15.2-17 Loss of Normal Feedwater With Power, Nuclear Power and Pressurizer Pres-sure Versus Time Figure 15.2-18 Loss of Normal Feedwater With Power, Pressurizer Water Volume and Pres-surizer Steam Relief Rate Versus Time Figure 15.2-19 Loss of Normal Feedwater With Power, Reactor Coolant Flow and Core Inlet/ | |||
Outlet Temperatures Versus Time Figure 15.2-20 Loss of Normal Feedwater With Power, Steam Generator Mass and Steam Generator Pressure Versus Time Figure 15.2-21 Feedline Break With Offsite Power; Nuclear Power and Pressurizer Pressure Versus Time Figure 15.2-22 Feedline Break With Offsite Power; Pressurizer Water Volume and Pressur-izer Steam Relief Rate Versus Time Figure 15.2-23 Feedline Break With Offsite Power; Cold Leg, Hot Leg and Saturation Tem-peratures Versus Time Figure 15.2-24 Feedline Break With Offsite Power; Steam Generator Mass and Steam Gener-ator Pressure Versus Time Figure 15.2-25 Feedline Break With Offsite Power; Feedwater Mass Flow Rates Versus Time Page 20 of 24 Revision 26 5/2016 | |||
Figure 15.2-26 Feedline Break Without Offsite Power; Nuclear Power and Pressurizer Pres-sure Versus Time Figure 15.2-27 Feedline Break Without Offsite Power; Pressurizer Water Volume and Pres-surizer Steam Relief Rate Versus Time Figure 15.2-28 Feedline Break Without Offsite Power; Cold Leg, Hot Leg and Saturation Temperatures Versus Time Figure 15.2-29 Feedline Break Without Offsite Power; Steam Generator Mass and Steam Generator Pressure Versus Time Figure 15.2-30 Feedline Break Without Offsite Power; Feedwater Mass Flow Rates Versus Time Figure 15.3-1 Full Loss of Flow (Undervoltage), Nuclear Power and RCS Flow Versus Time Figure 15.3-1a Full Loss of Flow (Underfrequency), Nuclear Power and RCS Flow Versus Time Figure 15.3-2 Full Loss of Flow (Undervoltage), Core Average and Hot Channel Heat Flux Versus Time Figure 15.3-2a Full Loss of Flow (Underfrequency), Core Average and Hot Channel Heat Flux Versus Time Figure 15.3-3 Full Loss of Flow (Undervoltage), RCS Pressures and DNBR Versus Time Figure 15.3-3a Full Loss of Flow (Underfrequency), DNBR and Reactor Coolant System Pressures Versus Time Figure 15.3-4 Partial Loss of Flow, Nuclear Power and RCS Flow Versus Time Figure 15.3-5 Partial Loss of Flow, RCS Pressures and RCS Loop Flows Versus Time Figure 15.3-6 Partial Loss of Flow, Core Average and Hot Channel Heat Flux Versus Time Figure 15.3-7 Partial Loss of Flow, DNBR Versus Time Figure 15.3-8 Locked Rotor, RCS Pressures and RCS Loop Flows Versus Time Figure 15.3-9 Locked Rotor, Nuclear Power and RCS Flow Versus Time Figure 15.3-10 Locked Rotor, Core Average Heat Flux and Cladding Inside Temperature Ver-sus Time Figure 15.4-1 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal From Subcrit-ical Conditions, Heat Flux and Nuclear Power Versus Time (422V+Fuel) | |||
Figure 15.4-2 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal From Subcrit-ical Conditions, Clad Inside and Fuel Average Temperature Versus Time(422V+Fuel) | |||
Figure 15.4-3 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Minimum Feedback, 100 pcm/sec, Nuclear Power and Heat Flux Versus Time Figure 15.4-4 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Minimum Feedback, 100 pcm/sec, Pressurizer Pressure and Pressurizer Water Volume Versus Time Figure 15.4-5 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Minimum Feedback, 100 pcm/sec, Tavg and DNBR Versus Time Figure 15.4-6 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Maximum Feedback, 5 pcm/sec, Nuclear Power and Heat Flux Versus Time Page 21 of 24 Revision 26 5/2016 | |||
Figure 15.4-7 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Maximum Feedback, 5 pcm/sec, Pressurizer Water Volume and Pressurizer Pressure Versus Time Figure 15.4-8 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Maximum Feedback, 5 pcm/sec, TAVG and DNBR Versus Time Figure 15.4-9 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Minimum and Maximum Feedback, DNBR Versus Reactivity Insertion Rate Figure 15.4-10 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from 60% | |||
Power, Minimum and Maximum Feedback, DNBR Versus Reactivity Inser-tion Rate Figure 15.4-11 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from 10% | |||
Power, Minimum and Maximum Feedback, DNBR Versus Reactivity Inser-tion Rate Figure 15.4-12 Startup of an Inactive Coolant Loop, Nuclear Power Versus Time Figure 15.4-13 Startup of an Inactive Coolant Loop, TAVG Versus Time Figure 15.4-14 Startup of an Inactive Coolant Loop, Core Inlet Temperature Versus Time Figure 15.4-15 Startup of an Inactive Coolant Loop, Pressurizer Pressure Versus Time Figure 15.4-16 Rod Cluster Control Assembly Ejection Beginning-of-Life, Full Power, Fuel and Clad Temperature and Nuclear Power Versus Time Figure 15.4-16a Rod Cluster Control Assembly Ejection, Beginning of Life, Full Power, Fuel and Clad Temperature and Nuclear Power Versus Time Figure 15.4-16b Rod Cluster Control Assembly Ejection, Beginning of Life, Zero Power, Fuel and Clad Temperature and Nuclear Power Versus Time Figure 15.4-17 Rod Cluster Control Assembly Ejection Beginning-of-Life, Zero Power, Fuel and Clad Temperature and Nuclear Power Versus Time Figure 15.4-17a Rod Cluster Control Assembly Ejection, End of Life, Full Power, Fuel and Clad Temperature and Nuclear Power Versus Time Figure 15.4-17b Rod Cluster Control Assembly Ejection, End of Life, Zero Power, Fuel and Clad Temperature and Nuclear Power Versus Time Figure 15.4-18 Rod Cluster Control Assembly Drop Heat Flux and Nuclear Power Versus Time Figure 15.4-19 Rod Cluster Control Assembly Drop Pressurizer Pressure and Core Average Temperature Versus Time Figure 15.4-20 Uncontrolled Rod Cluster Control Assembly Bank Withdrawl from 8% Power (RCS Pressure Case), Minimum Feedback, 55 pcm/sec, Nuclear Power and Heat Flux Versus Time Figure 15.4-21 Uncontrolled Rod Cluster Control Assembly Bank Withdrawl from 8% Power (RCS Pressure Case), Minimum Feedback, 55 pcm/sec, Pressurizer Pressure and Tavg Versus Time Figure 15.6-1 Steam Generator Tube Rupture (Overfill), Maximum Safety Injection Flow Versus Pressure Figure 15.6-1a RCS Depressurization, Nuclear Power Versus Time Figure 15.6-1b RCS Pressurization, Pressurizer Pressure Versus Time Figure 15.6-1c RCS Depressurization, Indicated Loop Average Temperature Versus Time Figure 15.6-1d RCS Depressurization, DNBR Versus Time Page 22 of 24 Revision 26 5/2016 | |||
Figure 15.6-2 SGTR (Overfill), Pressurizer Level and Pressurizer Pressure Versus Time Figure 15.6-3 SGTR (Overfill), Secondary Pressure and Steam Generator Liquid Mass Ver-sus Time Figure 15.6-4 SGTR (Overfill), Hot and Cold Leg Temperatures for Intact and Ruptured Steam Generators Versus Time Figure 15.6-5 SGTR (Overfill), Total Primary to Secondary Leakage and Total Integrated Primary to Secondary Leakage Versus Time Figure 15.6-6 SGTR (Overfill), Steam Generator Relief Flow and Integrated Steam Genera-tor Relief Flow Versus Time Figure 15.6-7 SGTR (Overfill), Steam Generator Water Volume Versus Time Figure 15.6-8 SGTR (Dose), Pressurizer Level and Pressurizer Pressure Versus Time Figure 15.6-9 SGTR (Dose), Secondary Pressure and Steam Generator Liquid Mass Versus Time Figure 15.6-10 SGTR (Dose), Hot and Cold Leg Temperatures for Intact and Ruptured Steam Generators Versus Time Figure 15.6-11 SGTR (Dose), Total Primary to Secondary Leakage and Total Integrated Pri-mary to Secondary Leakage Versus Time Figure 15.6-12 SGTR (Dose), Steam Generator Relief Flow and Integrated Steam Generator Relief Flow Versus Time Figure 15.6-13 SGTR (Dose), Steam Generator Water Volume Versus Time Figure 15.6-14 SGTR (Dose), Tube Rupture Flow Flashing Fraction and Integrated Flashed Break Versus Time Figure 15.6-15 Small Break LOCA Inch Break, Pressurizer Pressure Versus Time Figure 15.6-16 Small Break LOCA Inch Break, Core Mixture Level Versus Time Figure 15.6-17 Small Break LOCA Inch High Break, Peak Cladding Temperature at PCT Elevation Versus Time Figure 15.6-18 Small Break LOCA Inch High Break, Core Exit Vapor Flow Versus Time Figure 15.6-19 Small Break LOCA Inch Break, Hot Rod Heat Transfer Coefficient at PCT Elevation Versus Time Figure 15.6-20 Small Break LOCA Inch Break, Fluid Temperature at PCT Elevation Ver-sus Time Figure 15.6-21 Small Break LOCA - Axial Power Distribution, Heat Rate Versus Core Eleva-tion Figure 15.6-22 Small Break LOCA - 1.5-Inch Break, Pressurizer Pressure Versus Time Figure 15.6-23 Small Break LOCA Inch High Break, Pressurizer Pressure Versus Time Figure 15.6-24 Small Break LOCA - 1.5-Inch Break, Core Mixture Level Versus Time Figure 15.6-25 Small Break LOCA Inch Break, Core Mixture Level Versus Time Figure 15.6-26 Small Break LOCA - 1.5-Inch Break, Peal Cladding Temperature at PCT Ele-vation Versus Time Figure 15.6-27 Small Break LOCA Inch Break, Peak Cladding Temperature at PCT Ele-vation Versus Time Figure 15.6-28 Figure Deleted Figure 15.6-29 Figure Deleted Figure 15.6-30 Figure Deleted Page 23 of 24 Revision 26 5/2016 | |||
Figure 15.6-31 R.E. Ginna Vessel Model Noding Diagram1 Figure 15.6-32 R.E. Ginna Loop Model Noding Diagram Figure 15.6-33 R.E. Ginna Initial Transient Axial Power Distributions Figure 15.6-34 Containment Pressure Used for the R.E. Ginna Best-Estimate Large Break LOCA Initial Transient Figure 15.6-35 Peak Clad Temperature of the 5 rods for the Initial Transient Figure 15.6-36 Split Break Flow for the Initial Transient Figure 15.6-37 Total Flow at the Bottom of the Core for the Initial Transient Figure 15.6-38 Accumulator Injection Flow for the Initial Transient Figure 15.6-39 High Head Safety Injection Flow for the Initial Transient Figure 15.6-40 Low Head Safety Injection Flow for the Initial Transient Figure 15.6-41 Average Collapsed Liquid Level in the Downcomer for the Initial Transient Figure 15.6-42 Lower Plenum Collapsed Liquid Level for the Initial Transient Figure 15.6-43 Core Collapsed Liquid Levels for the Initial Transient Figure 15.6-44 Vessel Liquid Mass for the Initial Transient Figure 15.6-45 Pressurizer Pressure for the Initial Transient Figure 15.6-46 Hot Rod Peak Clad Temperature and Elevation for the Initial Transient Figure 15.6-47 R.E. Ginna PBOT/PMID Analysis and Operating Limits Figure 15.6-48 Lower Bound Containment Pressure for R.E. Ginna Analysis Page 24 of 24 Revision 26 5/2016 | |||
GINNA/UFSAR 17 QUALITY ASSURANCE 1 17.1 QUALITY ASSURANCE DURING DESIGN AND CONSTRUC- 2 TION 17.1.1 ORGANIZATION 2 17.1.2 QUALITY ASSURANCE PROGRAM 2 17.1.2.1 General 2 17.1.2.2 Rochester Gas and Electric Corporation 2 17.1.2.3 Westinghouse 3 17.1.2.3.1 General 3 17.1.2.3.2 Westinghouse Organization 4 17.1.2.3.3 Components Supplied By Westinghouse 4 17.1.2.3.4 Supplier Evaluation 5 17.1.2.3.5 Equipment Specifications 5 17.1.2.3.6 Purchase Order Review 6 17.1.2.3.7 Supplier Surveillance 6 17.1.2.3.8 Instrumentation and Control Equipment 7 17.1.2.3.9 Shipment of Components 8 17.1.2.3.10 Inspection and Installation of Equipment in the Field 9 17.1.2.3.11 Nonconforming Components or Material 10 17.1.2.3.12 Quality Control Records 10 17.1.2.4 Gilbert Associates, Inc 10 17.1.2.5 Bechtel Corporation 11 17.1.2.5.1 General 11 17.1.2.5.2 Material Certification 11 17.1.2.5.3 Backfill 12 17.1.2.5.4 Containment Liner 12 17.1.2.5.5 Piping 12 17.1.2.5.6 Stainless Steel Liners 12 17.1.2.5.7 Containment Tendons 12 17.1.2.5.8 Reinforcing Bar and Cadwelds 13 17.1.2.5.9 Machinery Setting and Alignment 13 Page 1 of 2 Revision 26 5 /2016 | |||
GINNA/UFSAR 17 QUALITY ASSURANCE 1 17.1.2.5.10 Concrete 13 17.1.2.5.11 Documentation 14 17.1.2.5.12 Site Cleanliness 14 17.1.3 QUALITY CONTROL ORGANIZATION FOR REACTOR CON- 14 TAINMENT STRUCTURE DESIGN AND ERECTION 17.1.3.1 General 14 17.1.3.2 Rochester Gas and Electric Corporation 14 17.1.3.3 Westinghouse Electric Corporation 15 17.1.3.4 Gilbert Associates, Inc 15 17.1.3.5 Pittsburgh Testing Laboratory 16 17.1.3.6 Dames & Moore 16 17.1.3.7 Bechtel Corporation 16 17.2 QUALITY ASSURANCE DURING THE OPERATIONS PHASE 18 FIGURES Figure 17.1-1 Quality Control Relationships During Construction Page 2 of 2 Revision 26 5 /2016 | |||
GINNA/UFSAR 18 UPDATED FINAL SAFETY ANALYSIS REPORT 1 SUPPLEMENT FOR LICENSE RENEWAL | |||
==18.1 INTRODUCTION== | |||
2 18.1.1 RENEWED FACILITY OPERATING LICENSE 2 18.1.2 UPDATED FINAL SAFETY ANALYSIS REPORT SUPPLEMENT 2 18.1.3 STRUCTURES, SYSTEMS, OR COMPONENTS ADDED SUBSE- 2 QUENT TO RENEWED LICENSE 18.2 Programs that Manage the Effects of Aging 5 18.2.1 Aging Management Programs 5 18.2.1.1 Aboveground Carbon Steel Tanks 5 18.2.1.2 ASME Section XI, Subsections IWB, IWC, & IWD Inservice Inspec- 5 tion 18.2.1.3 ASME Section XI, Subsections IWE & IWL Inservice Inspection 5 18.2.1.4 ASME Section XI, Subsection IWF Inservice Inspection 6 18.2.1.5 Bolting Integrity 6 18.2.1.6 Boric Acid Corrosion 6 18.2.1.7 Buried Piping and Tanks Inspection 6 18.2.1.8 Closed-Cycle (Component) Cooling Water System 6 18.2.1.9 Electrical Cables and Connections Not Subject to 10 CFR 50.49 Envi- 7 ronmental Qualification Requirements 18.2.1.10 Electrical Cables Not Subject to 10CFR50.49 Environmental Qualifi- 7 cation Requirements Used In Instrument Circuits 18.2.1.11 Fire Protection 7 18.2.1.12 Fire Water System 7 18.2.1.13 Flow-Accelerated Corrosion 7 18.2.1.14 Fuel Oil Chemistry 8 18.2.1.15 Inaccessible Medium-Voltage Cables Not Subject to 10CFR50.49 8 Environmental Qualification Requirements 18.2.1.16 Inspection of Overhead Heavy Load and Light Load (Related to Refu- 8 eling) Handling Systems 18.2.1.17 One-Time Inspection 8 18.2.1.18 Open-Cycle Cooling (Service) Water System 8 18.2.1.19 Periodic Surveillance and Preventive Maintenance 9 18.2.1.20 Nickel-Alloy Nozzles and Penetrations Inspection 9 Page 1 of 3 Revision 26 5 /2016 | |||
GINNA/UFSAR 18 UPDATED FINAL SAFETY ANALYSIS REPORT 1 SUPPLEMENT FOR LICENSE RENEWAL 18.2.1.21 Reactor Vessel Internals 9 18.2.1.22 Reactor Vessel Surveillance 9 18.2.1.23 Spent Fuel Pool Neutron Absorber Monitoring 10 18.2.1.24 Steam Generator Integrity 10 18.2.1.25 Structures Monitoring 10 18.2.1.26 Systems Monitoring 10 18.2.1.27 Thimble Tubes Inspection 10 18.2.1.28 Water Chemistry Control 11 18.3 Evaluation of Time-Limited Aging Analyses 12 18.3.1 Reactor Vessel Neutron Embrittlement 12 18.3.1.1 Upper Shelf Energy 12 18.3.1.2 Pressurized Thermal Shock 12 18.3.1.3 Pressure-Temperature Limits 13 18.3.2 Thermal Aging Embrittlement of Cast Austenitic Stainless Steel 13 (CASS) 18.3.3 Metal Fatigue 14 18.3.3.1 ASME Boiler and Pressure Vessel Code, Section III, Class 1 14 18.3.3.2 Reactor Vessel Underclad Cracking 15 18.3.3.3 ANSI B31.1 Piping 15 18.3.3.4 Accumulator Check Valves 16 18.3.3.5 Environmentally Assisted Fatigue 16 18.3.3.6 Reactor Vessel Nozzle-to-Weld Defect 17 18.3.3.7 Pressurizer Fracture Mechanics Analysis 17 18.3.4 Environmental Qualification of Electric Equipment 18 18.3.5 Concrete Containment Tendon Prestress 19 18.3.5.1 Containment Tendon Fatigue 19 18.3.5.2 Containment Tendon Bellows Fatigue 19 18.3.6 Containment Liner Plate and Penetration Fatigue 20 18.3.6.1 Containment Liner Anchorage Fatigue 20 18.3.7 Containment Liner Stress 20 18.3.8 Other Time-Limited Aging Analyses 21 Page 2 of 3 Revision 26 5 /2016 | |||
GINNA/UFSAR 18 UPDATED FINAL SAFETY ANALYSIS REPORT 1 SUPPLEMENT FOR LICENSE RENEWAL 18.3.8.1 Crane Load Cycle Limit 21 18.3.8.2 Reactor Coolant Pump (RCP) Flywheel 21 18.3.9 Exemptions 21 18.4 TLAA Supporting Activities 22 18.4.1 Concrete Containment Tendon Prestress 22 18.4.2 Environmental Qualification Program 22 18.4.3 Fatigue Monitoring Program 22 Page 3 of 3 Revision 26 5 /2016}} | |||
Latest revision as of 18:16, 10 November 2019
Text
GINNA/UFSAR 1 INTRODUCTION AND GENERAL DESCRIPTION OF THE 1 PLANT
1.1 INTRODUCTION
2 REFERENCES FOR SECTION 1.1 4 1.2 GENERAL PLANT DESCRIPTION 5 1.2.1 SITE AND ENVIRONMENT 5 1.2.2
SUMMARY
PLANT DESCRIPTION 5 1.2.3 STRUCTURES 6 1.2.3.1 General 6 1.2.3.2 Containment 6 1.2.3.3 Auxiliary Building 7 1.2.3.4 Intermediate Building (See Drawings 33013-2101, 33013-2102, 9 33013-2105, 33013-2106, 33013-2107, 33013-2113, 33013-2114, 33013-2115, and 33013-2121) 1.2.3.5 Turbine Building 10 1.2.3.6 Control Building 11 1.2.3.7 All-Volatile-Treatment Building 12 1.2.3.8 Standby Auxiliary Feedwater Pump Building 12 1.2.3.9 Screen House 12 1.2.3.10 Service Building 12 1.2.3.11 Diesel Generator Building 13 1.2.3.12 Old Steam Generator Storage Facility 13 1.2.3.13 Canister Preparation Building 13 1.2.3.14 ISFSI Transfer Path and Storage Pad 14 1.2.3.15 Administration Building 14 1.2.4 NUCLEAR STEAM SUPPLY SYSTEM 14 1.2.5 REACTOR AND PLANT CONTROL 15 1.2.6 WASTE DISPOSAL SYSTEM 15 1.2.7 FUEL HANDLING SYSTEM 15 1.2.8 TURBINE AND AUXILIARIES 16 1.2.9 ELECTRICAL SYSTEM 16 1.2.10 ENGINEERED SAFETY FEATURES PROTECTION SYSTEMS 16 1.2.11 DESIGN HIGHLIGHTS 17 1.2.11.1 Power Level 17 Page 1 of 9 Revision 26 5 /2016
GINNA/UFSAR 1.2.11.2 Reactor Coolant Loops 17 1.2.11.3 Peak Specific Power 17 1.2.11.4 Fuel Clad 17 1.2.11.5 Fuel Assembly Design 17 1.2.11.6 Engineered Safety Features 18 1.2.11.7 Emergency Power 18 1.2.12 STATION WATER USE 18 1.2.13 FACILITY SAFETY CONCLUSIONS 19 1.3 COMPARISON TABLES 20 1.3.1 COMPARISONS WITH SIMILAR FACILITY DESIGNS 20 1.3.2 COMPARISON OF FINAL AND PRELIMINARY SAFETY 20 ANALYSIS REPORT INFORMATION (Historical) 1.3.2.1 Partial Length Rod Cluster Control Assemblies 20 1.3.2.2 Burnable Shim Rods 20 1.3.2.3 Safety Injection System Trip Signal 20 1.3.2.4 Containment Spray System Signal 20 1.3.2.5 Safety Injection System Accumulators 21 1.3.2.6 Spray Additive 21 1.3.2.7 Rod Stop and Reactor Trip on Startup 21 1.3.2.8 Miniature Neutron Flux Detectors 21 1.3.2.9 Core Thermocouples 21 1.3.2.10 Initial Leak Rate Test Method 21 1.3.2.11 Auxiliary Building Ventilation Filters 21 1.3.2.12 Control Center Buses 21 1.3.2.13 Condenser Circulating Water Flow 21 1.3.2.14 Ramp Loading Range 21 1.3.2.15 Condensate Storage Tanks Capacity 22 1.3.2.16 Fuel Transfer System Drive 22 1.3.2.17 Steam Line Flow Nozzles 22 1.3.3 Comparison of Uprate Parameters 22 Table 1.3-1 COMPARISON OF DESIGN PARAMETERS WITH POINT 23 BEACH Page 2 of 9 Revision 26 5 /2016
GINNA/UFSAR Table 1.3-2 COMPARISON OF DESIGN PARAMETERS WITH SAN 33 ONOFRE AND CONNECTICUT YANKEE Table 1.3-3 COMPARISON OF GINNA AND KEWAUNEE UPRATE NSSS 35 DESIGN PARAMETERS 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 36 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 38 1.
5.1 INTRODUCTION
38 1.5.2 DEVELOPMENT OF THE FINAL CORE DESIGN AND FINAL 38 THERMAL-HYDRAULIC AND PHYSICS PARAMETERS 1.5.3 CORE STABILITY 39 1.5.3.1 Core Power Distribution 39 1.5.3.2 Out-of-Core Ion Chambers 39 1.5.3.3 In-Core Control Equipment 40 1.5.3.4 Startup Test Program 40 1.5.4 DEVELOPMENT OF LONG ION CHAMBERS 41 1.5.5 CONTROL ROD EJECTION AND DROPPED CONTROL ROD 42 ACCIDENT ANALYSES 1.5.6 CHARCOAL FILTERS 43 1.5.7 REACTOR COOLANT PUMP CONTROLLED LEAKAGE 43 SEALS 1.5.8 SAFETY INJECTION SYSTEM 44 1.5.8.l Development of Safety Injection System Design 44 1.5.8.2 Development of Core Cooling Analysis 44 1.5.9 DEVELOPMENT OF DESIGN, INSPECTION, AND ACCEP- 44 TANCE CRITERIA FOR PRESTRESSED REINFORCED-CON-CRETE PRESSURE VESSELS 1.5.9.1 Rock Anchors 45 1.5.9.1.1 Design Criteria and Assumptions 45 1.5.9.1.2 Test Verification and Results 45 1.5.9.2 Rock Anchor Grout 45 1.5.9.3 Tendon Inspection and Acceptance Criteria 46 1.5.9.4 Wall Tendons 46 1.5.9.4.1 Corrosion Protection 46 1.5.9.4.2 Inspection and Acceptance 47 Page 3 of 9 Revision 26 5 /2016
GINNA/UFSAR 1.5.10 DEVELOPMENT OF CONTAINMENT HYDROGEN RECOM- 47 BINER REFERENCES FOR SECTION 1.5 48 1.6 MATERIAL INCORPORATED BY REFERENCE 49 1.7 DRAWINGS AND OTHER DETAILED INFORMATION 59 1.7.1 ELECTRICAL, INSTRUMENTATION, AND CONTROL DRAW- 59 INGS 1.7.2 PIPING AND INSTRUMENTATION DIAGRAMS (P&ID) 59 1.7.3 OTHER DETAILED INFORMATION 59 Table 1.7-1 ELECTRICAL, INSTRUMENTATION, AND CONTROL DRAW- 60 INGS Table 1.7-2 PIPING AND INSTRUMENTATION DIAGRAMS (P &ID 61 1.8 CONFORMANCE TO NRC REGULATORY GUIDES 68 1.8.1 CONFORMANCE TO AEC SAFETY GUIDES 68 1.8.1.1 Safety Guide 1 -Net Positive Suction Head for Emergency Core 68 Cooling and Containment Heat Removal System Pumps 1.8.1.2 Safety Guide 2 -Thermal Shock to Reactor Pressure Vessels 68 1.8.1.3 Safety Guide 3 -Assumptions Used for Evaluating the Potential 69 Radiological Consequences of a Loss-of-Coolant Accident for Boil-ing Water Reactors 1.8.1.4 Safety Guide 4 -Assumptions Used for Evaluating the Potential 69 Radiological Consequences of a Loss-of-Coolant Accident for Pres-surized Water Reactors 1.8.1.5 Safety Guide 5 - Assumptions Used for Evaluating the Potential 69 Radiological Consequences of a Steam Line Break Accident for Boil-ing Water Reactors 1.8.1.6 Safety Guide 6 -Independence Between Redundant Standby (Onsite) 69 Power Sources and Between Their Distribution Systems 1.8.1.7 Safety Guide 7 -Control of Combustible Gas Concentrations in Con- 70 tainment Following a Loss-of-Coolant Accident 1.8.1.8 Safety Guide 8 -Personnel Selection and Training 70 1.8.1.9 Safety Guide 9 -Selection of Diesel-Generator Set Capacity for 70 Standby Power Supplies 1.8.1.10 Safety Guide 10 -Mechanical (Cadweld) Splices in Reinforcing Bars 71 of Concrete Containments 1.8.1.11 Safety Guide 11 -Instrument Lines Penetrating Primary Reactor Con- 72 tainment 1.8.1.12 Safety Guide 12 -Instrumentation for Earthquakes 72 Page 4 of 9 Revision 26 5 /2016
GINNA/UFSAR 1.8.1.13 Safety Guide 13 -Fuel Storage Facility Design Basis 72 1.8.1.14 Safety Guide 14 -Reactor Coolant Pump Flywheel Integrity 73 1.8.1.15 Safety Guide 15 -Testing of Reinforcing Bars for Concrete Structures 74 1.8.1.16 Safety Guide 16 -Reporting of Operating Information 75 1.8.1.17 Safety Guide 17 -Protection Against Industrial Sabotage 75 1.8.1.18 Safety Guide 18 - Structural Acceptance Test for Concrete Primary 76 Reactor Containments 1.8.1.18.1 Structural Integrity Test 76 1.8.1.18.2 Instrumentation 76 1.8.1.18.3 Displacement Measurements 77 1.8.1.18.4 Strain Measurements 78 1.8.1.18.5 Test Results 78 1.8.1.19 Safety Guide 19 -Nondestructive Examination of Primary Contain- 79 ment Liners 1.8.1.19.1 Test Provisions 79 1.8.1.19.2 Examination of Welds 79 1.8.1.19.3 Pressure Tests 80 1.8.1.19.4 Quality Control Provisions 80 1.8.1.20 Safety Guide 20 -Vibration Measurements on Reactor Internals 81 1.8.1.21 Safety Guide 21 -Measuring and Reporting Effluents From Nuclear 81 Power Plants 1.8.1.22 Safety Guide 22 -Periodic Testing of Protection System Actuation 82 Functions 1.8.1.23 Safety Guide 23 -Onsite Meteorological Programs 82 1.8.1.24 Safety Guide 24 -Assumptions Used for Evaluating the Potential 83 Radiological Consequences of a Pressurized Water Reactor Radioac-tive Gas Storage Tank Failure 1.8.1.25 Safety Guide 25 - Assumptions Used for Evaluating the Potential 83 Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors 1.8.1.26 Safety Guide 26 -Quality Group Classification and Standards 83 1.8.1.27 Safety Guide 27 -Ultimate Heat Sink 83 1.8.1.28 Safety Guide 28 -Quality Assurance Program Requirements 84 1.8.1.29 Safety Guide 29 -Seismic Design Classification 84 1.8.2 CONFORMANCE TO DIVISION I REGULATORY GUIDES 85 Page 5 of 9 Revision 26 5 /2016
GINNA/UFSAR 1.8.2.1 Regulatory Guide 1.4 -Assumptions Used for Evaluating the Poten- 85 tial Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors 1.8.2.2 Regulatory Guide 1.10 -Mechanical (Cadweld) Splices in Reinforc- 85 ing Bars of Category I Concrete Structures 1.8.2.3 Regulatory Guide 1.15 -Testing of Reinforcing Bars for Category I 85 Concrete Structures 1.8.2.4 Regulatory Guide 1.16 -Reporting of Operating Information 85 1.8.2.5 Regulatory Guide 1.17 -Protection of Nuclear Plants Against Indus- 85 trial Sabotage 1.8.2.6 Regulatory Guide 1.18 -Structural Acceptance Test for Concrete Pri- 85 mary Reactor Containments 1.8.2.7 Regulatory Guide 1.19 -Nondestructive Examination of Primary 85 Containment Liner Welds 1.8.2.8 Regulatory Guide 1.26, Revision 3 - Quality Group Classifications & 86 Standards for Water, Steam, and Radioactive -Waste Containing Components of Nuclear Power Plants 1.8.2.9 Regulatory Guide 1.29, Revision 3 -Seismic Design Classification 86 1.8.2.10 Regulatory Guide 1.30 - Quality Assurance Requirements for the 86 Installation, Inspection, and Testing of Instrumentation and Electrical Equipment 1.8.2.11 Regulatory Guide 1.31 -Control of Stainless Steel Welding 87 1.8.2.12 Regulatory Guide 1.32 -Use of IEEE Standard 308-1971, Criteria for 88 Class IEE Electric Systems for Nuclear Power Generating Stations 1.8.2.13 Regulatory Guide 1.33 -Quality Assurance Program Requirements 89 (Operation) 1.8.2.14 Regulatory Guide 1.34 -Control of Electroslag Weld Properties 89 1.8.2.15 Regulatory Guide 1.35 -Inservice Surveillance of Ungrouted Ten- 89 dons in Prestressed Concrete Containment Structures 1.8.2.16 Regulatory Guide 1.36, Revision 0 -Nonmetallic Thermal Insulation 90 for Austenitic Stainless Steel 1.8.2.17 Regulatory Guide 1.37, Revision 0 -Quality Assurance for Cleaning 90 of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants 1.8.2.18 Regulatory Guide 1.38, Revision 2 - Quality Assurance Require- 91 ments for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants 1.8.2.19 Regulatory Guide 1.39 -Housekeeping Requirements for Water- 92 Cooled Nuclear Power Plants Page 6 of 9 Revision 26 5 /2016
GINNA/UFSAR 1.8.2.20 Regulatory Guide 1.40 -Qualification Tests of Continuous-Duty 92 Motors Installed Inside the Containment of Water-Cooled Nuclear Power Plants 1.8.2.21 Regulatory Guide 1.41 - Preoperational Testing of Redundant Onsite 92 Electric Power Systems to Verify Proper Load Group Assignments 1.8.2.22 Regulatory Guide 1.42 -Interim Licensing Policy on As Low As 93 Practicable for Gaseous Radioiodine Releases from Light-Water-Cooled Nuclear Power Reactors 1.8.2.23 Regulatory Guide 1.43 -Control of Stainless Steel Weld Cladding of 93 Low-Alloy Steel Components 1.8.2.24 Regulatory Guide 1.44 - Control of the Use of Sensitized Stainless 94 Steel 1.8.2.25 Regulatory Guide 1.45 -Reactor Coolant Pressure Boundary Leakage 95 Detection System 1.8.2.26 Regulatory Guide 1.46 -Protection Against Pipe Whip Inside Con- 96 tainment 1.8.2.27 Regulatory Guide 1.47 -Bypassed and Inoperable Status Indication 96 for Nuclear Power Plant Safety Systems 1.8.2.28 Regulatory Guide 1.48 -Design Limits and Loading Combinations 96 for Seismic Category I Fluid System Components 1.8.2.29 Regulatory Guide 1.49 -Power Levels of Water-Cooled Nuclear 97 Power Plants 1.8.2.30 Regulatory Guide 1.50 -Control of Preheat Temperature for Welding 97 of Low-Alloy Steel 1.8.2.31 Regulatory Guide 1.51 -Inservice Inspection of ASME Code Class 2 98 and 3 Nuclear Power Plant Components 1.8.2.32 Regulatory Guide 1.52 - Design, Testing, and Maintenance Criteria 98 for Atmospheric Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants 1.8.2.33 Regulatory Guide 1.53 -Application of the Single-Failure Criterion 99 To Nuclear Power Plant Protection Systems 1.8.2.34 Regulatory Guide 1.54, Revision 0 -Quality Assurance Requirements 99 for Protective Coatings Applied to Water-Cooled Nuclear Power Plants 1.8.2.35 Regulatory Guide 1.55 -Concrete Placement in Seismic Category I 100 Structures 1.8.2.36 Regulatory Guide 1.57 -Design Limits and Loading Combinations 100 for Metal Primary Reactor Containment System Components 1.8.2.37 Regulatory Guide 1.59 -Design-Basis Floods for Nuclear Power 100 Plants Page 7 of 9 Revision 26 5 /2016
GINNA/UFSAR 1.8.2.38 Regulatory Guide 1.94, Revision 1 - Quality Assurance Installation, 101 Inspections, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants 1.8.2.39 Regulatory Guide 1.143, Revision 1 - Design Guidance for Radioac- 101 tive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants 1.8.3 CONFORMANCE TO IEEE CRITERIA 101 1.8.3.1 Criteria for Protection Systems for Nuclear Power Generating Sta- 101 tions (IEEE 279-1971) 1.8.3.2 Class 1E Electric Systems for Nuclear Power Generating Stations 101 (IEEE 308-1971) 1.8.3.2.1 Principal Design Criteria 101 1.8.3.2.2 Alternating Current Power Systems 102 1.8.3.2.2.1 General 102 1.8.3.2.2.2 Distribution Systems 103 1.8.3.2.2.3 Preferred Power Supply 103 1.8.3.2.2.4 Standby Power Supply 103 1.8.3.2.3 Direct Current Power Systems 104 1.8.3.2.3.1 General 104 1.8.3.2.3.2 Distribution System 104 1.8.3.2.3.3 Battery Supply 104 1.8.3.2.3.4 Battery Charger Supply 104 1.8.3.2.3.5 Protective Devices 105 1.8.3.2.3.6 Performance Discharge Test Provisions 105 1.8.3.2.4 Vital Instrumentation and Control Power Systems 105 1.8.3.2.5 Surveillance Requirements 105 1.8.3.3 Electrical Penetration Assemblies in Containment Structures for 106 Nuclear Fueled Power Generating Stations (IEEE 317 - April 1971) 1.8.3.4 Qualifying Class I Electric Equipment for Nuclear Power Generating 106 Stations (IEEE 323-April 1971) 1.8.3.5 Type Tests of Continuous Duty Class I Motors Installed Inside the 107 Containment of Nuclear Power Generating Stations (IEEE 334-1971) 1.8.3.6 Installation, Inspection, and Testing Requirements for Instrumenta- 107 tion and Electric Equipment During the Construction of Nuclear Power Generating Stations (IEEE 336-1971) 1.8.3.7 Trial Use Criteria for the Periodic Testing of Nuclear Power Generat- 108 ing Station Protection Systems (IEEE 338-1971)
Page 8 of 9 Revision 26 5 /2016
GINNA/UFSAR 1.8.3.8 Seismic Qualification of Class I Electrical Equipment for Nuclear 108 Power Generating Stations (IEEE 344-1971)
REFERENCES FOR SECTION 1.8 109 FIGURES Figure 1.2-1 Ginna Station Plot Plan Page 9 of 9 Revision 26 5 /2016
GINNA/UFSAR 2 SITE CHARACTERISTICS 1 2.1 GEOGRAPHY AND DEMOGRAPHY 2 2.1.1 SITE LOCATION AND DESCRIPTION 2 2.1.2 EXCLUSION AREA AUTHORITY AND CONTROL 2 2.1.3 POPULATION DISTRIBUTION 3 REFERENCES FOR SECTION 2.1 6 2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES 7 2.2.1 LOCATIONS AND ROUTES 7 2.
2.2 DESCRIPTION
7 Railroads 7 Pipelines 7 Waterways 7 Airports 8 Military Facilities 8 Toxic Chemicals 9 Onsite Toxic Chemicals 9 Offsite Toxic Chemicals 10 REFERENCES FOR SECTION 2.2 11 Table 2.2-1 TYPICAL INDUSTRIES IN WAYNE COUNTY (CIRCA 1969) 12 Table 2.2-2 TYPICAL INDUSTRIES IN THE ROCHESTER AREA OF MON- ROE COUNTY (CIRCA 1969) (LOCATED 18 MILES WEST OF THE SITE) 15 2.3 METEOROLOGY 16 2.3.1 REGIONAL CLIMATOLOGY 16 2.3.2 LOCAL METEOROLOGY 16 2.3.3 ONSITE METEOROLOGICAL MEASUREMENTS PROGRAM 17 2.3.4 DIFFUSION ESTIMATES 19 Long-Term Diffusion Characteristics 19 Meteorological Data 19 Airflow Trajectory and Terrain Influences 19 Atmospheric Diffusion Model 20 Source Configuration Considerations 21 Unobstructed Release Point 21 Obstructed Release Point 22 Removal Mechanisms 23 Summary of Plant Discharges 23 Input Assumptions 23 Results 24 Page 1 of 10 Revision 26 5/2016
GINNA/UFSAR Accident Analysis Diffusion Characteristics 24 Nuclear Regulatory Commission Evaluation (Historical) 24 Rochester Gas and Electric Corporation Evaluation (Historical) 26 Current Approved Evaluation 26 Conclusions 27 REFERENCES FOR SECTION 2.3 28 Table 2.3-1 WIND VELOCITY
SUMMARY
GINNA SITE TOWER, 50 FT.
TOWER (FEBRUARY 1965 - JANUARY 1967, INCLUSIVE) 30 Table 2.3-2 WIND VELOCITY
SUMMARY
GINNA SITE TOWER, 150 FT.
TOWER (FEBRUARY 1965 - JANUARY 1967, INCLUSIVE) 31 Table 2.3-3 WIND VELOCITY
SUMMARY
GINNA SITE TOWER, 250 FT.
TOWER (FEBRUARY 1965 - JANUARY 1967, INCLUSIVE) 32 Table 2.3-4 WIND VELOCITY
SUMMARY
(HOURS) ROCHESTER AIRPORT FIVE YEARS 33 Table 2.3-5 WIND VELOCITY
SUMMARY
(HOURS) DURING PRECIPITA TION ROCHESTER AIRPORT 34 Table 2.3-6 WIND VELOCITY
SUMMARY
(HOURS) ROCHESTER COAST GUARD STATION (1951 - 1955) 35 Table 2.3-7
SUMMARY
OF METEOROLOGICAL DATA GINNA SITE 36 Table 2.3-8a JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 37 Table 2.3-8b JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 38 Table 2.3-8c JOINT FREQUENCY TABLES OF WIND SPEED AND DIREC TION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 39 Table 2.3-8d JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 40 Table 2.3-8e JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 41 Table 2.3-8f JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 42 Page 2 of 10 Revision 26 5/2016
GINNA/UFSAR Table 2.3-8g JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 43 Table 2.3-8h JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 44 Table 2.3-8i JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 45 Table 2.3-8j JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 46 Table 2.3-8k JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 47 Table 2.3-8l JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 48 Table 2.3-8m JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION F ROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 49 Table 2.3-8n JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 33-FT LEVEL FOR 1975 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 33 FT) 50 Table 2.3-9a JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 51 Table 2.3-9b JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEM PERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 52 Table 2.3-9c JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 53 Table 2.3-9d JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 54 Page 3 of 10 Revision 26 5/2016
GINNA/UFSAR Table 2.3-9e JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 55 Table 2.3-9f JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 56 Table 2.3-9g JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 57 Table 2.3-9h JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 58 Table 2.3-9i JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 59 Table 2.3-9j JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 60 Table 2.3-9k JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 61 Table 2.3-9l JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 62 Table 2.3-9m JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 63 Table 2.3-9n JOINT FREQUENCY TABLES OF WIND SPEED AND DIRECTION FROM 50-FT LEVEL FOR 1966, 1967, AND 1973-74 (TEMPERATURE DIFFERENCE BETWEEN 150 FT AND 10 FT; ADJUSTED TO 150 FT TO 33 FT. SPEED ADJUSTED TO 33 FT.) 64 Page 4 of 10 Revision 26 5/2016
GINNA/UFSAR Table 2.3-10 GASEOUS DISCHARGE POINTS AT THE GINNA SITE 65 Table 2.3-11 VENT DESIGN INFORMATION FOR GINNA 66 Table 2.3-12 TABULATION OF INPUT ASSUMPTIONS FOR CALCULATIONS 67 Table 2.3-13 TOPOGRAPHIC ELEVATIONS FEET (MSL) FOR GINNA SITE PLANT GRADE IS 270 FEET 68 Table 2.3-14 ANNUAL DIFFUSION AND DEPOSITION ESTIMATES FOR ALL RECEPTOR LOCATIONS, RELEASE POINT: PLANT VENTS, WAKE-SPLIT 69 Table 2.3-15 GRAZING SEASON DIFFUSION AND DEPOSITION ESTIMATES FOR LIVESTOCK RECEPTOR LOCATIONS, RELEASE POINT:
PLANT VENTS, WAKE-SPLIT 70 Table 2.3-16 GRAZING SEASON DIFFUSION AND DEPOSITION ESTIMATES FOR ALL RECEPTOR LOCATIONS, RELEASE POINT: PLANT VENTS, WAKE-SPLIT 71 Table 2.3-17 ANNUAL DIFFUSION AND DEPOSITION ESTIMATES FOR ALL RECEPTOR LOCATIONS, RELEASE POINT: GROUND RELEASE IN BUILDING WAKE 72 Table 2.3-18 GRAZING SEASON DIFFUSION AND DEPOSITION ESTIMATES FOR LIVESTOCK RECEPTOR LOCATIONS, RELEASE POINT:
ASSUMED GROUND RELEASE IN BUILDING WAKE 73 Table 2.3-19 GRAZING SEASON DIFFUSION AND DEPOSITION ESTIMATES FOR ALL RECEPTOR LOCATIONS, RELEASE POINT: ASSUMED GROUND RELEASE IN BUILDING WAKE 74 Table 2.3-20 EXCLUSION AREA BOUNDARY DISTANCES 75 Table 2.3-21 KRPavan DIRECTION-DEPENDENT EAB x/Q
SUMMARY
76 Table 2.3-22 KRPavan DIRECTION-DEPENDENT LPZ X/Q
SUMMARY
78 2.4 HYDROLOGIC ENGINEERING 80 2.4.1 HYDROLOGIC DESCRIPTION 80 2.4.2 FLOODS 80 Flood Design Considerations 80 Effects of Local Intense Precipitation 81 2.4.3 PROBABLE MAXIMUM FLOOD ON STREAMS AND RIVERS 81 Flood Evaluation Summary 81 Derivation of Probable Maximum Flood 82 Deleted 82 2.4.4 LAKE ONTARIO SURGE FLOODING 82 2.4.5 ICE EFFECTS 82 2.4.6 COOLING WATER CANALS AND RESERVOIRS 83 2.4.7 FLOODING PROTECTION REQUIREMENTS 83 2.4.8 LOW WATER CONSIDERATIONS 83 2.4.9 DISPERSION, DILUTION, AND TRAVEL TIMES OF RELEASES OF LIQUID EFFLUENTS IN SURFACE WATERS 84 Page 5 of 10 Revision 26 5/2016
GINNA/UFSAR Near-Shore Lake Currents 84 Dispersion of Regulated Radioactive Liquid Releases 84 Regulated Radioactive Liquid Releases 84 Liquid Dispersion 85 Effect of Local Recirculation 86 Concentration of Nearest Public Water Supply Intake 87 Environmental Monitoring Program 87 Dispersion of Accidental Radioactive Liquid Releases 87 Accidental Releases to the Lake 87 Accidental Spills on the Ground 88 2.4.10 DISPERSION, DILUTION, AND TRAVEL TIMES OF RELEASES OF LIQUID EFFLUENTS IN SURFACE WATERS 89 Design-Basis Groundwater Level 89 Groundwater Protection Program 89 Water Use 90 REFERENCES FOR SECTION 2.4 92 Table 2.4-1 DEER CREEK OVERFLOW
SUMMARY
TABLE 94 Table 2.4-2 INDUSTRIAL AND MUNICIPAL WATER SUPPLIES 95 2.5 GEOLOGY, SEISMOLOGY, AND GEOTECHNICAL ENGINEER ING 97 2.5.1 BASIC GEOLOGIC AND SEISMIC INFORMATION 97 Regional Geology 97 Site Geology 98 2.5.2 VIBRATORY GROUND MOTION 99 Seismicity 99 Maximum Earthquake Potential 99 Surface Faulting 100 Nearby Regional Faulting 100 Ginna Site Vicinity Faulting 101 Ginna Excavation 102 2.5.3 STABILITY OF SLOPES 102 General 102 Onsite Slopes 103 Stability Analyses 103 Failure Evaluation 103 REFERENCES FOR SECTION 2.5 104 Table 2.5-1 EARTHQUAKE ACTIVITY NEAR ATTICA, NEW YORK 106 Table 2.5-2 MATERIAL PROPERTIES USED IN THE NRC STAFF ANALYSIS 107 OF SLOPE STABILITY Page 6 of 10 Revision 26 5/2016
GINNA/UFSAR FIGURES Figure 2.1-1 Location of the R. E. Ginna Nuclear Power Plant Figure 2.1-2 R. E. Ginna Site Figure 2.1-3 Figure Deleted Figure 2.1-4 Projection of Population Distribution 0-5 Miles Figure 2.1-5 1980 Population Estimates 0-5 Miles Figure 2.1-5a 1992 Population Estimates 0-10 Miles Figure 2.1-6 Projection of Population Distribution 0-40 Miles Figure 2.1-7 Location of Ginna Site Figure 2.1-8 Population Centers Over 2000 Figure 2.3-1 Climate of the Ginna Site Region Figure 2.3-2 Wind Direction Patterns Long Period Averages Figure 2.3-3 Sensor Placements, Primary Meteorological Tower Figure 2.3-4 Sensor Placements, Backup (Substation 13A) Meteorological Tower Figure 2.3-5 Ginna 1966, 50-Ft Wind Rose Figure 2.3-6 Ginna 1967, 50-Ft Wind Rose Figure 2.3-7 Ginna 1973-1974, 50-Ft Wind Rose Figure 2.3-8 Ginna 1975, 33-Ft Wind Rose Figure 2.3-9 Site Plan - Activity Release Points and Elevation Figure 2.4-1 Lake Ontario Levels Figure 2.4-2 General North-South Cross Section Ginna Site Figure 2.4-3 FIGURE DELETED Figure 2.4 Ginna Site Layout and Topography Figure 2.5-1 Plot Plan Showing Boring Locations Figure 2.5-2 Epicentral Location Map Figure 2.5-3 NRC Systematic Evaluation Program, Site Specific Spectrum, Ginna Site (5% Damping)
Page 7 of 10 Revision 26 5/2016
GINNA/UFSAR APPENDIX 2A PROBABLE MAXIMUM FLOOD AND LOW WATER CONDITIONS 1 2A.1 ESTIMATE OF WAVE RUNUP ON VERTICAL PLANT WALL ROBERT EMMETT GINNA NUCLEAR POWER PLANT ROCHESTER, NEW YORK 3 2A.1.1 GENERAL 3 2A.1.2 DISCUSSION OF FACTORS 3 2A.1.3 ANALYSIS 4 2A.1.4 DISCUSSION 4 2A.
1.5 CONCLUSION
S 4 REFERENCES 6 2A.2 MAXIMUM PROBABLE WATER LEVELS IN LAKE ONTARIO AT THE ROBERT EMMETT GINNA NUCLEAR POWER PLANT SITE 7 2A.
2.1 INTRODUCTION
7 2A.3 ANALYSES OF ALTERNATIVE HYDROLOGIC AND METEOROLOGIC CRITERIA 8 A. LAKE ONTARIO REGULATION 8 B. RAINFALL 10 C. EXTRATROPICAL CYCLONES 11 D. TROPICAL CYCLONES 12 2A.4 DESIGN STORM ANALYSIS FOR MAXIMUM PROBABLE WATER LEVEL 13 2A.4.1 EXTRATROPICAL STORM ANALYSIS 13 2A.4.2 TROPICAL STORM ANALYSIS 14 2A.4.3 EXTREME LOW WATER LEVEL 16 CONCLUSIONS 16 BIBLIOGRAPHY REFERENCES 18 APPENDIX 2A FIGURES Figure 1 Lake Ontario Bottom Profile Figure Exhibit 1 A typical northeaster with a single center and a single frontal system, occurring 0100 EST, April 2, 1958, four hours before a peak surge of 3.5 ft. at Boston, Mass.
Figure Exhibit 2 A complex northeaster with two low centers and a double frontal system, occurring 1900 EST, March 1, 1914, three hours before a peak surge of 4.1 ft. at Portland, Maine.
Figure Exhibit 3 Lake Ontario Stage Data Page 8 of 10 Revision 26 5/2016
GINNA/UFSAR APPENDIX 2B DRIFT AND DISPERSION CHARACTERISTICS OF LAKE ONTARIO NEARSHORE WATERS ROCHESTER, NEW YORK TO SODUS BAY, NEW YORK 20 2B.1 DRIFT AND DISPERSION CHARACTERISTICS OF LAKE ONTARIO NEARSHORE WATERS ROCHESTER, NEW YORK TO SODUS BAY, NEW YORK 21 2B.1.1 Summary 21 2B.1.2 Introduction 22 2B.1.3 General comment on effect of discharges on Lake Ontario 23 2B.1.4 Currents in Lake Ontario 23 2B.1.5 Tracer releases 25 2B.1.6 Discussion 30 2B.1.7 Point source 31 2B.1.8 Line source 31 2B.1.9 Diffuser source 31 2B.1.10 Jet source 32 Table 1 Temperature (C) at a station 6000 feet offshore of Brookwood 34 APPENDIX A to Observed Tracer Distributions (Parts per Billion) 35 APPENDIX 2B APPENDIX B to Wind Speed and Direction Observations Meteorological Tower on APPENDIX 2B Brookwood Site Anemometer - Elevation 150 Feet 36 Table 1 37 APPENDIX 2B THE EFFECT ON LAKE DILUTION OF MOMENT MIXING 47 ATTACHMENT I Table 1 52 Table 2 53 APPENDIX 2B FIGURES Figure 1 Upper - Lake Current, Lower - Wind Speed Figure 2 Figure 3 Figure 4 Maximum concentration versus distance, May 1965 East Distribution Figure 5 Maximum concentration versus distance, July 1965 East Distribution Figure 6 Maximum concentration versus distance, October 1965 East Distribution Figure 7 Maximum concentration versus distance, May 1965 West Distribution Figure 8 Maximum concentration versus distance, October 1965 West Distribution Figure 9 Figure 10 Probable temperature elevation (degrees F) in upper six feet of Lake Ontario for horizontal canal discharge at the Brookwood site.
Page 9 of 10 Revision 26 5/2016
GINNA/UFSAR APPENDIX A to APPENDIX 2B FIGURES Figure 1 Figure 2 Figure 3 APPENDIX 2B ATTACHMENT I Figure 3 APPENDIX 2C REPORT, SUPPLEMENTARY FOUNDATION STUDIES, 56 PROPOSED BROOKWOOD NUCLEAR POWER PLANT (R. E.
GINNA NUCLEAR POWER PLANT), ONTARIO, NEW YORK 2C.1 INTRODUCTION 58 2C.1.1 GENERAL 58 2C.1.2 PURPOSE 58 2C.1.3 SCOPE OF WORK 58 2C.1.4 SITE CONDITIONS 58 2C.2 DISCUSSION AND RECOMMENDATIONS 60 2C.2.1 GENERAL 60 2C.2.2 FOUNDATION INSTALLATION PROCEDURES 60 2C.2.3 FOUNDATION DESIGN CRITERIA 61 2C.2.4 TURBINE-GENERATOR FOUNDATION 61 2C.3 Report Appendix - Field Explorations and Laboratory Tests 63 2C.3.1 FIELD EXPLORATIONS 63 2C.3.2 LABORATORY TESTS 63 Table 1
SUMMARY
OF SOIL STRENGTH TEST DATA 65 APPENDIX 2C FIGURES Figure 1 Plate 1 - Plot Plan Figure 2 Plate 2 - Foundation Design Data Figure 3 Soil Sampler Type U Figure 4 Methods of Performing Unconfined Compression and Triaxial Compression Tests Figure 5 Plate A-1A - Log of Borings (Borings 201 through 202)
Figure 6 Plate A-1B - Log of Borings (Borings 203 through 207)
Figure 7 Plate A Unified Soil Classification System and Key to Test Data Page 10 of 10 Revision 26 5/2016
GINNA/UFSAR 3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA 2 3.1.1 ATOMIC INDUSTRIAL FORUM DESIGN CRITERIA 2 3.1.1.1 Overall Plant Requirements 2 3.1.1.1.1 Quality Standards 2 3.1.1.1.2 Performance Standards 4 3.1.1.1.3 Fire Protection 5 3.1.1.1.4 Sharing of Systems 5 3.1.1.1.5 Records Requirements 5 3.1.1.2 Protection by Multiple Fission Product Barriers 6 3.1.1.2.1 Reactor Core Design 6 3.1.1.2.2 Suppression of Power Oscillations 7 3.1.1.2.3 Overall Power Coefficient 7 3.1.1.2.4 Reactor Coolant Pressure Boundary 7 3.1.1.2.5 Reactor Containment 8 3.1.1.3 Nuclear and Radiation Controls 9 3.1.1.3.1 Control Room 9 3.1.1.3.2 Instrumentation and Controls Systems 9 3.1.1.3.3 Fission Process Monitors and Controls 10 3.1.1.3.4 Core Protection Systems 11 3.1.1.3.5 Engineered Safety Features Protection Systems 11 3.1.1.3.6 Monitoring Reactor Coolant Leakage 12 3.1.1.3.7 Monitoring Radioactivity Releases 13 3.1.1.3.8 Monitoring Fuel and Waste Storage 13 3.1.1.4 Reliability and Testability of Protection Systems 14 3.1.1.4.1 Protection Systems Reliability 14 3.1.1.4.2 Protection Systems Redundancy and Independence 15 3.1.1.4.2.1 Reactor Trip Circuits 15 3.1.1.4.2.2 Engineered Safety Features Initiation Circuits 15 3.1.1.4.3 Single-Failure Definition (Category B) 16 3.1.1.4.4 Separation of Protection and Control Instrumentation Systems 16 3.1.1.4.5 Protection Against Multiple Disability for Protection Systems 16 3.1.1.4.6 Emergency Power for Protection Systems 16 3.1.1.4.7 Demonstration of Functional Operability of Protection Systems 17 3.1.1.4.8 Protection Systems Failure Analysis Design 17 Page 1 of 39 Revision 26 5/2016
GINNA/UFSAR 3.1.1.5 Reactivity Control 18 3.1.1.5.1 Redundancy of Reactivity Control 18 3.1.1.5.2 Reactivity Hot Shutdown Capability 18 3.1.1.5.3 Reactivity Shutdown Capability 18 3.1.1.5.4 Reactivity Hold-Down Capability 19 3.1.1.5.5 Reactivity Control Systems Malfunction 19 3.1.1.5.6 Maximum Reactivity Worth of Control Rods 20 3.1.1.6 Reactor Coolant Pressure Boundary 20 3.1.1.6.1 Reactor Coolant Pressure Boundary Capability 20 3.1.1.6.2 Reactor Coolant Pressure Boundary Rapid Propagation Failure Preven- 21 tion 3.1.1.6.3 Reactor Coolant Pressure Boundary Brittle Fracture Prevention 22 3.1.1.6.4 Reactor Coolant Pressure Boundary Surveillance 22 3.1.1.7 Engineered Safety Features 23 3.1.1.7.1 Engineered Safety Features Basis for Design 23 3.1.1.7.2 Reliability and Testability of Engineered Safety Features 24 3.1.1.7.3 Emergency Power 24 3.1.1.7.4 Missile Protection 25 3.1.1.7.5 Engineered Safety Features Performance Capability 26 3.1.1.7.6 Engineered Safety Features Components Capability 26 3.1.1.7.7 Accident Aggravation Prevention 27 3.1.1.7.8 Emergency Core Cooling System (ECCS) Capability 27 3.1.1.7.9 Inspection of Emergency Core Cooling System (ECCS) 28 3.1.1.7.10 Testing of Emergency Core Cooling System (ECCS) Components 28 3.1.1.7.11 Testing of Emergency Core Cooling System (ECCS) 28 3.1.1.7.12 Testing of Operational Sequence of Emergency Core Cooling System 28 (ECCS) 3.1.1.7.13 Containment Design Basis 29 3.1.1.7.14 Nil Ductility Transition Temperature Requirement for Containment 29 Material 3.1.1.7.15 Reactor Coolant Pressure Boundary Outside Containment 30 3.1.1.7.16 Containment Heat Removal Systems 30 3.1.1.7.17 Containment Isolation Valves 30 3.1.1.7.18 Initial Leakage Rate Testing of Containment 30 3.1.1.7.19 Periodic Containment Leakage Rate Testing 31 3.1.1.7.20 Provisions for Testing of Penetrations 31 3.1.1.7.21 Provisions for Testing of Isolation Valves 31 Page 2 of 39 Revision 26 5/2016
GINNA/UFSAR 3.1.1.7.22 Inspection of Containment Pressure-Reducing Systems 32 3.1.1.7.23 Testing of Containment Pressure-Reducing Systems Components 32 3.1.1.7.24 Testing of Containment Spray Systems 32 3.1.1.7.25 Testing of Operational Sequence of Containment Pressure-Reducing 32 Systems 3.1.1.7.26 Inspection of Air Cleanup Systems 33 3.1.1.7.27 Testing of Air Cleanup Systems Components 33 3.1.1.7.28 Testing Air Cleanup System 33 3.1.1.7.29 Testing of Operational Sequence of Air Cleanup Systems 33 3.1.1.8 Fuel and Waste Storage Systems 34 3.1.1.8.1 Prevention of Fuel Storage Criticality 34 3.1.1.8.2 Fuel and Waste Storage Decay Heat 34 3.1.1.8.3 Fuel and Waste Storage Radiation Shielding 35 3.1.1.8.4 Protection Against Radioactivity Release From Spent Fuel and Waste 35 Storage 3.1.1.9 Control of Releases of Radioactivity to the Environment 35 3.1.2 GENERAL DESIGN CRITERIA 36 3.1.2.1 Overall Requirements 36 3.1.2.1.1 General Design Criterion 1 - Quality Standards and Records 37 3.1.2.1.2 General Design Criterion 2 - Design Bases for Protection Against Nat- 38 ural Phenomena 3.1.2.1.3 General Design Criterion 3 - Fire Protection 38 3.1.2.1.4 General Design Criterion 4 - Environmental and Missile Design Bases 39 3.1.2.1.5 General Design Criterion 5 - Sharing of Structures, Systems, and Com- 39 ponents 3.1.2.2 Protection by Multiple Fission Product Barriers 39 3.1.2.2.1 General Design Criterion 10 - Reactor Design 39 3.1.2.2.2 General Design Criterion 11 - Reactor Inherent Protection 40 3.1.2.2.3 General Design Criterion 12 - Suppression of Reactor Power Oscilla- 40 tions 3.1.2.2.4 General Design Criterion 13 - Instrumentation and Control 40 3.1.2.2.5 General Design Criterion 14 - Reactor Coolant Pressure Boundary 41 3.1.2.2.6 General Design Criterion 15 - Reactor Coolant System Design 41 3.1.2.2.7 General Design Criterion 16 - Containment Design 42 3.1.2.2.8 General Design Criterion 17 - Electrical Power Systems 42 3.1.2.2.9 General Design Criterion 18 - Inspection and Testing of Electrical 44 Power Systems 3.1.2.2.10 General Design Criterion 19 - Control Room 44 Page 3 of 39 Revision 26 5/2016
GINNA/UFSAR 3.1.2.3 Protection and Reactivity Control Systems 45 3.1.2.3.1 General Design Criterion 20 - Protection Systems Functions 45 3.1.2.3.2 General Design Criterion 21 - Protection System Reliability and Test- 45 ability 3.1.2.3.3 General Design Criterion 22 - Protection System Independence 46 3.1.2.3.4 General Design Criterion 23 - Protection System Failure Modes 46 3.1.2.3.5 General Design Criterion 24 - Separation of Protection and Control 47 Systems 3.1.2.3.6 General Design Criterion 25 - Protection System Requirements for 47 Reactivity Control Malfunctions 3.1.2.3.7 General Design Criterion 26 - Reactivity Control System Redundancy 48 and Capability 3.1.2.3.8 General Design Criterion 27 - Combined Reactivity Control System 48 Capability 3.1.2.3.9 General Design Criterion 28 - Reactivity Limits 49 3.1.2.3.10 General Design Criterion 29 - Protection Against Anticipated Opera- 49 tional Occurrences 3.1.2.4 Fluid Systems 49 3.1.2.4.1 General Design Criterion 30 - Quality of Reactor Coolant Pressure 49 Boundary 3.1.2.4.2 General Design Criterion 31 - Fracture Prevention of Reactor Coolant 50 Pressure Boundary 3.1.2.4.3 General Design Criterion 32 - Inspection of Reactor Coolant Pressure 51 Boundary 3.1.2.4.4 General Design Criterion 33 - Reactor Coolant Makeup 51 3.1.2.4.5 General Design Criterion 34 - Residual Heat Removal 52 3.1.2.4.6 General Design Criterion 35 - Emergency Core Cooling 52 3.1.2.4.7 General Design Criterion 36 - Inspection of Emergency Core Cooling 53 System (ECCS) 3.1.2.4.8 General Design Criterion 37 - Testing of Emergency Core Cooling 53 Systems (ECCS) 3.1.2.4.9 General Design Criterion 38 - Containment Heat Removal 54 3.1.2.4.10 General Design Criterion 39 - Inspection of Containment Heat 54 Removal System 3.1.2.4.11 General Design Criterion 40 - Testing of Containment Heat Removal 54 System 3.1.2.4.12 General Design Criterion 41 - Containment Atmosphere Cleanup 55 3.1.2.4.13 General Design Criterion 42 - Inspection of Containment Atmosphere 56 Cleanup Systems 3.1.2.4.14 General Design Criterion 43 - Testing of Containment Atmosphere 56 Cleanup Systems 3.1.2.4.15 General Design Criterion 44 - Cooling Water 57 Page 4 of 39 Revision 26 5/2016
GINNA/UFSAR 3.1.2.4.16 General Design Criterion 45 - Inspection of Cooling Water System 57 3.1.2.4.17 General Design Criterion 46 - Testing of Cooling Water System 58 3.1.2.5 Reactor Containment 58 3.1.2.5.1 General Design Criterion 50 - Containment Design Basis 58 3.1.2.5.2 General Design Criterion 51 - Fracture Prevention of Containment 59 Pressure Boundary 3.1.2.5.3 General Design Criterion 52 - Capability for Containment Leakage 59 Rate Testing 3.1.2.5.4 General Design Criterion 53 - Provisions for Containment Testing and 59 Inspection 3.1.2.5.5 General Design Criterion 54 - Piping Systems Penetrating Containment 60 3.1.2.5.6 General Design Criterion 55 - Reactor Coolant Pressure Boundary Pen- 60 etrating Containment 3.1.2.5.7 General Design Criterion 56 - Primary Containment Isolation 61 3.1.2.5.8 General Design Criterion 57 - Closed System Isolation Valves 62 3.1.2.6 Fuel and Radioactivity Control 62 3.1.2.6.1 General Design Criterion 60 - Control of Releases of Radioactive 62 Materials to the Environment 3.1.2.6.2 General Design Criterion 61 - Fuel Storage and Handling and Radioac- 62 tivity Control 3.1.2.6.3 General Design Criterion 62 - Prevention of Criticality in Fuel Storage 63 and Handling 3.1.2.6.4 General Design Criterion 63 - Monitoring Fuel and Waste Storage 63 3.1.2.6.5 General Design Criterion 64 - Monitoring Radioactivity Releases 64 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYS- 66 TEMS 3.
2.1 INTRODUCTION
66 3.2.2 SYSTEMATIC EVALUATION PROGRAM EVALUATION 66 3.2.2.1 Fracture Toughness 67 3.2.2.1.1 Pressurizer 67 3.2.2.1.2 Accumulators 68 3.2.2.1.3 Component Cooling Water (CCW) Pumps 68 3.2.2.1.4 Service Water Pumps 68 3.2.2.1.5 Main Steam Piping and Valves 69 3.2.2.1.6 Feedwater Piping and Valves 69 3.2.2.2 Radiography Requirements 69 3.2.2.2.1 Class 2 Pressure Vessels 69 3.2.2.2.2 Class 1 and 2 Welded Joints 70 3.2.2.2.3 Main Steam and Feedwater Piping 70 Page 5 of 39 Revision 26 5/2016
GINNA/UFSAR 3.2.2.3 Valve Design 71 3.2.2.4 Pump Design 71 3.2.2.5 Storage Tank Design 72 Table 3.2-1 CLASSIFICATION OF STRUCTURES, SYSTEMS, AND 74 COMPONENTS 3.3 WIND AND TORNADO LOADINGS 84 3.
3.1 INTRODUCTION
84 3.3.2 STRUCTURAL UPGRADE PROGRAM EVALUATION 84 3.3.2.1 Structural Evaluation Approach 84 3.3.2.1.1 Requirements 84 3.3.2.1.2 Structural Evaluation Process 84 3.3.2.1.3 Structural Evaluation Computer Program 85 3.3.2.1.4 Input Load Criteria 85 3.3.2.1.5 General Assumptions 86 3.3.2.1.6 Load Combinations and Acceptance Criteria 87 3.3.2.2 Structural Evaluation 88 3.3.2.2.1 Primary Member Evaluation 88 3.3.2.2.2 Secondary Member Evaluation 89 3.3.2.2.3 Connections and Anchorages Evaluation 89 3.3.2.2.4 Exterior Shell Evaluation 90 3.3.2.2.4.1 Siding 90 3.3.2.2.4.2 Concrete Masonry Block Walls 90 3.3.2.2.4.3 Architectural Items 91 3.3.2.3 Results of the Structural Evaluation 91 3.3.2.3.1 Primary Members 91 3.3.2.3.1.1 General 91 3.3.2.3.1.2 Severe Environmental Conditions 91 3.3.2.3.1.3 Extreme Snow Load Condition 92 3.3.2.3.1.4 132-mph Tornado 92 3.3.2.3.1.5 188-mph Tornado 92 3.3.2.3.1.6 250-mph Tornado 92 3.3.2.3.2 Secondary Members 93 3.3.2.3.3 Connections and Anchorages 93 3.3.2.3.4 Exterior Shell 94 3.3.2.3.4.1 Metal Siding 94 3.3.2.3.4.2 Roof Decking 94 Page 6 of 39 Revision 26 5/2016
GINNA/UFSAR 3.3.2.3.4.3 Block Walls 94 3.3.3 TORNADO MISSILES AND SAFE SHUTDOWN APPROACH 94 3.3.3.1 Background 94 3.3.3.2 Shutdown Methodology 95 3.3.3.2.1 Assumptions 95 3.3.3.2.2 Shutdown Details 95 3.3.3.3 Required Components 96 3.3.3.3.1 Refueling Water Storage Tank (RWST) 96 3.3.3.3.2 Electrical Buses 14, 17, and 18 96 3.3.3.3.3 Main Steam Lines A and B, and Main Feedwater Lines A and B 97 3.3.3.3.3.1 Results - Steel Rod 97 3.3.3.3.3.2 Results - Utility Pole 97 3.3.3.3.3.3 Failure of Block Walls 97 3.3.3.3.4 Surface of the Spent Fuel Pool 98 3.3.3.3.5 Diesel Generators and Their Fuel Supply 98 3.3.3.3.6 Relay Room 98 3.3.3.3.7 Service Water System 99 3.3.3.3.8 Standby Auxiliary Feedwater System 99 3.3.3.3.9 Instrumentation 99 3.3.3.3.10 Cable Tunnel 100 3.3.4 DESIGN TORNADO 100 3.3.4.1 Introduction 100 3.3.4.2 Safety Assessment 100 3.3.4.3 Reserve Plant Capacity 101 3.3.4.4 System Reserve Capacity 102 3.3.5 STRUCTURAL UPGRADE PROGRAM 103 3.3.5.1 Introduction 103 3.3.5.2 Criteria Changes 103 3.3.5.2.1 First Stage Review 103 3.3.5.2.2 Second Stage Review 104 3.3.5.3 Stability Evaluation 105 3.3.5.3.1 Primary Members 105 3.3.5.3.2 Connections and Anchorages 105 3.3.5.4 NRC Technical Evaluation Report (SEP Topic III-2) Open Items 106 3.3.5.4.1 Effective Tornado Loadings 106 3.3.5.4.2 Structural Loadings 107 Page 7 of 39 Revision 26 5/2016
GINNA/UFSAR 3.3.5.4.3 Structural Acceptance Criteria 107 3.3.5.4.4 Structural Systems 107 3.3.5.5 SEP Topic III-7.B, Loads, Load Combinations, and Design Criteria 108 3.3.5.6 Diesel Generator Component Operability 109 3.3.5.7 Conclusions 109 3.3.6 INTERMEDIATE BUILDING BLOCK WALL REINFORCEMENT 110 Table 3.3-1 PRIMARY MEMBER FAILURES PER LOADING COMBINATION 113 3.4 WATER LEVEL (FLOOD) DESIGN 114 3.4.1 FLOOD PROTECTION 114 3.4.1.1 Flood Protection Measures for Seismic Category I Structures 114 3.4.1.1.1 Introduction 114 3.4.1.1.2 Lake Ontario Flood Protection 114 3.4.1.1.3 Deer Creek Flood Protection 114 3.4.1.2 Permanent Dewatering System 115 3.4.2 FLOODING DUE TO FAILURE OF TANKS 115 3.4.3 ROOF DRAINAGE 116 3.5 MISSILE PROTECTION 119 3.5.1 INTERNALLY GENERATED MISSILES 119 3.5.1.1 Introduction 119 3.5.1.1.1 Design Criteria 119 3.5.1.1.2 Systematic Evaluation Program 119 3.5.1.2 Turbine Missiles 120 3.5.1.2.1 Introduction 120 3.5.1.2.2 Turbine Inspection Program 121 3.5.1.2.3 Systematic Evaluation Program Topic III-4 121 3.5.1.3 Effects of Internally Generated Missiles on Systems and Equipment 122 3.5.1.3.1 Systems Needed to Perform Safety Functions 122 3.5.1.3.1.1 Reactor Coolant System 122 3.5.1.3.1.2 Emergency Core Cooling System (ECCS) 123 3.5.1.3.1.3 Containment Heat Removal and Atmosphere Cleanup Systems 124 3.5.1.3.1.4 Chemical and Volume Control System 125 3.5.1.3.1.5 Residual Heat Removal System 126 3.5.1.3.1.6 Component Cooling Water System 126 3.5.1.3.1.7 Service Water System 126 3.5.1.3.1.8 Diesel-Generator Auxiliary Systems 127 Page 8 of 39 Revision 26 5/2016
GINNA/UFSAR 3.5.1.3.1.9 Main Steam System 127 3.5.1.3.1.10 Feedwater and Condensate Systems 128 3.5.1.3.1.11 Preferred Auxiliary Feedwater System 128 3.5.1.3.1.12 Standby Auxiliary Feedwater System (SAFW) 128 3.5.1.3.1.13 Ventilation Systems for Vital Areas 129 3.5.1.3.1.14 Combustible Gas Control System 129 3.5.1.3.2 Systems Whose Failure May Result in Activity Release 129 3.5.1.3.2.1 Spent Fuel Pool Cooling System 129 3.5.1.3.2.2 Sampling System 130 3.5.1.3.2.3 Waste Disposal System 130 3.5.1.3.2.4 Containment Shutdown Purge System 130 3.5.1.3.2.5 Instrument and Service Air Systems 130 3.5.1.3.3 Electrical Systems 131 3.5.1.3.3.1 Diesel Generators 131 3.5.1.3.3.2 Station Batteries 131 3.5.1.3.3.3 480-Volt Switchgear 131 3.5.1.3.3.4 Control Room 131 3.5.1.3.3.5 Cable Spreading/Relay Room 131 3.5.2 EXTERNALLY GENERATED MISSILES 132 3.5.2.1 Tornado Missiles 132 3.5.2.2 Site Proximity Missiles 132 3.5.2.2.1 Design Criteria 132 3.5.2.2.2 Nearby Hazardous Activities 132 3.5.2.2.3 Aircraft Hazards 133 3.6 PROTECTION AGAINST THE DYNAMIC EFFECTS ASSOCI- 135 ATED WITH THE POSTULATED RUPTURE OF PIPING 3.6.1 POSTULATED PIPING FAILURES IN FLUID SYSTEMS INSIDE 135 CONTAINMENT 3.6.1.1 Evaluation Procedure 135 3.6.1.1.1 Pipe Selection 135 3.6.1.1.2 Effects-Oriented Evaluation 136 3.6.1.1.3 Mechanistic Evaluation 136 3.6.1.2 Required Equipment 137 3.6.1.3 Safety Analysis 137 3.6.1.3.1 Single-Failure Considerations 137 3.6.1.3.1.1 Introduction 137 3.6.1.3.1.2 Containment Fan Coolers 138 Page 9 of 39 Revision 26 5/2016
GINNA/UFSAR 3.6.1.3.1.3 Low-Pressure Safety Injection Isolation Valves 138 3.6.1.3.2 High-Energy Line Break Effects 138 3.6.1.3.2.1 Introduction 138 3.6.1.3.2.2 Alternate Charging 139 3.6.1.3.2.3 Residual Heat Removal Pump Suction 139 3.6.1.3.2.4 Reactor Coolant Pump Seal-Water to Seals 140 3.6.1.3.2.5 Letdown Line 140 3.6.1.3.2.6 Charging Line 141 3.6.1.3.2.7 Steam Generator Blowdown Lines 142 3.6.1.3.2.8 Main Steam and Feedwater Lines 142 3.6.1.3.2.9 Residual Heat Removal Pump Discharge Line 145 3.6.1.3.2.10 Standby Auxiliary Feedwater Lines 145 3.6.1.3.2.11 Accumulator Lines and Branch Lines 145 3.6.1.3.2.12 Auxiliary Spray Line 148 3.6.1.3.2.13 Reactor Coolant System 149 3.6.1.3.2.14 Pressurizer Surge Line 149 3.6.1.3.2.15 Pressurizer Spray Lines 152 3.6.1.3.2.16 Pressurizer Safety and Relief Lines 152 3.6.2 POSTULATED PIPING FAILURES IN FLUID SYSTEMS OUTSIDE 153 CONTAINMENT 3.6.2.1 Introduction and Summary 153 3.6.2.1.1 Initial Evaluation 153 3.6.2.1.2 Systematic Evaluation Program Reevaluation 154 3.6.2.2 Evaluation Procedure 155 3.6.2.2.1 Initial Evaluation 155 3.6.2.2.2 Systematic Evaluation Program Reevaluation 156 3.6.2.3 Analysis Criteria 157 3.6.2.3.1 December 18, 1972, AEC Letter Evaluation Criteria 157 3.6.2.3.2 Systematic Evaluation Program Criteria 157 3.6.2.3.2.1 High-Energy Fluid Systems Piping 157 3.6.2.3.2.2 Moderate-Energy Fluid System Piping 159 3.6.2.3.2.3 Type of Breaks and Leakage Cracks in Fluid System Piping 160 3.4.1.1.1.1 Assumptions 161 3.4.1.1.1.2 Effects of Piping Failure 162 3.4.1.2 Analysis in Response to December 18, 1972, AEC Letter 162 Page 10 of 39 Revision 26 5/2016
GINNA/UFSAR 3.4.1.2.1 Rupture Load Analysis 162 3.4.1.2.2 Main Steam System Load Analysis 163 3.4.1.2.3 Feedwater System Load Analysis 163 3.4.1.2.4 Jet Impingement Load Analysis 163 3.4.1.2.5 Pipe Whip Analysis for Main Steam and Feedwater Piping 164 3.4.1.2.5.1 Analytical Methods 164 3.4.1.2.5.2 Results of Analysis 164 3.4.1.2.6 Blowdown Analysis 165 3.4.1.2.6.1 Main Steam Blowdown Analysis 165 3.4.1.2.6.2 Feedwater Blowdown Analysis 165 3.4.1.2.7 Compartment Pressurization Analysis 166 3.4.1.2.7.1 Main Steam Line Ruptures 166 3.4.1.2.7.2 Building Pressurization for a Branch Line Rupture 166 3.4.1.2.8 Flooding Analysis 166 3.4.1.2.8.1 Intermediate Building Flooding 166 3.4.1.2.8.2 Screen House and Turbine Building Flooding 167 3.4.1.3 Systematic Evaluation Program Analysis 167 3.4.1.3.1 Zone Reevaluation Performed as Part of the Systematic Evaluation 167 Program Review 3.4.1.3.1.1 Screen House 167 3.4.1.3.1.2 Intermediate Building 168 3.4.1.3.1.3 Turbine Building Main Steam and Main Feedwater Line Breaks 169 3.4.1.3.1.4 Structural Analysis of the Turbine Building for Pressurization 170 3.4.1.3.1.5 Battery Room/Mechanical Equipment Room Flooding 172 3.4.1.3.1.6 Auxiliary Feedwater Line Breaks on the 253-Ft Elevation of the Inter- 172 mediate Building 3.4.1.3.1.7 Relay Room and Air Handling Room 172 3.4.1.3.1.8 Auxiliary Building 173 3.4.1.3.2 Main Steam Safety and Relief Valves 174 3.4.1.3.2.1 Pipe Failures in the Intermediate Building 174 3.6.2.5.2.2 Pipe Failures in the Turbine Building 175 3.6.2.5.2.3 Decay Heat Removal Following Blowdown from Both Steam Genera- 176 tors 3.6.2.5.2.4 Conclusions 177 Table 3.6-1 LINES PENETRATING CONTAINMENT WHICH NORMALLY OR 181 OCCASIONALLY EXPERIENCE HIGH-ENERGY SERVICE CON-DITIONS Page 11 of 39 Revision 26 5/2016
GINNA/UFSAR Table 3.6-2 LINES INSIDE CONTAINMENT BUT NOT PENETRATING CON- 183 TAINMENT WHICH NORMALLY OR OCCASIONALLY EXPERI-ENCE HIGH-ENERGY SERVICE CONDITIONS Table 3.6-3 CONTAINMENT PIPE DATA 184 3.7 SEISMIC DESIGN 186 3.7.1 SEISMIC INPUT 186 3.7.1.1 Introduction 186 3.7.1.1.1 Original Seismic Classification 186 3.7.1.1.2 Seismic Reevaluation 187 3.7.1.1.2.1 Scope of Reevaluation 187 3.7.1.1.2.2 Reevaluation Criteria 187 3.7.1.2 Design Response Spectra 188 3.7.1.3 Design Time-History 188 3.7.1.4 Critical Damping Values 189 3.7.1.5 Supporting Media for Seismic Category I Structures 189 3.7.2 SEISMIC SYSTEM ANALYSIS 190 3.7.2.1 Seismic Analysis Methods 190 3.7.2.1.1 Original Seismic Analysis 190 3.7.2.1.2 Seismic Reevaluation 191 3.7.2.2 Natural Frequencies and Response Loads 192 3.7.2.3 Procedure Used for Mathematical Modeling 192 3.7.2.4 Soil-Structure Interaction 192 3.7.2.5 Development of Floor Response Spectra 192 3.7.2.6 Combination of Earthquake Directional Components 193 3.7.2.7 Combination of Modal Responses 193 3.7.2.8 Interaction of Nonseismic Structures with Seismic Category I Struc- 193 tures 3.7.2.9 Use of Constant Vertical Static Factors 194 3.7.3 SEISMIC SUBSYSTEM ANALYSIS 194 3.7.3.1 Seismic Analysis Methods 194 3.7.3.1.1 Original Design 194 3.7.3.1.1.1 Piping and Tanks 194 3.7.3.1.1.2 Steam Generator 195 3.7.3.1.1.3 Control Rod Drive Mechanisms 195 3.7.3.1.1.4 Reactor Internals 195 3.7.3.1.1.5 Reactor Vessel 196 Page 12 of 39 Revision 26 5/2016
GINNA/UFSAR 3.7.3.1.1.6 Pressurizer 196 3.7.3.1.2 Seismic Reevaluation 197 3.7.3.2 Basis for Selection of Frequencies 198 3.7.3.3 Use of Equivalent Static Analysis 198 3.7.3.4 Three Components of Earthquake Motion 198 3.7.3.5 Combination of Modal Responses 199 3.7.3.6 Analytical Procedures for Piping 199 3.7.3.6.1 Residual Heat Removal System Line from Reactor Coolant System 199 Loop A to Containment 3.7.3.6.2 Steam Line from Steam Generator B to Containment 200 3.7.3.6.3 Pressurizer Safety and Relief Lines 200 3.7.3.6.3.1 Analytical Methods 200 3.7.3.6.3.2 Transfer Matrix Method 201 3.7.3.6.3.3 Stiffness Matrix Formulation 202 3.7.3.7 Seismic Piping Upgrade Program 203 3.7.3.7.1 Program Scope 203 3.7.3.7.2 Piping Selection Criteria 203 3.7.3.7.3 Selected Lines 204 3.7.3.7.3.1 Reactor Coolant System 204 3.7.3.7.3.2 Main Steam 204 3.7.3.7.3.3 Main Feedwater 204 3.7.3.7.3.4 Auxiliary Feedwater 204 3.7.3.7.3.5 Safety Injection 205 3.7.3.7.3.6 Residual Heat Removal 205 3.7.3.7.3.7 Containment Spray 205 3.7.3.7.3.8 Chemical and Volume Control System 206 3.7.3.7.3.9 Steam Generator Blowdown 206 3.7.3.7.3.10 Service Water System 206 3.7.3.7.3.11 Component Cooling Water 207 3.7.3.7.3.12 Standby Auxiliary Feedwater 208 3.7.3.7.4 Codes and Standards 208 3.7.3.7.5 Analytical Procedures 208 3.7.3.7.5.1 General 208 3.7.3.7.5.2 Damping Values 208 3.7.3.7.5.3 Combination of Modal Responses 209 Page 13 of 39 Revision 26 5/2016
GINNA/UFSAR 3.7.3.7.5.4 Safe Shutdown Earthquake Stresses 211 3.7.3.7.5.5 Small Piping Analysis 212 3.7.3.7.5.6 Branch Line Analysis 212 3.7.3.7.5.7 Piping Beyond Scope of Upgrade Program 212 3.7.3.7.6 Piping System Models 213 3.7.3.7.7 Valve Model 214 3.7.3.7.8 Equipment Model 214 3.7.3.7.9 Interaction Effects 214 3.7.3.7.10 Support Model 214 3.7.3.7.10.1 Deviations 214 3.7.3.7.10.2 Support-Welded Attachments 215 3.7.4 SEISMIC INSTRUMENTATION 216 Table 3.7-1 ORIGINAL AND CURRENT RECOMMENDED DAMPING 218 VALUES Table 3.7-2 MODAL FREQUENCIES OF THE INTERCONNECTED 219 BUILDING MODEL Table 3.7-3 EQUIPMENT AND LOCATIONS WHERE IN-STRUCTURE 221 SPECTRA WERE GENERATED FOR THE SYSTEMATIC EVALUATION PROGRAM 3.8 DESIGN OF SEISMIC CATEGORY I STRUCTURES 222 3.8.1 CONTAINMENT 222 3.8.1.1 General Description 222 3.8.1.1.1 Containment Structure 222 3.8.1.1.2 Waterproofing 223 3.8.1.1.3 Rock Anchors 223 3.8.1.1.4 Construction Sequence 223 3.8.1.1.5 Steel Reinforcement 225 3.8.1.2 Mechanical Design Bases 226 3.8.1.2.1 General 226 3.8.1.2.2 Design Loads 226 3.8.1.2.3 Design Stress Criteria 227 3.8.1.2.3.1 Limiting Loads 227 3.8.1.2.3.2 Load Factors 228 3.8.1.2.3.3 Maximum Thermal Load 228 3.8.1.2.4 Load Capacity 229 Page 14 of 39 Revision 26 5/2016
GINNA/UFSAR 3.8.1.2.4.1 Reinforced Concrete 229 3.8.1.2.4.2 Prestressed Concrete 230 3.8.1.2.4.3 Liner 232 3.8.1.2.4.4 Rock 233 3.8.1.2.5 Codes and Standards 233 3.8.1.2.5 Codes and Standards Steam Generator Replacement (Dome 236 Opening Repairs 3.8.1.3 Seismic Design 238 3.8.1.3.1 Initial Seismic Design 238 3.8.1.3.2 Seismic Reanalysis 239 3.8.1.4 Containment Detailed Design 239 3.8.1.4.1 Stress Analysis 239 3.8.1.4.1.1 Analysis Methods 239 3.8.1.4.1.2 Analysis Results 240 3.8.1.4.1.3 Analysis for Steam Generator Replacement Dome Openings 241 3.8.1.4.2 Rock Anchors 241 3.8.1.4.2.1 Rock Anchor Design 241 3.8.1.4.2.2 Preinstallation Grouting Test 242 3.8.1.4.2.3 Previous Applications 243 3.8.1.4.2.4 Rock Hold-Down Capacity 243 3.8.1.4.2.5 Hold-Down Factor of Safety 245 3.8.1.4.2.6 Installation 245 3.8.1.4.3 Tendons 246 3.8.1.4.3.1 General Design 246 3.8.1.4.3.2 Seismic Considerations 248 3.8.1.4.3.3 Stressing Procedure 250 3.8.1.4.3.4 Corrosion Protection 251 3.8.1.4.4 Hinge Design 253 3.8.1.4.4.1 Tension Bars 253 3.8.1.4.4.2 Liner Knuckle 255 3.8.1.4.4.3 Elastomer Bearing Pads 256 3.8.1.4.5 Concrete 258 3.8.1.4.5.1 Radial Shear 258 3.8.1.4.5.2 Longitudinal Shears 258 3.8.1.4.5.3 Horizontal Shear 259 3.8.1.4.5.4 Anchorage Stresses 260 Page 15 of 39 Revision 26 5/2016
GINNA/UFSAR 3.8.1.4.5.5 Shell Stress Analytical Procedures 261 3.8.1.4.6 Insulation 266 3.8.1.4.7 Liner 267 3.8.1.4.7.1 Vibrations 267 3.8.1.4.7.2 Anchorage Fatigue Analysis 267 3.8.1.4.7.3 Base Slab Liner 267 3.8.1.4.7.4 Liner Stresses 268 3.8.1.4.7.5 Liner Buckling 269 3.8.1.4.7.6 Liner Corrosion Allowance 273 3.8.1.5 Penetrations 273 3.8.1.5.1 General 273 3.8.1.5.2 Electrical Penetrations 274 3.8.1.5.3 Piping Penetrations 275 3.8.1.5.4 Access Hatch and Personnel Locks 275 3.8.1.5.5 Fuel Transfer Penetration 276 3.8.1.5.6 Typical Penetration Analysis 277 3.8.1.5.6.1 Loss-of-Coolant Accident 277 3.8.1.5.6.2 Loss-of-Coolant Accident Plus Earthquake 279 3.8.1.5.7 Penetration Reinforcement Analyzed for Pipe Rupture 280 3.8.1.6 Quality Control and Material Specifications 281 3.8.1.6.1 Concrete 281 3.8.1.6.1.1 Ultimate Compressive Strength 281 3.8.1.6.1.2 Quality Control Measures 281 3.8.1.6.1.3 Concrete Suppliers 282 3.8.1.6.1.4 Concrete Specifications 283 3.8.1.6.1.5 Admixtures 285 3.8.1.6.2 Mild Steel Reinforcement 287 3.8.1.6.3 Cadwell Splices 288 3.8.1.6.4 Radial Tension Bars 289 3.8.1.6.5 Containment Liner 289 3.8.1.6.5.1 Fabrication and Workmanship 289 3.8.1.6.5.2 Penetrations 290 3.8.1.6.5.3 Welding 290 3.8.1.6.5.4 Erection Tolerances 291 3.8.1.6.5.5 Painting 291 3.8.1.6.6 Elastomer Pads 292 Page 16 of 39 Revision 26 5/2016
GINNA/UFSAR 3.8.1.6.7 Tendons 292 3.8.1.6.7.1 Materials 292 3.8.1.6.7.2 Tests and Inspection 293 3.8.1.6.8 Liner Insulation 293 3.8.1.7 Testing and Inservice Inspection Requirements 294 3.8.1.7.1 Construction Phase Testing 294 3.8.1.7.1.1 Liner 294 3.8.1.7.1.2 Prestressing Tendons 295 3.8.1.7.1.3 Concrete Reinforcement 295 3.8.1.7.1.4 Concrete 296 3.8.1.7.1.5 Elastomer Bearing Pads 297 3.8.1.7.1.6 Rock Anchor Tests 298 3.8.1.7.1.7 Large Opening Reinforcements 299 3.8.1.7.1.8 Liner Insulation 299 3.8.1.7.2 General Description of the Structural Integrity Test 299 3.8.1.7.2.1 Pressurization 299 3.8.1.7.2.2 Measurements 300 3.8.1.7.2.3 Test Pressure Justification 302 3.8.1.7.2.4 Test Results 302 3.8.1.7.2.5 Containment Return to Service Testing Post 1996 Steam Generator 302 Replacement 3.8.1.7.3 Postoperational Surveillance 303 3.8.1.7.3.1 Leakage Monitoring 303 3.8.1.7.3.2 Initial Tendon Surveillance Program 303 3.8.1.7.3.3 Current Tendon Surveillance Program 304 3.8.1.7.3.4 Current Tendon Surveillance Program Results 305 3.8.1.7.3.5 Test on Rock Anchors 306 3.8.1.7.3.6 Inservice Inspection 306 3.8.2 STRUCTURAL REANALYSIS PROGRAM 307 3.8.2.1 Design Codes, Criteria, and Load Combinations - SEP Topic III-7.B 307 3.8.2.1.1 Introduction 307 3.8.2.1.1.1 Seismic Category I Structures 307 3.8.2.1.1.2 Structural Codes 308 3.8.2.1.1.3 Code Comparison 310 3.8.2.1.2 Assessment of Design Codes and Load Changes for Concrete 310 Structures 3.8.2.1.2.1 Columns With Spliced Reinforcing 310 Page 17 of 39 Revision 26 5/2016
GINNA/UFSAR 3.8.2.1.2.2 Brackets and Corbels (Not on the Containment Shell) 311 3.8.2.1.2.3 Elements Loaded in Shear With No Diagonal Tension (Shear Friction) 312 3.8.2.1.2.4 Structural Walls - Primary Load Carrying 313 3.8.2.1.2.5 Elements Subject to Temperature Variations 314 3.8.2.1.2.6 Areas of Containment Shell Subject to Peripheral Shear 315 3.8.2.1.2.7 Areas of Containment Shell Subject to Torsion 316 3.8.2.1.2.8 Brackets and Corbels (On the Containment Shell) 316 3.8.2.1.2.9 Areas of Containment Shell Subject to Biaxial Tension 316 3.8.2.1.2.10 Steel Embedments Transmitting Loads to Concrete 317 3.8.2.1.3 Assessment of Design Codes and Load Changes for Steel Structures 317 3.8.2.1.3.1 Shear Connectors in Composite Beams 318 3.8.2.1.3.2 Composite Beams With Steel Deck 318 3.8.2.1.3.3 Hybrid Girders 318 3.8.2.1.3.4 Compression Elements 319 3.8.2.1.3.5 Tension Members 319 3.8.2.1.3.6 Coped Beams 319 3.8.2.1.3.7 Moment Connections 320 3.8.2.1.3.8 Lateral Bracing 320 3.8.2.1.3.9 Steel Embedments 320 3.8.2.1.4 Summary 322 3.8.2.2 Structural Reevaluation of Containment 322 3.8.2.2.1 Introduction 322 3.8.2.2.2 Containment Temperature 323 3.8.2.1.1 Containment Pressure 323 3.8.2.1.2 Seismic Loads 323 3.8.2.1.3 Design and Analysis Procedures 324 3.8.2.1.3.1 Containment Model 324 3.8.2.1.3.2 Seismic and Loss-of-Coolant Accident Loads 324 3.8.2.1.3.3 Pressure, Seismic, and Operating Temperature Loads 325 3.8.2.1.4 Structural Acceptance Criteria 326 3.8.2.1.5 Structural Evaluation of Containment 326 3.8.2.1.5.1 Seismic Analysis 326 3.8.2.1.5.2 Load Combinations 327 3.8.2.1.6 Structural Evaluation of Large Openings 328 3.8.2.1.7 Structural Evaluation of Tension Rods 328 Page 18 of 39 Revision 26 5/2016
GINNA/UFSAR 3.8.2.2 Dome Liner Reevaluation 328 3.8.2.2.1 Dome Liner Studs 328 3.8.2.2.2 Loads 328 3.8.2.2.2.1 Loss-of-Coolant Accident 328 3.8.2.2.2.2 Steam Line Break 329 3.8.2.2.3 Model Definition 329 3.8.2.2.3.1 General Dome Model 329 3.8.2.2.3.2 Insulation Termination Region Model 329 3.8.2.2.4 Analysis 330 3.8.2.2.4.1 Controlling Loads 330 3.8.2.2.4.2 Liner-Stud Interaction 330 3.8.2.2.4.3 Effect of Internal Pressure on Liner Buckling 332 3.8.2.2.5 Results and Conclusions 333 3.8.2.2.5.1 Insulation Termination Region 333 3.8.2.2.5.2 General Dome 334 3.8.2.2.5.3 Effect of Internal Pressure on Liner Buckling and Stud Integrity 335 3.8.2.2.6 Overall Conclusions 337 3.8.3 CONTAINMENT INTERNAL STRUCTURES 337 3.8.3.1 Description of the Internal Structures 337 3.8.3.2 Applicable Codes, Standards, and Specifications 338 3.8.3.3 Loads and Load Combinations 338 3.8.3.3.1 Load Combinations Considered 338 3.8.3.3.2 Applicable Load Combinations 338 3.8.3.4 Design and Analysis Procedures 339 3.8.3.4.1 Original Design 339 3.8.3.4.2 Systematic Evaluation Program Reevaluation 340 3.8.3.5 Method of Analysis 340 3.8.3.6 Structural Acceptance Criteria 341 3.8.3.7 Structural Evaluation 341 3.8.4 OTHER SEISMIC CATEGORY I STRUCTURES 341 3.8.4.1 Description of the Structures 341 3.8.4.1.1 Auxiliary Building 342 3.8.4.1.2 Control Building 342 3.8.4.1.3 Diesel Generator Building 343 3.8.4.1.4 Intermediate Building 343 Page 19 of 39 Revision 26 5/2016
GINNA/UFSAR 3.8.4.1.5 Standby Auxiliary Feedwater Building 344 3.8.4.1.6 Screen House 344 3.8.4.1.7 Turbine Building 345 3.8.4.1.8 Service Building 345 3.8.4.1.9 Interconnected Building Complex 346 3.8.4.1.10 Canister Preparation Building (CPB) 346 3.8.4.2 Applicable Codes, Standards, and Specifications 347 3.8.4.3 Loads and Load Combinations 347 3.8.4.4 Design and Analysis Procedures 347 3.8.4.4.1 Original Design and Analysis Procedures 347 3.8.4.4.2 SEP Reevaluation Design and Analysis Procedures 348 3.8.4.4.2.1 Mathematical Model 348 3.8.4.4.2.2 Method of Analysis 350 3.8.4.4.2.3 Structural Evaluation 351 3.8.4.5 Masonry Walls 352 3.8.4.5.1 Applicable Walls 352 3.8.4.5.2 Loads and Load Combinations 352 3.8.4.5.3 Stress Analysis 354 3.8.4.5.3.1 Computer Program 354 3.8.4.5.3.2 Seismic Analysis 354 3.8.4.5.4 Interstory Drift 355 3.8.4.5.5 Multi-Wythe Walls 355 3.8.4.5.6 Block Pullout 355 3.8.4.5.7 Structural Acceptance Criteria - Allowable Stresses 355 3.8.4.5.7.1 Normal Operating Conditions 355 3.8.4.5.7.2 Safe Shutdown Earthquake 356 3.8.4.5.8 Evaluation Results 356 3.8.4.5.8.1 General 356 3.8.4.5.8.2 Inelastic Analysis 357 3.8.4.5.8.3 Wall Modifications 357 3.8.4.5.9 Materials, Quality Control, and Special Construction 358 Techniques 3.8.5 FOUNDATIONS 359 Page 20 of 39 Revision 26 5/2016
GINNA/UFSAR Table 3.8-1a COMPUTER PROGRAM SAND INPUT FOR CONTAINMENT 364 SEISMIC ANALYSIS - DIMENSIONS AND FORMULA Table 3.8-1b COMPUTER PROGRAM SAND INPUT FOR CONTAINMENT 365 SEISMIC ANALYSIS - DIMENSION CALCULATIONS Table 3.8-1c COMPUTER PROGRAM SAND INPUT FOR CONTAINMENT 366 SEISMIC ANALYSIS - NATURAL FREQUENCIES AND
RESPONSE
Table 3.8-2 MAJOR STRUCTURES FOR WHICH PRESTRESSED ROCK 367 ANCHORS WERE USED Table 3.8-3 PROPERTIES AND TESTS FOR CONTAINMENT ANCHOR AND 369 TENDON CORROSION INHIBITOR Table 3.8-4 ALLOWABLE STRESSES 370 Table 3.8-5a CONTAINMENT STRUCTURE STRESSES - LOADING #1 DEAD 371 LOAD Table 3.8-5b CONTAINMENT STRUCTURE STRESSES - LOADING #2 FINAL 372 PRESTRESS - 636 K/TENDON Table 3.8-5c CONTAINMENT STRUCTURE STRESSES - LOADING #3 374 OPERATING TEMPERATURE - WINTER Table 3.8-5d CONTAINMENT STRUCTURE STRESSES - LOADING #4 376 OPERATING TEMPERATURE - SUMMER Table 3.8-5e CONTAINMENT STRUCTURE STRESSES - LOADING #5 377 INTERNAL PRESSURE Table 3.8-5f CONTAINMENT STRUCTURE STRESSES - LOADING #6 378 ACCIDENT TEMPERATURE - P = 60 PSIG, T = 286F Table 3.8-5g CONTAINMENT STRUCTURE STRESSES - LOADING #7 379 ACCIDENT TEMPERATURE - P = 90 PSIG, T = 312F Table 3.8-5h CONTAINMENT STRUCTURE STRESSES - LOADING #8 0.10G 381 EARTHQUAKE - HORIZONTAL + VERTICAL COMPONENT Table 3.8-6a CONTAINMENT STRUCTURE LOADING COMBINATIONS - 382 LOAD NUMBERS 1 THROUGH 48 Table 3.8-6b CONTAINMENT STRUCTURE LOADING COMBINATIONS - 384 KEY TO SYMBOLS Table 3.8-7 CONCRETE COVER REQUIRED FOR REINFORCING STEEL 385 Table 3.8-8 ELASTOMER PADS PROPERTIES 386 Table 3.8-9 ROCK ANCHOR A - UPLIFT TEST WITH JACKING FRAME, 387 MAY 19, 1966 Table 3.8-10 DESIGN CODE COMPARISON 388 Table 3.8-11 ACI 318-63 VERSUS ACI 349-76 CODE COMPARISONS 390 Table 3.8-12 ACI 301-63 VERSUS ACI 301-72 (REVISED 1975) COMPARISON 392 Table 3.8-13 ACI 318-63 VERSUS ASME B&PV CODE, SECTION III, 393 DIVISION 2, 1980 CODE COMPARISON Table 3.8-14 ASME B&PV CODE, SECTION III, DIVISION 2, 1980 (ACI 359-80) 394 VERSUS ACI 318-63 CODE COMPARISION Page 21 of 39 Revision 26 5/2016
GINNA/UFSAR Table 3.8-15 LIST OF STRUCTURAL ELEMENTS TO BE EXAMINED 395 Table 3.8-16 MASSES, MOMENT OF INERTIA (I), FLEXURAL AREA (A), 397 AND SHEAR AREA (As) FOR THE LLNL MODEL Table 3.8-17 MODAL FREQUENCIES FOR THE LAWRENCE LIVERMORE 398 NATIONAL LABORATORY CONTAINMENT SHELL MODEL Table 3.8-18 RESPONSE VALUES FOR REGULATORY GUIDE 1.60 399 HORIZONTAL (0.17g) AND VERTICAL (0.11g) SPECTRA INPUT Table 3.8-19 PEAK HARMONIC AMPLITUDES OF THE SEISMIC LOAD ON 400 CYLINDER AND DOME OF THE CONTAINMENT SHELL Table 3.8-20 MATERIAL PROPERTIES FOR STEEL, CONCRETE, AND FOAM 401 INSULATION Table 3.8-21 MAXIMUM DISPLACEMENTS OF 5/8-INCH S6L STUDS IN THE 402 INSULATION TERMINATION REGION Table 3.8-22 MAXIMUM DISPLACEMENT OF STUDS IN GENERAL DOME 403 Table 3.8-23 LOAD DEFINITIONS 404 3.9 MECHANICAL SYSTEMS AND COMPONENTS 406 3.9.1 SPECIAL TOPICS FOR MECHANICAL COMPONENTS 406 3.9.1.1 Design Transients 406 3.9.1.1.1 Load Combinations 406 3.9.1.1.2 Cyclic Loads 406 3.9.1.1.2.1 Thermal and Pressure Cyclic Loads 406 3.9.1.1.2.2 Pressurizer Surge Line 406 3.9.1.1.2.3 Unisolable Connections to the Reactor Coolant System 407 3.9.1.1.3 Transient Hydraulic Loads 408 3.9.1.1.4 Operating-Basis Earthquake 408 3.9.1.1.5 Safe Shutdown Earthquake 408 3.9.1.1.6 Secondary System Fluid Flow Instability (Water Hammer) 408 3.9.1.1.7 Loss-of-Coolant Accident 408 3.9.1.2 Computer Programs Used in Analysis 409 3.9.1.3 Experimental Stress Analysis 410 3.9.1.3.1 Plastic Model Analysis 410 3.9.1.3.2 Plastic Model Details 410 3.9.2 DYNAMIC TESTING AND ANALYSIS 412 3.9.2.1 Piping Systems 412 3.9.2.1.1 General 412 3.9.2.1.2 Seismic Category I Piping, 2-1/2 Inch Nominal Size and Larger 413 3.9.2.1.2.1 Static Analysis 413 3.9.2.1.2.2 Dynamic Analysis 413 Page 22 of 39 Revision 26 5/2016
GINNA/UFSAR 3.9.2.1.2.3 Residual Heat Removal System Line From Reactor Coolant System 414 Loop A to Containment 3.9.2.1.2.4 Steam Line From Steam Generator B to Containment 415 3.9.2.1.2.5 Charging Line 416 3.9.2.1.3 Seismic Category I Piping, 2-Inch Nominal Size and Under, Original 416 Design 3.9.2.1.4 Pressurizer Safety and Relief Valve Discharge Piping 416 3.9.2.1.4.1 1972 Analysis 416 3.9.2.1.4.2 NUREG 0737, Item II.D.1 Analysis 417 3.9.2.1.5 Main Steam Header Dynamic Load Factor Analysis 418 3.9.2.1.5.1 Extended Power Uprate Considerations 419 3.9.2.1.6 Secondary System Water Hammer 419 3.9.2.1.6.1 Analysis 419 3.9.2.1.6.2 Evaluation Results 420 3.9.2.1.6.3 Corrective Actions 420 3.9.2.1.6.4 Extended Power Uprate Considerations 421 3.9.2.1.7 Velan Swing Check Valves 421 3.9.2.1.8 Seismic Piping Upgrade Program 421 3.9.2.2 Safety-Related Mechanical Equipment 422 3.9.2.2.1 Original Seismic Input and Behavior Criteria 422 3.9.2.2.2 Current Seismic Input 423 3.9.2.2.3 Systematic Evaluation Program 423 3.9.2.2.4 Systematic Evaluation Program Reevaluation of Selected Mechanical 424 Components for Design Adequacy 3.9.2.2.4.1 Essential Service Water (SW) Pumps 424 3.9.2.2.4.2 Component Cooling Heat Exchanger 425 3.9.2.2.4.3 Component Cooling Surge Tank 425 3.9.2.2.4.4 Diesel-Generator Air Tanks 425 3.9.2.2.4.5 Boric Acid Storage Tank 426 3.9.2.2.4.6 Refueling Water Storage Tank (RWST) 426 3.9.2.2.4.7 Motor-Operated Valves 427 3.9.2.2.4.8 Steam Generators 427 3.9.2.2.4.9 Reactor Coolant Pumps 428 3.9.2.2.4.10 Pressurizer 428 3.9.2.2.4.11 Control Rod Drive Mechanism 429 3.9.2.3 Dynamic Response Analysis of Reactor Internals Under Operational 429 Flow Transients and Steady-State Conditions Page 23 of 39 Revision 26 5/2016
GINNA/UFSAR 3.9.2.3.1 Design Criteria 430 3.9.2.3.1.1 General 430 3.9.2.3.1.2 Critical Internals 430 3.9.2.3.1.3 Allowable Stress Criteria 431 3.9.2.3.2 Blowdown and Force Analysis 431 3.9.2.3.2.1 Computer Program 431 3.9.2.3.2.2 Blowdown Model 432 3.9.2.3.2.3 LATFORC MODEL 433 3.9.2.3.2.4 FORCE2 MODEL 433 3.9.2.3.3 Fuel Assembly Thimbles 434 3.9.2.3.4 Dynamic System Analysis of Reactor Internals Under Loss-of-Coolant 434 Accident (LOCA) 3.9.2.3.4.1 Mathematical Model of the Reactor Pressure Vessel (RPV) System 434 3.9.2.3.4.2 Analytical Methods 436 3.9.2.3.4.3 RPV Internal Hydraulic Loads 436 3.9.2.3.4.4 Reactor Coolant Loop Mechanical Loads 438 3.9.2.3.4.5 Results of the Analysis 438 3.9.2.3.5 Transverse Guide Tube Excitation by Blowdown Forces 438 3.9.2.3.5.1 General 438 3.9.2.3.6.1 Reactor Pressure Vessel System Thermal-Hydraulic Analysis 440 3.9.2.3.6.2 Bypass Flow Analysis 440 3.9.2.3.6.3 Thermal Analysis of the Baffle/Barrel Region 441 3.9.2.3.6.4 Pressure Drop Across the Baffle Plate Analyses 441 3.9.2.3.6.5 Flow Induced Vibration 441 3.9.2.3.6.6 Reactor Internals Structural Integrity 441 3.9.2.3.6.7 Control Rod Performance 441 3.9.2.3.6.8 Vessel/Internals/Fuel/Control Rod Response During Loca Conditions 442 3.9.2.3.6.9 Summary of Conclusions 442 3.9.2.4 Asymmetric Loss-of-Coolant Accident Loading Analysis 442 3.9.2.5 Seismic Evaluation of Reactor Vessel Internals 442 3.9.2.5.1 Analysis Procedure 442 3.9.2.5.2 Analysis Results 443 3.9.3 COMPONENT SUPPORTS AND CORE SUPPORT STRUCTURES 444 3.9.3.1 Loading Combinations, Design Transients, and Stress Limits 444 3.9.3.2 Component Supports 444 3.9.3.2.1 Reactor Vessel 444 Page 24 of 39 Revision 26 5/2016
GINNA/UFSAR 3.9.3.2.2 Steam Generators 445 3.9.3.2.3 Reactor Coolant Pumps 445 3.9.3.2.4 Pressurizer 446 3.9.3.2.5 Reactor Coolant Piping 446 3.9.3.3 Pipe Supports 446 3.9.3.3.1 Original Analysis 446 3.9.3.3.2 IE Bulletin Reanalysis 446 3.9.3.3.3 Seismic Piping Upgrade Program 447 3.9.3.3.3.1 Applicable Supports 447 3.9.3.3.3.2 Load Combinations and Stress Limits 447 3.9.3.3.3.3 Structural Requirements 447 3.9.3.3.4 Base Plate Flexibility 449 3.9.3.3.5 Snubbers 449 3.9.3.3.5.1 Design Loads 449 3.9.3.3.5.2 Surveillance Program 450 3.9.4 CONTROL ROD DRIVE SYSTEMS 450 3.9.4.1 Description 450 3.9.4.1.1 General 450 3.9.4.1.2 Latch Assembly 451 3.9.4.1.3 Pressure Vessel 452 3.9.4.1.4 Operating Coil Stack 452 3.9.4.1.5 Drive Shaft Assembly 452 3.9.4.1.6 Position Indicator Coil Stack 452 3.9.4.2 Design Loads, Stress Limits, and Allowable Deformation 452 3.9.4.3 Control Rod Drive Mechanism Housing Mechanical Failure Evaluation 453 3.9.4.3.1 Housing Description 453 3.9.4.3.2 Effects of Rod Travel Housing Longitudinal Failures 453 3.9.4.3.3 Effect of Rod Travel Housing Circumferential Failures 453 3.9.4.3.4 Summary 454 3.9.5 REACTOR PRESSURE VESSEL INTERNALS 454 3.9.5.1 Design Arrangements 454 3.9.5.1.1 Lower Core Support Structure 454 3.9.5.1.1.1 Support Structure Assembly 454 3.9.5.1.1.2 Lower Core Plate 454 3.9.5.1.1.3 Thermal Shield 455 3.9.5.1.1.4 Coolant Flow Passages 456 Page 25 of 39 Revision 26 5/2016
GINNA/UFSAR 3.9.5.1.1.5 Support and Alignment Arrangements 456 3.9.5.1.2 Upper Core Support Assembly 456 3.9.5.1.3 In-Core Instrumentation Support Structures 457 3.9.5.2 Loading Conditions 458 3.9.5.3 Design Bases 458 3.9.6 INSERVICE INSPECTION OF PUMPS AND VALVES 459 3.9.6.1 General 459 3.9.6.2 Inservice Testing of Pumps 459 3.9.6.3 Inservice Testing of Valves 460 3.9.7 Extended Power Uprate (EPU) 460 Table 3.9-1 ORIGINAL DESIGN LOADING COMBINATIONS AND STRESS 464 LIMITS Table 3.9-2 RESIDUAL HEAT REMOVAL LOOP A STRESS
SUMMARY
465 Table 3.9-3 MAIN STEAM LINE-LOOP B STRESS
SUMMARY
466 Table 3.9-4 CHARGING LINE STRESS
SUMMARY
467 Table 3.9-5 LOAD COMBINATIONS AND ACCEPTANCE CRITERIA FOR 468 PRESSURIZER SAFETY AND RELIEF VALVE PIPING AND SUPPORTS - UPSTREAM OF VALVES Table 3.9-6 LOAD COMBINATIONS AND ACCEPTANCE CRITERIA FOR 469 PRESSURIZER SAFETY AND RELIEF VALVE PIPING AND SUPPORTS - SEISMICALLY DESIGNED DOWNSTREAM PORTION Table 3.9-7 DEFINITIONS OF LOAD ABBREVIATIONS 470 Table 3.9-8 LOADING COMBINATIONS AND STRESS LIMITS FOR PIPING 471 FOR SEISMIC UPGRADE PROGRAMS Table 3.9-9 ALLOWABLE STEAM GENERATOR NOZZLE LOADS 472 Table 3.9-10 REACTOR COOLANT PUMP AUXILIARY NOZZLE UMBRELLA 473 LOADS Table 3.9-11 SYSTEMATIC EVALUATION PROGRAM STRUCTURAL 476 BEHAVIOR CRITERIA FOR DETERMINING SEISMIC DESIGN ADEQUACY Table 3.9-12 MECHANICAL COMPONENTS SELECTED FOR SEP SEISMIC 477 REVIEW Table 3.9-13 MAXIMUM STRESS HOT-LEG BREAK (ORIGINAL ANALYSIS) 478 Table 3.9-14 MAXIMUM STRESS COLD-LEG BREAK (ORIGINAL 479 ANALYSIS)
Table 3.9-15 MAXIMUM CORE BARREL STRESS AND DEFLECTION UNDER 480 HOT-LEG BLOWDOWN (ORIGINAL ANALYSIS)
Table 3.9-16a MAXIMUM STRESS INTENSITIES AND DEFLECTION COLD- 481 LEG BLOWDOWN (ORIGINAL ANALYSIS) - IN THE UPPER BARREL Page 26 of 39 Revision 26 5/2016
GINNA/UFSAR Table 3.9-16b MAXIMUM STRESS INTENSITIES AND DEFLECTION COLD- 482 LEG BLOWDOWN (ORIGINAL ANALYSIS) - AT THE UPPER BARREL ENDS Table 3.9-17 CORE BARREL STRESSES (ORIGINAL ANALYSIS) 483 Table 3.9-18 CORE BARREL STRESSES (ORIGINAL ANALYSIS) 484 Table 3.9-19 CORE BARREL STRESSES (ORIGINAL ANALYSIS) 485 Table 3.9-20 CORE BARREL STRESSES (ORIGINAL ANALYSIS) 486 Table 3.9-21 CORE BARREL STRESSES (ORIGINAL ANALYSIS) 487 Table 3.9-22 CORE BARREL STRESSES (ORIGINAL ANALYSIS) 489 Table 3.9-23a LOAD COMBINATIONS AND ALLOWABLE STRESS LIMITS 490 FOR PRIMARY EQUIPMENT SUPPORTS EVALUATION - FOR PLANT EVENTS Table 3.9-23b LOAD COMBINATIONS AND ALLOWABLE STRESS LIMITS 491 FOR PRIMARY EQUIPMENT SUPPORTS EVALUATION -
DEFINITION OF LOADING CONDITIONS FOR PRIMARY EQUIPMENT SUPPORTS EVALUATION IN TABLE 3.9-23a Table 3.9-24 RESIDUAL HEAT REMOVAL LOOP A SUPPORT LOADS1 492 CALCULATED FOR IE BULLETIN 79-07 Table 3.9-25a MAIN STEAM LINE LOOP B SUPPORT LOADS2 CALCULATED 495 FOR IE BULLETIN 79 SEISMIC SUPPORT Table 3.9-25b MAIN STEAM LINE LOOP B NOZZLE LOADS CALCULATED 496 FOR IE BULLETIN 79 NOZZLE LOADS Table 3.9-26 CHARGING LINE SUPPORT LOADSa CALCULATED FOR IE 497 BULLETIN 79-07 Table 3.9-27 LOADING COMBINATIONS AND STRESS LIMITS FOR 502 SUPPORTS ON PIPING SYSTEMS Table 3.9-28 ANALYSIS OF TYPICAL PIPE SUPPORT BASE PLATES 503 CALCULATED FOR IE BULLETIN 79-02 Table 3.9-29 INTERNALS DEFLECTIONS UNDER ABNORMAL OPERATION 504 3.10 SEISMIC QUALIFICATION OF SEISMIC CATEGORY I INSTRU- 505 MENTATION AND ELECTRICAL EQUIPMENT 3.10.1 SEISMIC QUALIFICATION CRITERIA 505 3.10.1.1 Original Criteria 505 3.10.1.2 Current Criteria 505 3.10.2 SEISMIC QUALIFICATION OF ELECTRICAL EQUIPMENT AND 506 INSTRUMENTATION 3.10.2.1 Introduction 506 3.10.2.2 Battery Racks 507 3.10.2.3 Motor Control Centers 1L and 1M 507 3.10.2.4 Switchgear 508 Page 27 of 39 Revision 26 5/2016
GINNA/UFSAR 3.10.2.5 Control Room Electrical Panels 508 3.10.2.6 Electrical Cable Raceways 509 3.10.2.7 Constant Voltage Transformers 509 3.10.3 SEISMIC QUALIFICATION OF SUPPORTS OF ELECTRICAL 509 EQUIPMENT AND INSTRUMENTATION 3.10.3.1 Equipment Addressed 510 3.10.3.2 Raceway Anchorages 510 3.10.3.2.1 Test Program 510 3.10.3.2.2 Test Loads 511 3.10.3.2.3 Expansion Anchor Test Results 512 3.10.3.2.4 Frictional Anchor Test Results 512 3.10.3.2.5 Embedded Anchor Test Results 513 3.10.3.3 Class 1E Equipment Anchorage Qualification Program 513 3.10.3.4 Conclusions 514 3.10.4 FUNCTIONAL CAPABILITY OF COMPONENTS 514 3.10.5 SEISMIC CATEGORY I TUBING 514 3.10.5.1 Codes and Standards 514 3.10.5.1.1 Tubing Design Requirements 515 3.10.5.1.2 Tubing Supports Design Requirements 515 3.10.5.2 Load Conditions 516 3.10.5.2.1 Tubing 516 3.10.5.2.2 Tubing Supports 516 3.10.5.3 Routing Requirements 517 Table 3.10-1 MAJOR CLASS 1E COMPONENTS AND THE BASIS FOR 520 SEISMIC QUALIFICATION Table 3.10-2 ELECTRICAL COMPONENTS SELECTED FOR SEISMIC 522 REVIEW Table 3.10-3 SHELL ANCHOR TEST
SUMMARY
523 Table 3.10-4 FRICTION BOLT TEST RESULT
SUMMARY
524 Table 3.10-5 CATEGORY 3 ANCHORS TEST
SUMMARY
525 Table 3.10-6 STRESS LIMITS FOR TUBING 526 3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRI- 527 CAL EQUIPMENT 3.
11.1 BACKGROUND
527 3.11.1.1 Initial Design Considerations 527 Page 28 of 39 Revision 26 5/2016
GINNA/UFSAR 3.11.1.2 Review of Environmental Qualification of Safety-Related Electrical 527 Equipment 3.11.2 Equipment Identification 528 3.11.3 IDENTIFICATION OF LIMITING ENVIRONMENTAL CONDI- 528 TIONS 3.11.3.1 Inside Containment 528 3.11.3.1.1 Post Loss-of-Coolant Accident Environment 528 3.11.3.1.2 Post Main Steam Line Break Environment 530 3.11.3.2 Auxiliary Building 530 3.11.3.2.1 Heating, Ventilation, and Air Conditioning 530 3.11.3.2.2 Loss of Ventilation 531 3.11.3.2.3 Radiation Levels 532 3.11.3.2.4 Flooding 532 3.11.3.3 Intermediate Building 532 3.11.3.4 Cable Tunnel 533 3.11.3.5 Control Building 533 3.11.3.6 Diesel Generator Rooms 534 3.11.3.7 Turbine Building 534 3.11.3.8 Auxiliary Building Annex 535 3.11.3.9 Screen House 535 3.11.4 EQUIPMENT QUALIFICATION INFORMATION 535 3.11.5 ENVIRONMENTAL QUALIFICATION PROGRAM 535 Table 3.11-1 ENVIRONMENTAL SERVICE CONDITIONS FOR EQUIPMENT 540 DESIGNED TO MITIGATE DESIGN-BASIS EVENTS Table 3.11-2 ESTIMATES FOR TOTAL AIRBORNE GAMMA DOSE 549 CONTRIBUTORS IN CONTAINMENT TO A POINT IN THE CONTAINMENT CENTER - GINNA STATION Table 3.11-3 ESTIMATES FOR TOTAL AIRBORNE BETA DOSE 551 CONTRIBUTORS IN CONTAINMENT TO A POINT IN THE CONTAINMENT CENTER - GINNA STATION Table 3.11-4 ESTIMATES FOR TOTAL AIRBORNE GAMMA DOSE 553 CONTRIBUTORS IN CONTAINMENT TO A POINT IN THE CONTAINMENT CENTER, REGULATORY GUIDE 1.89, REVISION 1 Table 3.11-5 ESTIMATES FOR TOTAL AIRBORNE BETA DOSE 555 CONTRIBUTORS IN CONTAINMENT TO A POINT IN THE CONTAINMENT CENTER, REGULATORY GUIDE 1.89, REVISION 1 Table 3.11-6 GINNA STATION/REGULATORY GUIDE 1.89, APPENDIX D, 557 COMPARISON OF POSTACCIDENT RADIATION ENVIRONMENT ASSUMPTIONS Page 29 of 39 Revision 26 5/2016
GINNA/UFSAR FIGURES Figure 3.7-1 Seismic Response Spectra, 8%g Housner Model Figure 3.7-2 Seismic Response Spectra, 20%g Housner Model Figure 3.7-3 NRC Systematic Evaluation Program Site Specific Spectrum, Ginna Site (5% Damping)
Figure 3.7-4 Comparison of the Housner Response Spectrum for 2% of Critical Damping with the 7% Regulatory Guide 1.60 Spectrum Figure 3.7-5 In-Structure Response Spectra for Interconnected Building, Half-Area and Full-Area Models Figure 3.7-6 Containment Building and Complex of Interconnected Seismic Cate- gory I and Nonseismic Structures, Plan View Figure 3.7-7 Horizontal Response Spectra - SEP Systematic Evaluation Program Figure 3.7-8 Steam Generator Mathematical Model Figure 3.7-9 Mathematical Model of Reactor Vessel Figure 3.7-10 Seismic Average Acceleration Spectrum Design Earthquake, 1% Damping Figure 3.7-11 Locations Where In-Structure Response Spectra Were Generated in Interconnected Building Complex Figure 3.7-12 SEP Response Spectra for Pressurizer PR-1 (Containment Building Elevation 253 ft) for 3%, 5%, and 7% Damping Figure 3.7-13 SEP Response Spectra for Control Rod Drive (Containment Building Elevation 253 ft) for 3%, 5%, 7% Damping Figure 3.7-14 SEP Response Spectra for Control Rod Drive (Containment Building Elevation 278 ft) for 3%, 5%, and 7% Damping Figure 3.7-15 SEP Response Spectra for Steam Generator SG-1A (Containment Building Elevation 250 ft) for 3%, 5%, and 7% Damping Figure 3.7-16 SEP Response Spectra for Steam Generator SG-1A (Containment Building Elevation 278 ft) for 3%, 5%, and 7% Damping Figure 3.7-17 SEP Response Spectra for Steam Generator SG-1B (Containment Building Elevation 250 ft) for 3%, 5%, and 7% Damping Figure 3.7-18 SEP Response Spectra for Steam Generator SG-1B (Containment Building Elevation 278 ft) for 3%, 5%, and 7% Damping Figure 3.7-19 SEP Response Spectra for Reactor Coolant Pump Rp-1A (Containment Building Elevation 247 ft) for 3%, 5%, and 7% Damping Figure 3.7-20 SEP Response Spectra for Reactor Coolant Pump RP-1B (Containment Building Elevation 247 ft) for 3%, 5%, and 7% Damping Figure 3.7-21 SEP Equipment Response Spectra for 3%, 5%, and 7% Damping at Auxiliary Building Platform (Elevation 281 ft 6 in)
Figure 3.7-22 SEP Equipment Response Spectra for 3%, 5%, and 7% Damping at Auxiliary Building Heat Exchanger 35 (Elevation 281 ft 6 in)
Page 30 of 39 Revision 26 5/2016
GINNA/UFSAR Figure 3.7-23 SEP Equipment Response Spectra for 3%, 5%, and 7% Damping at Auxiliary Building Surge Tank 34 Figure 3.7-24 SEP Equipment Response Spectra for 3%, 5%, and 7% Damping at Auxiliary Building Boric Acid Storage Tank 34 Figure 3.7-25 SEP Equipment Response Spectra for 3%, 5%, and 7% Damping at Auxiliary Building Operating Floor (Elevation 271 ft 6 in)
Figure 3.7-26 SEP Equipment Response Spectra for 3%, 5%, and 7% Damping at Control Building Basement Floor (Elevation 250 ft 0 in)
Figure 3.7-27 SEP Equipment Response Spectra for 3%, 5%, and 7% Damping at Control Building Relay Room Floor (Elevation 269 ft 9 in)
Figure 3.7-28 SEP Equipment Response Spectra for 3%, 5%, and 7% Damping at Control Room Floor (Elevation 289 ft 9 in)
Figure 3.7-29 Residual Heat Removal Line Inside Containment Figure 3.7-30 Lumped Mass Model - Steam Line B Figure 3.7-31 Structural Model, Pressurizer Safety and Relief Line Figure 3.8-1 Containment Cross Section and Details Figure 3.8-2 Containment Mat Foundation and Ring Girder Figure 3.8-3 Containment Mat Foundation, Reinforcement and Details Figure 3.8-4 Containment Wall Reinforcement and Details Figure 3.8-5 Containment Dome Reinforcement and Details Figure 3.8-6 Containment Miscellaneous Embedded Back-Up Steel Figure 3.8-7 Tendon Vent Cans and Grease Fill Connections Figure 3.8-8 Temperature Gradients - Operating Conditions Figure 3.8-9 Earthquake Meridional Forces Figure 3.8-10 Containment Dynamic Analysis Model Figure 3.8-11 Ginna Containment Mode Shapes Figure 3.8-12 Ginna Containment - Earthquake Response Figure 3.8-13 Moments, Shears, Deflection, Tensile Force, and Hoop Tension Dia-grams Load Combination A Figure 3.8-14 Moments, Shears, Deflection, Tensile Force, and Hoop Tension Dia-grams Load Combination B Figure 3.8-15 Moments, Shears, Deflection, Tensile Force, and Hoop Tension Dia-grams Load Combination C Figure 3.8-16 Tendon to Rock Coupling Figure 3.8-17 Containment - Top Tendon Access Figure 3.8-18 Containment Miscellaneous Steel Tendon Conduit - Hinge Detail Figure 3.8-19 Liner Knuckle Dimensions Figure 3.8-20 Containment Base to Cylinder Model Page 31 of 39 Revision 26 5/2016
GINNA/UFSAR Figure 3.8-21 Containment Dome to Cylinder Discontinuity Model Figure 3.8-22 Cracked Wall Shear Modulus Analysis Figure 3.8-23 Liner Shear Stress Analysis Figure 3.8-24 Windgirder, Shear Channels, and Shear Studs Figure 3.8-25 Cylinder Liner Plate Support Model Figure 3.8-26 Containment Penetration Details Figure 3.8-27 Containment Penetration Details (Typical)
Figure 3.8-28 Composite Drawing Electrical Penetration Figure 3.8-29 Containment Penetrations Section and Details Figure 3.8-30 Containment Equipment Hatch Figure 3.8-31 Containment Personnel Hatch Figure 3.8-32 Containment - Fuel Transfer Tube Penetration Figure 3.8-33 Containment Penetrations Arrangements and Location Figure 3.8-34 Test Coupon - Containment Concrete Shell Figure 3.8-35 Cadweld Splice Test Results Figure 3.8-36 Quality Control Chart for 5000 PSI Concrete Figure 3.8-37 Neoprene Base Hinge Load Deformation Specimen 1 Figure 3.8-38 Neoprene Base Hinge Load Deformation Specimen 2 Figure 3.8-39 Rock Anchor Test A-1 Figure 3.8-40 Containment - Rock Anchor A Test Figure 3.8-41 Containment - Rock Anchor B Test Figure 3.8-42 Containment - Rock Anchor C Test Figure 3.8-43 Accident Temperature Transient Inside the Containment Used for Liner Analysis Figure 3.8-44 Accident Pressure Transient Inside the Containment Used for Liner Analysis Figure 3.8-45 Plan View of the Facade Structure and Containment Figure 3.8-46 Accident Temperature Gradient Through the Uninsulated Containment Shell After 94 Seconds Figure 3.8-47 Accident Temperature Gradient Through the Uninsulated Containment Shell After 380 Seconds Figure 3.8-48 Ginna Containment Structure Figure 3.8-49 Liner Stud Interaction Models Figure 3.8-50 Accident Temperature Distribution in the Steel Liner Figure 3.8-51 Force Displacement Curve for 3/4 in. Headed Studs Figure 3.8-52 Force Displacement Curve for 5/8 in. S6L Studs Page 32 of 39 Revision 26 5/2016
GINNA/UFSAR Figure 3.8-53 Strut Buckling Under P and Delta T Figure 3.8-54 Pressure Effect on Liner Buckling Comparison With LOCA Figure 3.8-55 Reactor Containment Internal Structures Figure 3.8-56 Containment Interior Structures Model for STARDYNE Figure 3.8-57 Schematic Plan View of Major Ginna Structures Figure 3.8-58 Three-Dimensional View of Interconnected Building Complex Figure 3.8-59 Flow Chart of the Analysis of the Interconnected Building Complex Figure 3.8-60 Masonry Wall Reevaluation, Wall Location Plan, Lower Levels Figure 3.8-61 Masonry Wall Reevaluation, Wall Location Plan, Intermediate Levels Figure 3.8-62 Masonry Wall Reevaluation, Wall Location Plan, Operating Levels Figure 3.9-1 Steam-Generator Water Hammer Preliminary Forcing Function Figure 3.9-2 Plastic Model of Reactor Coolant System - Plan View Figure 3.9-3 Lumped Mass Dynamic Model of PCV 434 Figure 3.9-4 Lumped Mass Dynamic Model of PCV 435 Figure 3.9-5 Comparison of WHAM Results With LOFT Semi-Scale Blowdown Experiments, Test No. 519 Figure 3.9-6 Comparison of WHAM Results With LOFT Semi-Scale Blowdown Experiments, Test No. 560 Figure 3.9-6a Steam Generator Upper Support Systems Figure 3.9-7 Control Rod Drive Mechanism Assembly Figure 3.9-8 Control Rod Drive Mechanism Schematic Figure 3.9-9 Reactor Vessel Internals Figure 3.9-10 Detailed View of Reactor Vessel Internals Figure 3.10-1 Q-Deck Detail Figure 3.10-2 Unistrut Detail Figure 3.10-3 Threaded Insert Detail Poured in Place Anchor Figure 3.10-4 Tray Support Types for Friction Bolt Testing Figure 3.11-1 Containment Volume and Reactor Power LOCA Dose Corrections Page 33 of 39 Revision 26 5/2016
GINNA/UFSAR Appendix 3A INITIAL EVALUATION OF CAPABILITY TO WITHSTAND TOR- 558 NADOES 3A.1 INTRODUCTION AND CONCLUSIONS 559 3A.2 IDENTIFICATION OF CRITICAL SYSTEMS AND STRUCTURES 561 3A.3 TORNADO EFFECTS ON STRUCTURES 562 3A.3.1 GENERAL 562 3A.3.2 REACTOR CONTAINMENT 562 3A.3.3 AUXILIARY BUILDING 562 3A.3.4 INTERMEDIATE BUILDING 563 3A.3.5 DIESEL-GENERATOR ANNEX 563 3A.3.6 SCREEN HOUSE 563 3A.3.7 CONTROL ROOM 564 3A.3.8 SERVICE BUILDING 564 3A.3.9 CABLE TUNNELS 564 3A.4 TORNADO EFFECTS ON THE SYSTEMS REQUIRED FOR HOT 565 SHUTDOWN 3A.4.1 DECAY HEAT REMOVAL 565 3A.4.1.1 Steam Relief System 565 3A.4.1.2 Auxiliary Feedwater System 565 3A.4.1.3 Service Water System 566 3A.4.2 REACTIVITY CONTROL 567 3A.4.2.1 Boration System 567 3A.4.2.2 Boration Using Refueling Water 567 3A.4.3 CONTAINMENT VENTILATION SYSTEM 568 3A.4.4 EMERGENCY POWER SUPPLY SYSTEM 569 3A.4.5 CONTROL SYSTEM 569 3A.4.5.1 Control Room 569 3A.4.5.2 Systems of Batteries 569 3A.4.5.3 Steam-Generator Level and Pressure Indicators, Pressurizer Pressure 569 and Level Control 3A.5 TORNADO EFFECT ON SPENT FUEL POOL 571 Appendix 3A Figures Figure 1 Boration System Figure 2 Site Plot Plan Page 34 of 39 Revision 26 5/2016
GINNA/UFSAR Figure 3 Diesel Generator Annex - Elevation 253 ft 6 in.
Figure 4 Screen House Layout Figure 5 Steam Relief Valves Figure 6 Auxiliary Feedwater Pumps Figure 7 Component Cooling System Figure 8 Spent Fuel Storage Pool, Plan View Figure 9 Spent Fuel Storage Pool, Section View Appendix 3B DESIGN OF LARGE OPENING REINFORCEMENTS FOR CON- 572 TAINMENT VESSEL Table of Contents 573 Summary 576 I. Design Bases 576 II. GENERAL DESCRIPTION 576 III. STRESS DISTRIBUTION AROUND A CIRCULAR HOLE IN A 576 CIRCULAR CYLINDRICAL SHELL IV. ANALYSIS OF STRESSES AROUND LARGE OPENINGS 576 V. VERIFICATION OF REINFORCEMENT ADEQUACY 577
- 1. DESIGN BASES 579 1.1 General 579 1.2 Design Loads 579 1.3 Load Combinations 579 1.4 Material Stress/Strain Criteria 580 1.5 Test Condition 582 1.6 Operating Condition 582
- 2. GENERAL DESCRIPTION OF OPENING REINFORCEMENT 583 2.1 Introduction 583 2.2 Rebar for Discontinuity Stresses 583 2.3 Normal Shear at Edge of Opening 583 2.4 Prestressing 583
- 3. STRESS DISTRIBUTION AROUND A CIRCULAR HOLE IN A 584 CIRCULAR CYLINDRICAL SHELL 3.1 Introduction 584 3.2 Finite Element Method 585 3.3 Applications of Three-Dimensional Photoelasticity 586
- 4. ANALYSIS OF THE STRESSES AROUND LARGE OPENINGS IN 588 THE R. E. GINNA SECONDARY CONTAINMENT VESSEL Page 35 of 39 Revision 26 5/2016
GINNA/UFSAR 4.1 Verification of Finite-element Method of Analysis 588 4.2 General Considerations Concerning Methods of Analysis of Reinforced 589 Concrete Structures in the Cracked Condition 4.3 Stress Analysis in Cracked and Uncracked Conditions Under Operating 590 and Accident Loads 4.3.2 Basic Loading Conditions 592 4.3.3 Effect of Concrete Cracking 595 4.3.4 Effect of Creep and Shrinkage 597
- 5. Verification of Design Criteria 598 5.1 Basis For Verification of Shell Loading Capacity Due to Primary Loads 598 (Principal Stress-resultants and Principal Stress-couples) 5.2 Interaction Diagram 599 5.3 Reinforcing Steel 600 5.4 Maximum Liner Stresses 600 5.5 Penetration Barrel 600 5.6 Normal Shear 601 5.7 Rebar Anchorage 602 5.8 Tendon Losses 603 5.9 Summary of Design and Conclusions 604 Table 4-1 Load Combinations 608 Table 4-2 Stress Around Equipment Hatch-Loading (Uncracked Shell) 609 Table 4-3 Stress Around Equipment Hatch-Loading (Cracked Shell) 611 Table 5-1 Maximum Liner Stresses Stress tangent to the edge in Ksi 619 Appendix A to EFFECT OF CONCRETE CREEP AND THE SUSTAINED OPER- 620 APPENDIX 3B ATING STRESSES ON STRESS DISTRIBUTION AROUND OPEN-INGS IN A RAPIDLY PRESSURIZED REINFORCED CONCRETE VESSEL 3B.A EFFECT OF CONCRETE CREEP AND THE SUSTAINED OPER- 621 ATING STRESSES ON STRESS DISTRIBUTION AROUND OPEN-INGS IN A RAPIDLY PRESSURIZED REINFORCED CONCRETE VESSEL Appendix B TO EARTHQUAKE ANALYSIS 628 APPENDIX 3B 3B.B Earthquake Analysis 629 ADDENDUM TO ADDENDUM TO THE REPORT ON: DESIGN OF LARGE OPEN- 630 APPENDIX 3B ING REINFORCEMENTS FOR CONTAINMENT VESSEL 3B.C Introduction 631 1 Design 632 1.1 Concrete Shear 632 1.2 Interaction Diagrams 632 Page 36 of 39 Revision 26 5/2016
GINNA/UFSAR 1.3 Earthquake Design 632 1.4 Thermal Gradients 632 1.5 Penetration Material 633 1.6 Working Strength Design 633 1.7 Anchorage Plate Bearing Stress 633 1.8 Insulated Liner Temperature Increase 633 1.9 High Strength Rebar 633 1.10 Proof Test Instrumentation 633 1.11 Operating Conditions 634 1.12 Shear - Diagonal Tension 634 1.13 Normal Shears 635 1.14 Radial Shear at the Periphery of the Opening 635 1.15 Accident Temperature Effects 635 1.16 Analytical Model for Different Load Combinations 635 1.17 Shear Reinforcement 635 1.18 Equation (5.11) 636 1.19 Rebar Located Away from the Barrel 636 1.20 Verification of Analysis 637 1.21 Test Problem 638 1.22 Accident Temperature 638 2 Construction 639 2.1 Construction Schedule 639 2.2 Concrete Removal 639 2.3 Concrete Work 639 2.4 Retensioning Tendons 640 2.5 Rebar Splices 640 2.6 Tendon Conduit 640 Table I STRESS AROUND EQUIPMENT HATCH LOADING CONDITION 641 NO. 4 - Accident Temperature Appendix 3B Figures Figures Appendix 3B Figures Figure 1 Figure 2 Figure 3 Stress Distribution Around Openings in Cylindrical Shells Figure 4 Grid for Finite Element Analysis of the Stresses Around Openings Page 37 of 39 Revision 26 5/2016
GINNA/UFSAR Figure 5 Membrane Stress Around Opening Edge (Vessel Subject to Internal Pressure)
Figure 6 Surface Stresses Around Opening Edge (Vessel Subject to Internal Pressure)
Figure 7 Hoop Stresses Along Longitudinal Axis (Vessel Subject to Internal Pressure)
Figure 8 Axial Stresses Along Transverse Axis (vessel Subject to Inernal Pres-sure)
Figure 9 Hoop Stress-Resultant No Along Symmetry Axes (Test Problem)
Figure 10 Layer Thickness And Destination Figure 11 Nodal Forces Due to Curvature of Tendons in the Neighborhood of Opening Figure 12 Stress Distribution Around Openings (Thermal Gradient Near Equip-ment Opening)
Figure 13 Steady State Temperature Distributions - Winter Gradient Figure 14 Stress Distribution Around Openings (Effect of Bond Failure Along Terminated Rebars)
Figure 15 Hoop Stress-Resultant Along Horizontal And Vertical Symmetry Axes (Internal Pressure = 69 PSI)
Figure 16 Shell Displacements (Final Vertical Prestress)
Figure 17 Shell Displacements (69 PSI Internal Pressure)
Figure 18 Interaction Diagram for Axial Compression/Tension and Bending Figure 19 Interaction Diagram Ring Steel Direction Elements No. 73 & 74 Figure 20 Interaction Diagram Elements No. 97, 100, & 101 Figure 21 Interaction Diagram Elements No. 97, 100, & 101 Figure 22 Interaction Diagram Elements No. 33, 55, 66, & 77 Figure 23 Interaction Diagram Element No. 77 Figure 24 Interaction Diagram Element No. 55 Drawings Figure Drawing 1 Reactor Containment Vessel - Equipment/Personnel Access Reinforce-ment - Enlarged Sections Figure Drawing 2 Reactor Containment Vessel - Equipment Access Opening Reinforce-ment - Stretch-out & Sections Figure I Comparison of H.H. & GAI Results Hoop Stress Resultants Along Horizontal and Vertical Symmetry Axes (Internal Pressure = 69 PSI)
Figure Drawing 1 Reactor Containment Vessel - Equipment/Personnel Access Reinforcement - Enlarged Sections Page 38 of 39 Revision 26 5/2016
GINNA/UFSAR Figure Drawing 2 Reactor Containment Vessel - Equipment Access Opening Reinforcement - Stretch-out & Sections Figure Drawing 3 Large Openings - Pour Schedule Appendix 3C CONTAINMENT SHELL STRESS CALCULATION RESULTS 642 Table 3C-1 CONTAINMENT SHELL STRESS CALCULATION RESULTS 643 Appendix 3D CONTAINMENT TENDON ANCHORAGE HARDWARE CAPAC- 668 ITY TESTS Compressive Load Tests of 90 Wire Tendon Base Plate - Test on Con- 669 crete Stand Compressive Load Tests of 90 Wire Tendon Base Plate - Test on Con- 673 crete Stand Compression Tests of 90-Wire Anchor Head Assembly 681 Compression Tests of 90-Wire Anchor Head Assembly 683 Load Tests of Coupler and Adaptor 90-11 690 Load Tests of Coupler and Adaptor 90-11 692 90 Wire Tendon Test 696 90 Wire Tendon Test 697 90 Wire Tendon Test 698 Load Tests of 90-X7 Coupler 702 Appendix 3E CONTAINMENT LINER INSULATION PREOPERATIONAL 704 TESTS BM Containment Insulation SP-5290 Ginna Plant 705 Report No. E455-T-268, VINYLCEL (4 pcf) - Water Vapor Permeabil- 707 ity and Humid Aging Tests Report No. E455-T-266, VINYLCEL (4 pcf) - Effect of Heat and Pres- 711 sure Report No. E455-T-258, VINYLCEL - Resistance to Flame Exposure 718 Appendix 3F
SUMMARY
OF STRUCTURAL DESIGN CODE COMPARISON 740 Table of Contents 741 3F.1 INTRODUCTION 742 Table 3F.2-1 AISC 1963 VERSUS AISC 1980
SUMMARY
OF CODE 743 COMPARISON Table 3F.3-1 ACI 318-63 VERSUS ACI 349-76
SUMMARY
OF CODE 747 COMPARISON Table 3F.4-1 ACI 301-63 VERSUS ACI 301-72 (REVISED 1975)
SUMMARY
OF 756 CODE COMPARISON Table 3F.5-1 ACI 318-63 VERSUS ASME B&PV CODE, SECTION III, 762 DIVISION 2, 1980,
SUMMARY
OF CODE COMPARISON
- 1. Support load combination is seismic plus deadweight.
- 2. Support load combination is seismic plus deadweight.
Page 39 of 39 Revision 26 5/2016
GINNA/UFSAR 4 REACTOR 1 4.1
SUMMARY
DESCRIPTION 2 4.1.1 REACTOR CORE 2 4.1.2 WESTINGHOUSE OPTIMIZED FUEL ASSEMBLIES/422 VAN- 2 TAGE + FUEL ASSEMBLIES 4.1.3 RECONSTITUTED FUEL ASSEMBLIES 4 4.1.4 STARTUP REPORT 5
4.1 REFERENCES
FOR SECTION 4.1 6 4.2 FUEL SYSTEM DESIGN 7 4.2.1 DESIGN BASES 7 4.2.1.1 Performance Objectives 7 4.2.1.2 Principal Design Criteria 7 4.2.1.2.1 Reactor Core Design 7 4.2.1.2.2 Suppression of Power Oscillations 9 4.2.1.2.3 Redundancy of Reactivity Control 9 4.2.1.2.4 Reactivity MODE 3 (Hot Shutdown) Capability 9 4.2.1.2.5 Reactivity Shutdown Capability 10 4.2.1.2.6 Reactivity Holddown Capability 10 4.2.1.2.7 Reactivity Control Systems Malfunction 11 4.2.1.2.8 Maximum Reactivity Worth of Control Rods 12 4.2.1.2.9 Conformance With 1972 General Design Criteria 12 4.2.1.3 Safety Limits 13 4.2.1.3.1 Nuclear Limits 13 4.2.1.3.2 Reactivity Control Limits 13 4.2.1.3.3 Thermal and Hydraulic Limits 13 4.2.1.3.4 Mechanical Limits 14 4.2.1.3.4.1 Reactor Internals 14 4.2.1.3.4.2 Fuel Assemblies 15 4.2.1.3.4.3 Control Rods 16 4.2.1.3.4.4 Control Rod Drive Assembly 16 4.2.2 FUEL SYSTEM DESIGN DESCRIPTION 17 Page 1 of 5 Revision 26 5/2016
GINNA/UFSAR 4 REACTOR 1 4.2.3 CORE COMPONENTS DESIGN DESCRIPTION 17 4.2.3.1 Fuel Assembly 17 4.2.3.1.1 Top Nozzle, Springs, and Clamps 18 4.2.3.1.2 Bottom Nozzle 18 4.2.3.1.3 Guide Thimbles 19 4.2.3.1.4 Instrumentation Tube 19 4.2.3.1.5 Grid Assemblies 19 4.2.3.1.6 Fuel Rods 20 4.2.3.1.7 Fuel Assembly Joints and Connections 20 4.2.3.1.8 Fuel Assembly Identification 21 4.2.3.2 Control Rods 21 4.2.3.3 Neutron Source Assemblies 22 4.2.3.4 Plugging Devices 22 4.2.3.5 Fuel Pellet and Cladding Design Considerations 23 4.2.3.6 Reload Fuel Design 24 4.2.3.6.1 Reload Fuel Design - Westinghouse Optimized Fuel 24 4.2.3.6.2 Reload Fuel Design - Westinghouse OFA/VANTAGE + Fuel 24 4.2.3.6.3 Reload Fuel Design - Westinghouse 422V+ Fuel 24 4.2.3.7 Fuel Assembly and Rod Cluster Control Assembly Tests 24 4.2.3.7.1 Reactor Evaluation Center Tests 24 4.2.3.7.2 Loading and Handling Tests 24 4.2.3.7.3 Axial and Lateral Bending Tests 25 4.2.4 DESIGN EVALUATION 25 4.2.4.1 Fuel and Cladding Evaluation - Original Core 25 4.2.4.2 Design Evaluation - Reload Optimized Fuel Assembly, OFA/VAN- 26 TAGE+ Fuel Assembly, and 422 VANTAGE+ Fuel Assembly Designs 4.2.4.2.1 Introduction 26 4.2.4.2.2 Fuel Design 26 4.2.4.2.3 Design for Seismic and Loss-of-Coolant Accident Forces 26 4.2.4.2.4 Emergency Core Cooling System (ECCS) Calculation Loss-of-Coolant 26 Accident Cladding Models 4.2.4.2.5 Initial Fuel Conditions for Transient Analysis 27 Page 2 of 5 Revision 26 5/2016
GINNA/UFSAR 4 REACTOR 1 4.2.4.2.6 Predicted Clad Collapse Time 27 4.2.4.2.7 Nuclear Design 27 4.2.4.2.8 Fuel Assembly Hydraulic Lift-Off 28 4.2.4.2.9 Thermal-Hydraulic Analysis 28 4.2.4.2.9.1 Sensitivity Factors 28 4.2.4.2.9.2 WRB-1 Correlation 28 4.2.4.2.9.3 Rod Bow Penalties 29 4.2.4.2.9.4 DNBR Design Limits 30 4.2.4.3 Design Evaluation of Reconstituted Fuel Assemblies 30 4.2.5 CORE COMPONENTS TESTS AND INSPECTIONS 30
4.2 REFERENCES
FOR SECTION 4.2 31 Table 4.2-1 NUCLEAR DESIGN DATA 33 Table 4.2-2 CORE MECHANICAL DESIGN PARAMETERS 35 Table 4.2-3 FUEL DESIGN 38 Table 4.2-4 KINETIC PARAMETERS USED IN TRANSIENT ANALYSIS 39 (WESTINGHOUSE OFA/VANTAGE+ AND 422V+ GINNA FUEL ASSEMBLY 14 x 14 FUEL) 4.3 RELOAD CORE NUCLEAR DESIGN 40 4.3.1 PRELIMINARY DESIGN PHASE 40 4.3.2 DETERMINATION OF NUCLEAR-RELATED KEY SAFETY 41 PARAMETERS 4.3.2.1 Reactivity Control Aspects 41 4.3.2.1.1 Insertion Limits 42 4.3.2.1.2 Total Rod Worth 43 4.3.2.1.3 Trip Reactivity 43 4.3.2.1.4 Differential Rod Worths 43 4.3.2.1.5 Summary 44 4.3.2.2 Core Reactivity Parameters and Coefficients 44 4.3.2.2.1 Moderator Temperature Coefficient 44 4.3.2.2.2 Fuel Temperature Coefficient 45 4.3.2.2.3 Boron Worth 45 4.3.2.2.4 Delayed Neutrons 45 4.3.2.2.5 Prompt Neutron Lifetime 45 Page 3 of 5 Revision 26 5/2016
GINNA/UFSAR 4.3.2.2.6 Summary 46 4.3.2.3 Reactor Core Power Distribution 46 4.3.3 EVALUATION OF RELOADS WITH OFA/VANTAGE+ AND 422V+ 46 FUEL ASSEMBLIES 4.3.4 TESTS FOR REACTIVITY ANOMALIES 47
4.3 REFERENCES
FOR SECTION 4.3 48 4.4 THERMAL AND HYDRAULIC DESIGN 49 4.4.1 DESIGN BASIS 49 4.
4.2 DESCRIPTION
AND EVALUATION OF THE THERMAL-HYDRAU- 49 LIC DESIGN AND ANALYSIS OF RELOAD CORES 4.4.2.1 Hydraulic Evaluation 49 4.4.2.2 Thermal and Hydraulic Key Safety Parameters 49 4.4.2.2.1 Engineering Hot-Channel Factors 50 4.4.2.2.2 Axial Fuel Stack Shrinkage 50 4.4.2.2.3 Fuel Temperatures 50 4.4.2.2.4 Rod Internal Pressure 50 4.4.2.2.5 Core Thermal Limits 51 4.4.2.2.6 Key Safety Parameters for Specific Events 52 4.4.2.3 VIPRE Code 52 4.4.2.3.1 Steady-State Analysis 53 4.4.2.3.2 Transient Analysis 53 4.4.3 THERMAL-HYDRAULIC METHODOLOGY FOR OFA/VANTAGE+ 53 and 422V+ FUEL ASSEMBLY DESIGN EVALUATION 4.4.3.1 General 53 4.4.3.2 Rod Bow 55 4.4.4 THERMAL AND HYDRAULIC TESTS AND INSPECTIONS 55 4.4.5 REACTOR COOLANT FLOW MEASUREMENT 55 4.4.5.1 Pump Power 56 4.4.5.2 Secondary Heat Balance 56 4.4.5.3 Elbow Tap Differential Pressure 56 4.4.5.4 Core Exit Thermocouple 56 4.4.5.5 Pump Power-Differential Pressure 57 4.4.5.6 Experience 58 4.4.5.7 Low Flow Trip Setpoint 59 4.4.5.8 Precision Calorimetric Measurement for Reactor Coolant System Flow 59
4.4 REFERENCES
FOR SECTION 4.4 63 Page 4 of 5 Revision 26 5/2016
GINNA/UFSAR Table 4.4-1 THERMAL AND HYDRAULIC DESIGN PARAMETERS 65 4.5 REACTOR MATERIALS 67 4.5.1 CONTROL ROD DRIVE SYSTEM STRUCTURAL MATERIALS 67 4.5.2 REACTOR INTERNALS MATERIALS 67 4.6 FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEM 68 FIGURES Figure 4.2-1 Typical Rod Cluster Control Assembly Figure 4.2-2 Fuel Assembly and Control Cluster Cross Section Figure 4.2-3 14 x 14 OFA and 422V+ Fuel Assemblies Figure 4.2-4 OFA and 422V+ Top Nozzle Assemblies Figure 4.2-5 Debris Filter Bottom Nozzle Figure 4.2-6 Optimized Guide Thimble Assembly Figure 4.2-7 Optimized Instrumentation Tube Figure 4.2-8 Mid-Grid Connection Figure 4.2-9 Removable Top Nozzle and Top Grid Connection Figure 4.3-1 Control Rod Cluster Groups Figure 4.4-1 Typical Pump Power Versus Flow Curves Page 5 of 5 Revision 26 5/2016
GINNA/UFSAR 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 1 5.1
SUMMARY
DESCRIPTION 2 5.1.1 GENERAL 2 5.1.2 PERFORMANCE OBJECTIVES 2 5.1.3 DESIGN CRITERIA 2 5.1.3.1 Quality Standards 3 5.1.3.2 Performance Standards 3 5.1.3.3 Records Requirements 4 5.1.3.4 Missile Protection 4 5.1.3.5 Reactor Coolant Pressure Boundary 4 5.1.3.6 Monitoring Reactor Coolant Leakage 5 5.1.3.7 Reactor Coolant Pressure Boundary Capability 6 5.1.3.8 Reactor Coolant Pressure Boundary Rapid Propagation Failure Prevention 6 5.1.3.9 Reactor Coolant Pressure Boundary Surveillance 7 5.1.3.10 Adequacy of Reactor Coolant System Design Relative to 1972 10 CFR 50, 8 Appendix A, Criteria 5.1.4 DESIGN CHARACTERISTICS 8 5.1.4.1 Design Pressure 8 5.1.4.2 Design Temperature 9 5.1.5 CYCLIC LOADS 9 5.1.6 SERVICE LIFE 9 5.1.7 RELIANCE ON INTERCONNECTED SYSTEMS 10 5.1.8 SYSTEM INCIDENT POTENTIAL 10 Table 5.1-1 REACTOR COOLANT SYSTEM PRESSURE SETTINGS 12 Table 5.1-2 REACTOR COOLANT PIPING DESIGN DATA 13 Table 5.1-3 REACTOR COOLANT SYSTEM DESIGN PRESSURE DROP 14 Table 5.1-4 THERMAL AND LOADING CYCLES 15 5.2 INTEGRITY OF THE REACTOR COOLANT PRESSURE BOUNDARY 16 5.2.1 COMPLIANCE WITH CODES 16 5.2.1.1 System Integrity 16 5.2.1.2 Codes and Classifications 17 5.2.1.2.1 Code Requirements 17 5.2.1.2.2 Quality Control 17 5.2.1.2.3 Field Erection Procedures 18 5.2.1.3 Seismic Loads 18 Page 1 of 8 Revision 26 5/2016
GINNA/UFSAR 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.2.2 OVERPRESSURIZATION PROTECTION 19 5.2.2.1 Normal Operation 19 5.2.2.2 Low Temperature Overpressure Protection (LTOP) System 19 5.2.2.2.1 Design Bases 20 5.2.2.2.2 System Description 20 5.2.2.2.3 System Evaluation 21 5.2.2.2.3.1 General 21 5.2.2.2.3.2 Mass Addition Case 22 5.2.2.2.3.3 Heat Addition at 60F 22 5.2.2.2.3.4 Heat Addition at 320F 23 5.2.2.2.3.5 Administrative Controls 23 5.2.2.2.4 Tests and Inspections 24 5.2.3 REACTOR COOLANT PRESSURE BOUNDARY MATERIALS 24 5.2.3.1 Material Specifications 24 5.2.3.1.1 Nondestructive Examination of Materials and Components Prior to Operation 24 5.2.3.1.1.1 Quality Assurance Program 24 5.2.3.1.1.2 Welding and Heat Treatment 25 5.2.3.1.2 Quality Assurance for Electroslag Welds 26 5.2.3.1.2.1 Piping Elbows 26 5.2.3.1.2.2 Reactor Coolant Pump Casings 26 5.2.3.1.2.3 Reactor Coolant Pump Field Erection and Welding 28 5.2.3.2 Compatibility With Reactor Coolant 28 5.2.4 INSERVICE INSPECTION AND TESTING OF THE REACTOR COOLANT SYS- 29 TEM PRESSURE BOUNDARY 5.2.4.1 Inservice Inspection Program 29 5.2.4.2 Inspection Areas and Components 29 5.2.4.2.1 Accessible Components and Areas 29 5.2.4.2.2 Accessible Areas During Refueling 31 5.2.4.3 Accessibility 31 5.2.4.4 Examination Methods 32 5.2.4.5 Evaluation of Examination Results 33 5.2.4.6 Repair Requirements 33 5.2.4.7 Pressure Testing 33 5.2.4.8 Exemptions 33 Page 2 of 8 Revision 26 5/2016
GINNA/UFSAR 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.2.5 DETECTION OF LEAKAGE THROUGH REACTOR COOLANT PRESSURE 34 BOUNDARY 5.2.5.1 Leakage Detection Methods 34 5.2.5.2 Leakage Limitations 35 5.2.5.3 Locating Leaks 36 5.2.5.4 Leakage Detection System Descriptions 36 5.2.5.4.1 Containment Air Particulate and Radiogas Monitor 36 5.2.5.4.1.1 Air Particulate Monitor 36 5.2.5.4.1.2 Sensitivity Assumptions 36 5.2.5.4.1.3 Leakage Detection Threshold 38 5.2.5.4.1.4 Radiogas Monitor 39 5.2.5.4.2 Humidity Detector 39 5.2.5.4.3 Condensate Measuring System 39 5.2.5.4.4 Liquid Inventory in Process Systems and Containment Sumps 40 5.2.5.5 Leakage Detection System Evaluation 40 Table 5.2-1 REACTOR COOLANT SYSTEM CODE REQUIREMENTS 44 Table 5.2-2 MATERIALS OF CONSTRUCTION OF THE REACTOR COOLANT SYSTEM 45 COMPONENTS Table 5.2-3 REACTOR COOLANT SYSTEM QUALITY ASSURANCE PROGRAM 46 Table 5.2-4 Table DELETED 49 Table 5.2-5 REACTOR COOLANT PRESSURE BOUNDARY TO CONTAINMENT LEAK- 50 AGE DETECTION SYSTEMS Table 5.2-6 REACTOR COOLANT PRESSURE BOUNDARY INTERSYSTEM LEAKAGE 51 DETECTION SYSTEMS Table 5.2-7 SEQUENCE OF EVENTS - MASS ADDITION CASE 52 Table 5.2-8 HEAT ADDITION AT 60F - SEQUENCE OF EVENTS 53 Table 5.2-9 HEAT ADDITION AT 320F - SEQUENCE OF EVENTS 54 5.3 REACTOR VESSEL 55 5.3.1 REACTOR VESSEL MATERIALS 55 5.3.1.1 Reactor Vessel Description 55 5.3.1.2 Material Specifications 56 5.3.1.3 Testing and Surveillance 57 5.3.2 PRESSURE-TEMPERATURE LIMITS 57 Page 3 of 8 Revision 26 5/2016
GINNA/UFSAR 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.3.2.1 Thermal and Pressure Loadings 57 5.3.2.2 Pressure-Temperature Limits 58 5.3.2.3 Pressure-Temperature Limit Calculation 59 5.3.2.4 Irradiation Effect on Pressure-Temperature Limit 59 5.3.2.5 Heatup and Cooldown Rates 60 5.3.3 REACTOR VESSEL INTEGRITY 60 5.3.3.1 Safety Factors 60 5.3.3.2 Material Surveillance Program 61 5.3.3.3 Surveillance Program Analysis 62 5.3.3.3.1 Results Summary 63 5.3.3.3.2 Charpy V-Notch Impact Test Results 65 5.3.3.3.3 Tension Test Results 66 5.3.3.3.4 Radiation Analysis and Neutron Dosimetry 66 5.3.3.4 Analysis of Effects of Loss of Coolant and Safety Injection on the Reactor Vessel 66 5.3.3.4.1 Reactor Vessel 66 5.3.3.4.2 Safety Injection Nozzles 68 5.3.3.4.3 Fuel Assembly Grid Springs 68 5.3.3.4.4 Core Barrel and Thermal Shield 68 5.3.3.4.5 Subsequent Analyses of Reactor Vessel 68 5.3.3.5 Pressurized Thermal Shock 69 Table 5.3-1 REACTOR VESSEL SPECIFICATIONS 74 Table 5.3-2 REACTOR VESSEL DESIGN DATA 75 Table 5.3-3 REACTOR VESSEL MATERIALS 76 Table 5.3-4 IDENTIFICATION OF BELTLINE MATERIALS 77 Table 5.3-5 BELTLINE MATERIAL CHEMICAL COMPOSITION (WEIGHT PERCENT) 78 Table 5.3-6a MECHANICAL PROPERTIES OF BELTLINE MATERIALS - FORGINGS 79 Table 5.3-6b MECHANICAL PROPERTIES OF BELTLINE MATERIALS 80 Table 5.3-7
SUMMARY
OF PRIMARY-PLUS-SECONDARY STRESS INTENSITY FOR 81 COMPONENTS OF THE REACTOR VESSEL Table 5.3-8
SUMMARY
OF CUMULATIVE FATIGUE USAGE FACTORS FOR COMPO- 83 NENTS OF THE REACTOR VESSEL Table 5.3-9
SUMMARY
OF SURVEILLANCE CAPSULE RESULTS 84 Table 5.3-10 COMPARISON OF SURVEILLANCE MATERIAL 30 FT-LB TRANSITION 85 TEMPERATURE SHIFTS AND UPPER SHELF ENERGY DECREASES WITH REGULATORY GUIDE 1.99, REVISION 2, PREDICTIONS Page 4 of 8 Revision 26 5/2016
GINNA/UFSAR 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.4 COMPONENT AND SUBSYSTEM DESIGN 87 5.4.1 REACTOR COOLANT PUMPS 87 5.4.1.1 General Description 87 5.4.1.1.1 Centrifugal Pump 87 5.4.1.1.2 Controlled Leakage Shaft Seal 87 5.4.1.1.3 Pump Motor 88 5.4.1.1.4 Vibration Measurement 88 5.4.1.1.5 Lube Oil Leakage Collection System 89 5.4.1.2 Pump Flywheel Integrity 89 5.4.1.2.1 Pump Overspeed 89 5.4.1.2.2 Pump Flywheel Design and Fabrication 89 5.4.1.2.3 Flywheel Design Evaluation 90 5.4.1.2.4 Pump Seismic Design 90 5.4.1.2.5 Inservice Inspection Program 91 5.4.1.2.6 Conclusion 91 5.4.2 STEAM GENERATORS 91 5.4.2.1 Replacement Steam Generator Materials 92 5.4.2.2 Steam Generator Inservice Inspection 92 5.4.2.3 Replacement Steam Generator Design Evaluation 92 5.4.2.4 High Cycle Fatigue Failure of Original Steam Generator Tubes 93 5.4.3 REACTOR COOLANT PIPING 93 5.4.3.1 General 93 5.4.3.1.1 General Description 93 5.4.3.1.2 Pressure Isolation of Low-Pressure Systems 94 5.4.3.2 Reactor Coolant System Vents 94 5.4.3.2.1 General 94 5.4.3.2.2 Reactor Head Vent System Description 95 5.4.4 MAIN STEAM LINE ISOLATION SYSTEM 96 5.4.5 RESIDUAL HEAT REMOVAL (RHR) SYSTEM 97 5.4.5.1 Design Bases 97 5.4.5.2 System Design 98 5.4.5.2.1 Codes and Classifications 99 5.4.5.2.2 Components 99 Page 5 of 8 Revision 26 5/2016
GINNA/UFSAR 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.4.5.2.2.1 Heat Exchangers 99 5.4.5.2.2.2 Pumps 99 5.4.5.2.2.3 Valves 99 5.4.5.2.2.4 Piping 100 5.4.5.3 Performance Evaluation 100 5.4.5.3.1 Isolation Requirement 100 5.4.5.3.1.1 Isolation Valve Description 100 5.4.5.3.1.2 Deviations From Branch Technical Position RSB 5-1 101 5.4.5.3.2 Residual Heat Removal Overpressure Protection 102 5.4.5.3.2.1 Design Basis 102 5.4.5.3.2.2 Analysis 102 5.4.5.3.2.3 Effect of Stuck Open Relief Valve 103 5.4.5.3.3 Residual Heat Removal Pump Protection 104 5.4.5.3.4 Single-Failure Considerations 105 5.4.5.3.5 Leakage Provisions 106 5.4.5.3.6 Boron Concentration 107 5.4.5.4 Residual Heat Removal at Reduced Coolant Inventory 107 5.4.5.4.1 Generic Letter 88-17 Requirements 107 5.4.5.4.2 Containment Closure 108 5.4.5.4.3 Instrumentation for Reduced Inventory Operation 109 5.4.5.4.4 Available Equipment to Mitigate Loss of Residual Heat Removal Cooling 110 5.4.5.4.5 Reduced Inventory Procedures 110 5.4.5.4.6 Analyses 111 5.4.5.5 Tests and Inspections 112 5.4.6 MAIN STEAM AND FEEDWATER PIPING 112 5.4.7 PRESSURIZER 113 5.4.7.1 System Description 113 5.4.7.2 Seismic Evaluation 114 5.4.8 PRESSURIZER RELIEF DISCHARGE SYSTEM 115 5.4.8.1 System Description 115 5.4.8.2 System Analysis 116 5.4.9 VALVES 116 5.4.9.1 Original Valve Design 116 5.4.9.2 Valve Wall Thickness 117 5.4.9.3 Motor-Operated Valve Program 117 Page 6 of 8 Revision 26 5/2016
GINNA/UFSAR 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.4.10 SAFETY AND PRESSURIZER POWER OPERATED RELIEF VALVES (PORVs) 119 5.4.10.1 System Description 119 5.4.10.2 Performance Testing and Evaluation 120 5.4.11 COMPONENT SUPPORTS 121 5.4.11.1 Design Criteria 121 5.4.11.1.1 General 121 5.4.11.1.2 Asymmetric Loss-of-Coolant Accident Loading 121 5.4.11.1.3 Lamellar Tearing 122 5.4.11.2 Support Structures 122 5.4.11.2.1 Reactor Vessel Supports 122 5.4.11.2.2 Steam Generator Supports 123 5.4.11.2.3 Reactor Coolant Pump Supports 123 5.4.11.2.4 Pressurizer Supports 123 5.4.11.2.5 Reactor Coolant Piping Supports 123 5.4.11.2.6 Inspection and Testing 123 Table 5.4-1 REACTOR COOLANT PUMP DESIGN DATA 129 Table 5.4-2 REPLACEMENT STEAM GENERATOR DESIGN DATA 130 Table 5.4-3 REACTOR COOLANT PUMP COMPOSITE HOT PERFORMANCE CURVE 131 DATA Table 5.4-4 REACTOR COOLANT PUMPS COLD PERFORMANCE CURVE DATA FOR 133 INDIVIDUAL IMPELLERS Table 5.4-5 REACTOR VESSEL HEAD VENT EQUIPMENT PARAMETERS 134 Table 5.4-6 RESIDUAL HEAT REMOVAL SYSTEM COMPONENT DESIGN DATA 136 Table 5.4-7 PRESSURIZER DESIGN DATA 138 Table 5.4-8 PRESSURIZER RELIEF TANK DESIGN DATA 139 Table 5.4-9 VALVE AND PIPING INFORMATION 140 Page 7 of 8 Revision 26 5/2016
GINNA/UFSAR 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS FIGURES Figure 5.2-1 Figure DELETED Figure 5.2-2 Figure DELETED Figure 5.2-3 Reactor Coolant Leak Detection Sensitivity Figure 5.3-1 Reactor Vessel Schematic Figure 5.3-2 Identification and Location of Beltline Region Material Figure 5.3-3 Arrangement of Surveillance Capsules in the Reactor Vessel Figure 5.4-1 Reactor Coolant Pump Figure 5.4-2 Reactor Coolant Pump Estimated Performance Characteristics Figure 5.4-2a Reactor Coolant Pump Composite Curve, Calculated Hot Performance, Total Head and Hydraulic Efficiency Versus Flow Figure 5.4-2b Reactor Coolant Pump Composite Curve, Calculated Hot Performance, Brake Horse- power Versus Flow Figure 5.4-2c Reactor Coolant Pump Composite Curve, Calculated Cold Performance, Total Head and Hydraulic Efficiency Versus Flow Figure 5.4-2d Reactor Coolant Pump Composite Curve, Calculated Cold Performance, Brake Horsepower Versus Flow Figure 5.4-3 Reactor Coolant Pressure Shaft Seal Arrangement Figure 5.4-4 Reactor Coolant Pump Flywheel Figure 5.4-5 Reactor Coolant Pump Flywheel Primary Stress at Operating Speed Figure 5.4-6 Replacement Steam Generator Figure 5.4-7 Figure DELETED Figure 5.4-8 Pressurizer Figure 5.4-9 Pressurizer Relief Tank Page 8 of 8 Revision 26 5/2016
GINNA/UFSAR 6 ENGINEERED SAFETY FEATURES 1 6.1 ENGINEERED SAFETY FEATURES INTRODUCTION AND MATERIALS 2 6.
1.1 INTRODUCTION
2 6.1.2 ENGINEERED SAFETY FEATURES MATERIALS 2 6.1.2.1 Postaccident Environmental Conditions 2 6.1.2.1.1 General 2 6.1.2.1.2 Design-Basis Accident Temperature-Pressure Cycle 3 6.1.2.1.3 Design-Basis Accident Radiation Environment 4 6.1.2.1.4 Design Chemical Composition of the Emergency Core Cooling Solution 4 6.1.2.1.5 Trace Composition of Emergency Core Cooling Solution 5 6.1.2.2 Materials of Construction in the Containment 5 6.1.2.3 Corrosion of Metals of Construction in Design-Basis Emergency Core Cooling 6 Solu- tion 6.1.2.3.1 Corrosion Resistance 6 6.1.2.3.2 Caustic Stress Cracking Resistance 7 6.1.2.4 Corrosion of Metals of Construction by Trace Contaminants in Emergency Core 7 Cool- ing Solution 6.1.2.4.1 Low Temperature of Emergency Core Cooling Solution 8 6.1.2.4.2 Low Chloride Concentration of Emergency Core Cooling Solution 8 6.1.2.4.3 Alkaline Nature of the Emergency Core Cooling Solution 9 6.1.2.4.4 Summary 9 6.1.2.5 Corrosion of Aluminum Alloys 9 6.1.2.6 Compatibility of Protective Coatings With the Postaccident Environment 10 6.1.2.7 Evaluation of the Compatibility of Concrete-Emergency Core Cooling Solution in 10 the Postaccident Environment 6.1.2.8 Miscellaneous Materials of Construction 12 6.1.2.8.1 Sealants 12 6.1.2.8.2 Containment Recirculation Fan Cooler (CRFC) Materials 12 6.1.2.8.3 Polyvinyl Chloride Protective Coating 13 6.1.2.8.4 Vinylcel Insulation 14 6.1.2.9 Organic Materials Evaluation 15 Table 6.1-1 REVIEW OF SOURCES OF VARIOUS ELEMENTS IN CONTAINMENT 18 AND THEIR EFFECTS ON MATERIALS OF CONSTRUCTION Table 6.1-2 MATERIALS OF CONSTRUCTION IN GINNA STATION CONTAINMENT 20 Table 6.1-3 INVENTORIES OF ALUMINUM INSIDE CONTAINMENT BUILDING 21 Table 6.1-4 CORROSION OF MATERIALS IN SODIUM BORATE SOLUTION 22 Page 1 of 14 Revision 26 5/2016
GINNA/UFSAR 6 ENGINEERED SAFETY FEATURES 1 Table 6.1-5 GINNA Post-LOCA CONTAINMENT TEMPERATURES 23 Table 6.1-6 CONCRETE SPECIMEN TEST DATA 24 Table 6.1-7 EVALUATION OF SEALANT MATERIALS FOR USE IN THE CONTAINMENT 25 6.2 CONTAINMENT SYSTEMS 26 6.2.1 CONTAINMENT SYSTEM STRUCTURE 26 6.2.1.1 Design Basis 26 6.2.1.1.1 Principal Design Criteria 27 6.2.1.1.1.1 General 27 6.2.1.1.1.2 Quality Standards 27 6.2.1.1.1.3 Performance Standards 27 6.2.1.1.1.4 Fire Protection 28 6.2.1.1.1.5 Records Requirement 28 6.2.1.1.1.6 Reactor Containment 29 6.2.1.1.1.7 Reactor Containment Design Basis 30 6.2.1.1.1.8 Nil Ductility Transition Requirement for Containment Material 30 6.2.1.1.2 Supplementary Accident Criteria 30 6.2.1.1.3 Energy and Material Release 31 6.2.1.2 Containment Integrity Evaluation 31 6.2.1.2.1 Systematic Evaluation Program (SEP) Evaluation 31 6.2.1.2.1.1 Introduction 31 6.2.1.2.1.2 NRC Analyses 32 6.2.1.2.1.3 Summary 32 6.2.1.2.2 Mass and Energy Release Safety Analysis 32 6.2.1.2.2.1 Loss-of-Coolant (LOCA) Mass and Energy Releases 32 6.2.1.2.2.2 Input Parameters and Assumptions 34 6.2.1.2.2.3 Desription of Analyses 38 6.2.1.2.2.4 Mass and Energy Release Data 41 6.2.1.2.2.5 Long-Term Mass and Energy Releases 44 6.2.1.2.2.6 Long-Term LOCA Containment Response 45 6.2.1.2.2.7 Description of the LOCA GOTHIC Containment Model 49 6.2.1.2.2.8 Results 53 6.2.1.2.3 Secondary System Pipe Break Analysis 55 6.2.1.2.3.1 Event Analysis 55 6.2.1.2.3.2 Protective Features 55 6.2.1.2.3.3 Single Failures Assumed 56 6.2.1.2.3.4 Operator Actions Assumed 56 6.2.1.2.3.5 Chronological Description of Event 56 6.2.1.2.3.6 Impact on Fission Product Barriers 57 Page 2 of 14 Revision 26 5/2016
GINNA/UFSAR 6.2.1.2.3.7 Reactor Core and Plant System Evaluation 57 6.2.1.2.3.8 Input Parameters and Initial Conditions 58 6.2.1.2.3.9 Methodology 58 6.2.1.2.3.10 Acceptance Criteria 58 6.2.1.2.3.11 Results 58 6.2.1.2.3.12 Radiological Consequences 58 6.2.1.2.3.13 Conclusion 59 6.2.1.3 Evaluation of Containment Internal Structures 59 6.2.1.3.1 Introduction 59 6.2.1.3.2 Reactor Coolant Loop Compartment Pressure 59 6.2.1.3.3 Thermal Gradients 60 6.2.1.3.4 Reactor Vessel and Steam Generator Annulus Pressure 60 6.2.1.3.5 Seismic Evaluation 61 6.2.1.3.6 Technical Evaluation for Extended Power Uprate (EPU) Conditions 61 6.2.1.3.6.1 Introduction 61 6.2.1.3.6.2 Input Parameters and Assumptions 61 6.2.1.3.6.3 Acceptance Criteria 62 6.2.1.3.6.4 Description of Analysis 63 6.2.1.3.6.5 Short-Term LOCA M&E Releases Results 63 6.2.1.4 Minimum Operating Conditions 64 6.2.1.5 Instrumentation Requirements 64 6.2.1.5.1 Pressure 64 6.2.1.5.2 Sump Level 64 6.2.1.5.3 Radiation 65 6.2.1.5.4 Containment Temperature and Dewpoint 65 6.2.2 CONTAINMENT HEAT REMOVAL SYSTEMS 65 6.2.2.1 Containment Recirculation Fan Cooler (CRFC) System 66 6.2.2.1.1 Design Bases 66 6.2.2.1.1.1 Capacity 66 6.2.2.1.1.2 Design Objectives 67 6.2.2.1.1.3 Special Features 68 6.2.2.1.2 System Design 68 6.2.2.1.2.1 System Description 68 6.2.2.1.2.2 Design Analysis 69 6.2.2.1.2.3 Redundancy Provisions 70 6.2.2.1.2.4 Actuation Provisions 71 6.2.2.1.2.5 Environmental Protection 71 6.2.2.1.3 Design Evaluation 72 6.2.2.1.4 Tests and Inspections 73 6.2.2.1.5 Instrumentation 73 Page 3 of 14 Revision 26 5/2016
GINNA/UFSAR 6.2.2.2 Containment Spray System 74 6.2.2.2.1 Design Bases 74 6.2.2.2.1.1 Design Criteria 74 6.2.2.2.1.2 Performance Objectives 75 6.2.2.2.1.3 Service Life 76 6.2.2.2.1.4 Codes and Classifications 76 6.2.2.2.2 System Design 76 6.2.2.2.2.1 Operational Requirements 76 6.2.2.2.2.2 Refueling Water Storage Tank (RWST) 77 6.2.2.2.2.3 Containment Spray Pumps 77 6.2.2.2.2.4 Liquid Jet Eductor 77 6.2.2.2.2.5 Spray Ring Headers 77 6.2.2.2.2.6 Spray Nozzles 78 6.2.2.2.2.7 Environmental Qualification 78 6.2.2.2.2.8 System Tests 78 6.2.2.2.3 Design Evaluation 78 6.2.2.2.3.1 Design Basis 78 6.2.2.2.3.2 Heat Transfer Calculations 79 6.2.2.2.3.3 Reliance on Interconnected Systems 81 6.2.2.2.3.4 Reliability Considerations 81 6.2.2.2.3.5 Containment Spray Pump Net Positive Suction Head Requirements 81 6.2.2.2.3.6 Equipment Protection 81 6.2.2.2.4 Minimum Operating Conditions 82 6.2.2.2.5 Tests and Inspections 82 6.2.2.2.6 Instrumentation 82 6.2.2.2.6.1 Interlock and Control Features 82 6.2.2.2.6.2 Control Room and Local Indication 82 6.2.3 SECONDARY CONTAINMENT 83 6.2.4 CONTAINMENT ISOLATION SYSTEM 83 6.2.4.1 Design Criteria 83 6.2.4.2 Design Basis 84 6.2.4.2.1 Functional Requirements 84 6.2.4.2.2 Seismic Design 85 6.2.4.3 System Design 85 6.2.4.3.1 Isolation Valve Parameters Tabulation 86 6.2.4.3.2 Isolation Valve Operability 86 6.2.4.4 Design Evaluation 87 6.2.4.4.1 Current Safety Criteria 87 6.2.4.4.2 Class 1 Penetrations (Outgoing Lines, Reactor Coolant System) 88 6.2.4.4.2.1 Applicable Lines 88 Page 4 of 14 Revision 26 5/2016
GINNA/UFSAR 6.2.4.4.2.2 Class 1 Penetration Evaluation 88 6.2.4.4.3 Class 2 (Outgoing Lines) 89 6.2.4.4.3.1 Applicable Lines 89 6.2.4.4.3.2 Class 2 Evaluation 89 6.2.4.4.4 Class 3 (Incoming Lines) 90 6.2.4.4.4.1 Class 3A Penetrations 91 6.2.4.4.4.2 Class 3B Penetrations 91 6.2.4.4.5 Class 4 Penetrations (Closed System, Missile Protected) 92 6.2.4.4.5.1 Applicable Lines 92 6.2.4.4.5.2 Class 4 Evaluation 93 6.2.4.4.6 Class 5 Penetrations (Special Service) 94 6.2.4.4.6.1 Applicable Lines 94 6.2.4.4.6.2 Class 5 Evaluation 94 6.2.4.4.7 Special Cases 95 6.2.4.4.8 Instrumentation and Controls Evaluation 95 6.2.4.4.9 Containment Purging During Normal Plant Operation 95 6.2.5 COMBUSTIBLE GAS CONTROL IN THE CONTAINMENT 96 6.2.5.1 Introduction 97 6.2.5.2 Hydrogen Recombiner System 98 6.2.5.2.1 Description 98 6.2.5.2.2 Containment Venting 99 6.2.5.3 Design Evaluation 99 6.2.5.3.1 Hydrogen Production and Accumulation 99 6.2.5.3.1.1 Zirconium-Water Reaction 100 6.2.5.3.1.2 Radiolytic Hydrogen Generation 100 6.2.5.3.1.3 Corrosion of Materials 101 6.2.5.3.1.4 Initial Inventory in the RCS and Pressurizer 102 6.2.5.3.2 Effect of Recombiners 102 6.2.6 CONTAINMENT LEAKAGE TESTING 103 6.2.6.1 Containment Design Leakage 103 6.2.6.2 Tests and Inspections 103 6.2.6.2.1 Design Criteria 103 6.2.6.2.2 Initial Containment Leakage Rate Testing 103 6.2.6.2.3 Periodic Containment Leakage Rate Testing (Type A Tests) 104 6.2.6.2.4 Provisions for Testing of Type B Penetrations 104 6.2.6.2.5 Provisions for Testing of Isolation Valves (Type C) 105 6.2.6.3 Leakage Test Compliance with 10 CFR 50, Appendix J 105 Table 6.2-1 SYSTEM PARAMETERS INITIAL CONDITIONS 111 Table 6.2-2 SAFETY INJECTION FLOW - MINIMUM SAFEGUARDS 112 Table 6.2-3 SAFETY INJECTION FLOW - MAXIMUM SAFEGUARDS 113 Page 5 of 14 Revision 26 5/2016
GINNA/UFSAR Table 6.2-4 LOCA M&E RELEASE ANALYSIS - CORE DECAY HEAT FRACTION 114 Table 6.2-5 DOUBLE-ENDED HOT LEG BREAK BLOWDOWN M&E RELEASE 116 Table 6.2-6 DOUBLE-ENDED HOT LEG BREAK - MASS BALANCE 120 Table 6.2-7 DOUBLE-ENDED HOT LEG BREAK - ENERGY BALANCE 121 Table 6.2-8 DOUBLE-ENDED PUMP SUCTION BREAK MIN SI BLOWDOWN M&E 122 RELEASE Table 6.2-9 DOUBLE-ENDED PUMP SUCTION BREAK MIN SI 126 Reflood M&E Release Table 6.2-10 DOUBLE-ENDED PUMP SUCTION BREAK 132 Min SI Principle Parameters During Reflood Table 6.2-11 DOUBLE-ENDED PUMP SUCTION BREAK 134 Post Reflood M&E Release-Minimum Safeguards Table 6.2-12 DOUBLE-ENDED PUMP SUCTION BREAK MASS BALANCE - MIN SI 141 Table 6.2-13 DOUBLE-ENDED PUMP SUCTION BREAK ENERGY BALANCE - 142 MINIMUM SAFEGUARDS Table 6.2-14 Table DELETED 143 Table 6.2-15 Table DELETED 144 Table 6.2-16 CONTAINMENT RESPONSE ANALYSIS PARAMETERS 145 Table 6.2-17 CONTAINMENT RECIRCULATION FAN COOLER HEAT REMOVAL 146 CAPABIL- ITY AS A FUNCTION OF CONTAINMENT STEAM SATURATION TEMPERA- TURE Table 6.2-18 LOCA CONTAINMENT RESPONSE ANALYSIS RECIRCULATION 147 SYSTEM ALIGNMENT PARAMETERS Table 6.2-19 CONTAINMENT STRUCTURAL HEAT SINK INPUT 148 Table 6.2-20 MATERIAL PROPERTIES FOR CONTAINMENT STRUCTURAL HEAT SINKS 151 Table 6.2-21 DOUBLE-ENDED HOT LEG BREAK SEQUENCE OF EVENTS 152 Table 6.2-22 DOUBLE-ENDED PUMP SUCTION BREAK SEQUENCE OF EVENTS 153 (Minimum Safeguards)
Table 6.2-23 LOCA CONTAINMENT RESPONSE RESULTS 155 Table 6.2-24 INITIAL CONDITIONS AND MAJOR ASSUMPTIONS FOR THE STEAMLINE 156 BREAK MASS AND ENERGY RELEASE MODEL (LIMITING CONTAINMENT PRESSURE CASE)
Table 6.2-25 MAJOR CONTAINMENT ANALYSIS ASSUMPTIONS 157 Table 6.2-26 SEQUENCE OF EVENTS 158 STEAMLINE BREAK, VITAL BUS FAILURE Table 6.2-27 CONTAINMENT SPRAY PUMP DESIGN PARAMETERS 159 Table 6.2-28 SINGLE FAILURE ANALYSIS - CONTAINMENT SPRAY SYSTEM 160 Table 6.2-29 CONTAINMENT PIPING PENETRATIONS AND ISOLATION BOUNDARIES 161 Table 6.2-30 CONTAINMENT PIPING PENETRATIONS AND ISOLATION BOUNDARIES 178
- NOTES FOR TABLE 6.2-29 Table 6.2-31 CONTAINMENT PIPING PENETRATIONS AND ISOLATIONBOUNDARIES - 181 LEGEND FOR Table 6.2-29 Table 6.2-32 EFFECT OF LOSS OF AIR OR POWER SUPPLY TO AIR-OPERATED VALVES 182 Page 6 of 14 Revision 26 5/2016
GINNA/UFSAR Table 6.2-33 ESSENTIAL AND NONESSENTIAL SYSTEM CONTAINMENT PENETRATIONS 184 Table 6.2-34 PARAMETERS AND ASSUMPTIONS USED TO DETERMINE HYDROGEN GENERATION (HISTORICAL) 188 Table 6.2-35 FISSION PRODUCT DECAY ENERGY IN SUMP SOLUTION (HISTORICAL) 189 Table 6.2-36 FISSION PRODUCT DECAY ENERGY IN THE CORE (HISTORICAL) 190 6.3 EMERGENCY CORE COOLING SYSTEM (ECCS) 191 6.3.1 DESIGN CRITERIA 191 6.3.1.1 Emergency Core Cooling System (ECCS) Capability 191 6.3.1.2 Inspection of Emergency Core Cooling System (ECCS) 192 6.3.1.3 Testing of Emergency Core Cooling System (ECCS) and Components 192 6.3.1.4 Testing of Operational Sequence of Emergency Core Cooling System (ECCS) 192 6.3.1.5 Service Life 193 6.3.1.6 Codes and Classifications 193 6.3.2 SYSTEM DESIGN AND OPERATION 193 6.3.2.1 System Description 193 6.3.2.1.1 General 193 6.3.2.1.2 Injection Phase 196 6.3.2.1.3 Recirculation Phase 197 6.3.2.2 Component Description 197 6.3.2.2.1 Accumulators 197 6.3.2.2.2 Safety Injection Pumps 198 6.3.2.2.2.1 Operation 198 6.3.2.2.2.2 Pump Design and Fabrication 199 6.3.2.2.3 Refueling Water Storage Tank (RWST) 199 6.3.2.2.4 Heat Exchangers 200 6.3.2.2.5 Boric Acid Storage Tanks 200 6.3.2.2.6 Containment Sump B 201 6.3.2.2.7 Valves 201 6.3.2.2.7.1 General 201 6.3.2.2.7.2 Motor-Operated Valves 202 6.3.2.2.7.3 Manual Valves 203 6.3.2.2.7.4 Accumulator Check Valves 203 6.3.2.2.7.5 Leakage Limitations 204 6.3.2.2.8 Piping 205 6.3.2.2.8.1 General 205 6.3.2.2.8.2 Design Criteria 205 6.3.2.2.8.3 Design Review 205 6.3.2.2.8.4 Materials 206 6.3.2.2.8.5 Welding and Fabrication 206 Page 7 of 14 Revision 26 5/2016
GINNA/UFSAR 6.3.2.2.8.6 Packaging 207 6.3.2.2.9 Motors 207 6.3.2.3 System Operation 207 6.3.2.3.1 Separation 207 6.3.2.3.2 System Actuation 207 6.3.2.3.3 Injection Phase 208 6.3.2.3.4 Recirculation Phase 208 6.3.2.3.5 Steam Line Break Protection 209 6.3.2.3.6 Safety Injection System Leakage Outside Containment 210 6.3.3 DESIGN EVALUATION 210 6.3.3.1 Range of Core Protection 210 6.3.3.1.1 Safety Injection Requirements Versus Break Size 210 6.3.3.1.2 Makeup System Capacity 211 6.3.3.1.3 System Evaluation 211 6.3.3.2 System Response 212 6.3.3.3 Safety Injection System Switchover From Injection to Recirculation 212 6.3.3.4 Boron Precipitation During Long-Term Cooling 214 6.3.3.5 Single Failure Analysis 214 6.3.3.6 Passive Systems 215 6.3.3.7 Emergency Flow to the Core 215 6.3.3.8 Recirculation Loop Leakage 216 6.3.3.9 Safety Injection Pump Net Positive Suction Head Requirements 216 6.3.3.10 Seismic Analysis 218 6.3.3.11 MODE 4 (Hot Standby) LOCA Evaluation 218 6.3.3.12 Alternate RCS Injection (BDB) 219 6.3.4 MINIMUM OPERATING CONDITIONS 220 6.3.5 TESTS AND INSPECTIONS 220 6.3.5.1 Inspection 220 6.3.5.2 System Testing 220 6.3.5.3 Components Testing 221 6.3.5.4 Operational Sequence Testing 222 6.3.5.5 Gas Intrusion Management Program 223 6.3.6 INSTRUMENTATION 223 6.3.6.1 Containment Sump Level 223 6.3.6.2 Refueling Water Storage Tank (RWST) Level 224 6.3.6.3 Accumulator Pressure and Level 224 6.3.6.4 Boric Acid Storage Tank Level 224 6.3.6.5 Residual Heat Exchanger Flow and Temperature 224 Page 8 of 14 Revision 26 5/2016
GINNA/UFSAR 6.3.6.6 Safety Injection Line Flow 224 6.3.6.7 Safety Injection Pumps Discharge Pressure 224 6.3.6.8 Pump Energization 225 6.3.6.9 Valve Position 225 Table 6.3-1 QUALITY STANDARDS OF SAFETY INJECTION SYSTEM COMPONENTS 228 Table 6.3-2 ACCUMULATOR DESIGN PARAMETERS 232 Table 6.3-3 SAFETY INJECTION SYSTEM PUMPS DESIGN PARAMETERS 233 Table 6.3-4 REFUELING WATER STORAGE TANK (RWST) DESIGN PARAMETERS 234 Table 6.3-5 RESIDUAL HEAT REMOVAL HEAT EXCHANGERS DESIGN PARAMETERS 235 Table 6.3-6 RECIRCULATION LOOP LEAKAGE INFORMATION USED IN 236 ORIGINAL ANALYSIS Table 6.3-7 INSTRUMENTATION READOUTS ON THE CONTROL BOARD FOR 237 OPERA- TOR MONITORING DURING RECIRCULATION Table 6.3-8 SAFETY INJECTION VALVE OPERATION AND INTERLOCKS 239 Table 6.3-9 SINGLE FAILURE ANALYSIS - SAFETY INJECTION SYSTEM 241 6.4 HABITABILITY SYSTEMS 243 6.4.1 DESIGN CRITERION 243 6.4.2 SYSTEM DESIGN 244 6.4.2.1 Definition of Control Room Envelope (CRE) 244 6.4.2.2 Ventilation System Design 244 6.4.2.2.1 Normal HVAC System - NORMAL and PURGE Modes of Operation 244 6.4.2.2.2 CREATS System - EMERGENCY Mode of Operation 244 6.4.2.3 Leak Tightness 245 6.4.2.4 Interaction with Other Zones and Pressure-containing Equipment. 245 6.4.2.4.1 Interaction with the Turbine Building 246 6.4.2.4.2 Interaction with the Relay Room 246 6.4.2.5 Shielding Design 246 6.4.2.6 System Operational Procedures 246 6.4.3 DESIGN EVALUATIONS 247 6.4.3.1 Radiological Analysis 247 6.4.3.2 Protection from Toxins 248 6.4.3.2.1 Chlorine 249 6.4.3.2.2 Ammonia 249 6.4.3.2.3 Halon 249 6.4.3.2.4 Refrigerant 249 6.4.3.2.5 Sodium Hypochlorite 249 6.4.3.2.6 Carbon Dioxide 250 6.4.3.3 Protection from Smoke and Fire 250 Page 9 of 14 Revision 26 5/2016
GINNA/UFSAR 6 ENGINEERED SAFETY FEATURES 1 6.4.3.3.1 Internal Sources of Smoke and Fire 250 6.4.3.3.2 External Sources of Smoke and Fire 250 6.4.3.4 Protection from Temperature Extremes 251 6.4.4 TESTS AND INSPECTIONS 251 6.4.5 INSTRUMENTATION REQUIREMENT 252 Table 6.4-1 Control Room Habitability Radiological Evaluation - Assumptions and Results 255 Table 6.4-2 CORE ACTIVITIES 257 6.5 FISSION PRODUCT REMOVAL SYSTEMS 258 6.5.1 ENGINEERED SAFETY FEATURE FILTER SYSTEMS 258 6.5.1.1 Introduction 258 6.5.1.2 Containment Air Filtration System 258 6.5.1.2.1 Design Basis 258 6.5.1.2.2 System Design 259 6.5.1.2.2.1 General Description 259 6.5.1.2.2.2 Charcoal Filters 259 6.5.1.2.2.3 HEPA Filters 260 6.5.1.2.2.4 Protection From Sodium Hydroxide Attack 260 6.5.1.2.2.5 Fire Protection 261 6.5.1.2.3 Design Evaluation 262 6.5.1.2.3.1 Decay Heat Generation in the Charcoal Filters 262 6.5.1.2.3.2 Decay Heat Dissipation With Normal Air Flow 262 6.5.1.2.3.3 Decay Heat Dissipation With Loss of Air Flow 262 6.5.1.2.4 Tests and Inspections 263 6.5.1.2.4.1 HEPA Filter Tests 263 6.5.1.2.4.2 Charcoal Filter Tests 264 6.5.1.2.4.3 System Tests 264 6.5.1.2.5 Instrumentation Requirements 264 6.5.1.3 Generic Letter 96-06 Requirements 265 6.5.1.4 Generic Letter 99-02 Requirements 266 6.5.2 CONTAINMENT SPRAY AND NaOH SYSTEMS 266 6.5.2.1 System Design and Operation 266 6.5.2.1.1 Spray Additive Tank 266 6.5.2.1.2 Effect of Sodium Hydroxide and Boric Acid Mixing 267 6.5.2.1.3 Iodine Retention 269 6.5.2.2 Iodine Effectiveness Evaluation of the Containment Spray and NaOH Systems 270 Page 10 of 14 Revision 26 5/2016
GINNA/UFSAR 6 ENGINEERED SAFETY FEATURES 1 6.5.2.2.1 Purpose of Chemical Modification 270 6.5.2.2.1.1 Thermal Capacity 270 6.5.2.2.1.2 Absorption of Iodine in Refueling Water Spray 270 6.5.2.2.1.3 Iodine Absorption with Sodium Hydroxide Addition 271 6.5.2.2.1.4 Spray Absorption Process for Iodine Removal 272 6.5.2.2.2 Technical Basis for Iodine Removal Factor 272 6.5.2.2.2.1 Analytical Model and Assumptions 272 6.5.2.2.2.2 Removal of Elemental Iodine 275 6.5.2.2.2.3 Removal of Other Airborne Forms of Iodine 276 6.5.2.2.2.4 Experimental Verification 276 Table 6.5-1 DATA FOR CHARCOAL FILTER EVALUATION 280 6.6 INSERVICE INSPECTION OF CLASS 2 AND 3 COMPONENTS 281 6.
6.1 INTRODUCTION
281 6.6.2 INSERVICE INSPECTION PROGRAM
SUMMARY
281 6.6.2.1 Scope 281 6.6.2.2 Inspection Intervals 281 6.6.2.3 Extent and Frequency 281 6.6.2.4 Examination Methods 282 6.6.2.5 Evaluation of Examination Results 282 6.6.2.6 System Pressure Testing 282 6.6.2.7 Records and Reports 282 6.6.2.8 Exemptions 282 FIGURES Figure 6.1-1 Design-Basis Accident, Containment Temperature Profile Figure 6.1-2 Design-Basis Accident, Containment Pressure Profile Figure 6.1-3 Postaccident Core Materials Design Conditions Figure 6.1-4 Containment Atmosphere Total Gamma Dose Figure 6.1-5 Containment Atmosphere Total Beta Dose Figure 6.1-6 pH of Unadjusted Boric Acid Solutions Figure 6.1-7 Titration Curve for Boric Acid With Sodium Hydroxide Figure 6.1-8 Temperature-Concentration Relation For Caustic Corrosion of Austenitic Stainless Steel Figure 6.1-9 Aluminum Corrosion Rates in LOCA Environment Figure 6.1-10 Post LOCA Containment Hydrogen Production Rate Figure 6.1-11 Boron Loss From Boron-Concrete Reaction Following a Design-Basis Accident Figure 6.1-12 Post LOCA Containment Hydrogen Production Figure 6.2-1 Containment Atmosphere Pressure, Double-Ended Hot Leg Break Figure 6.2-2 Containment Atmosphere Temperature, Double-Ended Hot Leg Break Figure 6.2-3 Containment Sump Temperature, Double-Ended Hot Leg Break Page 11 of 14 Revision 26 5/2016
GINNA/UFSAR Figure 6.2-4 Containment Atmosphere Pressure, Double-Ended Pump Suction Break -
Minimum Safeguards Figure 6.2-5 Containment Atmosphere Temperature, Double-Ended Pump Suction Break -
Mini- mum Safeguards Figure 6.2-6 Containment Sump Temperature, Double-Ended Pump Suction Break -
Minimum Safeguards Figure 6.2-7 1.1-Ft2 Break Case 25B With 102% Power and Diesel Failure Assumed, Containment Pressure Versus Time Figure 6.2-8 1.4-Ft2 Break Case 13A With 102% Power and Vital Bus Failure Assumed, Contain- ment Steam Temperature Versus Time Figure 6.2-9 1.1-Ft2 Break Case 25B With 102% Power and Diesel Failure Assumed, Containment Steam Temperature Versus Time Figure 6.2-10 Reactor Containment Fan Cooler, Accident Versus Normal Air Flow Figure 6.2-11 Figure DELETED Figure 6.2-12 Figure Deleted Figure 6.2-13 Steam Generator Inspection and Maintenance Cabling Access Penetration 2 Figure 6.2-13a Fuel Transfer Tube Penetration 29 Figure 6.2-14 Reactor Coolant System Charging Line Penetration 100 Figure 6.2-15 Safety Injection System Pentrations 101, 110b, and 113 Figure 6.2-16 Alternate Charging Line Penetration 102 Figure 6.2-17 Construction Fire Service Water Penetration 103 Figure 6.2-18 Containment Spray Header A Penetration 105 Figure 6.2-19 Reactor Coolant Pump A Seal Water Line Penetration 106 Figure 6.2-20 Sump A Discharge Penetration 107 Figure 6.2-21 Reactor Coolant Pump Seal Water Return and Excess Letdown Penetration 108 Figure 6.2-22 Containment Spray Header B Penetration 109 Figure 6.2-23 Reactor Coolant Pump B Seal Water Line Penetration 110a Figure 6.2-24 Residual Heat Removal to Loop B Cold Leg Penetration 111 Figure 6.2-25 Letdown Line from Reactor Coolant System Penetration 112 Figure 6.2-26 Standby Auxiliary Feedwater to Steam Generators A and B Penetrations 119 and 123b Figure 6.2-27 Nitrogen to Accumulators Penetration 120a Figure 6.2-28 Pressurizer Relief Tank Gas Analyzer Penetration 120b Figure 6.2-29 Pressurizer Relief Tank Makeup Water Penetration 121a Figure 6.2-30 Pressurizer Relief Tank N2 Penetration 121b Figure 6.2-31 Containment Pressure Transmitters PT-945 and PT-946 Penetration 121c Figure 6.2-32 Reactor Coolant Drain Tank to Gas Analyzer Penetration 123a Figure 6.2-33 Component Cooling Water to and from Excess Letdown Heat Exchanger Penetrations 124a and 124c Figure 6.2-34 Containment Postaccident Air Sample (C Fan) Penetrations 124b and 124d Figure 6.2-35 Component Cooling Water from Reactor Coolant Pump 1B Penetration 125 Figure 6.2-36 Component Cooling Water from Reactor Coolant Pump 1A Penetration 126 Page 12 of 14 Revision 26 5/2016
GINNA/UFSAR Figure 6.2-37 Component Cooling Water to Reactor Coolant Pump 1A Penetration 127 Figure 6.2-38 Component Cooling Water to Reactor Coolant Pump B Penetration 128 Figure 6.2-39 Reactor Coolant Drain Tank Gas Header Penetration 129 Figure 6.2-40 Component Cooling Water from and to Reactor Support Coolers Penetrations 130 and 131 Figure 6.2-41 Mini-Purge Exhaust Penetration 132 Figure 6.2-42 Residual Heat Removal from Loop A Hot Leg Penetration 140 Figure 6.2-43 Sump B to Reactor Coolant Drain Tank Pump A Penetration 141 Figure 6.2-44 Sump B to Reactor Coolant Drain Tank Pump B Penetration 142 Figure 6.2-45 Reactor Coolant Drain Tank Discharge Penetration 143 Figure 6.2-46 Reactor Compartment Cooling Unit A Supply and Return Penetrations 201a and 209b Figure 6.2-47 Reactor Compartment Cooling Unit B Supply and Return Penetrations 201b and 209a Figure 6.2-48 Hydrogen Recombiner B (Main and Pilot) Penetrations 202a and 202b Figure 6.2-49 Containment Pressure Transmitters PT-947 and PT-948 Penetration 203a Figure 6.2-50 Figure DELETED Figure 6.2-51 Purge Supply Penetration 204 Figure 6.2-52 Reactor Coolant System Loop B Hot Leg Sample Penetration 205 Figure 6.2-53 Pressurizer Liquid Sample Penetration 206a Figure 6.2-54 Steam Generator A Sample Penetration 206b Figure 6.2-55 Pressurizer Steam Sample Penetration 207a Figure 6.2-56 Steam Generator B Sample Penetration 207b Figure 6.2-57 Hydrogen Recombiner A and B Oxygen Makeup Penetration 210 Figure 6.2-58 Purge Exhaust Penetration 300 Figure 6.2-59 Figure Deleted Figure 6.2-60 Hydrogen Recombiner A (Main and Pilot) Penetrations 304a and 304b Figure 6.2-61 Containment Postaccident Air Sample Penetrations 305a, 305c, and 305d Figure 6.2-62 Containment Air Sample (Return) Penetration 305b Figure 6.2-63 Containment Air Sample Outlet Penetration 305e Figure 6.2-64 Fire Service Water Penetration 307 Figure 6.2-65 Service Water for Containment Fan Coolers, Penetrations 308, 311, 312, 315, 316, 319, 320, and 323 Figure 6.2-66 Mini-Purge Supply Penetration 309 Figure 6.2-67 Instrument Air Penetration 310a Figure 6.2-68 Service Air Penetration 310b Figure 6.2-69 Leakage Test Depressurization Penetration 313 Figure 6.2-70 Leakage Test Supply Penetration 317 Figure 6.2-71 Steam Generator A Blowdown Penetration 321 Figure 6.2-72 Steam Generator B Blowdown Penetration 322 Page 13 of 14 Revision 26 5/2016
GINNA/UFSAR Figure 6.2-73 Demineralized Water Penetration 324 Figure 6.2-74 Containment H2 Monitors Penetrations 332a, 332b, and 332d Figure 6.2-75 Containment Pressure Transmitters PT-944, PT-949, and PT-950 Penetration 332c Figure 6.2-76 Main Steam from Steam Generator A Penetration 401 Figure 6.2-77 Main Steam from Steam Generator B Penetration 402 Figure 6.2-78 Main and Auxiliary Feedwater to Steam Generators A and B Penetrations 403 and 404 Figure 6.2-79 Figure DELETED Sheet 1 -
Figure 6.2-79 Figure DELETED Sheet 2 -
Figure 6.2-80 Containment Hydrogen Production With and Without Recombiner Figure 6.3-1 Sheet Figure DELETED 1-Figure 6.3-1 Sheet Figure DELETED 2-Figure 6.3-2 Safety Injection Pump Performance Characteristics Figure 6.3-3 Residual Heat Removal Pump Reactor Injection Capability Figure 6.3-4 Range of Core Protection Provided by Various Components of the Safety Injection Sys- tem Figure 6.5-1 Carbon Cell Banking Arrangement Figure 6.5-2 Filters - Containment Unit Figure 6.5-3 Iodine Partition Coefficient and pH in the Containment Versus Time Figure 6.5-4 Pressure Dependence of the Ratio vG/t Page 14 of 14 Revision 26 5/2016
GINNA/UFSAR 7 INSTRUMENTATION AND CONTROLS 1
7.1 INTRODUCTION
2 7.1.1 IDENTIFICATION OF SAFETY-RELATED SYSTEMS 2 7.1.2 IDENTIFICATION OF SAFETY CRITERIA 3 7.1.2.1 General Design Criteria 3 7.1.2.2 Compliance with IEEE 279-1971 3 7.1.2.2.1 Design Basis 3 7.1.2.2.2 Requirements 4 7.1.2.2.2.1 Operability 4 7.1.2.2.2.2 Testability 5 7.1.2.2.2.3 Control of Protective Actions 5 7.2 REACTOR TRIP SYSTEM (RTS) 8 7.2.1 DESIGN BASES 8 7.2.1.1 Design Criteria 8 7.2.1.1.1 Fuel Damage Limits 8 7.2.1.1.2 Reliability and Testability 9 7.2.1.1.3 Redundancy and Independence 10 7.2.1.1.4 Effects of Adverse Conditions 10 7.2.1.1.5 Testing While In Operation 11 7.2.1.1.6 Fail Safe Design 11 7.2.1.1.7 Single Failure Criterion 11 7.2.1.2 Seismic Design 12 7.2.1.3 Operating Environment 12 7.
2.2 DESCRIPTION
13 7.2.2.1 Logic Train 14 7.2.2.1.1 Sensors 14 7.2.2.1.2 Process and Nuclear Instrumentation 15 7.2.2.1.3 Protection Cabinets 15 7.2.2.1.4 Logic Relay Cabinets 15 7.2.2.1.5 Trip Breakers 16 7.2.2.2 Reactor Trips 17 7.2.2.2.1 General 17 7.2.2.2.2 Manual Trip 17 7.2.2.2.3 High-Nuclear-Flux (Power Range) Trip 17 7.2.2.2.4 High-Nuclear-Flux (Intermediate Range) Trip 18 Page 1 of 9 Revision 26 5/2016
GINNA/UFSAR 7 INSTRUMENTATION AND CONTROLS 1 7.2.2.2.5 High-Nuclear-Flux (Source Range) Trip 18 7.2.2.2.6 Overtemperature Delta T Trip 18 7.2.2.2.7 Overpower Delta T Trip 18 7.2.2.2.8 Low Pressurizer Pressure Trip 18 7.2.2.2.9 High Pressurizer Pressure Trip 19 7.2.2.2.10 High Pressurizer Water Level Trip 19 7.2.2.2.11 Low Reactor Coolant Flow Trip 19 7.2.2.2.12 Safety Injection System Actuation Trip 20 7.2.2.2.13 Turbine Trip/Reactor Trip 20 7.2.2.2.14 Low-Low Steam-Generator Water Level Trip 20 7.2.2.3 Interlocks 20 7.2.2.4 Permissive Circuits 21 7.2.2.4.1 P-1 Permissive 21 7.2.2.4.2 P-2 Permissive 21 7.2.2.4.3 P-3 Permissive 21 7.2.2.4.4 P-4 Permissive 21 7.2.2.4.5 P-6 Permissive 21 7.2.2.4.6 P-7 Permissive 22 7.2.2.4.7 P-8 Permissive 22 7.2.2.4.8 P-9 Permissive 22 7.2.2.4.9 P-10 Permissive 22 7.2.2.5 Alarms 22 7.2.2.6 Design Features 23 7.2.2.6.1 Isolation of Redundant Protection Channels 23 7.2.2.6.1.1 Channelized Design 23 7.2.2.6.1.2 Separation 24 7.2.2.6.2 Channel Bypass or Removal from Operation 24 7.2.2.6.3 Capability for Test and Calibration 24 7.2.2.6.4 Information Readout and Indication of Bypass 25 7.2.2.6.5 Physical Isolation 25 7.2.2.6.6 Sensor Line Separation 25 7.2.2.6.7 Instrument Line Identification 26 7.2.3 ANALYSIS 26 7.2.3.1 Reactor Trip System (RTS) and Departure From Nucleate Boiling 26 7.2.3.2 Core Protection System 26 Page 2 of 9 Revision 26 5/2016
GINNA/UFSAR 7 INSTRUMENTATION AND CONTROLS 1 7.2.3.2.1 Overpower Protection 26 7.2.3.2.2 Overtemperature Protection 27 7.2.4 REACTOR TRIP SIGNAL TESTING 27 7.2.4.1 Analog Channel Testing 27 7.2.4.2 Logic Channel Testing 28 7.2.4.2.1 Planned Tests 28 7.2.4.2.2 Test Procedure 29 7.2.4.2.3 Logic Channel Test Panels 29 7.2.4.3 Trip Breaker Testing and Preventive Maintenance 30 7.2.5 INTERACTION OF CONTROL AND PROTECTION SYSTEMS 30 7.2.5.1 Introduction 30 7.2.5.2 Specific Control and Protection Interactions 30 7.2.5.2.1 Nuclear Flux 30 7.2.5.2.2 Coolant Temperature 31 7.2.5.2.3 Pressurizer Pressure 31 7.2.5.2.4 Pressurizer Level 32 7.2.6 ANTICIPATED-TRANSIENT-WITHOUT-SCRAM MITIGATION 33 SYSTEM ACTUATION CIRCUITRY Table 7.2-1 Table DELETED 36 Table 7.2-2 PERMISSIVE CIRCUITS 37 Table 7.2-3 REACTOR TRIP FUNCTION SETPOINTS 38 7.3 ENGINEERED SAFETY FEATURES SYSTEMS 40 7.3.1 DESIGN CRITERIA 40 7.3.1.1 Protection Systems 40 7.3.1.2 Redundancy and Independence 41 7.3.1.3 Testing While In Operation 42 7.3.1.4 Fail Safe Design 42 7.3.2 SYSTEM DESCRIPTION 43 7.3.2.1 Initiating Circuitry 43 7.3.2.2 System Functions 44 7.3.2.2.1 Steam Line Isolation 44 7.3.2.2.2 Feedwater Line Isolation 44 7.3.2.3 Sensing and Display Instrumentation 45 7.3.2.3.1 Reactor Vessel Level Indication System 45 7.3.2.3.2 Containment Pressure 45 7.3.2.3.3 Containment Sump Level 46 Page 3 of 9 Revision 26 5/2016
GINNA/UFSAR 7.3.2.3.4 Accumulator Level and Pressure 46 7.3.2.3.5 Refueling Water Storage Tank Level (RWST) 46 7.3.2.3.6 Sodium Hydroxide Tank Level and Flow 46 7.3.2.3.7 Safety Injection Pumps Discharge Pressure and Flow 46 7.3.2.3.8 Residual Heat Removal (Low-Head Safety Injection) Flow 46 7.3.2.3.9 Pump Energization 46 7.3.2.3.10 Valve Position 46 7.3.2.3.11 Residual Heat Exchangers 47 7.3.2.3.12 Alarms 47 7.3.2.3.13 Air Coolers 47 7.3.2.3.14 Local Instrumentation 47 7.3.2.4 Engineered Safety Features Reset Controls 47 7.3.3 DESIGN EVALUATION 48 7.3.3.1 Engineered Safety Features Systems Isolation 48 7.3.3.2 Loss of Voltage or Degraded Voltage on Engineered Safety Features Bus 48 7.3.4 TESTING 48 7.3.4.1 Analog Channel Testing 48 7.3.4.2 Logic Channel Testing 49 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 51 7.
4.1 DESCRIPTION
51 7.4.1.1 Reactor Trip System (RTS) 51 7.4.1.2 Auxiliary Feedwater Systems 52 7.4.1.3 Main Steam System 52 7.4.1.4 Service Water System 53 7.4.1.5 Chemical and Volume Control System 53 7.4.1.6 Component Cooling Water System (CCW) 54 7.4.1.7 Residual Heat Removal System 54 7.4.1.8 Electrical Instrumentation and Power Systems 55 7.4.2 EVALUATION 55 7.4.3 EMERGENCY SHUTDOWN CONTROL 55 7.4.3.1 General 55 7.4.3.2 Residual Heat Removal 56 7.4.3.3 Reactivity Control 57 7.4.3.4 Pressurizer Pressure and Level Control 57 7.4.3.5 Electrical Systems 57 7.4.3.6 Startup of Other Equipment 58 7.4.3.7 Indication and Controls Provided Outside the Control Room 58 7.4.3.7.1 Local Panel Indication 58 7.4.3.7.2 Local Motor Controls 59 Page 4 of 9 Revision 26 5/2016
GINNA/UFSAR 7.4.3.7.3 Valve Control 59 7.4.3.7.4 Pressurizer Heater Control 60 7.4.3.7.5 Lighting 60 7.4.3.7.6 Communications 60 7.4.3.7.7 Electrical Systems 60 7.4.4 ALTERNATIVE SHUTDOWN SYSTEM 60 7.4.4.1 System Description 60 7.4.4.2 Alternative Shutdown Stations 61 7.4.4.2.1 Charging Pump Room (Primary Station) (see Section 7.4.3.7.1 F) 61 7.4.4.2.2 Intermediate Building North (Primary Station) (see Section 7.4.3.7.1 E) 62 7.4.4.2.3 Emergency Diesel Generator Area (Support Station) (see Section 7.4.3.7.7) 62 7.4.4.2.4 480-Volt Alternating Current Bus 14 (Support Station) 62 7.4.4.2.5 Battery Rooms 1A and 1B (Support Station) 62 7.4.4.2.6 Motor Control Centers 1C and 1D (Support Station) 62 7.4.4.2.7 480-Volt Alternating Current Bus 18 (Support Station) 63 7.4.4.2.8 Selected Safe Shutdown Systems 63 Table 7.4-1 FUNCTIONS FOR SHUTDOWN AND COOLDOWN 65 Table 7.4-2 SAFE SHUTDOWN INSTRUMENTS 66 Table 7.4-3 SAFE SHUTDOWN SYSTEMS POWER SOURCE AND LOCATION 69 Table 7.4-4 APPENDIX R ALTERNATIVE SHUTDOWN METHODS AND 71 CONTROL LOCATIONS 7.5 SAFETY-RELATED DISPLAY INSTRUMENTATION 74 7.5.1 CONTROL ROOM 74 7.5.1.1 Description 74 7.5.1.1.1 General 74 7.5.1.1.2 Main Control Board 74 7.5.1.1.3 Other Control Room Displays 75 7.5.1.2 Design Review 76 7.5.2 SAFETY PARAMETER DISPLAY 76 Table 7.5-1 COMPARISON OF GINNA STATION POSTACCIDENT 78 INSTRUMENTATION TO REGULATORY GUIDE 1.97, REVISION 3, CRITERIA 7.6 OTHER INSTRUMENTATION SYSTEMS REQUIRED FOR SAFETY 94 7.6.1 OVERPRESSURE PROTECTION DURING LOW POWER OPERATION 94 7.6.2 AUXILIARY FEEDWATER SYSTEM AUTOMATIC INITIATION AND 95 FLOW INDICATION 7.6.3 SUBCOOLING METER 95 7.6.4 DIRECT CURRENT POWER SYSTEM BUS VOLTAGE MONITORING 95 AND ANNUNCIATION Page 5 of 9 Revision 26 5/2016
GINNA/UFSAR 7 INSTRUMENTATION AND CONTROLS 1 7.6.5 REACTOR VESSEL LEVEL INDICATION SYSTEM 96 7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY 99 7.
7.1 DESCRIPTION
99 7.7.1.1 General 99 7.7.1.1.1 Reactor Control System 99 7.7.1.1.2 Steam Dump Control System 99 7.7.1.1.3 Reactivity Control 100 7.7.1.1.4 Reactor Control System Operation 100 7.7.1.1.5 Pressurizer Pressure and Water Level Control System 101 7.7.1.1.6 Steam Dump System 101 7.7.1.2 Rod Control System 102 7.7.1.2.1 Control Group Control 102 7.7.1.2.1.1 General 102 7.7.1.2.1.2 Rod Control Input Signals 102 7.7.1.2.1.3 Rod Control Program 103 7.7.1.2.2 Shutdown Group Control 104 7.7.1.2.3 Control Rod Drive Performance 104 7.7.1.2.4 Control Rod Power Supply System 104 7.7.1.2.4.1 General 104 7.7.1.2.4.2 Control Rod Power Supply Connections 105 7.7.1.2.5 Control Rod Power Supply Evaluation 106 7.7.1.2.5.1 Alternating Current Power Connections 106 7.7.1.2.5.2 Direct Current Power Connections 107 7.7.1.2.5.3 Evaluation Summary 108 7.7.1.2.6 Rod Position Indication System 109 7.7.1.2.6.1 Microprocessor System 109 7.7.1.2.6.2 Digital System 110 7.7.1.2.6.3 Actual Position Indication 110 7.7.1.2.6.4 Demand Position Indication 110 7.7.1.2.6.5 Rod Deviation Alarm 110 7.7.1.2.7 Pulse-to-Analog Converter 111 7.7.1.2.8 Interlocks and Rod Stops 111 7.7.1.2.9 Rod Insertion Limit Circuit 112 7.7.1.2.10 Rod Drop Protection 112 7.7.1.2.11 Asymmetric Rod Cluster Control Assembly Withdrawal 113 Page 6 of 9 Revision 26 5/2016
GINNA/UFSAR 7 INSTRUMENTATION AND CONTROLS 1 7.7.1.2.12 Rod Control Cabinet Cooling 114 7.7.1.3 Pressurizer Pressure and Level Control 114 7.7.1.3.1 Pressure Control 114 7.7.1.3.2 Level Control 115 7.7.1.4 Turbine Bypass 115 7.7.1.5 Steam Generator Level Control 116 7.7.1.6 Steam Generator Overfill Protection 118 7.7.2 CONTROL SYSTEM EVALUATION 118 7.7.2.1 Plant Stability 118 7.7.2.2 Step Load Changes Without Turbine Bypass 119 7.7.2.3 Loading and Unloading 119 7.7.2.4 Loss of Load With Turbine Bypass 120 7.7.2.5 Turbine Trip With Reactor Trip 120 7.7.2.6 Control Rod Misalignment 121 7.7.2.6.1 General 121 7.7.2.6.2 Consequences of Rod Misalignment 121 7.7.2.6.3 Analysis of Control Rod Misalignment 122 7.7.2.6.4 Redundant Checks for Control Rod Malfunction 122 7.7.2.6.4.1 Operator Checks 122 7.7.2.6.4.2 Additional Periodic Tests 123 7.7.2.6.4.3 Details of Instrumentation System 123 7.7.2.6.4.4 Power Range Nuclear Instrumentation 123 7.7.2.6.4.5 Thermocouples 124 7.7.2.6.4.6 In-Core Movable Detectors 124 7.7.2.6.4.7 Summary 125 7.7.2.6.5 Expected Instrument Response to Control Rod Misalignment Ginna Station 125 7.7.2.6.6 Plant Startup Tests 126 7.7.3 NUCLEAR INSTRUMENTATION SYSTEM 126 7.7.3.1 Design Basis 126 7.7.3.2 System Design 127 7.7.3.2.1 Source Range Description 128 7.7.3.2.2 Intermediate Range Description 129 7.7.3.2.3 Power Range Description 130 7.7.3.2.4 Dropped Rod Protection 131 7.7.3.2.5 Audio Count Rate Channel 131 Page 7 of 9 Revision 26 5/2016
GINNA/UFSAR 7 INSTRUMENTATION AND CONTROLS 1 7.7.3.2.6 Recorders 131 7.7.3.2.7 Power Supply 132 7.7.3.2.8 Equipment Locations 132 7.7.3.3 System Evaluation 132 7.7.4 IN-CORE INSTRUMENTATION 132 7.7.4.1 Design Basis 132 7.7.4.2 System Design 132 7.7.4.2.1 General 132 7.7.4.2.2 Thermocouples 133 7.7.4.2.3 Movable Miniature Neutron Flux Detectors 133 7.7.4.2.4 Control and Readout System 134 7.7.5 REACTOR COOLANT TEMPERATURE INDICATION 135 7.7.6 PLANT PROCESS COMPUTER SYSTEM AND SAFETY ASSESSMENT SYS- 136 TEM 7.7.6.1 General 136 7.7.6.2 Plant Process Computer System 137 7.7.6.3 Safety Parameter Display System 138 Table 7.7-1 OUT-OF-PHASE CURRENTS (AMPS) 140 Table 7.7-2 ROD STOPS 141 Table 7.7-3 EXPECTED MAXIMUM VARIATIONS BETWEEN 142 SYMMETRICALLY LOCATED DETECTORS FIGURES Figure 7.2-1 Reactor Protection Systems Figure 7.2-2 Figure DELETED Figure 7.2-3 Figure DELETED Figure 7.2-4 Figure DELETED Figure 7.2-5 Figure DELETED Figure 7.2-6 Figure DELETED Figure 7.2-7 Figure DELETED Figure 7.2-8 Figure DELETED Figure 7.2-9 Figure DELETED Figure 7.2-10 Figure DELETED Figure 7.2-11 Figure DELETED Figure 7.2-12 Design Philosophy to Achieve Isolation Between Channels Figure 7.2-13 Figure DELETED Figure 7.2-14 Channel Configuration (Channel 1 Typical) Tavg / T Control and Protection System Page 8 of 9 Revision 26 5/2016
GINNA/UFSAR Figure 7.2-15 Analog System Symbols Figure 7.2-16 Analog Channel Testing Arrangement Figure 7.2-17 Trip Logic Channels Figure 7.2-18 Analog Channels Figure 7.2-19 Logic Channel Test Panels Figure 7.2-20 Electrical Diagram - Undervoltage Coil and Shunt Trip Assembly Figure 7.3-1 Sheet 1 - Figure DELETED Figure 7.3-1 Sheet 2 - Figure DELETED Figure 7.3-2 Figure DELETED Figure 7.3-3 Figure DELETED Figure 7.3-4 Actuation Circuits of Engineered Safety Features Figure 7.3-5 Figure DELETED Figure 7.3-6 Analog and Logic Channel Testing Figure 7.5-1 Control Room Layout Figure 7.7-1 Reactor Control System Figure 7.7-2 Simplified Block Diagram of Reactor Control Systems Figure 7.7-3 Control Group - Rod Drive System Figure 7.7-4 Power Supply to Rod Control Equipment and Control Rod Drive Mechanisms Figure 7.7-4a Illustration of MRPI Indication Figure 7.7-5 Figure DELETED Figure 7.7-6 Nuclear Protection System Figure 7.7-7 Power Range Nuclear Detector Locations Figure 7.7-8 Location of Control Rods and Instrumentation Figure 7.7-9 Power Range Nuclear Instrumentation System Figure 7.7-10 Neutron Detectors and Range of Operation Figure 7.7-11 In-Core Instrumentation - Assembly Figure 7.7-12 Sheet 1 - In-Core Instrumentation, Details Figure 7.7-12 Sheet 2 - In-Core Instrumentation, Details Figure 7.7-13 Typical Arrangement of Moveable Miniature Neutron Flux Detector System Figure 7.7-14 Advanced Digital Feedwater Control System Input Signal Validation Figure 7.7-15 Advanced Digital Feedwater Control System Flow Controller and Cv Demand Figure 7.7-16 Advanced Digital Feedwater Control System Valve Sequence and Tracking Logic Page 9 of 9 Revision 26 5/2016
GINNA/UFSAR 8 ELECTRIC POWER 1
8.1 INTRODUCTION
2 8.1.1 GENERAL 2 8.1.2 OFFSITE POWER DESCRIPTION 2 8.1.3 ONSITE POWER DESCRIPTION 3 8.1.4 PRINCIPAL DESIGN CRITERIA 3 8.1.4.1 Performance Standards 3 8.1.4.2 Emergency Power 4 8.1.4.3 Adequacy of Electrical Design Relative to 1972 Criteria 5 8.1.4.4 Potential Risk of Station Blackout 6 8.1.4.5 Station Blackout Program 7 8.1.4.5.1 Assumptions 7 8.1.4.5.2 Ventilation 8 8.1.4.5.3 Plant Classification 8 8.1.4.5.4 Diesel Generator Reliability 8 8.1.4.5.5 Diesel Generator Cold Starts 9 8.1.4.6 Fukushima - Diverse and Flexible Coping Strategies (FLEX) 9 8.1.4.6.1 Regulatory Requirements 9 8.1.4.6.2 Site Response to NRC Orders 9 8.1.4.6.3 Summary of FLEX Strategies 10 8.1.4.6.3.1 Reactor Core Cooling and Heat Removal 10 8.1.4.6.3.2 RCS Inventory and Reactivity Control 10 8.1.4.6.3.3 Containment Integrity 11 8.1.4.6.3.4 Spent Fuel Pool Cooling 12 8.1.4.6.3.5 Electric Power 12 8.2 OFFSITE POWER SYSTEM 15 8.
2.1 DESCRIPTION
15 8.2.1.1 Transmission System 15 8.2.1.1.1 Step-up Transformers 15 8.2.1.1.2 Transmission Lines 15 8.2.1.1.3 Circuit Breakers 16 Page 1 of 4 Revision 26 5/2016
GINNA/UFSAR 8.2.1.1.4 Protective Relay Circuits 16 8.2.1.2 Station Auxiliary (Startup) Transformers 12A and 12B 17 8.2.2 ANALYSIS 19 8.2.2.1 Transmission System 19 8.2.2.1.1 Loss of Ginna Station Output 19 8.2.2.1.2 Switchyard Direct Current Power System 19 8.2.2.1.3 Transmission Network Protective Features 19 8.2.2.1.4 Northeast Power Coordinating Council Load-Shedding Practice 20 8.2.2.2 Station Auxiliary (Startup) Transformers 12A and 12B 21 8.2.2.2.1 Original Ginna Station Design 21 8.2.2.2.2 Transformer Failure Rates 21 8.2.2.2.3 Backup Auxiliary Transformers 22 8.2.2.3 Radiation Exposure During Restoration of Power 23 Table 8.2-1 DIRECT RADIATION DOSE RATES1 (REM/HR) 25 Table 8.2-2 INHALATION DOSE RATES2 (REM/SEC) 26 8.3 ONSITE POWER SYSTEM 27 8.3.1 ALTERNATING CURRENT POWER SYSTEM 27 8.3.1.1 Description 27 8.3.1.1.1 Single-Line Diagrams 27 8.3.1.1.2 Station Unit Transformer 27 8.3.1.1.3 The 4160-Volt System 28 8.3.1.1.4 The 480-Volt System 28 8.3.1.1.4.1 480-Volt Buses 28 8.3.1.1.4.2 Class 1E Trains 29 8.3.1.1.5 The 120-Volt Alternating Current System 30 8.3.1.1.5.1 Instrument Bus 1A 30 8.3.1.1.5.2 Instrument Bus 1B 31 8.3.1.1.5.3 Instrument Bus 1C 31 8.3.1.1.5.4 Instrument Bus 1D 31 8.3.1.1.6 Emergency Power 32 8.3.1.1.6.1 Emergency Power Sources 32 Page 2 of 4 Revision 26 5/2016
GINNA/UFSAR 8.3.1.1.6.2 Diesel-Generator Rapid Startup and Loading 32 8.3.1.1.6.3 Diesel-Generator Protective Trips 33 8.3.1.1.6.4 Fuel Oil Supply 33 8.3.1.1.6.5 Diesel-Generator Startup Logic 34 8.3.1.1.6.6 Emergency Power Supply 34 8.3.1.1.6.7 Alternative Shutdown Provisions 36 8.3.1.1.6.8 Regulatory Review of Diesel-Generator Capability 36 8.3.1.2 Analysis 36 8.3.1.2.1 Evaluation of Layout and Load Distribution 36 8.3.1.2.2 Diesel Generators 37 8.3.1.2.3 Normal Power Sources 37 8.3.1.2.4 Reliability Assurance 37 8.3.1.2.4.1 Redundancy 37 8.3.1.2.4.2 Sequencing Circuits 37 8.3.1.2.4.3 Sequencing Relays 38 8.3.1.2.4.4 Engineered Safety Features Actuation 39 8.3.1.2.4.5 Separation 39 8.3.1.2.4.6 Fuse Coordination 40 8.3.1.2.4.7 Overload and Short Circuit Protection 40 8.3.1.2.5 Instrument Bus Evaluation 40 8.3.1.2.6 Loss of Offsite Power Under Accident Conditions 41 8.3.1.2.6.1 Operator Actions 41 8.3.1.2.6.2 Reliability Assurance 42 8.3.1.2.7 Degraded Grid Voltage 43 8.3.1.2.7.1 Susceptibility to Degraded Grid Voltage Conditions 43 8.3.1.2.7.2 Adequacy of Onsite Power System Voltages 44 8.3.1.3 Containment Electrical Penetrations 44 8.3.1.4 Independence of Redundant Systems 45 8.3.1.4.1 Criteria Relating to Cable-Tray Loading and Separation 45 8.3.1.4.2 Separation of Redundant Circuits 46 8.3.1.4.3 Quality Assurance 47 Page 3 of 4 Revision 26 5/2016
GINNA/UFSAR 8.3.2 DIRECT CURRENT POWER SYSTEMS 47 8.3.2.1 Description 47 8.3.2.1.1 Direct Current System 47 8.3.2.1.2 Battery Room 48 8.3.2.1.3 Battery Chargers 48 8.3.2.1.4 Technical Support Center Battery 49 8.3.2.2 Analysis 49 8.3.2.3 Direct Current Fuse Coordination 50 8.3.3 FIRE PROTECTION FOR CABLE SYSTEMS 52 Table 8.3-1a ENGINEERED SAFETY FEATURES ACTUATION (ESFAS) 56 SEQUENCE ACTION (TRAIN A)
Table 8.3-1b ENGINEERED SAFETY FEATURES ACTUATION (ESFAS) 57 SEQUENCE ACTION - DIESEL GENERATOR LOADING Table 8.3-2a DIESEL GENERATOR LOADING (TRAIN A) 58 Table 8.3-2b DIESEL GENERATOR LOADING (TRAIN B) 59 Table 8.3-3 CONTAINMENT ELECTRICAL PENETRATIONS 60 Table 8.3-4 MAJOR BATTERY LOADS 62
- 1. At listed locations following a loss-of-coolant accident.
- 2. Within 200 ft of containment following a loss-of-coolant accident.
FIGURES Figure 8.1-1 Electrical Distribution System Figure 8.2-1 Transmission Connections Page 4 of 4 Revision 26 5/2016
GINNA/UFSAR 9 AUXILIARY SYSTEMS 1 9.1 FUEL STORAGE AND HANDLING 2 9.1.1 NEW FUEL STORAGE 2 9.1.2 SPENT FUEL STORAGE 2 9.1.2.1 Design Criteria 3 9.1.2.1.1 General 3 9.1.2.1.2 Effective Multiplication Factor 4 9.1.2.1.3 Protection Against Damage 4 9.1.2.1.4 Storage Capacity 4 9.1.2.1.5 Fuel Pool Cooling System Instrumentation 5 9.1.2.1.6 Seismic Design 5 9.1.2.1.7 Fuel Handling System 6 9.1.2.1.8 Minimum Center-to-Center Spacing 6 9.1.2.1.9 Stability of Fuel Storage Racks 6 9.1.2.1.10 Fuel Pool Leakage Prevention 7 9.1.2.1.11 Depth of Water Over Fuel 7 9.1.2.1.12 Fixed Neutron Poisons 7 9.1.2.1.13 Bearing Loads on Pool Liner 8 9.1.2.2 Description 8 9.1.2.2.1 Spent Fuel Pool (SFP) 8 9.1.2.2.2 Spent Fuel Storage Racks 9 9.1.2.3 Design Evaluation 10 9.1.2.4 Nuclear Analysis 11 9.1.2.4.1 Methods of Analysis 11 9.1.2.4.1.1 Criticality Methodology 14 9.1.2.4.1.2 Criticality Analysis of Consolidated Rod Storage Canisters in Spent 16 Fuel Racks 9.1.2.4.1.3 Summary of Criticality Results 17 9.1.2.4.2 Accident Analysis 19 9.1.2.4.2.1 Fresh Fuel Storage Racks 19 Page 1 of 15 Revision 26 4/2016
GINNA/UFSAR 9.1.2.4.2.2 Spent Fuel Storage Racks 19 9.1.2.5 Thermal-Hydraulic Analysis 20 9.1.2.6 Radiological Evaluation 20 9.1.2.7 Radiological Consequences of Tornado Missile Accident (TMA) 21 9.1.2.8 Radiological Consequences of a Dropped Consolidated Canister 21 9.1.3 SPENT FUEL POOL COOLING 22 9.1.3.1 Design Bases 22 9.1.3.2 System Design and Operation 24 9.1.3.2.1 System Design 24 9.1.3.2.2 System Operation 24 9.1.3.2.3 Suction Lineup Using Spent Fuel Pool (SFP) Pump A 25 9.1.3.3 Spent Fuel Pool (SFP) Cooling System Components 25 9.1.3.3.1 Spent Fuel Pool (SFP) Heat Exchangers 26 9.1.3.3.2 Spent Fuel Pool (SFP) Pumps 26 9.1.3.3.3 Spent Fuel Pool (SFP) Filter 26 9.1.3.3.4 Spent Fuel Pool (SFP) Strainer 26 9.1.3.3.5 Spent Fuel Pool (SFP) Demineralizer 26 9.1.3.3.6 Spent Fuel Pool (SFP) Skimmer 26 9.1.3.3.7 Spent Fuel Pool (SFP) Valves 26 9.1.3.3.8 Spent Fuel Pool (SFP) Piping 27 9.1.3.4 System Evaluation 27 9.1.3.4.1 Thermal-Hydraulic Analysis 27 9.1.3.4.1.1 Heat Removal Requirements 27 9.1.3.4.1.2 Service Water Temperature 27 9.1.3.4.1.3 Analysis of Heat Removal System 28 9.1.3.4.1.4 Cooling Water Flow in Fuel Pool 28 9.1.3.4.1.5 Cooling Analysis of Individual Fuel Assemblies 28 9.1.3.4.1.6 Cooling Analysis of Consolidated Fuel Canisters 30 9.1.3.4.1.7 Normal 1/3 Core Off-Load Cooling Capability 31 9.1.3.4.1.8 Full Core Off-Load Cooling Capability 32 9.1.3.4.2 Leakage Provisions 33 Page 2 of 15 Revision 26 4/2016
GINNA/UFSAR 9.1.3.4.3 Interruption of Spent Fuel Pool (SFP) Cooling 34 9.1.3.5 Minimum Operating Conditions 36 9.1.4 FUEL HANDLING SYSTEMS 36 9.1.4.1 Reactor Cavity 36 9.1.4.2 Refueling Canal 37 9.1.4.3 Fuel Handling Equipment 37 9.1.4.3.1 Auxiliary Building Crane 37 9.1.4.3.2 New Fuel Elevator 38 9.1.4.3.3 Spent Fuel Pool (SFP) Bridge 38 9.1.4.3.4 Fuel Transfer System 38 9.1.4.3.5 Manipulator Crane 39 9.1.4.3.6 Reactor Vessel Head Lifting Device 40 9.1.4.3.7 Reactor Internals Lifting Device 41 9.1.4.3.8 Rod Cluster Control Assembly Changing Fixture 41 9.1.4.3.9 Upper Internals Storage Stand 42 9.1.4.3.10 125 Ton Cask Handling Crane 42 9.1.4.4 Fuel Handling/Refueling Tools 42 9.1.4.4.1 New Fuel Assembly Handling Tool 42 9.1.4.4.2 Spent Fuel Handling Tool 42 9.1.4.4.2.1 Spent Fuel Cask Loading Tool (SFCLT) 43 9.1.4.4.3 Burnable Poison Rod Assembly Handling Tool 43 9.1.4.4.4 Control Rod Drive Shaft Tool 44 9.1.4.4.5 Thimble Plug Handling Tool 44 9.1.4.4.6 Irradiation Sample Handling Tool 44 9.1.4.4.7 Stud Tensioners 45 9.1.4.4.8 Portable Rod Cluster Control Assemble (RCCA) Tool 45 9.1.4.5 Fuel Handling System Operation During MODE 6 (Refueling) 45 9.1.4.5.1 Introduction 45 9.1.4.5.2 Preparation Phase 45 9.1.4.5.3 MODE 6 (Refueling) Phase 46 9.1.4.5.4 Reactor Reassembly 47 Page 3 of 15 Revision 26 4/2016
GINNA/UFSAR 9.1.4.6 Fuel Handling System Evaluation 47 9.1.4.6.1 Incident Protection 48 9.1.4.6.2 Malfunction Analysis 48 9.1.4.7 Minimum Operating Conditions 48 9.1.4.8 Tests and Inspections 48 9.1.5 CONTROL OF HEAVY LOADS 48 9.1.5.1 Conduct of Heavy Loads Movements 49 Table 9.1-1 FUEL PARAMETERS EMPLOYED IN THE CRITICALITY 57 ANALYSIS Table 9.1-2 TORNADO MISSILE ACCIDENT DOSE ANALYSIS 58 ASSUMPTIONS Table 9.1-2 TORNADO MISSILE ACCIDENT DOSE ANALYSIS 59 ASSUMPTIONS (OFFSITE X/Q)
Table 9.1-2 TORNADO MISSILE ACCIDENT DOSE ANALYSIS 60 ASSUMPTIONS (OFFSITE BREATHING RATES)
Table 9.1-3 SPENT FUEL POOL (SFP) COOLING SYSTEM RATING 61 Table 9.1-4 SPENT FUEL POOL (SFP) COOLING SYSTEM COMPONENT 62 DATA Table 9.1-5 OFFSITE AND CONTROL ROOM DOSES FOR THE SPENT FUEL 64 POOL TORNADO MISSILE ACCIDENT Table 9.1-6 HEAT-UP TIMES ASSOCIATED WITH LOSS OF SPENT FUEL 65 POOL COOLING 9.2 WATER SYSTEMS 66 9.2.1 SERVICE WATER (SW) SYSTEM 66 9.2.1.1 Design Bases 66 9.2.1.2 Description 66 9.2.1.2.1 General Description 66 9.2.1.2.2 Service Water System Design 68 9.2.1.2.3 Service Water System Initiation on Loss of Offsite Power 68 9.2.1.2.4 Containment Cooling Coils 69 9.2.1.2.5 Radiation Monitors 70 9.2.1.2.6 Service Water Fouling 70 9.2.1.3 Design Evaluation 71 9.2.1.4 Postaccident Conditions 72 9.2.1.4.1 Recirculation Phase 72 9.2.1.4.2 Limiting Steam Line Break Events 73 9.2.1.4.3 Accident Considerations With Offsite Power Available 73 9.2.1.4.4 Postulated Service Water Pump Discharge Check Valve Failure 74 Page 4 of 15 Revision 26 4/2016
GINNA/UFSAR 9.2.1.5 Tests and Inspections 74 9.2.2 COMPONENT COOLING WATER (CCW) SYSTEM 75 9.2.2.1 Design Bases 75 9.2.2.2 System Design and Operation 75 9.2.2.3 Component Description 77 9.2.2.4 System Evaluation 78 9.2.2.4.1 Availability and Reliability 78 9.2.2.4.1.1 Accessibility 78 9.2.2.4.1.2 Seismic Design 78 9.2.2.4.1.3 Loss of Component Cooling Water System 78 9.2.2.4.1.4 Component Cooling Water Surge Tank 79 9.2.2.4.1.5 Safety-Related Functions 79 9.2.2.4.1.6 Flow-Induced Vibration 80 9.2.2.4.2 Leakage Provisions 81 9.2.2.4.2.1 Introduction 81 9.2.2.4.2.2 Leakage Detection 81 9.2.2.4.2.3 Relief Valves 82 9.2.2.4.3 Incident Control 82 9.2.2.4.4 Malfunction Analysis 83 9.2.2.5 Instrumentation Requirements 84 9.2.2.6 Minimum Operating Conditions 84 9.2.2.7 Tests and Inspections 84 9.2.3 DEMINERALIZED WATER MAKEUP SYSTEM 84 9.2.4 CONDENSATE STORAGE FACILITIES 85 Table 9.2-1 LOADS SUPPLIED BY SERVICE WATER (SW) SYSTEM 88 Table 9.2-2 MAJOR SERVICE WATER SYSTEM FLOWS 90 Table 9.2-3 COMPONENT COOLING LOOP COMPONENT DATA 93 Table 9.2-4 FAILURE ANALYSIS OF PUMPS, HEAT EXCHANGERS, AND 95 VALVES Table 9.2-5 MINIMUM ALLOWED COMPONENTS FOR THE COMPONENT 96 COOLING WATER (CCW) SYSTEM 9.3 PROCESS AUXILIARIES 97 9.3.1 INSTRUMENT AND SERVICE AIR SYSTEMS 97 9.3.1.1 System Description 97 9.3.1.2 Component Description 98 Page 5 of 15 Revision 26 4/2016
GINNA/UFSAR 9.3.1.2.1 Compressors 98 9.3.1.2.2 Aftercoolers 98 9.3.1.2.3 Air Receivers 99 9.3.1.2.4 Filters and Dryers 99 9.3.2 SAMPLING SYSTEMS 100 9.3.2.1 Nuclear Sampling System 100 9.3.2.1.1 Design Bases 100 9.3.2.1.1.1 Functional Requirements 100 9.3.2.1.1.2 Operational Requirements 101 9.3.2.1.2 System Design and Operation 101 9.3.2.1.2.1 Sampling System 101 9.3.2.1.2.2 Reactor Coolant Samples 102 9.3.2.1.2.3 Chemical and Volume Control System Samples 102 9.3.2.1.2.4 Steam Generator Liquid Samples 102 9.3.2.1.2.5 Sample Sink 103 9.3.2.1.2.6 Instrumentation 103 9.3.2.1.2.7 Steam Generator Blowdown 103 9.3.2.1.3 Component Description 103 9.3.2.1.3.1 Sample Heat Exchangers 103 9.3.2.1.3.2 Delay Coil 103 9.3.2.1.3.3 Sample Pressure Vessels 103 9.3.2.1.3.4 Sample Sink 104 9.3.2.1.3.5 Piping and Fittings 104 9.3.2.1.3.6 Instrumentation 104 9.3.2.1.3.7 Valves 104 9.3.2.1.4 System Evaluation 105 9.3.2.1.4.1 Availability and Reliability 105 9.3.2.1.4.2 Leakage Provisions 105 9.3.2.1.4.3 Malfunction Analysis 105 9.3.2.1.5 Minimum Operating Conditions 105 9.3.2.1.6 Tests and Inspections 105 9.3.2.2 Nonnuclear Sampling System 105 9.3.2.2.1 Steam Generator Blowdown Sampling 105 9.3.2.2.2 Hotwell Sampling 106 Page 6 of 15 Revision 26 4/2016
GINNA/UFSAR 9.3.2.2.3 Condensate Sampling 106 9.3.2.2.4 Feedwater Sampling 106 9.3.2.2.5 Main Steam Sampling 106 9.3.2.2.6 Heater Drain Tank Sampling 106 9.3.2.2.7 Sampling Cooling 106 9.3.2.3 Postaccident Sampling System 107 9.3.2.3.1 Design Bases 107 9.3.2.3.2 System Description 108 9.3.2.3.3 Component Description and Operation 108 9.3.2.3.3.1 Liquid and Gas Sample Panel 108 9.3.2.3.3.2 Gas Sampling 109 9.3.2.3.3.3 Liquid Sampling 110 9.3.2.3.3.4 Instrument Panel 111 9.3.2.3.3.5 Electrical Control Panel 111 9.3.2.3.3.6 Postaccident Sampling System Coolers 111 9.3.2.3.3.7 Postaccident Sampling System Waste Tank 112 9.3.2.3.3.8 Postaccident Sampling System Waste Transfer Pump 112 9.3.2.3.3.9 Postaccident Sampling System Waste Tank Evacuating Compressor 112 9.3.2.3.3.10 Containment Sump A Sample Pump 113 9.3.3 EQUIPMENT AND FLOOR DRAINS SYSTEMS 113 9.3.4 CHEMICAL AND VOLUME CONTROL SYSTEM 113 9.3.4.1 Design Bases 113 9.3.4.1.1 Redundancy of Reactivity Control 113 9.3.4.1.2 Reactivity Holddown Capability 114 9.3.4.1.3 Reactivity Hot Shutdown Capability 114 9.3.4.1.4 Reactivity Shutdown Capability 115 9.3.4.1.5 Codes and Classifications 115 9.3.4.2 System Design and Operation 115 9.3.4.2.1 General 115 9.3.4.2.2 Letdown and Charging Systems 116 9.3.4.2.2.1 General 116 9.3.4.2.2.2 Charging Pump Control 117 9.3.4.2.3 Seal-Water Injection System 118 9.3.4.2.4 Reactor Makeup Control System 118 9.3.4.2.4.1 System Description 118 Page 7 of 15 Revision 26 4/2016
GINNA/UFSAR 9.3.4.2.4.2 Automatic Makeup 120 9.3.4.2.4.3 Dilution 120 9.3.4.2.4.4 Boration 120 9.3.4.2.5 Boron Recycle System 121 9.3.4.2.5.1 System Description 121 9.3.4.2.5.2 Alarm Functions 122 9.3.4.2.6 Heat Tracing System 122 9.3.4.3 Component Description 123 9.3.4.3.1 Letdown and Charging Systems 123 9.3.4.3.1.1 Regenerative Heat Exchanger 123 9.3.4.3.1.2 Letdown Orifices 123 9.3.4.3.1.3 Nonregenerative Heat Exchanger 123 9.3.4.3.1.4 Mixed-Bed Demineralizers 124 9.3.4.3.1.5 Cation Bed Demineralizer 124 9.3.4.3.1.6 Deborating Demineralizers 124 9.3.4.3.1.7 Resin Fill Tank 124 9.3.4.3.1.8 Reactor Coolant Filter 125 9.3.4.3.1.9 Volume Control Tank 125 9.3.4.3.1.10 Charging Pumps 125 9.3.4.3.1.11 Charging Pump Leakoff Tank 126 9.3.4.3.1.12 Charging Pump Dampener 126 9.3.4.3.1.13 Excess Letdown Heat Exchanger 126 9.3.4.3.2 Seal-Water Injection System 126 9.3.4.3.2.1 Seal-Water Heat Exchanger 126 9.3.4.3.2.2 Seal-Water Filter 127 9.3.4.3.2.3 Seal-Water Injection Filters 127 9.3.4.3.3 Reactor Makeup Control System 127 9.3.4.3.3.1 Boric Acid Filter 127 9.3.4.3.3.2 Boric Acid Storage Tanks 127 9.3.4.3.3.3 Batching Tank 128 9.3.4.3.3.4 Boric Acid Tank Heaters 128 9.3.4.3.3.5 Boric Acid Transfer Pumps 128 9.3.4.3.3.6 Boric Acid Blender 128 9.3.4.3.3.7 Chemical Mixing Tank 128 9.3.4.3.3.8 Heat Tracing 129 Page 8 of 15 Revision 26 4/2016
GINNA/UFSAR 9.3.4.3.3.9 Reactor Makeup Water Pumps 129 9.3.4.3.3.10 Reactor Makeup Water Tank 129 9.3.4.3.4 Boron Recycle System 129 9.3.4.3.4.1 Holdup Tanks 129 9.3.4.3.4.2 Holdup Tank Recirculation Pump 129 9.3.4.3.4.3 Gas Stripper Feed Pumps 130 9.3.4.3.4.4 Base Removal Ion Exchanger 130 9.3.4.3.4.5 Cation Ion Exchanger 130 9.3.4.3.4.6 Ion Exchanger Filter 130 9.3.4.3.4.7 Gas Stripper Equipment 130 9.3.4.3.4.8 Boric Acid Evaporator Equipment 131 9.3.4.3.4.9 Evaporator Condensate Demineralizers 131 9.3.4.3.4.10 Condensate Filter 131 9.3.4.3.4.11 Concentrates Filter 131 9.3.4.3.4.12 Concentrates Holding Tank 132 9.3.4.3.4.13 Concentrates Holding Tank Transfer Pumps 132 9.3.4.3.4.14 Monitor Tanks 132 9.3.4.3.4.15 Monitor Tank Pump 132 9.3.4.3.5 Valves 132 9.3.4.3.6 Piping 133 9.3.4.4 System Evaluation 133 9.3.4.4.1 Availability and Reliability 133 9.3.4.4.2 Seismic Analysis 134 9.3.4.4.3 Leakage Prevention 134 9.3.4.4.4 Incident Control 134 9.3.4.4.5 Malfunction Analysis 135 9.3.4.4.5.1 System Failures 135 9.3.4.4.5.2 Inadvertent Dilution 135 9.3.4.4.5.3 Alternative Methods of Boration 135 9.3.4.4.5.4 Inadvertent Dilution of Boric Acid Storage Tanks 136 9.3.4.4.5.5 Loss of Seal Injection Water 136 9.3.4.4.6 Overpressurization Protection 136 9.3.4.4.6.1 Suction Lines 136 Page 9 of 15 Revision 26 4/2016
GINNA/UFSAR 9.3.4.4.6.2 Discharge Lines 137 9.3.4.4.7 Galvanic Corrosion 137 9.3.4.4.8 Control of Tritium 137 9.3.4.4.9 Reactor Coolant Activity Concentration Calculations 138 9.3.4.4.9.1 Computation Method 138 9.3.4.4.9.2 Tritium Production 140 9.3.4.4.9.3 Radioactivity Monitoring 140 9.3.4.4.9.4 Technical Specifications Limits 140 9.3.4.4.9.5 Tritium Limit 140 9.3.4.4.9.6 R.E. Ginna Normal Operation RCS and Secondary Coolant Sources 140 9.3.4.5 Minimum Operating Conditions 141 Table 9.3-1 NUCLEAR PROCESS SAMPLING SYSTEM CODE 143 REQUIREMENTS Table 9.3-2 NUCLEAR PROCESS SAMPLING SYSTEM COMPONENTS 144 Table 9.3-3 MALFUNCTION ANALYSIS OF NUCLEAR PROCESS 146 SAMPLING SYSTEM Table 9.3-4 POSTACCIDENT SAMPLING SYSTEM FUNCTIONAL 147 REQUIREMENTS Table 9.3-5 LIQUID AND GAS SAMPLE PANEL ANALYTICAL EQUIPMENT 148 REQUIREMENTS Table 9.3-6 CHEMICAL AND VOLUME CONTROL SYSTEM 149 PERFORMANCE PARAMETERS Table 9.3-7 PRINCIPAL COMPONENT DATA
SUMMARY
151 Table 9.3-8 MALFUNCTION ANALYSIS OF CHEMICAL AND VOLUME 153 CONTROL SYSTEM Table 9.3-9 REACTOR COOLANT SYSTEM EQUILIBRIUM ACTIVITIES 154 Table 9.3-10 ARAMETERS USED IN THE 1811 MWT UPRATE CALCULATION 156 OF REACTOR COOLANT FISSION PRODUCT ACTIVITIES Table 9.3-11a PARAMETERS USED IN THE CALCULATION OF TRITIUM 157 PRODUCTION IN THE REACTOR COOLANT - BASIC ASSUMPTIONS Table 9.3-11b CALCULATION OF TRITIUM PRODUCTION IN THE REACTOR 158 COOLANT Table 9.3-12 ANSI/ANS 18.1-1999 NORMAL SOURCE INPUT PARAMETERS 159 Table 9.3-12 ANSI/ANS 18.1-1999 NORMAL SOURCE INPUT PARAMETERS 160 Page 10 of 15 Revision 26 4/2016
GINNA/UFSAR 9.4 AIR CONDITIONING, HEATING, COOLING, AND VENTILATION SYSTEMS 162 9.4.1 CONTAINMENT VENTILATION SYSTEM 162 9.4.1.1 Design Bases 162 9.4.1.1.1 Design Objectives 162 9.4.1.1.2 Design Criteria 163 9.4.1.2 System Design 164 9.4.1.2.1 Introduction 164 9.4.1.2.2 Containment Recirculation Cooling and Filtration System 164 9.4.1.2.3 Control Rod Drive Mechanism Cooling System 165 9.4.1.2.4 Reactor Compartment Cooling System 166 9.4.1.2.5 Refueling Water Surface and Purge System 166 9.4.1.2.6 Containment Auxiliary Charcoal Filter System 166 9.4.1.2.7 Containment Post-accident Charcoal Filter System 167 9.4.1.2.8 Containment Shutdown Purge System 167 9.4.1.2.9 Containment Mini-Purge System 168 9.4.1.2.10 Penetration Cooling System 168 9.4.2 AUXILIARY BUILDING VENTILATION SYSTEM 169 9.4.2.1 Design Basis 169 9.4.2.2 System Design and Operation 169 9.4.2.2.1 System Design Objective 169 9.4.2.2.2 Charcoal Filter Circuit 169 9.4.2.2.3 System Operation 170 9.4.2.3 System Components 171 9.4.2.3.1 Auxiliary Building Air Handling Unit 171 9.4.2.3.2 Auxiliary Building Exhaust Fan 1C 171 9.4.2.3.3 Auxiliary Building Exhaust Fans 1A and 1B 171 9.4.2.3.4 Auxiliary Building Exhaust Fan 1G 171 9.4.2.3.5 Auxiliary Building Charcoal Filter Fans 1A and 1B 172 9.4.2.3.6 Penetration Cooling Fans 1A and 1B 172 9.4.2.3.7 Pump Area Coolers 172 9.4.2.3.8 Intermediate Building Supply and Exhaust Fans 172 9.4.2.3.9 Steam Isolation Dampers 173 9.4.2.4 System Evaluation 173 9.4.2.4.1 Effect of Loss of Cooling on Pumps and Valves 173 9.4.2.4.2 Revised Auxiliary Building Loss of Cooling Analysis 174 Page 11 of 15 Revision 26 4/2016
GINNA/UFSAR 9.4.2.4.2.1 AUXILIARY BUILDING TEMPERATURE WITH MINIMUM 175 SERVICE WATER FLOW 9.4.2.4.3 Effect of Loss of Offsite Power on Ventilation Flow 176 9.4.3 CONTROL ROOM AREA VENTILATION SYSTEM 176 9.4.4 SPENT FUEL POOL AREA VENTILATION SYSTEM 176 9.4.5 TURBINE BUILDING VENTILATION SYSTEM 176 9.4.6 SERVICE BUILDING VENTILATION SYSTEM 177 9.4.7 ALL-VOLATILE-TREATMENT BUILDING VENTILATION 177 SYSTEM 9.4.7.1 Introduction 177 9.4.7.2 Summary Description of the System 178 9.4.7.2.1 Compressor and Booster Pump Area Ventilation System 178 9.4.7.2.2 Demineralizer Area Ventilation System 179 9.4.7.2.3 Demineralizer Area Control Room System 179 9.4.7.2.4 Heating System 179 9.4.8 TECHNICAL SUPPORT CENTER VENTILATION SYSTEM 179 9.4.8.1 System Description 179 9.4.8.2 System Operation 180 9.4.8.2.1 Cooling Systems 180 9.4.8.2.2 Heating Systems 181 9.4.9 ENGINEERED SAFETY FEATURES VENTILATION SYSTEMS 181 9.4.9.1 Engineered Safety Features Equipment Ventilation and Cooling 182 9.4.9.2 Relay Room 182 9.4.9.3 Battery Rooms 183 9.4.9.4 Essential Auxiliary Systems 183 9.4.9.5 Diesel Generators 184 9.4.9.6 Standby Auxiliary Feedwater System (SAFW) 185 9.4.9.6.1 System Operation 185 9.4.9.6.2 Controls and Instrumentation 186 9.4.9.7 Post-accident Fan Coolers and Charcoal Filters 187 9.4.10 STATION HEATING STEAM SYSTEM 187 Table 9.4-1 CONTAINMENT VENTILATION SYSTEM PRINCIPAL 189 COMPONENT DATA
SUMMARY
9.5 OTHER AUXILIARY SYSTEMS 192 Page 12 of 15 Revision 26 4/2016
GINNA/UFSAR 9.5.1 FIRE PROTECTION 192 9.5.1.1 Design Basis Summary 192 9.5.1.1.1 Defense-in-Depth 192 9.5.1.1.2 NFPA 805 Performance Criteria 193 9.5.1.1.3 Codes of Record 194 9.5.1.1.4 Required Systems 194 9.5.1.1.5 Definition of Power Block Structures 195 9.5.1.2 System Design 195 9.5.1.2.1 General 195 9.5.1.2.2 Fire Detection and Signaling Systems 196 9.5.1.2.3 Fire Suppression Systems 197 9.5.1.2.3.1 Water Supply 197 9.5.1.2.3.2 Fire Pumps 197 9.5.1.2.3.3 Piping and Valves 198 9.5.1.2.3.4 Fire Hydrants 199 9.5.1.2.3.5 Yard Loop 199 9.5.1.2.3.6 Interior Hose Stations 199 9.5.1.2.3.7 Water Suppression Systems 200 9.5.1.2.3.8 Gas Suppression Systems 201 9.5.1.2.3.9 Portable Fire Extinguishers 202 9.5.1.2.3.10 Wet Chemical Suppression System 202 9.5.1.2.4 Other Design Considerations 202 9.5.1.2.4.1 Smoke Removal 202 9.5.1.2.4.2 Breathing Equipment 202 9.5.1.2.4.3 Control Building Ventilation 202 9.5.1.2.4.4 Reactor Coolant Pump Motor Oil Collection System 203 9.5.1.2.4.5 Floor Drains and Curbs 203 9.5.1.2.4.6 Lighting Systems 203 9.5.1.2.4.7 Communications 203 9.5.1.2.4.8 Electrical Cable Insulation 203 9.5.1.2.4.9 Fire Barriers 204 9.5.1.2.4.10 Electrical Cable Penetrations 205 9.5.1.2.4.11 Piping and Duct Penetrations 205 9.5.1.2.4.12 Cable Separation 205 Page 13 of 15 Revision 26 4/2016
GINNA/UFSAR 9.5.1.2.4.13 Spray Shields 206 9.5.1.2.4.14 Construction Joints 206 9.5.1.3 Safety Evaluation 206 9.5.1.4 Fire Protection Program Documentation, Configuration Control 206 and Quality Assurance 9.5.2 COMMUNICATIONS SYSTEMS 207 9.5.2.1 Public Address System 207 9.5.2.2 Telephone Systems 207 9.5.2.3 Radio Systems 208 9.5.2.4 Offsite Communications 208 9.5.2.5 Emergency Communications With the NRC 209 9.5.3 LIGHTING SYSTEMS 209 9.5.4 DIESEL GENERATOR FUEL OIL STORAGE AND TRANSFER 210 SYSTEM 9.5.5 DIESEL GENERATOR COOLING SYSTEM 211 9.5.6 DIESEL GENERATOR STARTING SYSTEM 211 9.5.7 DIESEL GENERATOR LUBRICATION SYSTEM 212 9.5.8 DIESEL GENERATOR COMBUSTION AIR INTAKE AND 212 EXHAUST Table 9.5-1 FIRE SERVICE WATER HOSE REEL LOCATIONS 215 Table 9.5-2 Power Block Buildings 217 FIGURES Figure 9.1-1 Fuel Handling Structures Figure 9.1-2 Figure DELETED Figure 9.1-3 Arrangement of Spent Fuel Storage Racks Figure 9.1-4 Figure DELETED Figure 9.1-5 Figure DELETED Figure 9.1-6 Figure DELETED Figure 9.1-7 Spent Fuel Pool Cooling Cycle Figure 9.1-8 Figure DELETED Figure 9.1-9 Figure DELETED Figure 9.1-10 Reactor Vessel Head Lifting Device Figure 9.1-11 Fuel Handling Devices Figure 9.3-1 Maximum Tritium Activity Released to Primary Coolant Figure 9.5-1 Figure DELETED Page 14 of 15 Revision 26 4/2016
GINNA/UFSAR Figure 9.5-2 Figure DELETED Figure 9.5-3 Figure DELETED Figure 9.5-4 Figure DELETED Figure 9.5-5 Figure DELETED Figure 9.5-6 Diesel Engine Lubricating Oil Systems (Simplified)
Page 15 of 15 Revision 26 4/2016
GINNA/UFSAR 10 STEAM AND POWER CONVERSION SYSTEM 1
10.1 INTRODUCTION
2 10.1.1
SUMMARY
DESCRIPTION 2 10.1.1.1 Functional Description 2 10.1.1.2 Radioactivity 3 10.1.1.3 Major Systems 3 10.1.2 DESIGN BASES 4 10.1.2.1 System Design 4 10.1.2.2 Codes and Classifications 5 10.1.3 SYSTEM EVALUATION 5 10.1.3.1 Variables Limits Functions 5 10.1.3.2 Transient Effects 6 10.1.3.3 Secondary-Primary Interactions 7 Table 10.1-1 STEAM AND POWER CONVERSION SYSTEM COMPONENT 8 DESIGN PARAMETERS 10.2 TURBINE GENERATOR AND CONTROLS 11 10.2.1 MAIN TURBINE 11 10.2.1.1 Description 11 10.2.1.2 Turbine Controls 11 10.2.1.2.1 Description 11 10.2.1.2.2 Automatic Load Reduction 12 10.2.1.3 Turbine Disk Integrity 12 10.2.1.4 Turbine Supervisory Instrumentation 13 10.2.2 MAIN GENERATOR 13 10.2.3 ELECTROHYDRAULIC CONTROL SYSTEM 14 10.2.3.1 Function 14 10.2.3.2 Components 14 10.2.3.3 Alarms and Controls 15 10.2.3.4 Turbine Trip Devices 16 10.2.3.4.1 Overspeed Trip Mechanism 16 Page 1 of 5 Revision 26 5/2016
GINNA/UFSAR 10.2.3.4.2 Auxiliary Governor 17 10.2.3.4.3 Protective Trip Devices 17 10.2.3.4.4 Testing and Inspection 18 10.3 MAIN STEAM SYSTEM 21 10.3.1 DESIGN BASIS 20 10.3.2 SYSTEM DESCRIPTION 20 10.3.2.1 Flow Path 20 10.3.2.2 Steam Generators 21 10.3.2.3 Steam Piping 22 10.3.2.4 Main Steam Safety Valves (MSSV) 22 10.3.2.5 Atmospheric Relief Valves (ARV) 22 10.3.2.6 Main Steam Isolation Valves 23 10.3.2.7 Main Steam Non-Return Check Valves 23 10.3.2.8 Main Steam Header 24 10.3.2.9 Main Turbine Stop Valves and Control Valves 24 10.3.2.10 Moisture Separator Reheaters 24 10.3.2.11 Reheater Stop and Intercept Valves 25 10.3.3 INSTRUMENTATION REQUIREMENTS 25 10.4 CONDENSATE AND FEEDWATER SYSTEMS 27 10.
4.1 DESCRIPTION
27 10.4.2 FLOW PATH 27 10.4.3 MAIN CONDENSERS 27 10.4.4 CONDENSATE SYSTEM 28 10.4.4.1 Condensate Pumps 28 10.4.4.2 Condensate Booster Pumps 29 10.4.4.3 Low-Pressure Heaters 29 10.4.4.4 Condensate Bypass Valve 30 10.4.5 FEEDWATER SYSTEM 30 10.4.5.1 Main Feedwater Pumps 30 10.4.5.2 High-Pressure Heaters 31 Page 2 of 5 Revision 26 5/2016
GINNA/UFSAR 10.4.5.3 Feedwater Flow Control 31 10.4.5.4 Feedwater Flow Measurement 31 10.5 AUXILIARY FEEDWATER SYSTEMS 33 10.
5.1 INTRODUCTION
33 10.5.2 DESIGN BASES 33 10.5.2.1 Functional Requirements 33 10.5.2.2 Preferred Auxiliary Feedwater System 34 10.5.2.3 Standby Auxiliary Feedwater System (SAFW) 34 10.5.3 SYSTEMS OPERATION AND DESCRIPTION 35 10.5.3.1 Preferred Auxiliary Feedwater System 35 10.5.3.1.1 Normal Lineup 35 10.5.3.1.2 Startup and Cooldown Operations 35 10.5.3.1.3 Transient Operations 36 10.5.3.1.4 System Description 36 10.5.3.2 Standby Auxiliary Feedwater System (SAFW) 38 10.5.4 DESIGN EVALUATION 38 10.5.4.1 System Evaluation 38 10.5.4.2 Alternating Current Independence of the Turbine-Driven Auxiliary 39 Feedwater Pump (TDAFW) 10.5.5 INSTRUMENTATION 40 10.5.5.1 Motor-Driven Auxiliary Feedwater Pump (MDAFW) Controls 40 10.5.5.2 Preferred Auxiliary Feedwater System Initiation 40 10.5.5.3 Auxiliary Feedwater System Alarms 41 10.5.5.4 Auxiliary Feedwater Performance Indications 42 10.5.5.5 Control From Outside the Control Room 42 10.6 CIRCULATING WATER SYSTEM 45 10.6.1 DESIGN BASES 45 10.6.2 SYSTEM DESCRIPTION 45 10.6.2.1 Intake Structure 45 10.6.2.2 Inlet Tunnel 46 10.6.2.3 Traveling Screens 46 Page 3 of 5 Revision 26 5/2016
GINNA/UFSAR 10.6.2.4 Circulating Water Pumps 46 10.6.2.5 Condenser Inlet and Outlet Valves 46 10.6.2.6 Condensate Cooler 46 10.6.2.7 Screen House 46 10.6.2.8 Piping and Discharge Canal 47 10.6.2.9 Flooding Protection 47 10.6.3 INSTRUMENTATION AND CONTROL 48 10.6.4 INSERVICE INSPECTION 48 10.7 OTHER FEATURES OF THE STEAM AND POWER CONVER- 50 SION SYSTEM 10.7.1 STEAM DUMP SYSTEM 50 10.7.2 HEATER DRAIN SYSTEM 51 10.7.3 EXTRACTION STEAM SYSTEM 51 10.7.4 CONDENSATE STORAGE SYSTEM 52 10.7.5 STEAM-GENERATOR BLOWDOWN AND BLOWDOWN 53 RECOVERY SYSTEM 10.7.5.1 Steam-Generator Blowdown System 53 10.7.5.2 Blowdown Recovery System 54 10.7.5.3 Blowdown System Operation 54 10.7.6 MAIN TURBINE AND GENERATOR AUXILIARY SYSTEMS 55 10.7.6.1 Gland Sealing Steam and Exhaust System 56 10.7.6.2 Air Ejectors 57 10.7.6.3 Vacuum Priming System 57 10.7.6.4 Exhaust Hood Spray System 57 10.7.6.5 Turbine Lube-Oil System 58 10.7.6.6 Generator Hydrogen Cooling System 59 10.7.6.7 Generator Seal-Oil System 60 Page 4 of 5 Revision 26 5/2016
GINNA/UFSAR 10.7.6.8 Generator Exciter Cooling 60 10.7.7 SECONDARY CHEMISTRY CONTROL 60 10.7.7.1 All-Volatile-Treatment Chemistry 60 10.7.7.1.1 Background 60 10.7.7.1.2 All-Volatile-Treatment Chemistry Control 60 10.7.7.2 Nonnuclear Sampling System 62 10.7.7.3 Water Chemistry Monitoring Program 62 10.7.7.4 Catalytic Oxygen Removal System 62 10.7.7.5 Condensate Polishing Demineralizer System 63 10.7.7.5.1 System Description 63 10.7.7.5.2 Resin Transfer 64 10.7.7.5.3 Regeneration (Drawing 33013-1910, Sheets 1 and 2) 64 10.7.7.5.4 Waste Disposal (Drawing 33013-1912) 65 10.7.7.6 Chemical Dispersant 65 10.7.8 EROSION/CORROSION MONITORING PROGRAM 66 Table 10.7-1 COMPUTERIZED SECONDARY WATER CHEMISTRY MONI- 67 TORING SYSTEM FIGURES Figure 10.2-1 Turbine Control and Protection System Figure 10.6-1 Figure DELETED Sheet 1 -
Figure 10.6-1 Figure DELETED Sheet 2 -
Figure 10.6-2 Screen House Area Plot Plan Figure 10.6-3 Circulating Water Intake Cross Section Page 5 of 5 Revision 26 5/2016
GINNA/UFSAR 11 RADIOACTIVE WASTE MANAGEMENT 1 11.1 DESIGN CRITERIA AND SOURCE TERMS 2 11.1.1 GENERAL DESIGN CRITERIA 2 11.1.1.1 AIF General Design Criterion 70 (1967) 2 11.1.1.2 Appendix A General Design Criteria (1972) 3 11.1.2 SOURCE TERMS 3 11.1.2.1 Liquid Sources 4 11.1.2.2 Gaseous Sources 4 11.1.2.3 Radioactivity Inputs 4 11.2 LIQUID WASTE MANAGEMENT SYSTEM 7 11.2.1 DESIGN BASES 7 11.2.2 SYSTEM DESCRIPTION 7 11.2.2.1 Laundry and Hot Shower Tanks 7 11.2.2.2 Chemical Drain Tank 7 11.2.2.3 Reactor Coolant Drain Tank and Pumps 7 11.2.2.4 Waste Holdup Tank 8 11.2.2.4.1 Liquid Waste Sources 8 11.2.2.4.2 Waste Holdup Tank Discharge 9 11.2.2.5 Auxiliary Building Sump Tank, Sump Tank Pumps, and Sump Pumps 9 11.2.2.6 Waste Evaporator 10 11.2.2.7 Evaporator Feed Tank 10 11.2.2.8 Evaporator Feed Tank Pumps 11 11.2.2.9 Concentrator 11 11.2.2.10 Distillate Tank 12 11.2.2.11 Waste Condensate Demineralizers 12 11.2.2.12 Waste Condensate Tanks 12 11.2.2.13 Vendor Supplied Demineralization System 12 11.2.2.14 High Conductivity Waste Tank 13 11.2.2.15 Retention Tank 13 11.2.2.16 Neutralizing Tank 13 Page 1 of 4 Revision 26 5/2016
GINNA/UFSAR 11.2.2.17 Monitor Tanks 13 11.2.2.18 Radwaste Control System 13 11.2.2.19 Dry Cleaning Unit 13 11.2.2.20 Piping and Valves 14 11.2.3 LIQUID EFFLUENT RELEASE CONCENTRATIONS AND DOSES 14 11.2.3.1 Liquid Release Control 14 11.2.3.2 Dose Calculations 14 11.2.3.3 Accidental Spill of Liquid Radwastes 16 Table 11.2-1 WASTE DISPOSAL COMPONENTS CODE REQUIREMENTS 18 Table 11.2-2a LIQUID WASTE SYSTEM COMPONENT
SUMMARY
DATA - 19 TANKS Table 11.2-2b LIQUID WASTE SYSTEM COMPONENT
SUMMARY
DATA - 20 PUMPS Table 11.2-2c LIQUID WASTE SYSTEM COMPONENT
SUMMARY
DATA - 21 COMPRESSORS Table 11.2-3 LIQUID WASTE DISPOSAL SYSTEM PERFORMANCE DATA 22 Table 11.2-4a LIQUID EFFLUENTS, 10 CFR 50, APPENDIX I CALCULATIONS, 23 ASSUMPTIONS (HISTORICAL)
Table 11.2-4b LIQUID EFFLUENTS, 10 CFR 50, APPENDIX I CALCULATIONS, 24 ASSUMPTIONS1 (HISTORICAL)
Table 11.2-5 LIQUID EFFLUENTS, 10 CFR 50, APPENDIX I CALCULATIONS, 25 RESULTSa. (HISTORICAL)
Table 11.2-6 MAXIMUM INDIVIDUAL DOSES FROM LIQUID EFFLUENTS 28 (MREM/YEAR) (HISTORICAL)
Table 11.2-7 ESTIMATED ANNUAL DOSES TO THE PUBLIC DUE TO NOR- 30 MAL OPERATION LIQUID RADWASTE EFFLUENTS - CORE POWER LEVEL 1811 MWt 11.3 GASEOUS WASTE MANAGEMENT SYSTEM 31 11.3.1 DESIGN BASES 31 11.3.2 SYSTEM DESCRIPTION 31 11.3.2.1 Operation 31 11.3.2.2 Components 31 11.3.2.2.1 Waste Gas Compressors 32 11.3.2.2.2 Gas Decay Tanks 32 Page 2 of 4 Revision 26 5/2016
GINNA/UFSAR 11.3.2.2.3 Waste Disposal Panel 33 11.3.2.2.4 Gas Analyzer 33 11.3.2.2.5 Nitrogen Manifold 34 11.3.2.2.6 Hydrogen Manifold 34 11.3.2.2.7 Valves 34 11.3.3 GASEOUS RADIOACTIVE RELEASES 34 Table 11.3-1 GASEOUS EFFLUENTS, 10 CFR 50, APPENDIX I 37 CALCULATIONS, ASSUMPTIONS2 (HISTORICAL)
Table 11.3-2 GASEOUS EFFLUENTS, 10 CFR 50, APPENDIX I 39 CALCULATIONS, RESULTS3 (HISTORICAL)
Table 11.3-3 MAXIMUM INDIVIDUAL DOSES FROM GASEOUS EFFLUENTS 41 (HISTORICAL)
Table 11.3-4 ESTIMATED ANNUAL DOSES TO THE PUBLIC DUE TO 43 NORMAL OPERATION GASEOUS RADWASTE EFFLUENTS -
CORE POWER LEVEL 1811 MWt 11.4 SOLID WASTE MANAGEMENT SYSTEM 44 11.
4.1 DESCRIPTION
44 11.4.1.1 General 44 11.4.1.1.1 Types of Solid Waste 44 11.4.1.1.2 Sludge 44 11.4.1.1.3 Oily Waste 44 11.4.1.1.4 Bead Resin 44 11.4.1.1.5 Spent Filters 45 11.4.1.1.6 Dry Active Waste 45 11.4.1.2 Spent Resin Storage Tanks 45 11.4.1.3 Storage Facilities 45 11.4.2 SOLID WASTE ESTIMATES 46 11.4.3 PROCESS CONTROL PROGRAM 46 Table 11.4-1 ANNUAL SHIPMENT OF SOLID WASTE (JULY 1990-JUNE 1991) 48 11.5 PROCESS AND EFFLUENT RADIATION MONITORING AND 49 SAMPLING SYSTEMS 11.5.1 DESIGN BASES 49 11.5.2 SYSTEM DESCRIPTION 49 Page 3 of 4 Revision 26 5/2016
GINNA/UFSAR 11.5.2.1 General 49 11.5.2.2 Process Radiation Monitoring System 50 11.5.2.2.1 General Description 50 11.5.2.2.2 Containment Iodine Monitor 50 11.5.2.2.3 Plant Vent Iodine Monitor 50 11.5.2.2.4 Containment Particulate and Noble Gas Monitors 51 11.5.2.2.5 Containment Vent High-Range Effluent Monitor 51 11.5.2.2.6 Plant Vent Particulate Monitors 51 11.5.2.2.7 Plant Vent Noble Gas and High-Range Effluent Monitor 51 11.5.2.2.8 Air Ejector and Gland Steam Exhaust Monitors 52 11.5.2.2.9 Containment Service Water Monitor 53 11.5.2.2.10 Component Cooling Water Monitor 53 11.5.2.2.11 Liquid Waste Disposal Monitor 53 11.5.2.2.12 Steam-Generator Blowdown Monitor 53 11.5.2.2.13 Spent Fuel Pool (SFP) Heat Exchanger Service Water Monitors 54 11.5.2.2.14 Retention Tank Monitor 54 11.5.2.2.15 High Conductivity Waste Tank Monitor 54 11.5.2.2.16 Control Room Radiation Monitors 55 11.5.2.3 Tritium Sampling 55 11.5.3 DESIGN EVALUATION 55 11.5.4 ENVIRONMENTAL RADIOACTIVITY MONITORING PRO- 55 GRAM Figures Figure 11.5-1 Gaseous Radwaste Treatment Systems Effluent Paths and Controls Figure 11.5-2 Liquid Radwaste Treatment Systems Effluent Paths and Controls Page 4 of 4 Revision 26 5/2016
GINNA/UFSAR 12 RADIATION PROTECTION 1 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES 2 ARE AS LOW AS IS REASONABLY ACHIEVABLE 12.1.1 ALARA PROGRAM 2 12.1.2 ORGANIZATIONAL RESPONSIBILITIES 2 12.1.3 RADIATION PROTECTION PROGRAM 2 12.2 RADIATION SOURCES 3 Table 12.2-1 SHIELDING SOURCE TERMS (T=0) (HISTORICAL) 6 Table 12.2-2 SHIELDING SOURCE ACTIVITY at T=0 hrs - Power Level 1811 10 MWt 12.3 RADIATION PROTECTION DESIGN FEATURES 16 12.3.1 DESIGN CRITERIA 16 12.3.1.1 Conformance to 1967 Design Criteria 16 12.3.1.2 Conformance to 1972 Design Criteria 16 12.3.2 SHIELDING 17 12.3.2.1 Design Basis 17 12.3.2.2 Shielding Design 19 12.3.2.2.1 Primary Shield 19 12.3.2.2.2 Secondary Shield 19 12.3.2.2.3 Containment Structure 20 12.3.2.2.4 Fuel Handling (Water and Pool) Shield 20 12.3.2.2.5 Auxiliary Shielding 21 12.3.2.2.6 Shielding Design Modifications 21 12.3.2.2.7 Containment Accessibility Procedure 22 12.3.3 VENTILATION 22 12.3.3.1 Gas Collection and Decay Tank System 23 12.3.3.2 Plant Ventilation System 23 12.3.3.3 Containment Ventilation System 24 12.3.3.4 Air Ejector and Gland Seal Exhaust System 24 12.3.4 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONI- 24 TORING INSTRUMENTATION 12.3.4.1 Introduction 24 Page 1 of 3 Revision 26 5/2016
GINNA/UFSAR 12.3.4.2 Description 25 12.3.4.3 Radiation Monitoring System Detectors 26 12.3.5 EQUIPMENT AND SYSTEM DECONTAMINATION 27 12.3.5.1 Design Basis 27 12.3.5.2 Methods of Decontamination 27 12.3.5.3 Decontamination Facilities 27 Table 12.3-1 PLANT ZONE CLASSIFICATIONS 30 Table 12.3-2a PRIMARY SHIELD NEUTRON FLUXES AND DESIGN PARAME- 31 TERS - CALCULATED NEUTRON FLUXES (HISTORICAL)
Table 12.3-2b PRIMARY SHIELD NEUTRON FLUXES AND DESIGN PARAME- 32 TERS - DESIGN PARAMETERS (HISTORICAL)
Table 12.3-3 SECONDARY SHIELD DESIGN PARAMETERS (HISTORICAL) 33 Table 12.3-4 CONTAINMENT STRUCTURE DESIGN PARAMETERS (HISTOR- 34 ICAL)
Table 12.3-5 REFUELING CANAL AND SPENT FUEL POOL DESIGN PARAM- 35 ETERS (HISTORICAL)
Table 12.3-6 PRINCIPAL AUXILIARY SHIELDING (HISTORICAL) 36 12.4 DOSE ASSESSMENT 37 12.4.1 OPERATION IN MODES 1 AND 2 37 12.4.2 FUEL HANDLING OPERATIONS 37 12.4.3 POSTACCIDENT CONDITIONS 38 12.4.3.1 Summary 38 12.4.3.2 Methodology 39 12.4.3.2.1 Calculation of Dose Rates 39 12.4.3.2.2 Doses to Personnel During Postaccident Access to Vital Areas 39 12.4.3.3 Areas That May Require Access for Postaccident Operations 40 12.4.3.3.1 Hydrogen Recombiner Control Panel (Area A) 40 12.4.3.3.2 Postaccident Containment Air Sample Penetration No. 203 (Area B) 40 12.4.3.3.3 Nuclear Sample Room (Area C) 40 12.4.3.3.4 Primary Chemistry Laboratory (Area D) 41 12.4.3.3.5 Count Room (Area E) 41 12.4.3.3.6 Postaccident Containment Sample Penetration No. 305 (Area F) 41 12.4.3.3.7 Radwaste Control Panel (Area G) 41 12.4.3.3.8 Safeguards Bus 16 (Area H) 41 Page 2 of 3 Revision 26 5/2016
GINNA/UFSAR 12.4.3.3.9 Safeguards Bus 14 (Area I) 42 12.4.3.3.10 Postaccident Containment Air Sample Penetration (Area J) 42 12.4.3.3.11 Auxiliary Building Heating, Ventilation, and Air Conditioning (Area 43 K) 12.4.3.3.12 Spent Fuel Pool (SFP) and Auxiliary Building Heating, Ventilation, 43 and Air Conditioning Filters (Area L and Area M) 12.4.3.3.13 Control Access High Efficiency Particulate Air and Charcoal Filters 43 (Area N) 12.4.3.3.14 Control Room 43 12.4.3.3.15 Control Room 43 Table 12.4-1 RADIATION MONITORING SYSTEM READINGS (1983) 45 Table 12.4-2 RADIATION SURVEY READINGS IN PLANT AREAS (1983) 46 Table 12.4-3 Table DELETED 48 Table 12.4-4 EXPOSURE RATES FOR VITAL AREAS AS A FUNCTION OF 49 TIME (R/hr)
Table 12.4-5 VITAL AREA RADIATION DOSE
SUMMARY
50 12.5 RADIATION PROTECTION PROGRAM ADMINISTRATION 51 12.5.1 ORGANIZATION 51 12.5.2 EXPOSURE CONTROL PROGRAM 51 12.5.2.1 External Exposure 51 12.5.2.2 Internal Exposure 51 12.5.2.3 Respiratory Protection 52 12.5.2.4 Radioactive Sources Control 52 12.5.2.5 Medical Examinations 52 12.5.3 SURVEILLANCE PROGRAM 53 12.5.3.1 Surveys 53 12.5.3.2 Radiation Work Permits 53 12.5.3.3 Access Control, Posting, and Labeling 53 12.5.3.3.1 Restricted Areas 53 12.5.3.3.2 Access Control 53 12.5.3.3.3 Protective Apparel 54 12.5.4 RADIATION PROTECTION FACILITIES AND EQUIPMENT 54 Table 12.5-1 Table DELETED 56 FIGURES Figure 12.5-1 Table DELETED Page 3 of 3 Revision 26 5/2016
GINNA/UFSAR 13 CONDUCT OF OPERATIONS 1 13.1 ORGANIZATIONAL STRUCTURE OF GINNA NUCLEAR GENERATING 2 STATION 13.1.1 ORIGINAL CONSTRUCTION ORGANIZATION 2 13.1.1.1 Design and Construction Activities (Project Phase) 2 13.1.1.2 Preoperational Activities 2 13.1.2 CORPORATE NUCLEAR OPERATIONS ORGANIZATION 2 13.1.2.1 President and Chief Executive Officer, CENG (CEO) 3 13.1.2.2 Senior Vice President - Nuclear Operations Chief Nuclear Officer CENG 3 (COO/ CNO) 13.1.2.3 DELETED 13.1.3 GINNA OPERATING ORGANIZATION 3 13.1.3.1 Site Vice President (SVP) 3 13.1.3.2 Plant Manager 3 13.1.3.3 Director, Site Engineering 4 13.1.3.3.1 Engineering Supervision 4 13.1.3.3.1.1 Department Engineer(s) 4 13.1.3.3.1.2 Reactor Engineer(s) 4 13.1.3.4 Director, Site Work Management 4 13.1.3.5 Director, Site Training 5 13.1.3.6 Manager, Site Security 5 13.1.3.7 Director, Site Operations 5 13.1.3.8 Manager, Site Radiation Protection 6 13.1.3.9 Manager, Site Chemistry, Environment and Radwaste 7 13.1.3.10 Director, Site Maintenance 8 13.1.4 QUALIFICATIONS OF PLANT PERSONNEL 8 13.1.4.1 Qualifications of Plant Staff 8 13.1.4.2 Qualifications of Incumbent Plant Personnel 8 13.2 TRAINING PROGRAM 10 13.2.1 PLANT STAFF TRAINING PROGRAM 10 13.2.1.1 Objectives 10 13.2.1.2 Initial Training Programs (Historical) 10 13.2.1.2.1 Personnel Selection 10 13.2.1.2.2 Nuclear Theory 10 13.2.1.2.3 Plant Systems and Operations 11 13.2.1.2.4 Final Phase 11 13.2.1.3 Onsite Training Prior to Startup 12 Page 1 of 3 Revision 26 5/2016
GINNA/UFSAR 13.2.2 REPLACEMENT AND RETRAINING OF PERSONNEL 13 13.2.2.1 Licensed Operator Replacement and Requalification Training 13 13.2.2.2 Replacement and Retraining of Unlicensed Personnel 13 13.2.2.3 General Employee Training 13 REFERENCES FOR SECTION 13.2 15 Table 13.2-1 INITIAL GINNA STATION PERSONNEL TRAINING 16 13.3 EMERGENCY PLANNING 17 13.4 REVIEW AND AUDIT 18 13.4.1 ONSITE REVIEW 18 13.4.2 INDEPENDENT REVIEW 18 13.4.3 AUDIT PROGRAM 18 13.4.3.1 Nuclear Safety Review Board 18 13.4.3.2 Quality Performance & Assessment Group 18 13.5 PLANT PROCEDURES 19 13.5.1 ADMINISTRATIVE PROCEDURES 19 13.5.1.1 Conformance With Regulatory Guide 1.33 19 13.5.1.2 Preparation of Procedures 19 13.5.1.3 Description of Administrative Procedures 19 13.5.2 OPERATING AND MAINTENANCE PROCEDURES 19 13.5.2.1 Control Room Operating Procedures 19 13.5.2.2 Site Contingency Procedures 20 13.5.2.2.1 General 20 13.5.2.2.2 Adverse Weather Conditions 20 13.5.2.2.2.1 High Winds 20 13.5.2.2.2.2 Tornadoes 20 13.5.2.2.2.3 Ice Storms 20 13.5.2.2.3 High Water or Flood Emergency Plan 21 13.5.2.2.4 Earthquake Emergency Plan 21 13.5.2.2.5 Fire Emergency Plan 21 13.5.2.3 Other Procedures 22 REFERENCES FOR SECTION 13.5 24 13.6 INDUSTRIAL SECURITY 25 FIGURES Figure 13.1-1 Site Vice President Figure 13.1-2 Plant Manager Figure 13.1-3 Site Engineering Figure 13.1-4 Site Work Management Page 2 of 3 Revision 26 5/2016
GINNA/UFSAR Figure 13.1-5 Site Training Figure 13.1-6 Site Operations Figure 13.1-7 Site Radiation Protection Figure 13.1-8 Site Chemistry, Environment and Radwaste Figure 13.1-9 Site Maintenance Page 3 of 3 Revision 26 5/2016
GINNA/UFSAR 14 INITIAL TEST PROGRAM 1 14.1
SUMMARY
OF TEST PROGRAM AND OBJECTIVES 2 14.1.1 STARTUP AND POWER TESTING AT 1300 MEGAWATTS THER- 2 MAL 14.1.1.1 Summary 2 14.1.1.2 Tests Prior to Reactor Fueling 3 14.1.1.2.1 Summary 3 14.1.1.2.2 Test Objectives 4 14.1.1.3 Final Plant Preparation 10 14.1.1.3.1 Core Loading 10 14.1.1.3.2 Postloading Tests 11 14.1.1.4 Initial Testing in the Operating Reactor 12 14.1.1.4.1 General 12 14.1.1.4.2 Initial Criticality 12 14.1.1.4.3 Zero Power Testing 12 14.1.1.4.4 Power Level Escalation 13 14.1.1.4.5 Post Startup Surveillance and Testing Requirements 13 14.1.2 POWER TEST PROGRAM TO 1520 MEGAWATTS THERMAL 14 REFERENCES FOR SECTION 14.1 15 Table 14.1-1 INITIAL TESTING
SUMMARY
- INITIAL CRITICALITY 16 THROUGH 100 - HOUR ACCEPTANCE TEST 14.2 INITIAL ORGANIZATION AND STAFFING 19 14.2.1 INITIAL STARTUP AND OPERATING STRUCTURE 19 14.2.2 ONSITE TRAINING PRIOR TO STARTUP 19 14.3 TEST PROCEDURES 21 14.3.1 PRE FUEL LOADING TESTS 21 14.3.2 POST FUEL LOADING TESTS 21 14.4 CONDUCT OF TEST PROGRAM 22 14.4.1 CONDUCT OF INITIAL TEST PROGRAM 22 14.4.2 REVIEW, EVALUATION, AND APPROVAL OF TEST RESULTS 23 14.4.3 TEST RECORDS 23 Page 1 of 6 Revision 26 5/2016
GINNA/UFSAR 14.5 TEST PROGRAM SCHEDULE 24 14.5.1 INITIAL CRITICALITY TO ACCEPTANCE 24 14.5.2 1520 MEGAWATTS THERMAL POWER TEST PROGRAM 25 14.6 INDIVIDUAL TEST DESCRIPTIONS 26 14.6.1 INITIAL STARTUP AND POWER TEST PROGRAM 26 14.6.1.1 Safety Injection Systems Preoperational Tests 26 14.6.1.1.1 Safety Injection Test 26 14.6.1.1.2 Accumulator Blowdown Test 30 14.6.1.1.3 Safety Injection Flow Test 33 14.6.1.1.4 Containment Spray System Test 34 14.6.1.1.5 Residual Heat Removal System Test 34 14.6.1.1.6 Safeguards Systems Operational Test 34 14.6.1.1.7 Emergency Diesel Generator Test 35 14.6.1.1.8 Direct Current Test 36 14.6.1.2 Preoperational Instrumentation and Control Tests 36 14.6.1.2.1 Reactor Coolant System Pressure Comparison Test 36 14.6.1.2.2 Resistance Temperature Detector Cross Calibration Test 36 14.6.1.2.3 Steam Generator Manual Control and Level Instrumentation Test 37 14.6.1.2.4 Rod Position Indication System Test 37 14.6.1.2.5 Rod Stepping Test 38 14.6.1.2.6 Rod Cluster Control Assembly Drop Time and Partial Length Rods 38 Operational Tests 14.6.1.2.7 In-Core Thermocouples Test 38 14.6.1.2.8 Movable In-Core Detector System Test 38 14.6.1.2.9 Reactor Makeup Blender and Boric Acid Transfer Pumps Operational 38 Test 14.6.1.2.10 Pressurizer Level Control Test 39 14.6.1.2.11 Pressurizer Pressure Control Test 40 14.6.1.2.12 Steam Dump Test 41 14.6.1.2.13 Radiation Monitoring System Operational Test 41 14.6.1.2.14 Reactor Coolant System Flow Measurement Test 42 14.6.1.2.15 Nuclear Instrumentation Test 42 14.6.1.3 Safety and Relief Valve Tests 42 Page 2 of 6 Revision 26 5/2016
GINNA/UFSAR 14.6.1.3.1 Pressurizer Safety Valve Test 42 14.6.1.3.2 Main Steam Safety Valve Test 42 14.6.1.4 Waste Systems Tests 43 14.6.1.4.1 Liquid Waste Concentration Demonstration Test 43 14.6.1.4.2 Waste Disposal System Gaseous Waste Test 43 14.6.1.4.3 Liquid Waste Processing Test 44 14.6.1.5 Reactor Coolant System Measurement Tests 44 14.6.1.5.1 Reactor Vessel Internals Measurement Test 44 14.6.1.5.2 Reactor Coolant System Vibration Test 46 14.6.1.5.3 Preoperational Reactor Coolant System Leakage Test 46 14.6.1.5.4 Reactor Coolant System Thermal Expansion Test 47 14.6.1.5.5 Flow Coastdown Test 47 14.6.1.5.6 Natural Circulation Test 47 14.6.1.6 Miscellaneous Safety-Related Tests 48 14.6.1.6.1 Backfeed from the 115-Kilovolt Grid Test 48 14.6.1.6.2 Blackout Test Without Safety Injection Test 49 14.6.1.6.3 Main Steam Isolation Valve Test 49 14.6.1.6.4 Fire Service Water Test 50 14.6.1.6.5 Electrical System Logic Test 50 14.6.1.6.6 Reactor Trip System (RTS) Operational Test 50 14.6.1.6.7 Reactor Coolant System Hydro Test 51 14.6.1.6.8 Ventilation Systems Test 51 14.6.1.6.9 Preoperational Containment Vessel Leak Rate Test 51 14.6.1.6.10 Structural Integrity Test 52 14.6.1.6.11 Reactor Trip System (RTS) Operation Time Response Test 53 14.6.1.7 Operational and Transient Tests 53 14.6.1.7.1 Ten Percent Load Swing Test at Thirty Percent Power 53 14.6.1.7.2 Generator Trip Test 54 14.6.1.7.3 Ten Percent Load Swing Test at Seventy-Five Percent Power Level 54 14.6.1.7.4 Fifty Percent Load Reduction from Seventy-Five Percent Power Level 54 Test 14.6.1.7.5 One Hundred Percent Power Level Transient Tests 55 Page 3 of 6 Revision 26 5/2016
GINNA/UFSAR 14.6.1.7.6 Operational Dynamic Rod Drop Test 57 14.6.1.7.7 Delta T Zero Power Alignment and Delta T Channel Span Adjustment 58 Tests 14.6.1.7.8 Nuclear Instrumentation Calibration and Reactor Coolant System Flow 58 Confirmation 14.6.1.7.9 Ex-Core In-Core Calibration Test 61 14.6.1.8 Startup Physics Testing 61 14.6.1.8.1 Introduction 61 14.6.1.8.2 Power Distribution Measurements 62 14.6.1.8.3 Zero Power Critical Boron Concentrations 62 14.6.1.8.4 Reactivity Coefficients and Shutdown Margin 62 14.6.1.8.5 Ejected and Dropped Rod Worths 62 14.6.1.8.6 Xenon Oscillation Test 62 14.6.2 POWER TEST PROGRAM TO 1520 MEGAWATTS THERMAL 63 14.6.2.1 Test Description 63 14.6.2.2 Steam Generator Moisture Carryover Tests 65 14.6.2.3 Assembly Delta T Measurements 65 14.6.2.4 Plant Radiation Surveys 65 14.6.2.5 Reactor Physics Measurements 66 14.6.2.5.1 Zero Power Measurements 66 14.6.2.5.2 At-Power Measurements 66 Table 14.6-1a ACCUMULATOR BLOWDOWN TEST RESULTS - OBSERVED 69 RESULTS Table 14.6-1b ACCUMULATOR BLOWDOWN TEST RESULTS - PREDICTION 70 OF FINAL PRESSURE Table 14.6-1c ACCUMULATOR BLOWDOWN TEST RESULTS - PIPE RESIS- 71 TANCE Table 14.6-2 BEGINNING OF CYCLE ZERO POWER CRITICAL BORON CON- 72 CENTRATIONS Table 14.6-3 REACTIVITY COEFFICIENTS AND SHUTDOWN MARGIN 73 Table 14.6-4 EJECTED AND DROPPED ROD WORTHS 74 Table 14.6-5 AVERAGE INCREASE IN CONTAINMENT RADIATION LEVELS 75 DURING UPRATING PROGRAM TO 1520 MEGAWATTS THER-MAL Table 14.6-6
SUMMARY
OF MEASURED PARAMETERS AT HOT ZERO 76 POWER PRIOR TO UPRATING TO 1520 MEGAWATTS THERMAL Page 4 of 6 Revision 26 5/2016
GINNA/UFSAR Table 14.6-7 SELECTED DATA FOR FLUX MAPS 77 FIGURES Figure 14.2-1 Initial Plant Organization Figure 14.6-1 Original Safety Injection System Functional Test Data Sheet (Typical)
Sheet 1 -
Figure 14.6-1 Original Safety Injection System Functional Test Data Sheet (Typical)
Sheet 2 -
Figure 14.6-2 Rod Cluster Control Assembly Drop Time Test Results (Typical)
Sheet 1 -
Figure 14.6-2 Rod Cluster Control Assembly Drop Time Test Results (Typical)
Sheet 2 -
Figure 14.6-2 Rod Cluster Control Assembly Drop Time Test Results (Typical)
Sheet 3 -
Figure 14.6-2 Rod Cluster Control Assembly Drop Time Test Results (Typical)
Sheet 4 -
Figure 14.6-3 Partial-Length Rods Operation Test Results (Typical)
Figure 14.6-4 Radiation Monitoring System Operational Test Data Sheet (Typical)
Sheet 1 -
Figure 14.6-4 Radiation Monitoring System Operational Test Data Sheet (Typical)
Sheet 2 -
Figure 14.6-5 Pressurizer Safety Valve Test Data Sheet Figure 14.6-6 Reactor Vessel Internals Displacement Indicators Figure 14.6-7 Reactor Vessel Accelerometers Sheet 1 -
Figure 14.6-7 Reactor Vessel Accelerometers Sheet 2 -
Figure 14.6-7 Reactor Vessel Accelerometers Sheet 3 -
Figure 14.6-7 Reactor Vessel Accelerometers Sheet 4 -
Figure 14.6-8 Reactor Coolant System Vibration Test Data Sheet Sheet 1 -
Figure 14.6-8 Reactor Coolant System Vibration Test Data Sheet Sheet 2 -
Figure 14.6-9 Reactor Coolant Pump Flow Coastdown Test Sheet 1 -
Figure 14.6-9 Reactor Coolant Pump Flow Coastdown Test Sheet 2 -
Figure 14.6-9 Reactor Coolant Pump Flow Coastdown Test Sheet 3 -
Page 5 of 6 Revision 26 5/2016
GINNA/UFSAR Figure 14.6-9 Reactor Coolant Pump Flow Coastdown Test Sheet 4 -
Figure 14.6-9 Reactor Coolant Pump Flow Coastdown Test Sheet 5 -
Figure 14.6-10 Natural Circulation Test Data Sheet 1 -
Figure 14.6-10 Natural Circulation Test Data Sheet 2 -
Figure 14.6-11 Preoperational Containment Leak Rate Test Data Figure 14.6-12 Preoperational Containment Leak Rate Test, Pressurization System Figure 14.6-13 50% Load Reduction From 75% Power Level Test, Recorded Process Sheet 1 - Variables Figure 14.6-13 50% Load Reduction From 75% Power Level Test, Recorded Process Sheet 2 - Variables Figure 14.6-13 50% Load Reduction From 75% Power Level Test, Recorded Process Sheet 3 - Variables Figure 14.6-13 50% Load Reduction From 75% Power Level Test, Recorded Process Sheet 4 - Variables Figure 14.6-13 50% Load Reduction From 75% Power Level Test, Recorded Process Sheet 5 - Variables Figure 14.6-14 Operational Dynamic Rod Drop Test, Nuclear Power Channel Signals Figure 14.6-15 Ginna 1520 Megawatts Power Test Program, March 1 Through April 15, 1972 Figure 14.6-16 Ginna Uprating, Control Bank D Differential and Integral Worth, Cycle 1B 7800 MWd/MTU, Hot Zero Power Figure 14.6-17 Ginna Uprating, Control Bank C Differential and Integral Worth, Cycle 1B 7800 MWd/MTU, Hot Zero Power Figure 14.6-18 Ginna Uprating, Boron Concentration Versus Reactivity Insertion, Cycle 1B 7800 MWd/MTU, Hot Zero Power Figure 14.6-19 Ginna Uprating, Isothermal Temperature Coefficient Versus Boron Concentration, Cycle 1B 7800 MWd/MTU, Hot Zero Power Figure 14.6-20 Relative Power During Uprating: 1300, 1380, 1455 MWt Figure 14.6-21 Relative Power During Uprating: 1455 MWt Figure 14.6-22 Ginna Uprating, Power Range Output Versus Core Power, Channel NE41 Figure 14.6-23 Ginna Uprating, Axial Offset Calibration, Channel 41 Output Normal-ized to 1520 MWt Figure 14.6-24 Relative Power at 1520 MWt, April 12, 1972 Page 6 of 6 Revision 26 5/2016
15 ACCIDENT ANALYSES 1 15.0 GENERAL 2 15.0.1 INITIAL CONDITIONS 2 15.0.1.1 Assumed Values of Initial Conditions 2 15.0.2 POWER DISTRIBUTION 3 15.0.3 REACTIVITY COEFFICIENTS ASSUMED IN THE ACCIDENT 4 ANALY- SES 15.0.4 ROD CLUSTER CONTROL ASSEMBLY INSERTION 4 CHARACTERIS- TICS 15.0.5 TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN THE 4 ACCI- DENT ANALYSES 15.0.6 INSTRUMENTATION DRIFT AND CALORIMETRIC ERRORS - 5 POWER RANGE NEUTRON FLUX 15.0.7 COMPUTER CODES 5 15.0.7.1 FACTRAN 6 15.0.7.2 RETRAN 6 15.0.7.3 TWINKLE 6 15.0.7.4 VIPRE 7 15.0.7.5 ADVANCED NODAL CODE (ANC) 7 15.0.8 CLASSIFICATION OF PLANT CONDITIONS 7 15.0.8.1 Condition I - Normal Operation 8 15.0.8.2 Condition II - Faults of Moderate Frequency 8 15.0.8.3 Condition III - Infrequent Faults 8 15.0.8.4 Condition IV - Limiting Faults 8 15.0.9 UFSAR Re-write 9 15.0.9.1 General Layout 9 15.0.9.2 Interpretation of Operator Action Times 9
15.0 REFERENCES
FOR SECTION 15.0 10 Table 15.0-1 NSSS PCWG Parameters for Ginna Station Uprate Program 11 Table 15.0-1 NSSS PCWG Parameters for Ginna Station Uprate Program 12 Table 15.0-2 Non-LOCA Analysis Limits and Analysis Results 13 Table 15.0-3 Non-LOCA Plant Initial Condition Assumptions 16 Table 15.0-4 Pressurizer and Main Steam System (MSS) Pressure Relief 17 Assump- tions Table 15.0-5 Core Kinetics Parameters and Reactivity Feedback Coefficients 21 Table 15.0-6 Summary of RPS and ESFAS Functions Actuated 22 Page 1 of 24 Revision 26 5/2016
Table 15.0-7 Overtemperature and Overpower T Setpoints 25 Table 15.0-8 DETERMINATION OF MAXIMUM OVERPOWER TRIP POINT 26
- POWER RANGE NEUTRON FLUX CHANNEL - BASED ON NOMINAL SETPOINT CONSIDERING INHERENT INSTRU-MENT ERRORS Table 15.0-9 Summary of Initial Conditions and Computer Codes Used 27 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 30 15.1.1 DECREASE IN FEEDWATER TEMPERATURE 30 15.1.1.1 Description of Event 30 15.1.1.2 Frequency of Event 31 15.1.1.3 Event Analysis 31 15.1.1.3.1 Protective Features 31 15.1.1.3.2 Single Failures Assumed 31 15.1.1.3.3 Operator Actions Assumed 31 15.1.1.3.4 Chronological Description of Event 31 15.1.1.3.5 Impact on Fission Product Barriers 31 15.1.1.4 Reactor Core and Plant System Evaluation 32 15.1.1.4.1 Input Parameters and Initial Conditions 32 15.1.1.4.2 Methodology 32 15.1.1.4.3 Acceptance Criteria 32 15.1.1.4.4 Results 32 15.1.1.5 Radiological Consequences 32 15.1.1.6 Conclusions 32 15.1.2 INCREASE IN FEEDWATER FLOW 33 15.1.2.1 Increase in Feedwater Flow at Full Power 33 15.1.2.1.1 Description of Event 33 15.1.2.1.2 Frequency of Event 33 15.1.2.1.3 Event Analysis 33 15.1.2.1.3.1 Protective Features 34 15.1.2.1.3.2 Single Failures Assumed 34 15.1.2.1.3.3 Operator Actions Assumed 35 15.1.2.1.3.4 Chronological Description of Event 35 15.1.2.1.3.5 Impact on Fission Product Barriers 35 15.1.2.1.4 Reactor Core and Plant System Evaluation 35 15.1.2.1.4.1 Input Parameters and Initial Conditions 35 15.1.2.1.4.2 Method of Analysis 36 Page 2 of 24 Revision 26 5/2016
15.1.2.1.4.3 Acceptance Criteria 36 15.1.2.1.4.4 Results 36 15.1.2.1.5 Radiological Consequences 37 15.1.2.1.6 Conclusion 37 15.1.2.2 Increase in Feedwater Flow at Zero Power 37 15.1.2.2.1 Description of Event 37 15.1.2.2.2 Frequency of Event 37 15.1.2.2.3 Event Analysis 37 15.1.2.2.3.1 Protective Features 37 15.1.2.2.3.2 Single Failures Assumed 38 15.1.2.2.3.3 Operator Actions Assumed 38 15.1.2.2.3.4 Chronological Description of Event 38 15.1.2.2.3.5 Impact on Fission Product Barriers 38 15.1.2.2.4 Reactor Core and Plant System Evaluation 38 15.1.2.2.4.1 Input Parameters and Initial Conditions 38 15.1.2.2.4.2 Methodology 39 15.1.2.2.4.3 Acceptance Criteria 39 15.1.2.2.5 Radiological Consequences 39 15.1.2.2.6 Conclusion 39 15.1.3 EXCESSIVE LOAD INCREASE INCIDENT 39 15.1.3.1 Description of Event 39 15.1.3.2 Frequency of Event 40 15.1.3.3 Event Analysis 40 15.1.3.3.1 Protective Features 40 15.1.3.3.2 Single Failures Assumed 41 15.1.3.3.3 Operator Actions Assumed 41 15.1.3.3.4 Chronological Description of Event 41 15.1.3.3.5 Impact on Fission Product Barriers 41 15.1.3.4 Reactor Core and Plant System Evaluation 41 15.1.3.4.1 Input Parameters and Initial Conditions 41 15.1.3.4.2 Methodology 42 15.1.3.4.3 Acceptance Criteria 42 15.1.3.5 Radiological Consequences 42 15.1.3.6 Conclusions 42 15.1.4 INADVERTENT OPENING OF A STEAM GENERATOR RELIEF/ 43 SAFETY VALVE 15.1.5 SPECTRUM OF STEAM SYSTEM PIPING FAILURES INSIDE AND 43 OUTSIDE OF CONTAINMENT Page 3 of 24 Revision 26 5/2016
15.1.5.1 Description of Event 43 15.1.5.2 Frequency of Event 43 15.1.5.3 Event Analysis 44 15.1.5.3.1 Protective Features 44 15.1.5.3.2 Single Failures Assumed 45 15.1.5.3.3 Operator Actions Assumed 45 15.1.5.3.4 Chronological Description of Event 45 15.1.5.3.5 Impact on Fission Product Barriers 45 15.1.5.4 Reactor Core and Plant System Evaluation 46 15.1.5.4.1 Input Parameters and Initial Conditions 46 15.1.5.4.2 Methodology 47 15.1.5.4.3 Acceptance Criteria 48 15.1.5.4.4 Results 48 15.1.5.5 Radiological Consequences 49 15.1.5.6 Conclusions 50 15.1.5.7 Supplemental Evaluations 50 15.1.5.7.1 SEV-1073 50 15.1.5.7.2 HZP 6 Inch Steamline Break 50 15.1.5.7.3 High Steam Flow Setpoint Increase Evaluation 50 15.1.5.7.4 Steamline Rupture a Full Power 51 15.1.5.8 Potential for Containment Overpressurization 51 15.1.6 COMBINED STEAM GENERATOR ATMOSPHERIC RELIEF VALVE 51 (ARV) AND MAIN FEEDWATER REGULATING VALVE (MFRV)
FAIL- URES 15.1.6.1 Description of Event 51 15.1.6.2 Frequency of Event 52 15.1.6.3 Event Analysis 52 15.1.6.3.1 Protective Features 53 15.1.6.3.2 Single Failures Assumed 53 15.1.6.3.3 Operator Actions Assumed 53 15.1.6.3.4 Chronological Description of Event 54 15.1.6.3.5 Impact on Fission Product Barriers 54 15.1.6.4 Reactor Core and Plant System Evaluation 54 15.1.6.4.1 Input Parameters and Initial Conditions 54 15.1.6.4.2 Methodology 55 15.1.6.4.3 Acceptance Criteria 56 15.1.6.4.4 Results 56 15.1.6.5 Radiological Consequences 56 Page 4 of 24 Revision 26 5/2016
15.1.6.6 Conclusions 57
15.1 REFERENCES
FOR SECTION 15.1 58 Table 15.1-1 TIME SEQUENCE OF EVENTS FOR FEEDWATER MALFUNC- 59 TION TRANSIENTS HOT FULL POWER - SINGLE LOOP - WITH ROD CONTROL Table 15.1-2 TIME SEQUENCE OF EVENTS FOR FEEDWATER MALFUNC- 60 TION TRANSIENTS HOT FULL POWER - SINGLE LOOP -
WITH- OUT ROD CONTROL Table 15.1-3 TIME SEQUENCE OF EVENTS FOR FEEDWATER MALFUNC- 61 TION TRANSIENTS HOT FULL POWER - MULTI LOOP -
WITH ROD CONTROL Table 15.1-4 TIME SEQUENCE OF EVENTS FOR FEEDWATER MALFUNC- 62 TION TRANSIENTS HOT FULL POWER - MULTI LOOP -
WITH- OUT ROD CONTROL Table 15.1-5 Table DELETED 63 Table 15.1-6 TIME SEQUENCE OF EVENTS FOR STEAM LINE RUPTURE 64 Table 15.1-7
SUMMARY
OF MAIN FEEDWATER REGULATING VALVES 65 (MFRV)/STEAM GENERATOR ATMOSPHERIC RELIEF VALVE (ARV) COMBINATION FAILURE CASES EVALUATED Table 15.1-8 MSLB DOSE ANALYSIS ASSUMPTIONS 66 Table 15.1-9 RESULTS FOR MAIN STEAM LINE BREAK, REM TEDE 68 Table 15.1-10 TIME SEQUENCE OF EVENTS FOR THE COMBINED FAILURE 69 OF TWO MFRV's AND TWO ARV's AT HOT FULL POWER 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 70 15.2.1 STEAM PRESSURE REGULATOR MALFUNCTION OR FAILURE 70 THAT RESULTS IN DECREASING STEAM FLOW 15.2.2 LOSS OF EXTERNAL ELECTRICAL LOAD 70 15.2.2.1 Description of Event 70 15.2.2.2 Frequency of Event 70 15.2.2.3 Event Analysis 70 15.2.2.3.1 Protective Features 71 15.2.2.3.2 Single Failures Assumed 71 15.2.2.3.3 Operator Actions Assumed 71 15.2.2.3.4 Chronological Description of Event 72 15.2.2.3.5 Impact on Fission Product Barriers 72 15.2.2.4 Reactor Core and Plant System Evaluation 72 15.2.2.4.1 Input Parameters and Initial Conditions 72 15.2.2.4.2 Method of Analysis 73 15.2.2.4.3 Acceptance Criteria 74 15.2.2.4.4 Results 74 15.2.2.5 Radiological Consequences 75 Page 5 of 24 Revision 26 5/2016
15.2.2.6 Conclusions 75 15.2.2.7 Supplemental Evaluations 75 15.2.3 TURBINE TRIP 75 15.2.4 LOSS OF CONDENSER VACUUM 75 15.2.5 LOSS OF ALL ALTERNATING CURRENT POWER TO THE STATION 76 AUXILIARIES 15.2.5.1 Description of the event 76 15.2.5.2 Frequency of Event 76 15.2.5.3 Event Analysis 77 15.2.5.3.1 Protective Features 77 15.2.5.3.2 Single Failures Assumed 78 15.2.5.3.3 Operator Actions Assumed 78 15.2.5.3.4 Chronological Description of Event 78 15.2.5.3.5 Impact on Fission Product Barriers 78 15.2.5.4 Reactor Core and Plant System Evaluation 79 15.2.5.4.1 Input Parameters and Initial Conditions 79 15.2.5.4.2 Method of Analysis 80 15.2.5.4.3 Acceptance Criteria 80 15.2.5.4.4 Results 81 15.2.5.5 Radiological Consequences 81 15.2.5.6 Conclusions 81 15.2.5.7 Supplemental Evaluations 82 15.2.6 LOSS OF NORMAL FEEDWATER FLOW 82 15.2.6.1 Description of Event 82 15.2.6.2 Frequency of Event 83 15.2.6.3 Event Analysis 83 15.2.6.3.1 Protective Features 83 15.2.6.3.2 Single Failures Assumed 84 15.2.6.3.3 Operator Actions Assumed 84 15.2.6.3.4 Chronological Description of Event 84 15.2.6.3.5 Impact on Fission Product Barriers 84 15.2.6.4 Reactor Core and Plant System Evaluation 84 15.2.6.4.1 Input Parameters and Initial Conditions 84 15.2.6.4.2 Method of Analysis 86 15.2.6.4.3 Acceptance Criteria 86 15.2.6.4.4 Results 86 15.2.6.5 Radiological Consequences 87 15.2.6.6 Conclusions 87 Page 6 of 24 Revision 26 5/2016
15.2.6.7 Supplemental Evaluations 87 15.2.7 FEEDWATER SYSTEM PIPE BREAKS 88 15.2.7.1 Description of Event 88 15.2.7.2 Frequency of Event 88 15.2.7.3 Event Analysis 88 15.2.7.3.1 Protective Features 88 15.2.7.3.2 Single Failures Assumed 89 15.2.7.3.3 Operator Actions Assumed 90 15.2.7.3.4 Chronological Description of Event 90 15.2.7.3.5 Impact on Fission Product Barriers 90 15.2.7.4 Reactor Core and Plant System Evaluation 90 15.2.7.4.1 Input Parameters and Initial Conditions 90 15.2.7.4.2 Method of Analysis 92 15.2.7.4.3 Acceptance Criteria 92 15.2.7.4.4 Results 93 15.2.7.5 Radiological Consequences 94 15.2.7.6 Conclusions 94
15.2 REFERENCES
FOR SECTION 15.2 95 Table 15.2-1 TIME SEQUENCE OF EVENTS FOR LOSS OF EXTERNAL ELEC- 96 TRICAL LOAD Table 15.2-2 TIME SEQUENCE OF EVENTS FOR LOSS OF OFFSITE ALTER- 97 NATING CURRENT POWER TO THE STATION AUXILIARIES Table 15.2-3 Table DELETED 98 Table 15.2-4 TIME SEQUENCE OF EVENTS FOR LOSS OF NORMAL FEED- WATER 99 FLOW Table 15.2-5 TIME SEQUENCE OF EVENTS FOR THE FEEDWATER LINE PIPE 100 BREAK (0.3 FT2 BREAK AREA) 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE 101 15.3.1 FLOW COASTDOWN ACCIDENTS 101 15.3.1.1 Description of Event 101 15.3.1.2 Frequency of Event 101 15.3.1.3 Event Analysis 101 15.3.1.3.1 Protective Features 102 15.3.1.3.2 Single Failures Assumed 103 15.3.1.3.3 Operator Actions Assumed 103 15.3.1.3.4 Chronological Description of Event 103 15.3.1.3.5 Impact on Fission Product Barriers 103 15.3.1.4 Reactor Core and Plant System Evaluation 103 Page 7 of 24 Revision 26 5/2016
15.3.1.4.1 Input Parameters and Initial Conditions 103 15.3.1.4.2 Method of Analysis 104 15.3.1.4.3 Acceptance Criteria 104 15.3.1.4.4 Results 104 15.3.1.5 Radiological Consequences 105 15.3.1.6 Conclusions 105 15.3.2 LOCKED ROTOR ACCIDENT 106 15.3.2.1 Description of Event 106 15.3.2.2 Frequency of Event 106 15.3.2.3 Event Analysis 106 15.3.2.3.1 Protective Features 106 15.3.2.3.2 Single Failures Assumed 107 15.3.2.3.3 Operator Actions Assumed 107 15.3.2.3.4 Chronological Description of Event 107 15.3.2.3.5 Impact on Fission Product Barriers 107 15.3.2.4 Reactor Core and Plant System Evaluation 108 15.3.2.4.1 Input Parameters and Initial Conditions 108 15.3.2.4.2 Method of Analysis 108 15.3.2.4.3 Acceptance Criteria 109 15.3.2.4.4 Results 110 15.3.2.5 Radiological Consequences 110 15.3.2.6 Conclusions 110
15.3 REFERENCES
FOR SECTION 15.3 111 Table 15.3-1 TIME SEQUENCE OF EVENTS FOR LOSS OF REACTOR COOL- 112 ANT FLOW Table 15.3-2
SUMMARY
OF LIMITING RESULTS FOR LOCKED ROTOR 113 ACCIDENT Table 15.3-3 TIME SEQUENCE OF EVENTS FOR LOCKED ROTOR INCIDENT 114 Table 15.3-4 LR Dose Analysis Assumptions 115 Table 15.13-5 RESULTS FOR LOCKED ROTOR 116 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 117 15.4.1 UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY WITH- 117 DRAWAL FROM A SUBCRITICAL CONDITION 15.4.1.1 Description of Event 117 15.4.1.2 Frequency of Event 117 15.4.1.3 Event Analysis 117 15.4.1.3.1 Protective Features 117 15.4.1.3.2 Single Failures Assumed 118 15.4.1.3.3 Operator Actions Assumed 118 15.4.1.3.4 Chronological Description of Event 118 Page 8 of 24 Revision 26 5/2016
15.4.1.3.5 Impact on Fission Product Barriers 118 15.4.1.4 Reactor Core and Plant System Evaluation 118 15.4.1.4.1 Input Parameters and Initial Conditions 118 15.4.1.4.2 Methodology 120 15.4.1.4.3 Acceptance Criteria 120 15.4.1.4.4 Results 120 15.4.1.5 Radiological Evaluation 121 15.4.1.6 Conclusions 121 15.4.2 UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY WITH- 121 DRAWAL AT POWER 15.4.2.1 Description of Event 121 15.4.2.2 Frequency of Event 121 15.4.2.3 Event Analysis 121 15.4.2.3.1 Protective Features 122 15.4.2.3.2 Single Failures Assumed 122 15.4.2.3.3 Operator Actions Assumed 122 15.4.2.3.4 Chronological Description of Event 122 15.4.2.3.5 Impact on Fission Product Barriers 122 15.4.2.4 Reactor Core and Plant System Evaluation 123 15.4.2.4.1 Input Parameters and Initial Conditions 123 15.4.2.4.2 Methodology 124 15.4.2.4.3 Acceptance Criteria 124 15.4.2.4.4 Results 125 15.4.2.5 Radiological Evaluation 126 15.4.2.6 Conclusions 126 15.4.3 STARTUP OF AN INACTIVE REACTOR COOLANT LOOP 126 15.4.3.1 Description of Event 126 15.4.3.2 Frequency of Event 127 15.4.3.3 Event Analysis 127 15.4.3.3.1 Protective Features 127 15.4.3.3.2 Single Failures Assumed 127 15.4.3.3.3 Operator Actions Assumed 127 15.4.3.3.4 Chronological Description of Event 127 15.4.3.3.5 Impact on Fission Product Barriers 128 15.4.3.4 Reactor Core and Plant System Evaluation 128 15.4.3.4.1 Input Parameters and Initial Conditions 128 15.4.3.4.2 Methodology 128 15.4.3.4.3 Acceptance Criteria 129 Page 9 of 24 Revision 26 5/2016
15.4.3.4.4 Results 129 15.4.3.4.5 Effect of 18 Month Fuel Cycle Changes 130 15.4.3.5 Radiological Evaluation 130 15.4.3.6 Conclusions 130 15.4.4 CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION 130 15.4.4.1 Description of Event 130 15.4.4.2 Frequency of Event 131 15.4.4.3 Event Analysis 131 15.4.4.3.1 Protective Features and Single Failures Assumed 131 15.4.4.3.1.1 Reactor in Mode 1 or Mode 2 131 15.4.4.3.1.2 Reactor in MODES 3 to 6 132 15.4.4.3.1.3 Indication and Alarms 132 15.4.4.3.2 Operator Actions Assumed 132 15.4.4.3.3 Chronological Description of Event 133 15.4.4.3.4 Impact on Fission Product Barriers 133 15.4.4.4 Reactor Core and Plant System Evaluation 133 15.4.4.4.1 Methodology 133 15.4.4.4.2 Acceptance Criteria 133 15.4.4.4.3 Dilution During Refueling (MODE 6) 134 15.4.4.4.3.1 Input Parameters and Initial Conditions 134 15.4.4.4.3.2 Results 135 15.4.4.4.4 Dilution During Cold Shutdown (MODE 5) 135 15.4.4.4.5 Dilution at Startup (MODE 2) 135 15.4.4.4.5.1 Input Parameters and Initial Conditions 135 15.4.4.4.5.2 Results 136 15.4.4.4.6 Dilution at Power (MODE 1) 136 15.4.4.4.6.1 Input Parameters and Initial Conditions 136 15.4.4.4.6.2 Results 137 15.4.4.4.7 Dilution from a Single Failure While in Residual Heat Removal Mode - 137 Inad- vertent Draining of the Spray Additive Tank.
15.4.4.4.8 Dilution from a Single Failure While in Residual Heat Removal 137 Mode (MODE 5) -Boron Dilution from the Reactor Coolant Drain Tank.
15.4.4.4.8.1 Input Parameters and Initial Conditions 137 15.4.4.4.8.2 Results 138 15.4.4.4.9 Dilution from a Single Failure While in Residual Heat Removal Mode 138 (MODE 5) -Boron Dilution Due to Resin Changing in the Purification Sys- tem.
15.4.4.4.9.1 Input Parameters and Initial Conditions 138 15.4.4.4.9.2 Results 139 Page 10 of 24 Revision 26 5/2016
15.4.4.4.10 Dilution from a Single Failure While in Residual Heat Removal Mode 139 (MODE 6) -Boron Dilution from Reactor Coolant Drain Tank After Refuel- ing.
15.4.4.4.10.1 Input Parameters and Initial Conditions 139 15.4.4.4.10.2 Results 139 15.4.4.5 Radiological Evaluation 140 15.4.4.6 Conclusions 140 15.4.5 RUPTURE OF A CONTROL ROD DRIVE MECHANISM HOUSING 140
- ROD CLUSTER CONTROL ASSEMBLY EJECTION 15.4.5.1 Description of Event 140 15.4.5.1.1 Nuclear Design 141 15.4.5.1.2 Effects on Adjacent Housings 141 15.4.5.2 Frequency of Event 141 15.4.5.3 Event Analysis 141 15.4.5.3.1 Protective Features 141 15.4.5.3.2 Single Failures Assumed 142 15.4.5.3.3 Operator Actions Assumed 142 15.4.5.3.4 Chronological Description of Event 142 15.4.5.3.5 Impact on Fission Product Barriers 142 15.4.5.4 Reactor Core and Plant System Evaluation 143 15.4.5.4.1 Input Parameters and Initial Conditions 143 15.4.5.4.2 Methodology 143 15.4.5.4.2.1 Average Core Analysis 144 15.4.5.4.2.2 Ejected Rod Worths and Hot Channel Factors 144 15.4.5.4.2.3 Hot Spot Analysis 144 15.4.5.4.2.4 Reactivity Feedback Weighting Factors 145 15.4.5.4.2.5 System Overpressure Analysis 145 15.4.5.4.3 Acceptance Criteria 146 15.4.5.4.4 Results 146 15.4.5.4.4.1 Beginning of Life, Full Power - Case (1) 147 15.4.5.4.4.2 Beginning of Life, Zero Power - Case (2) 147 15.4.5.4.4.3 End of Life, Full Power - Case (3) 147 15.4.5.4.4.4 End of Life, Zero Power - Case (4) 147 15.4.5.4.4.5 Pressure Surge 147 15.4.5.4.4.6 Lattice Deformations 148 15.4.5.5 Radiological Evaluation 148 15.4.5.6 Conclusions 148 15.4.6 ROD CLUSTER CONTROL ASSEMBLY DROP 148 Page 11 of 24 Revision 26 5/2016
15.4.6.1 Description of Event 148 15.4.6.2 Frequency of Event 149 15.4.6.3 Event Analysis 149 15.4.6.3.1 Protective Features 149 15.4.6.3.2 Single Failures Assumed 150 15.4.6.3.3 Operator Actions Assumed 150 15.4.6.3.4 Chronological Description of Event 150 15.4.6.3.5 Impact on Fission Product Barriers 150 15.4.6.4 Reactor Core and Plant System Evaluation 150 15.4.6.4.1 Input Parameters and Initial Conditions 150 15.4.6.4.2 Methodology 151 15.4.6.4.2.1 One or More Dropped Rod Cluster Control Assemblies From the Same Group 151 15.4.6.4.2.2 Dropped Rod Cluster Control Assembly Bank 151 15.4.6.4.2.3 Statically Misaligned Rod Cluster Control Assembly 151 15.4.6.4.3 Acceptance Criteria 151 15.4.6.4.4 Results 152 15.4.6.4.4.1 One or More Dropped Rod Cluster Control Assemblies 152 15.4.6.4.4.2 Dropped Rod Cluster Control Assembly Bank 152 15.4.6.4.4.3 Statically Misaligned Rod Cluster Control Assembly 152 15.4.6.5 Radiological Evaluation 153 15.4.6.6 Conclusions 153
15.4 REFERENCES
FOR SECTION 15.4 154 Table 15.4-1 TIME SEQUENCE OF EVENTS FOR UNCONTROLLED ROD CLUSTER 156 CONTROL ASSEMBLY WITHDRAWAL FROM A SUB- CRITICAL Table 15.4-2 TIME SEQUENCE OF EVENTS FOR UNCONTROLLED ROD CLUSTER 157 CONTROL ASSEMBLY WITHDRAWAL AT POWER Table 15.4-3 PARAMETERS USED IN THE ANALYSIS OF THE ROD CLUS- TER 158 CONTROL ASSEMBLY EJECTION ACCIDENT Table 15.4-4 TIME SEQUENCE OF EVENTS FOR ROD CLUSTER CONTROL 159 ASSEMBLY EJECTION Table 15.4-5 REA CONTAINMENT ASSUMPTIONS 160 Table 15.4-6 RESULTS FOR REA DOSE, REM TEDE 162 15.5 INCREASE IN REACTOR COOLANT INVENTORY 163
15.5 REFERENCES
FOR SECTION 15.5 164 15.6 DECREASE IN REACTOR COOLANT INVENTORY 165 Page 12 of 24 Revision 26 5/2016
15.6.1 INADVERTENT OPENING OF A PRESSURIZER SAFETY VALVE OR 165 PRESSURIZER POWER OPERATED RELIEF VALVE (PORV) 15.6.1.1 Description of Event 165 15.6.1.2 Frequency of Event 165 15.6.1.3 Event Analysis 165 15.6.1.3.1 Protective Features 165 15.6.1.3.2 Single Failures Assumed 165 15.6.1.3.3 Operator Actions Assumed 165 15.6.1.3.4 Chronological Description of Event 165 15.6.1.3.5 Impact on Fission Product Barriers 165 15.6.1.4 Reactor Core and Plant System Evaluation 166 15.6.1.4.1 Input Parameters and Initial Conditions 166 15.6.1.4.2 Methodology 166 15.6.1.4.3 Acceptance Criteria 166 15.6.1.4.4 Results 167 15.6.1.5 Radiological Consequences 167 15.6.1.6 Conclusions 167 15.6.2 RADIOLOGICAL CONSEQUENCES OF SMALL LINES 167 CARRYING PRIMARY COOLANT OUTSIDE CONTAINMENT 15.6.3 Steam Generator Tube Rupture 168 15.6.3.1 Description of Event 168 15.6.3.2 Frequency of Event 168 15.6.3.3 Event Analysis 169 15.6.3.3.1 Protective Features 169 15.6.3.3.2 Single Failures Assumed 170 15.6.3.3.2.1 Single Failure - Margin to Overfill 170 15.6.3.3.2.2 Single Failure - Mass Release 171 15.6.3.3.3 Operator Actions Assumed 171 15.6.3.3.3.1 Operator Actions to Terminate Tube Rupture Flow 171 15.6.3.3.3.2 Operator Actions Due to Single Failures 173 15.6.3.3.3.3 Operator Actions for Cooldown to MODE 5 (Cold Shutdown) 173 15.6.3.3.4 Chronological Description of Event 174 15.6.3.3.5 Impact on Fission Product Barriers 174 15.6.3.4 Reactor Core and Plant System Evaluation 175 15.6.3.4.1 Input Parameters and Initial Conditions 175 15.6.3.4.2 Methodology 176 15.6.3.4.3 Acceptance Criteria 177 15.6.3.4.4 Results 177 Page 13 of 24 Revision 26 5/2016
15.6.3.4.4.1 SGTR Margin to Overfill Transient Analysis 177 15.6.3.4.4.2 SGTR Mass Release Transient Analysis 179 15.6.3.5 Radiological Consequences 180 15.6.3.6 Conclusions 181 15.6.4 PRIMARY SYSTEM PIPE RUPTURES 181 15.6.4.1 Loss of Reactor Coolant from Small Ruptured Pipes or From Cracks in Large 181 Pipes Which Actuates Emergency Core Cooling System (ECCS) 15.6.4.1.1 Description of Event 181 15.6.4.1.2 Frequency of Event 182 15.6.4.1.3 Event Analysis 182 15.6.4.1.3.1 Protective Features 182 15.6.4.1.3.2 Single Failures Assumed 183 15.6.4.1.3.3 Operator Actions Assumed 183 15.6.4.1.3.4 Chronological Description of Event 183 15.6.4.1.3.5 Impact on Fission Product Barriers 184 15.6.4.1.4 Reactor Core and Plant System Evaluation 184 15.6.4.1.4.1 Input Parameters and Initial Conditions 184 15.6.4.1.4.2 Methodology 185 15.6.4.1.4.3 Acceptance Criteria 186 15.6.4.1.4.4 Results 186 15.6.4.1.4.5 Effect of Emergency Core Cooling System (ECCS) Evaluation Model Modi- 187 fications 15.6.4.1.5 Radiological Evaluation 187 15.6.4.1.6 Conclusions 187 15.6.4.2 Major Reactor Coolant System Pipe Ruptures (Loss-of-Coolant Accident) 187 15.6.4.2.1 Description of Event 187 15.6.4.2.2 Frequency of Event 189 15.6.4.2.3 Event Analysis 189 15.6.4.2.3.1 Protective Features 189 15.6.4.2.3.2 Single Failures Assumed 190 15.6.4.2.3.3 Operator Actions Assumed 190 15.6.4.2.3.4 Chronological Description of Event 190 15.6.4.2.3.5 Impact on Fission Product Barriers 192 15.6.4.2.4 Reactor Core and Plant System Evaluation 192 15.6.4.2.4.1 Input Parameters and Initial Conditions 192 15.6.4.2.4.2 Methodology 195 15.6.4.2.4.3 Acceptance Criteria 200 Page 14 of 24 Revision 26 5/2016
15.6.4.2.4.4 Results 200 15.6.4.2.5 Radiological Evaluation 201 15.6.4.2.6 Conclusions 202
15.6 REFERENCES
FOR SECTION 15.6 203 Table 15.6-1 COMPARISON OF NOMINAL AND PLANT PARAMETERS USED 207 IN STEAM GENERATOR TUBE RUPTURE (SGTR) ANALYSIS Table 15.6-2 OPERATOR ACTION TIMES 208 Table 15.6-3 SEQUENCE OF EVENTS - MARGIN TO OVERFILL ANALYSIS 209 Table 15.6-4 OPERATOR ACTION TIMES FOR DESIGN BASIS STEAM GEN- 210 ERATOR TUBE RUPTURE ANALYSIS Table 15.6-5 SEQUENCE OF EVENTS - OFFSITE RADIATION DOSE ANALY- 211 SIS Table 15.6-6 SGTR DOSE ANALYSIS ASSUMPTIONS 212 Table 15.6-7 STEAM RELEASES AND RUPTURE FLOW 214 Table 15.6-8 RESULTS FOR SGTR, REM TEDE 215 Table 15.6-9 TIME SEQUENCE OF EVENTS - ACCIDENTAL DEPRESSURIZA- 216 TION OF THE RCS Table 15.6-10 TOTAL SMALL BREAK LOSS-OF-COOLANT ACCIDENT 217 SAFETY INJECTION AND SPILL FLOW Table 15.6-11 SMALL BREAK LOSS-OF-COOLANT ACCIDENT KEY ASSUMP- 218 TIONS Table 15.6-12 SMALL BREAK LOSS-OF-COOLANT ACCIDENT MAIN STEAM 220 SAFETY VALVE (MSSV) ASSUMPTIONS Table 15.6-13 SMALL BREAK LOSS-OF-COOLANT ACCIDENT TIME 221 SEQUENCE OF EVENTS Table 15.6-14 SMALL BREAK LOSS-OF-COOLANT ACCIDENT FUEL 222 CLADDING RESULTS Table 15.6-15 LARGE BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS 223 TIME SEQUENCE OF EVENTS FOR DECLG BREAK Table 15.6-16 Key LBLOCA Parameters and Initial Transient Assumptions for R. 224 E. Ginna Analysis Table 15.6-17 LARGE BREAK LOCA ANALYSIS SAFETY INJECTION 227 FLOW VERSUS PRESSURE Table 15.6-18a PARAMETERS FOR CONTAINMENT PRESSURE - DRY 229 CON- TAINMENT DATA Table 15.6-18b STRUCTURAL HEAT SINK DATA 230 Table 15.6-19 PLANT OPERATING RANGE ALLOWED BY THE BEST- 232 ESTIMATE LARGE BREAK LOCA ANALYSIS (R. E. GINNA)
Page 15 of 24 Revision 26 5/2016
Table 15.6-20 LIMITING LARGE BREAK PCT AND OXIDATION RESULTS 234 FOR R. E. GINNA Table 15.6-21 ASSUMPTIONS FOR ANALYSIS OF RADIOLOGICAL CONSE- 235 QUENCES OF THE LOSS-OF-COOLANT ACCIDENT Table 15.6-21A LBLOCA DOSE
SUMMARY
, REM TEDE 237 Table 15.6-22 Total Core Activity (Curies) at End of 525-day Fuel Cycle - including 238 Decay Table 15.6-23 Core Inventory Fraction Released into Containment 241 Table 15.6-24 TABLE DELETED 243 15.7 RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 244 15.7.1 RADIOACTIVE GAS WASTE SYSTEM FAILURE 244 15.7.1.1 Gas Decay Tank Rupture 244 15.7.1.1.1 Description of Event 244 15.7.1.1.2 Frequency of Event 244 15.7.1.1.3 Event Analysis 244 15.7.1.1.3.1 Single Failures Assumed 245 15.7.1.1.3.2 Operator Actions Assumed 245 15.7.1.1.3.3 Chronological Description of Event 245 15.7.1.1.3.4 Impact on Fission Product Barriers 245 15.7.1.1.4 Reactor Core and Plant System Evaluation 245 15.7.1.1.4.1 Input Parameters and Initial Conditions 245 15.7.1.1.4.2 Methodology 246 15.7.1.1.4.3 Acceptance Criteria 246 15.7.1.1.4.4 Results 246 15.7.1.1.5 Radiological Evaluation 246 15.7.1.1.6 Conclusions 246 15.7.1.2 Volume Control Tank Rupture 247 15.7.1.2.1 Description of Event 247 15.7.1.2.2 Frequency of Event 247 15.7.1.2.3 Event Analysis 247 15.7.1.2.3.1 Single Failures Assumed 247 15.7.1.2.3.2 Operator Actions Assumed 247 15.7.1.2.3.3 Chronological Description of Event 247 15.7.1.2.3.4 Impact on Fission Product Barriers 248 15.7.1.2.4 Reactor Core and Plant System Evaluation 248 15.7.1.2.4.1 Input Parameters and Initial Conditions 248 15.7.1.2.4.2 Methodology 248 Page 16 of 24 Revision 26 5/2016
15.7.1.2.4.3 Acceptance Criteria 249 15.7.1.2.4.4 Results 249 15.7.1.2.5 Radiological Evaluation 249 15.7.1.2.6 Conclusions 249 15.7.2 RADIOACTIVE LIQUID WASTE SYSTEM FAILURE 249 15.7.2.1 Description of Event 249 15.7.2.2 Frequency of Event 250 15.7.2.3 Event Analysis 250 15.7.2.3.1 Single Failures Assumed 251 15.7.2.3.2 Operator Actions Assumed 251 15.7.2.3.3 Chronological Description of Event 251 15.7.2.3.4 Impact on Fission Product Barriers 251 15.7.2.4 Reactor Core and Plant System Evaluation 251 15.7.2.4.1 Input Parameters and Initial Conditions 251 15.7.2.4.2 Methodology 252 15.7.2.4.3 Acceptance Criteria 252 15.7.2.4.4 Results 252 15.7.2.4.4.1 Accidental Release of Liquid Waste Assessment 252 15.7.2.4.4.2 Spent Resin Storage Tank Assessment 253 15.7.2.4.5 Effects of 18-month Fuel Cycle 254 15.7.2.5 Radiological Evaluation 254 15.7.2.6 Conclusions 254 15.7.3 FUEL HANDLING ACCIDENTS 254 15.7.3.1 Description of Event 254 15.7.3.1.1 MODE 6 (Refueling) Preparations 254 15.7.3.1.2 Fuel Handling Equipment Safety Features 255 15.7.3.1.3 Fuel Handling Operations Precautions 256 15.7.3.1.4 Consequence of Dropped Fuel Assembly 256 15.7.3.2 Frequency of Event 257 15.7.3.3 Event Analysis 257 15.7.3.3.1 Protective Features 258 15.7.3.3.2 Single Failures Assumed 258 15.7.3.3.3 Operator Actions Assumed 258 15.7.3.3.4 Chronological Description of Event 258 15.7.3.3.5 Impact on Fission Product Barriers 258 15.7.3.4 Reactor Core and Plant System Evaluation 259 15.7.3.4.1 Input Parameters and Initial Conditions 259 Page 17 of 24 Revision 26 5/2016
15.7.3.4.2 Methodology 259 15.7.3.4.3 Acceptance Criteria 259 15.7.3.4.4 Results 259 15.7.3.5 Radiological Evaluation 259 15.7.3.6 Conclusions 260
15.7 REFERENCES
FOR SECTION 15.7 261 Table 15.7-1 FISSION PRODUCT INVENTORY AND ACTIVITY RELEASED 264 FROM POOL Table 15.7-2 FHA DOSE ANALYSIS ASSUMPTIONS 265 Table 15.7-3 FHA DOSE. REM TEDE 266 Table 15.7-4 Table DELETED 267 Table 15.7-5 Table DELETED 268 Table 15.7-6 Table DELETED 269 15.8 ANTICIPATED TRANSIENTS WITHOUT SCRAM 270 15.8.1 ANTICIPATED TRANSIENTS WITHOUT SCRAM (atws) 270 15.8.2 frequency of event 270 15.8.3 Event Analysis 270 15.8.3.1 Single Failures Assumed 270 15.8.3.2 Operator Actions Assumed 270 15.8.3.3 Chronological Description of Event 271 15.8.3.4 Impact on Fission Product Barriers 271 15.8.4 Reactor Core and Plant System Evaluation 272 15.8.4.1 Input Parameters and Initial Conditions 272 15.8.4.2 Methodology 273 15.8.4.3 Acceptance Criteria 273 15.8.4.4 Results 273 15.8.5 Radiological Evaluation 274 15.8.6 Conclusions 274
15.8 REFERENCES
FOR SECTION 15.8 275 FIGURES Figure 15.0-1 Core Limits and Overpower-Overtemperature Delta T Setpoints (Tref =
576.0F)
Figure 15.0-2 Reactivity Coefficients Used in Non-LOCA Safety Analysis Figure 15.0-3 Reactivity Insertion Scram Curves Figure 15.1-1 Feedwater Flow Increase at Full Power, Nuclear Power and Loop Average Temperature Versus Time Figure 15.1-2 Feedwater Flow Increase at Full Power, Pressurizer Pressure and Steam Gen-erator Pressure Versus Time Page 18 of 24 Revision 26 5/2016
Figure 15.1-3 Feedwater Flow Increase at Full Power, Steam Generator Mass Versus Time Figure 15.1-4 Steam Line Rupture, Multiplication Factor Versus Core Average Temperature (Calculated at 1050 psia)
Figure 15.1-5 Steam Line Rupture, Integrated Doppler Defect Versus Fraction of Power Figure 15.1-6 Steam Line Rupture, 1.4ft2 Break with Power, Two Loops in Service, Core Heat Flux and Nuclear Power Versus Time Figure 15.1-7 Steam Line Rupture, 1.4ft2 Break with Power, Two Loops in Service, Pressur-izer Water Volume and Pressurizer Pressure Versus Time Figure 15.1-8 Steam Line Rupture, 1.4ft2 Break with Power, Two Loops in Service, Loop TAVG and Cold Leg Loop Temperature Versus Time Figure 15.1-9 Steam Line Rupture, 1.4ft2 Break with Power, Two Loops in Service, Faulted Loop Steam Flow and Total Feedwater Flow Versus Time Figure 15.1-10 Steam Line Rupture, 1.4ft2 Break with Power, Two Loops in Service, Core Averaged Boron and Reactivity Versus Time Figure 15.1-11 Steam Line Rupture, 1.4ft2 Break Without Power, Two Loops in Service, Core Heat Flux and Nuclear Power Versus Time Figure 15.1-12 Steam Line Rupture, 1.4ft2 Break Without Power, Two Loops in Service, Pressurizer Water Volume and Pressurizer Pressure Versus Time Figure 15.1-13 Steam Line Rupture, 1.4ft2 Break Without Power, Two Loops in Service, Loop TAVG and Cold Leg Loop Temperatures Versus Time Figure 15.1-14 Steam Line Rupture, 1.4ft2 Break Without Power, Two Loops in Service, Faulted Loop Steam Flow and Total Feedwater Flow Versus Time Figure 15.1-15 Steam Line Rupture, 1.4ft2 Break without Power, Two Loops in Service, Core Averaged Boron and Reactivity Versus Time Figure 15.1-16 Steam Line Rupture, 1.4ft2 Break with Power, One Loop in Service, Core Heat Flux and Nuclear Power Versus Time Figure 15.1-17 Steam Line Rupture, 1.4ft2 Break with Power, One Loop in Service, Pressur-izer Water Volume and Pressurizer Pressure Versus Time Figure 15.1-18 Steam Line Rupture, 1.4ft2 Break with Power, One Loop in Service, Loop TAVG and Cold Leg Loop Temperatures Versus Time Figure 15.1-19 Steam Line Rupture, 1.4ft2 Break with Power, One Loop in Service, Faulted Loop Steam Flow and Total Feedwater Flow Versus Time Figure 15.1-20 Steam Line Rupture, 1.4ft2 Break with Power, One Loop in Service, Core Averaged Boron and Reactivity Versus Time Figure 15.1-21 Combined Atmospheric Relief Valve and Main Feedwater Regulating Valve Failure, Nuclear Power and Core Heat Flux Versus Time Figure 15.1-22 Combined Atmospheric Relief Valve and Main Feedwater Regulating Valve Failure, Loop Average Temperature and Pressurizer Pressure Versus Time Figure 15.1-23 Combined Atmospheric Relief Valve and Main Feedwater Regulating Valve Failure, DNBR Versus Time Figure 15.1-24 Combined Atmospheric Relief Valve and Main Feedwater Regulating Valve Failure, Steam Generator Level and Steam Generator Mass Versus Time Page 19 of 24 Revision 26 5/2016
Figure 15.2-1 Loss of Load, with Automatic Pressure Control, Nuclear Power and DNBR Versus Time Figure 15.2-2 Loss of Load, with Automatic Pressure Control, RCS Average Temperature and Pressurizer Water Volume Versus Time Figure 15.2-3 Loss of Load, with Automatic Pressure Control, Steam Generator Pressure and Pressurizer Pressure Versus Time Figure 15.2-4 Loss of Load, Without Pressure Control, Nuclear Power Versus Time Figure 15.2-5 Loss of Load, Without Pressure Control, RCS Average Temperature and Pres-surizer Water Volume Versus Time Figure 15.2-6 Loss of Load, Without Pressure Control, Steam Generator Pressure and Reac-tor Coolant System Pressures Versus Time Figure 15.2-7 Loss of Load, Peak MSS Pressure Case, Nuclear Power Versus Time Figure 15.2-8 Loss of Load, Peak MSS Pressure Case, RCS Average Temperature and Pres-surizer Water Volume Versus Time Figure 15.2-9 Loss of Load, Peak MSS Pressure Case, Steam Generator Pressure and Pres-surizer Pressure Versus Time Figure 15.2-10 Figure Deleted Figure 15.2-11 Figure Deleted Figure 15.2-12 Figure Deleted Figure 15.2-13 Loss of Offsite Alternating Current Power to the Station Auxiliaries, Nuclear Power and Pressurizer Pressure Versus Time Figure 15.2-14 Loss of Offsite Alternating Current Power to the Station Auxiliaries, Pressur-izer Water Volume and Pressurizer Steam Relief Rate Versus Time Figure 15.2-15 Loss of Offsite Alternating Current Power to the Station Auxiliaries, Reactor Coolant Flow and Core Inlet/Outlet Temperatures Versus Time Figure 15.2-16 Loss of Offsite Alternating Current Power to the Station Auxiliaries, Steam Generator Mass and Steam Generator Pressure Versus Time Figure 15.2-17 Loss of Normal Feedwater With Power, Nuclear Power and Pressurizer Pres-sure Versus Time Figure 15.2-18 Loss of Normal Feedwater With Power, Pressurizer Water Volume and Pres-surizer Steam Relief Rate Versus Time Figure 15.2-19 Loss of Normal Feedwater With Power, Reactor Coolant Flow and Core Inlet/
Outlet Temperatures Versus Time Figure 15.2-20 Loss of Normal Feedwater With Power, Steam Generator Mass and Steam Generator Pressure Versus Time Figure 15.2-21 Feedline Break With Offsite Power; Nuclear Power and Pressurizer Pressure Versus Time Figure 15.2-22 Feedline Break With Offsite Power; Pressurizer Water Volume and Pressur-izer Steam Relief Rate Versus Time Figure 15.2-23 Feedline Break With Offsite Power; Cold Leg, Hot Leg and Saturation Tem-peratures Versus Time Figure 15.2-24 Feedline Break With Offsite Power; Steam Generator Mass and Steam Gener-ator Pressure Versus Time Figure 15.2-25 Feedline Break With Offsite Power; Feedwater Mass Flow Rates Versus Time Page 20 of 24 Revision 26 5/2016
Figure 15.2-26 Feedline Break Without Offsite Power; Nuclear Power and Pressurizer Pres-sure Versus Time Figure 15.2-27 Feedline Break Without Offsite Power; Pressurizer Water Volume and Pres-surizer Steam Relief Rate Versus Time Figure 15.2-28 Feedline Break Without Offsite Power; Cold Leg, Hot Leg and Saturation Temperatures Versus Time Figure 15.2-29 Feedline Break Without Offsite Power; Steam Generator Mass and Steam Generator Pressure Versus Time Figure 15.2-30 Feedline Break Without Offsite Power; Feedwater Mass Flow Rates Versus Time Figure 15.3-1 Full Loss of Flow (Undervoltage), Nuclear Power and RCS Flow Versus Time Figure 15.3-1a Full Loss of Flow (Underfrequency), Nuclear Power and RCS Flow Versus Time Figure 15.3-2 Full Loss of Flow (Undervoltage), Core Average and Hot Channel Heat Flux Versus Time Figure 15.3-2a Full Loss of Flow (Underfrequency), Core Average and Hot Channel Heat Flux Versus Time Figure 15.3-3 Full Loss of Flow (Undervoltage), RCS Pressures and DNBR Versus Time Figure 15.3-3a Full Loss of Flow (Underfrequency), DNBR and Reactor Coolant System Pressures Versus Time Figure 15.3-4 Partial Loss of Flow, Nuclear Power and RCS Flow Versus Time Figure 15.3-5 Partial Loss of Flow, RCS Pressures and RCS Loop Flows Versus Time Figure 15.3-6 Partial Loss of Flow, Core Average and Hot Channel Heat Flux Versus Time Figure 15.3-7 Partial Loss of Flow, DNBR Versus Time Figure 15.3-8 Locked Rotor, RCS Pressures and RCS Loop Flows Versus Time Figure 15.3-9 Locked Rotor, Nuclear Power and RCS Flow Versus Time Figure 15.3-10 Locked Rotor, Core Average Heat Flux and Cladding Inside Temperature Ver-sus Time Figure 15.4-1 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal From Subcrit-ical Conditions, Heat Flux and Nuclear Power Versus Time (422V+Fuel)
Figure 15.4-2 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal From Subcrit-ical Conditions, Clad Inside and Fuel Average Temperature Versus Time(422V+Fuel)
Figure 15.4-3 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Minimum Feedback, 100 pcm/sec, Nuclear Power and Heat Flux Versus Time Figure 15.4-4 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Minimum Feedback, 100 pcm/sec, Pressurizer Pressure and Pressurizer Water Volume Versus Time Figure 15.4-5 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Minimum Feedback, 100 pcm/sec, Tavg and DNBR Versus Time Figure 15.4-6 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Maximum Feedback, 5 pcm/sec, Nuclear Power and Heat Flux Versus Time Page 21 of 24 Revision 26 5/2016
Figure 15.4-7 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Maximum Feedback, 5 pcm/sec, Pressurizer Water Volume and Pressurizer Pressure Versus Time Figure 15.4-8 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Maximum Feedback, 5 pcm/sec, TAVG and DNBR Versus Time Figure 15.4-9 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Minimum and Maximum Feedback, DNBR Versus Reactivity Insertion Rate Figure 15.4-10 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from 60%
Power, Minimum and Maximum Feedback, DNBR Versus Reactivity Inser-tion Rate Figure 15.4-11 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from 10%
Power, Minimum and Maximum Feedback, DNBR Versus Reactivity Inser-tion Rate Figure 15.4-12 Startup of an Inactive Coolant Loop, Nuclear Power Versus Time Figure 15.4-13 Startup of an Inactive Coolant Loop, TAVG Versus Time Figure 15.4-14 Startup of an Inactive Coolant Loop, Core Inlet Temperature Versus Time Figure 15.4-15 Startup of an Inactive Coolant Loop, Pressurizer Pressure Versus Time Figure 15.4-16 Rod Cluster Control Assembly Ejection Beginning-of-Life, Full Power, Fuel and Clad Temperature and Nuclear Power Versus Time Figure 15.4-16a Rod Cluster Control Assembly Ejection, Beginning of Life, Full Power, Fuel and Clad Temperature and Nuclear Power Versus Time Figure 15.4-16b Rod Cluster Control Assembly Ejection, Beginning of Life, Zero Power, Fuel and Clad Temperature and Nuclear Power Versus Time Figure 15.4-17 Rod Cluster Control Assembly Ejection Beginning-of-Life, Zero Power, Fuel and Clad Temperature and Nuclear Power Versus Time Figure 15.4-17a Rod Cluster Control Assembly Ejection, End of Life, Full Power, Fuel and Clad Temperature and Nuclear Power Versus Time Figure 15.4-17b Rod Cluster Control Assembly Ejection, End of Life, Zero Power, Fuel and Clad Temperature and Nuclear Power Versus Time Figure 15.4-18 Rod Cluster Control Assembly Drop Heat Flux and Nuclear Power Versus Time Figure 15.4-19 Rod Cluster Control Assembly Drop Pressurizer Pressure and Core Average Temperature Versus Time Figure 15.4-20 Uncontrolled Rod Cluster Control Assembly Bank Withdrawl from 8% Power (RCS Pressure Case), Minimum Feedback, 55 pcm/sec, Nuclear Power and Heat Flux Versus Time Figure 15.4-21 Uncontrolled Rod Cluster Control Assembly Bank Withdrawl from 8% Power (RCS Pressure Case), Minimum Feedback, 55 pcm/sec, Pressurizer Pressure and Tavg Versus Time Figure 15.6-1 Steam Generator Tube Rupture (Overfill), Maximum Safety Injection Flow Versus Pressure Figure 15.6-1a RCS Depressurization, Nuclear Power Versus Time Figure 15.6-1b RCS Pressurization, Pressurizer Pressure Versus Time Figure 15.6-1c RCS Depressurization, Indicated Loop Average Temperature Versus Time Figure 15.6-1d RCS Depressurization, DNBR Versus Time Page 22 of 24 Revision 26 5/2016
Figure 15.6-2 SGTR (Overfill), Pressurizer Level and Pressurizer Pressure Versus Time Figure 15.6-3 SGTR (Overfill), Secondary Pressure and Steam Generator Liquid Mass Ver-sus Time Figure 15.6-4 SGTR (Overfill), Hot and Cold Leg Temperatures for Intact and Ruptured Steam Generators Versus Time Figure 15.6-5 SGTR (Overfill), Total Primary to Secondary Leakage and Total Integrated Primary to Secondary Leakage Versus Time Figure 15.6-6 SGTR (Overfill), Steam Generator Relief Flow and Integrated Steam Genera-tor Relief Flow Versus Time Figure 15.6-7 SGTR (Overfill), Steam Generator Water Volume Versus Time Figure 15.6-8 SGTR (Dose), Pressurizer Level and Pressurizer Pressure Versus Time Figure 15.6-9 SGTR (Dose), Secondary Pressure and Steam Generator Liquid Mass Versus Time Figure 15.6-10 SGTR (Dose), Hot and Cold Leg Temperatures for Intact and Ruptured Steam Generators Versus Time Figure 15.6-11 SGTR (Dose), Total Primary to Secondary Leakage and Total Integrated Pri-mary to Secondary Leakage Versus Time Figure 15.6-12 SGTR (Dose), Steam Generator Relief Flow and Integrated Steam Generator Relief Flow Versus Time Figure 15.6-13 SGTR (Dose), Steam Generator Water Volume Versus Time Figure 15.6-14 SGTR (Dose), Tube Rupture Flow Flashing Fraction and Integrated Flashed Break Versus Time Figure 15.6-15 Small Break LOCA Inch Break, Pressurizer Pressure Versus Time Figure 15.6-16 Small Break LOCA Inch Break, Core Mixture Level Versus Time Figure 15.6-17 Small Break LOCA Inch High Break, Peak Cladding Temperature at PCT Elevation Versus Time Figure 15.6-18 Small Break LOCA Inch High Break, Core Exit Vapor Flow Versus Time Figure 15.6-19 Small Break LOCA Inch Break, Hot Rod Heat Transfer Coefficient at PCT Elevation Versus Time Figure 15.6-20 Small Break LOCA Inch Break, Fluid Temperature at PCT Elevation Ver-sus Time Figure 15.6-21 Small Break LOCA - Axial Power Distribution, Heat Rate Versus Core Eleva-tion Figure 15.6-22 Small Break LOCA - 1.5-Inch Break, Pressurizer Pressure Versus Time Figure 15.6-23 Small Break LOCA Inch High Break, Pressurizer Pressure Versus Time Figure 15.6-24 Small Break LOCA - 1.5-Inch Break, Core Mixture Level Versus Time Figure 15.6-25 Small Break LOCA Inch Break, Core Mixture Level Versus Time Figure 15.6-26 Small Break LOCA - 1.5-Inch Break, Peal Cladding Temperature at PCT Ele-vation Versus Time Figure 15.6-27 Small Break LOCA Inch Break, Peak Cladding Temperature at PCT Ele-vation Versus Time Figure 15.6-28 Figure Deleted Figure 15.6-29 Figure Deleted Figure 15.6-30 Figure Deleted Page 23 of 24 Revision 26 5/2016
Figure 15.6-31 R.E. Ginna Vessel Model Noding Diagram1 Figure 15.6-32 R.E. Ginna Loop Model Noding Diagram Figure 15.6-33 R.E. Ginna Initial Transient Axial Power Distributions Figure 15.6-34 Containment Pressure Used for the R.E. Ginna Best-Estimate Large Break LOCA Initial Transient Figure 15.6-35 Peak Clad Temperature of the 5 rods for the Initial Transient Figure 15.6-36 Split Break Flow for the Initial Transient Figure 15.6-37 Total Flow at the Bottom of the Core for the Initial Transient Figure 15.6-38 Accumulator Injection Flow for the Initial Transient Figure 15.6-39 High Head Safety Injection Flow for the Initial Transient Figure 15.6-40 Low Head Safety Injection Flow for the Initial Transient Figure 15.6-41 Average Collapsed Liquid Level in the Downcomer for the Initial Transient Figure 15.6-42 Lower Plenum Collapsed Liquid Level for the Initial Transient Figure 15.6-43 Core Collapsed Liquid Levels for the Initial Transient Figure 15.6-44 Vessel Liquid Mass for the Initial Transient Figure 15.6-45 Pressurizer Pressure for the Initial Transient Figure 15.6-46 Hot Rod Peak Clad Temperature and Elevation for the Initial Transient Figure 15.6-47 R.E. Ginna PBOT/PMID Analysis and Operating Limits Figure 15.6-48 Lower Bound Containment Pressure for R.E. Ginna Analysis Page 24 of 24 Revision 26 5/2016
GINNA/UFSAR 17 QUALITY ASSURANCE 1 17.1 QUALITY ASSURANCE DURING DESIGN AND CONSTRUC- 2 TION 17.1.1 ORGANIZATION 2 17.1.2 QUALITY ASSURANCE PROGRAM 2 17.1.2.1 General 2 17.1.2.2 Rochester Gas and Electric Corporation 2 17.1.2.3 Westinghouse 3 17.1.2.3.1 General 3 17.1.2.3.2 Westinghouse Organization 4 17.1.2.3.3 Components Supplied By Westinghouse 4 17.1.2.3.4 Supplier Evaluation 5 17.1.2.3.5 Equipment Specifications 5 17.1.2.3.6 Purchase Order Review 6 17.1.2.3.7 Supplier Surveillance 6 17.1.2.3.8 Instrumentation and Control Equipment 7 17.1.2.3.9 Shipment of Components 8 17.1.2.3.10 Inspection and Installation of Equipment in the Field 9 17.1.2.3.11 Nonconforming Components or Material 10 17.1.2.3.12 Quality Control Records 10 17.1.2.4 Gilbert Associates, Inc 10 17.1.2.5 Bechtel Corporation 11 17.1.2.5.1 General 11 17.1.2.5.2 Material Certification 11 17.1.2.5.3 Backfill 12 17.1.2.5.4 Containment Liner 12 17.1.2.5.5 Piping 12 17.1.2.5.6 Stainless Steel Liners 12 17.1.2.5.7 Containment Tendons 12 17.1.2.5.8 Reinforcing Bar and Cadwelds 13 17.1.2.5.9 Machinery Setting and Alignment 13 Page 1 of 2 Revision 26 5 /2016
GINNA/UFSAR 17 QUALITY ASSURANCE 1 17.1.2.5.10 Concrete 13 17.1.2.5.11 Documentation 14 17.1.2.5.12 Site Cleanliness 14 17.1.3 QUALITY CONTROL ORGANIZATION FOR REACTOR CON- 14 TAINMENT STRUCTURE DESIGN AND ERECTION 17.1.3.1 General 14 17.1.3.2 Rochester Gas and Electric Corporation 14 17.1.3.3 Westinghouse Electric Corporation 15 17.1.3.4 Gilbert Associates, Inc 15 17.1.3.5 Pittsburgh Testing Laboratory 16 17.1.3.6 Dames & Moore 16 17.1.3.7 Bechtel Corporation 16 17.2 QUALITY ASSURANCE DURING THE OPERATIONS PHASE 18 FIGURES Figure 17.1-1 Quality Control Relationships During Construction Page 2 of 2 Revision 26 5 /2016
GINNA/UFSAR 18 UPDATED FINAL SAFETY ANALYSIS REPORT 1 SUPPLEMENT FOR LICENSE RENEWAL
18.1 INTRODUCTION
2 18.1.1 RENEWED FACILITY OPERATING LICENSE 2 18.1.2 UPDATED FINAL SAFETY ANALYSIS REPORT SUPPLEMENT 2 18.1.3 STRUCTURES, SYSTEMS, OR COMPONENTS ADDED SUBSE- 2 QUENT TO RENEWED LICENSE 18.2 Programs that Manage the Effects of Aging 5 18.2.1 Aging Management Programs 5 18.2.1.1 Aboveground Carbon Steel Tanks 5 18.2.1.2 ASME Section XI, Subsections IWB, IWC, & IWD Inservice Inspec- 5 tion 18.2.1.3 ASME Section XI, Subsections IWE & IWL Inservice Inspection 5 18.2.1.4 ASME Section XI, Subsection IWF Inservice Inspection 6 18.2.1.5 Bolting Integrity 6 18.2.1.6 Boric Acid Corrosion 6 18.2.1.7 Buried Piping and Tanks Inspection 6 18.2.1.8 Closed-Cycle (Component) Cooling Water System 6 18.2.1.9 Electrical Cables and Connections Not Subject to 10 CFR 50.49 Envi- 7 ronmental Qualification Requirements 18.2.1.10 Electrical Cables Not Subject to 10CFR50.49 Environmental Qualifi- 7 cation Requirements Used In Instrument Circuits 18.2.1.11 Fire Protection 7 18.2.1.12 Fire Water System 7 18.2.1.13 Flow-Accelerated Corrosion 7 18.2.1.14 Fuel Oil Chemistry 8 18.2.1.15 Inaccessible Medium-Voltage Cables Not Subject to 10CFR50.49 8 Environmental Qualification Requirements 18.2.1.16 Inspection of Overhead Heavy Load and Light Load (Related to Refu- 8 eling) Handling Systems 18.2.1.17 One-Time Inspection 8 18.2.1.18 Open-Cycle Cooling (Service) Water System 8 18.2.1.19 Periodic Surveillance and Preventive Maintenance 9 18.2.1.20 Nickel-Alloy Nozzles and Penetrations Inspection 9 Page 1 of 3 Revision 26 5 /2016
GINNA/UFSAR 18 UPDATED FINAL SAFETY ANALYSIS REPORT 1 SUPPLEMENT FOR LICENSE RENEWAL 18.2.1.21 Reactor Vessel Internals 9 18.2.1.22 Reactor Vessel Surveillance 9 18.2.1.23 Spent Fuel Pool Neutron Absorber Monitoring 10 18.2.1.24 Steam Generator Integrity 10 18.2.1.25 Structures Monitoring 10 18.2.1.26 Systems Monitoring 10 18.2.1.27 Thimble Tubes Inspection 10 18.2.1.28 Water Chemistry Control 11 18.3 Evaluation of Time-Limited Aging Analyses 12 18.3.1 Reactor Vessel Neutron Embrittlement 12 18.3.1.1 Upper Shelf Energy 12 18.3.1.2 Pressurized Thermal Shock 12 18.3.1.3 Pressure-Temperature Limits 13 18.3.2 Thermal Aging Embrittlement of Cast Austenitic Stainless Steel 13 (CASS) 18.3.3 Metal Fatigue 14 18.3.3.1 ASME Boiler and Pressure Vessel Code,Section III, Class 1 14 18.3.3.2 Reactor Vessel Underclad Cracking 15 18.3.3.3 ANSI B31.1 Piping 15 18.3.3.4 Accumulator Check Valves 16 18.3.3.5 Environmentally Assisted Fatigue 16 18.3.3.6 Reactor Vessel Nozzle-to-Weld Defect 17 18.3.3.7 Pressurizer Fracture Mechanics Analysis 17 18.3.4 Environmental Qualification of Electric Equipment 18 18.3.5 Concrete Containment Tendon Prestress 19 18.3.5.1 Containment Tendon Fatigue 19 18.3.5.2 Containment Tendon Bellows Fatigue 19 18.3.6 Containment Liner Plate and Penetration Fatigue 20 18.3.6.1 Containment Liner Anchorage Fatigue 20 18.3.7 Containment Liner Stress 20 18.3.8 Other Time-Limited Aging Analyses 21 Page 2 of 3 Revision 26 5 /2016
GINNA/UFSAR 18 UPDATED FINAL SAFETY ANALYSIS REPORT 1 SUPPLEMENT FOR LICENSE RENEWAL 18.3.8.1 Crane Load Cycle Limit 21 18.3.8.2 Reactor Coolant Pump (RCP) Flywheel 21 18.3.9 Exemptions 21 18.4 TLAA Supporting Activities 22 18.4.1 Concrete Containment Tendon Prestress 22 18.4.2 Environmental Qualification Program 22 18.4.3 Fatigue Monitoring Program 22 Page 3 of 3 Revision 26 5 /2016