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| number = ML081820238
| number = ML081820238
| issue date = 05/23/2008
| issue date = 05/23/2008
| title = Joseph M. Farley, Units 1 and 2, Updated Final Safety Analysis Report, Revision 21, Chapter 3, Table of Contents Through Appendix 3J, Page 3J-7
| title = Updated Final Safety Analysis Report, Revision 21, Chapter 3, Table of Contents Through Appendix 3J, Page 3J-7
| author name =  
| author name =  
| author affiliation = Southern Nuclear Operating Co, Inc
| author affiliation = Southern Nuclear Operating Co, Inc
Line 15: Line 15:


=Text=
=Text=
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3-i REV 21  5/08 3.0  DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS TABLE OF CONTENTS Page
 
3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA..................................3.1-1
 
3.1.1 Criterion 1 - Quality Standards and Records.............................................3.1-1 3.1.2 Criterion 2 - Design Bases for Protection Against Natural  Phenomena...............................................................................................3.1-2  3.1.3 Criterion 3 - Fire Protection.......................................................................3.1-2 3.1.4 Criterion 4 - Environmental and Missile Design Bases..............................3.1-3 3.1.5 Criterion 5 - Sharing of Structures, Systems, and Components................3.1-4 3.1.6 Criterion 10 - Reactor Design....................................................................3.1-5 3.1.7 Criterion 11 - Reactor Inherent Protection.................................................3.1-5 3.1.8 Criterion 12 - Suppression of Reactor Power Oscillations.........................3.1-6 3.1.9 Criterion 13 - Instrumentation and Control................................................3.1-6 3.1.10 Criterion 14 - Reactor Coolant Pressure Boundary...................................3.1-7 3.1.11 Criterion 15 - Reactor Coolant System Design..........................................3.1-8 3.1.12 Criterion 16 - Containment Design............................................................3.1-9 3.1.13 Criterion 17 - Electric Power Systems.......................................................3.1-9 3.1.14 Criterion 18 - Inspection and Testing of Electric Power Systems............3.1-11 3.1.15 Criterion 19 - Control Room.....................................................................3.1-11 3.1.16 Criterion 20 - Protection System Functions.............................................3.1-12 3.1.17 Criterion 21 - Protection System Reliability and Testability.....................3.1-13 3.1.18 Criterion 22 - Protection System Independence......................................3.1-14 3.1.19 Criterion 23 - Protection System Failure Modes......................................3.1-15 3.1.20 Criterion 24 - Separation of Protection and Control Systems..................3.1-16 3.1.21 Criterion 25 - Protection System Requirements for Reactivity Control Malfunctions............................................................................................3.1-17  3.1.22 Criterion 26 - Reactivity Control System Redundancy  and  Capability.................................................................................................3.1-17  3.1.23 Criterion 27 - Combined Reactivity Control Systems Capability..............3.1-18 3.1.24 Criterion 28 - Reactivity Limits.................................................................3.1-18 3.1.25 Criterion 29 - Protection Against Anticipated Operational  Occurrences............................................................................................3.1-19  3.1.26 Criterion 30 - Quality of Reactor Coolant Pressure Boundary.................3.1-19 3.1.27 Criterion 31 - Fracture Prevention of Reactor Coolant  Pressure Boundary..................................................................................3.1-20  3.1.28 Criterion 32 - Inspection of Reactor Coolant Pressure Boundary............3.1-21 3.1.29 Criterion 33 - Reactor Coolant Makeup...................................................3.1-21 3.1.30 Criterion 34 - Residual Heat Removal.....................................................3.1-22 3.1.31 Criterion 35 - Emergency Core Cooling...................................................3.1-22 3.1.32 Criterion 36 - Inspection of Emergency Core Cooling System................3.1-24 3.1.33 Criterion 37 - Testing of Emergency Core Cooling System.....................3.1-24 3.1.34 Criterion 38 - Containment Heat Removal...............................................3.1-25 3.1.35 Criterion 39 - Inspection of Containment Heat Removal System............3.1-26 FNP-FSAR-3
 
3-ii REV 21  5/08 TABLE OF CONTENTS        Page
 
3.1.36 Criterion 40 - Testing of Containment Heat Removal System.................3.1-26 3.1.37 Criterion 41 - Containment Atmosphere Cleanup....................................3.1-27 3.1.38 Criterion 42 - Inspection of Containment Atmosphere Cleanup  Systems...................................................................................................3.1-28  3.1.39 Criterion 43 - Testing of Containment Atmosphere Cleanup  Systems...................................................................................................3.1-28  3.1.40 Criterion 44 - Cooling Water....................................................................3.1-29 3.1.41 Criterion 45 - Inspection of Cooling Water System..................................3.1-29 3.1.42 Criterion 46 - Testing of Cooling Water System......................................3.1-30 3.1.43 Criterion 50 - Containment Design Basis................................................3.1-30 3.1.44 Criterion 51 - Fracture Prevention of Containment Pressure  Boundary.................................................................................................3.1-31  3.1.45 Criterion 52 - Capability for Containment Leakage Rate Testing............3.1-32 3.1.46 Criterion 53 - Provisions for Containment Testing and Inspection..........3.1-32 3.1.47 Criterion 54 - Piping Systems Penetrating Containment.........................3.1-32 3.1.48 Criterion 55 - Reactor Coolant Pressure Boundary Penetrating  Containment............................................................................................3.1-33  3.1.49 Criterion 56 - Primary Containment Isolation...........................................3.1-34 3.1.50 Criterion 57 - Closed System Isolation Valves........................................3.1-34 3.1.51 Criterion 60 - Control of Release of Radioactive Materials to the  Environment............................................................................................3.1-35  3.1.52 Criterion 61 - Fuel Storage and Handling and Radioactivity Control.......3.1-35 3.1.53 Criterion 62 - Prevention of Criticality in Fuel Storage and Handling......3.1-37 3.1.54 Criterion 63 - Monitoring Fuel and Waste Storage..................................3.1-37 3.1.55 Criterion 64 - Monitoring Radioactivity Releases.....................................3.1-38
 
3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS AND SYSTEMS.................3.2-1
 
3.2.1 Seismic Classification................................................................................3.2-1
 
3.2.1.1 Definitions................................................................................3.2-1 3.2.1.2 Category I Structures...............................................................3.2-1 3.2.1.3 Category I Mechanical Components and Systems..................3.2-2 3.2.1.4 Category I Electrical Equipment...............................................3.2-2 3.2.1.5 Category I Instrumentation and Control Systems  Equipment................................................................................3.2-4  3.2.1.6 Structures and Systems of Mixed Category.............................3.2-4
 
3.2.2 System Quality Group Classification.........................................................3.2-4
 
3.3 WIND AND TORNADO LOADINGS............................................................................3.3-1
 
3.3.1 Wind Loadings...........................................................................................3.3-1
 
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3-iii REV 21  5/08 TABLE OF CONTENTS        Page
 
3.3.1.1 Design Wind Velocity...............................................................3.3-1 3.3.1.2 Basis for Wind Velocity Selection............................................3.3-1 3.3.1.3 Vertical Velocity Distribution and Gust Factor..........................3.3-1 3.3.1.4 Determination of Applied Forces..............................................3.3-2
 
3.3.2 Tornado Loadings......................................................................................3.3-2
 
3.3.2.1 Applicable Design Parameters.................................................3.3-2 3.3.2.2 Determination of Forces on Structures....................................3.3-3 3.3.2.3 Ability of Category I Structures to Perform Despite Failure  of Structures Not Designed for Tornado Loads.......................3.3-4
 
3.4 WATER LEVEL (FLOOD) DESIGN.............................................................................3.4-1
 
3.4.1 Flood Protection........................................................................................3.4-1 3.4.2 Analysis Procedures .................................................................................3.4-2
 
3.5 MISSILE PROTECTION..............................................................................................3.5-1
 
3.5.1 Missile Barriers and Loadings...................................................................3.5-1
 
3.5.1.1 Accident/Incident Generated Missiles Inside Containment......3.5-1 3.5.1.2 Environmental Load Generated Missiles.................................3.5-1 3.5.1.3 Site Proximity Missiles.............................................................3.5-4 3.5.1.4 Accident/Incident Generated Missiles Inside Category I  Structures Other Than Containment........................................3.5-4
 
3.5.2 Missile Selection........................................................................................3.5-4
 
3.5.2.1 Missile Selection Within the Containment................................3.5-4 3.5.2.2 Missiles Selected Outside the Containment............................3.5-6 3.5.2.3 Missile Selection Within Category I Structures Other  Than Containment...................................................................3.5-6
 
3.5.3 Selected Missiles.......................................................................................3.5-6 3.5.4 Barrier Design Procedures........................................................................3.5-6 3.5.5 Missile Barrier Features.............................................................................3.5-8
 
3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING...............................................................3.6-1
 
3.6.1 Systems in Which Design Basis Piping Breaks are Postulated  to Occur.....................................................................................................3.6-1 FNP-FSAR-3
 
3-iv REV 21  5/08 TABLE OF CONTENTS        Page
 
3.6.2 Design Basis Methods and Piping Break Criteria......................................3.6-2
 
3.6.2.1 Criteria.....................................................................................3.6-2 3.6.2.2 Reactor Coolant Loops............................................................3.6-4 3.6.2.3 Class 1 Branch Lines...............................................................3.6-4 3.6.2.4 Class 2 and 3 Lines.................................................................3.6-6 3.6.2.5 Break Types.............................................................................3.6-7
 
3.6.3 Design Loading Combinations ..................................................................3.6-8
 
3.6.3.1 Reactor Coolant Piping............................................................3.6-8 3.6.3.2 Class 1 Branch Lines...............................................................3.6-8 3.6.3.3 Class 2 and 3 Lines ................................................................3.6-8
 
3.6.4 Dynamic Analysis......................................................................................3.6-9
 
3.6.4.1 Postulated Break Locations.....................................................3.6-9
 
3.6.5 Protective Measures ...............................................................................3.6-11
 
3.6.5.1 Pipe Whip Restraints.............................................................3.6-11 3.6.5.2 Jet Impingement ...................................................................3.6-12 3.6.5.3 Separation and Redundancy.................................................3.6-13
 
3.6.6 Structural Analysis ..................................................................................3.6-13
 
3.6.6.1 Outside Containment ............................................................3.6-13 3.6.6.2 Inside Containment. ..............................................................3.6-13 3.6.6.3 Pipe Whip Restraint Design...................................................3.6-14
 
3.7 SEISMIC DESIGN.......................................................................................................3.7-1
 
3.7.1 Seismic Input.............................................................................................3.7-1
 
3.7.1.1 Design Response Spectra ......................................................3.7-2 3.7.1.2 Design Response Spectra Derivation......................................3.7-2 3.7.1.3 Critical Damping Values ..........................................................3.7-3 3.7.1.4 Bases for Site Dependent Analysis..........................................3.7-3 3.7.1.5 Soil Supported Category I Structures......................................3.7-3 3.7.1.6 Soil Structure Interaction..........................................................3.7-4
 
3.7.2 Seismic System Analysis ..........................................................................3.7-5
 
3.7.2.1 Seismic Analysis Methods.......................................................3.7-6 3.7.2.2 Natural Frequencies and Response Loads..............................3.7-9 FNP-FSAR-3
 
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3.7.2.3 Procedures Used to Lump Masses..........................................3.7-9 3.7.2.4 Rocking and Translational Response Summary......................3.7-9 3.7.2.5 Methods Used to Couple Soil with Seismic System  Structures...............................................................................3.7-10  3.7.2.6 Development of Floor Response Spectra..............................3.7-10 3.7.2.7 Differential Seismic Movement of Interconnected  Components...........................................................................3.7-10  3.7.2.8 Effects of Variations on Floor Response Spectra .................3.7-10 3.7.2.9 Use of Constant Vertical Load Factors..................................3.7-10 3.7.2.10 Methods Used to Account for Torsional Effects ....................3.7-11 3.7.2.11 Comparison of Responses ....................................................3.7-11 3.7.2.12 Methods for Seismic Analysis of Dams..................................3.7-11 3.7.2.13 Methods to Determine Category I Structure Overturning  Moment..................................................................................3.7-11  3.7.2.14 Analysis Procedure for Dampings..........................................3.7-11 3.7.3 Seismic Subsystem Analysis...................................................................3.7-12
 
3.7.3.1 Determination of Number of Earthquake Cycles....................3.7-12 3.7.3.2 Basis for Selection of Forcing Frequencies...........................3.7-12 3.7.3.3 Root Mean Square Basis.......................................................3.7-12 3.7.3.4 Procedure for Combining Modal Responses ........................3.7-13 3.7.3.5 Significant Dynamic Response Modes..................................3.7-14 3.7.3.6 Design Criteria and Analytical Procedures for Piping............3.7-15 3.7.3.7 Basis for Computing Combined Response............................3.7-15 3.7.3.8 Amplified Seismic Responses ...............................................3.7-15 3.7.3.9 Use of Simplified Dynamic Analysis.......................................3.7-15 3.7.3.10 Modal Period Variation...........................................................3.7-16 3.7.3.11 Torsional Effects of Eccentric Masses...................................3.7-16 3.7.3.12 Piping Outside Containment..................................................3.7-16 3.7.3.13 Interaction of Other Piping With Category I Piping................3.7-16 3.7.3.14 Field Location of Supports and Restraints.............................3.7-17 3.7.3.15 Seismic Analysis for Fuel Elements, Control Assemblies  and Control Rod Drives..........................................................3.7-17
 
3.7.4 Seismic Instrumentation Program ..........................................................3.7-17
 
3.7.4.1 Comparison with NRC Regulatory Guide 1.12......................3.7-17 3.7.4.2 Location and Description of Instrumentation .........................3.7-18 3.7.4.3 Control Room Operator Notification.......................................3.7-19 3.7.4.4 Comparison of Measured and Predicted Responses............3.7-19
 
FNP-FSAR-3
 
3-vi REV 21  5/08 TABLE OF CONTENTS        Page
 
3.7.5 Seismic Design Control...........................................................................3.7-20
 
3.7.5.1 Seismic Design Control - Construction  Phase......................3.7-20 3.7.5.2 Seismic Design Control - Operational Phase.........................3.7-23
 
3.8 DESIGN OF CATEGORY I STRUCTURES................................................................3.8-1
 
3.8.1 Concrete Containment...............................................................................3.8-1
 
3.8.1.1 Description of the Containment................................................3.8-1 3.8.1.2 Applicable Codes, Standards, and Specifications...................3.8-2 3.8.1.3 Loads and Loading Combinations...........................................3.8-7 3.8.1.4 Design and Analysis Procedures...........................................3.8-16 3.8.1.5 Structural Acceptance Criteria...............................................3.8-21 3.8.1.6 Materials, Quality Control, and Special Construction  Techniques ...........................................................................3.8-21  3.8.1.7 Testing and Inservice Surveillance Requirements.................3.8-34
 
3.8.2 Steel Containment System (ASME Class MC Components) ..................3.8-39  3.8.3 Internal Structures...................................................................................3.8-39
 
3.8.3.1 Description of the Internal Structures.....................................3.8-40 3.8.3.2 Applicable Codes, Standards and Specifications..................3.8-41 3.8.3.3 Loads and Loading Combinations.........................................3.8-41 3.8.3.4 Design and Analysis Procedures...........................................3.8-44 3.8.3.5 Structural Acceptance Criteria...............................................3.8-47 3.8.3.6 Materials, Quality Control and Special Construction  Techniques............................................................................3.8-47  3.8.3.7 Testing and Inservice Surveillance Requirements.................3.8-49
 
3.8.4 Other Category I Structures ....................................................................3.8-49
 
3.8.4.1 Description of the Structures..................................................3.8-50 3.8.4.2 Applicable Codes, Standards and Specifications..................3.8-53 3.8.4.3 Loads and Loading Combinations.........................................3.8-55 3.8.4.4 Design and Analysis Procedures...........................................3.8-58 3.8.4.5 Structural Acceptance Criteria...............................................3.8-58 3.8.4.6 Materials, Quality Control, and Special Construction  Techniques ...........................................................................3.8-59  3.8.4.7 Testing and Inservice Surveillance Requirements.................3.8-60
 
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3-vii REV 21  5/08 TABLE OF CONTENTS        Page
 
3.8.5 Foundations and Concrete Supports.......................................................3.8-60
 
3.8.5.1 Description of the Foundations and Supports........................3.8-60 3.8.5.2 Applicable Codes, Standards and Specifications..................3.8-62 3.8.5.3 Loads and Loading Combinations.........................................3.8-62 3.8.5.4 Design and Analysis Procedures...........................................3.8-62 3.8.5.5 Structural Acceptance Criteria...............................................3.8-63 3.8.5.6 Materials, Quality Control, and Special Construction  Techniques ...........................................................................3.8-63  3.8.5.7 Testing and Inservice Surveillance Requirements.................3.8-63
 
3.8.6 Masonry Walls.........................................................................................3.8-70
 
3.9 MECHANICAL SYSTEMS AND COMPONENTS ......................................................3.9-1
 
3.9.1 Dynamic System Analysis and Testing ....................................................3.9-1
 
3.9.1.1 Vibrational Operational Test Program......................................3.9-1 3.9.1.2 Dynamic Testing Procedures...................................................3.9-1 3.9.1.3 Dynamic System Analysis Methods for Reactor Internals.......3.9-2 3.9.1.4 Correlation of Test and Analytical Results...............................3.9-7 3.9.1.5 Analysis Methods Under LOCA Loadings................................3.9-8 3.9.1.6 Analytical Methods for ASME Code Class 1  Components...........................................................................3.9-12
 
3.9.2 ASME Code Class 2 and 3 Components................................................3.9-12
 
3.9.2.1 Plant Conditions and Design Loadings Combinations...........3.9-12 3.9.2.2 Design Loading Combination.................................................3.9-12 3.9.2.3 Design Stress Limits..............................................................3.9-12 3.9.2.4 Analytical and Empirical Methods for Design of Pumps  and Valves.............................................................................3.9-13  3.9.2.5 Design and Installation Criteria, Pressure-Relieving  Devices..................................................................................3.9-13  3.9.2.6 Stress Levels for Category I Components.............................3.9-13 3.9.2.7 Field Run Piping System........................................................3.9-13 3.9.2.8 Class 2 and 3 Component Supports......................................3.9-14
 
3.9.3 Components Not Covered by ASME Code..............................................3.9-14
 
3.9.3.1 Faulted Conditions.................................................................3.9-15 3.9.3.2 Structural Response of Reactor Vessel  Internals  During LOCA and Seismic  Conditions..................................3.9-15  3.9.3.3 Results and Acceptance Criteria............................................3.9-18 FNP-FSAR-3
 
3-viii REV 21  5/08 TABLE OF CONTENTS        Page
 
3.9.3.4 Method of Analysis.................................................................3.9-20 3.9.3.5 Evaluation of Reactor Internals for  Accumulator Line  Cold Leg and Pressurizer Surge Line Hot Leg Breaks..........3.9-21  3.9.3.6 Baffle-Former Bolt Replacement Analysis.............................3.9-21 3.9.3.7 Heating, Ventilation, and Air Conditioning (HVAC)  Equipment..............................................................................3.9-23
 
3.9.4 Operability Assurance.............................................................................3.9-23
 
3.9.4.1 ASME Code Class Valves.....................................................3.9-23 3.9.4.2 ASME Code Class Pumps.....................................................3.9-24 3.9.4.3 Qualification of Vital Appurtenances......................................3.9-24
 
3.10 SEISMIC DESIGN OF CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT......................................................................................3.10-1
 
3.10.1 Seismic Design Criteria...........................................................................3.10-1 3.10.2 Seismic Analyses, Testing Procedures and Restraint Measures............3.10-4
 
3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRICAL EQUIPMENT.............................................................................................................3.11-1
 
3.11.1 Equipment Identification and Environmental Conditions.........................3.11-1 3.11.2 Qualification Tests and Analyses............................................................3.11-3
 
3.11.2.1 Equipment Inside Containment..............................................3.11-3 3.11.2.2 Equipment Outside Containment...........................................3.11-3 3.11.2.3 Equipment Supplied by Bechtel and Southern Company  Services.................................................................................3.11-4  3.11.2.4 Equipment Supplied by Westinghouse..................................3.11-4
 
3.11.3 Qualification Test Results........................................................................3.11-7 3.11.4 Loss of Ventilation...................................................................................3.11-8
 
APPENDIX 3A CONFORMANCE WITH NRC REGULATORY GUIDES...........................3A-1
 
APPENDIX 3B CONTAINMENT PROOF TESTS..............................................................3B-1
 
APPENDIX 3C MECHANICAL SPLICING REINFORCING BAR  USING THE  CADWELD PROCESS..............................................................................3C-1
 
APPENDIX 3D JUSTIFICATION FOR LOAD FACTORS AND LOAD COMBINATIONS USED IN DESIGN EQUATIONS FOR THE CONTAINMENT..............................3D-1
 
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3-ix REV 21  5/08 TABLE OF CONTENTS        Page
 
APPENDIX 3E JUSTIFICATION FOR CAPACITY REDUCTION  FACTORS 
( - FACTORS) USED IN DETERMINING CAPACITY OF  CONTAINMENTS......................................................................................3E-1
 
APPENDIX 3F COMPUTER PROGRAMS USED IN STRUCTURAL  ANALYSIS.............3F-1
 
APPENDIX 3G QUALITY CONTROL PROCEDURES FOR FIELD  WELDING AND NONDESTRUCTIVE EXAMINATIONS  OF CONTAINMENT LINER PLATE.......................................................................................................3G-1
 
APPENDIX 3H CONTAINMENT STRUCTURAL ACCEPTANCE TEST...........................3H-1
 
APPENDIX 3I LINER PLATE STABILITY...........................................................................3I-1
 
APPENDIX 3J MAIN STEAM LINE BREAK OUTSIDE CONTAINMENT PRESSURE  AND TEMPERATURE ANALYSIS.............................................................3J-1
 
APPENDIX 3K HIGH-ENERGY LINE PIPE BREAK (OUTSIDE CONTAINMENT)...........3K-1
 
APPENDIX 3L ASME SECTION III NUCLEAR CLASS AUXILIARY PIPING  STRUCTURAL ANALYSIS.........................................................................3L-1
 
APPENDIX 3M REACTOR PRESSURE VESSEL SUPPORT LOADS..............................3M-1
 
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3-x REV 21  5/08 LIST OF TABLES
 
3.2-1 Summary of Criteria - Mechanical System Components 
 
3.2-2 Summary of Quality Class Require ments - Mechanical System Components 
 
3.2-3 Listing of P&IDs
 
3.2-4 Component Coding 
 
3.2-5 ASME Code Cases for Class 1 Components
 
3.3-1 Wind Loads with Gust Factor 
 
3.5-1 Missile Barriers Inside Containment 
 
3.5-2 CRDM - Missile Characteristics 
 
3.5-3 Valve - Missile Characteristics 
 
3.5-4 Piping Temperature Element Assembly - Missile Characteristics 
 
3.5-5 Characteristics of Other Missiles Postulated Within Containment 
 
3.5-6 Missile Barriers Away From Containment 
 
3.5-7 Rod Drive Power Supply Motor-Generator Set Flywheel Missile Characteristics 
 
3.6-2 Thrust Loads Due To a Full Area Pipe Rupture (Class 2 and 3 Piping) 
 
3.6-7 Analysis Results 
 
3.7-1 Percentage of Critical Damping Factors 
 
3.7-2 System Period Interval 
 
3.7-3 Methods Used for Seismic Analyses of Category I Structures 
 
3.7-4 Methods Used for Seismic Analyses of Category I Systems and Components 
 
3.7-5 Natural Frequencies for Category I Structures 
 
3.7-6 Comparison of Translational and Torsional Frequencies 
 
3.7-7 Containment Shell Comparison of Response Spectrum and Time History Analysis, Safe Shutdown Earthquake (East -West Direction) 
 
FNP-FSAR-3
 
3-xi REV 21  5/08 LIST OF TABLES
 
3.7-8 Maximum Allowable Span Between Seismic Restraints for Pipes 2 In. and Under 
 
3.8-1 Post-Tensioning System - BBRV (170) 
 
3.8-2 Stress Analysis Results 
 
3.8-3 Containment Strains (x 1006) 
 
3.8-4 Containment Stresses In Equipment Hatch Area 
 
3.8-5 Aggregate Tests 
 
3.8-6 Cement Tests 
 
3.8-7 Fly Ash Tests 
 
3.8-8 Prestressing Sequences 
 
3.8-9 Calculated Results - Internal Structures 
 
3.8-10 Calculated Results - Auxiliary Building 
 
3.8-11 Calculated Results - Diesel Generator Building 
 
3.8-12 Calculated Results - River Intake Structure 
 
3.8-13 Calculated Results - Intake Structure at Storage Pond 
 
3.8-14 Calculated Results - Electrical Cable Tunnels 
 
3.9-1 Design Criteria for ASME Class 2 and 3 Components 
 
3.9-2 Maximum Deflections Specified for Reactor Internal Support Structures 
 
3.9-3 Design Criteria for Components Not Covered by ASME Code 
 
3.9-4 Comparison of Best Estimate and Design Values of Peak Seismic Accelerations 
 
3.9-5 Summary of Stress and Margin of Safety to Code Allowables 
 
3.11-1 EQ Program Environmental Conditions 
 
FNP-FSAR-3
 
3-xii REV 21  5/08 LIST OF FIGURES
 
3.3-1 Wind Forces and Distribution on Category I Structures 
 
3.4-1 Water Pressure on Structures 
 
3.6-1 Loss of Reactor Coolant Accident Boundary Limits 
 
3.7-1 One-half Safe Shutdown Earthquake Ground Spectra 0.05 g (Horizontal and Vertical) 
 
3.7-2 Safe Shutdown Earthquake Ground Spectra 0.10 g (Horizontal and Vertical) 
 
3.7-3 Synthesized Time History (1/2 SSE and SSE) 
 
3.7-4 Time History Spectrum Envelope on Response Spectrum (1/2 SSE) 
 
3.7-5 Time History Spectrum Envelope on Response Spectrum (SSE) 
 
3.7-6 A Lumped-Mass Model of Structure Foundation System 
 
3.7-7 Constants, x, , and z for Rectangular Bases 
 
3.7-8 through Containment - Seismic Results 
 
3.7-13
 
3.7-14 Internal Structure - Seismic Results 
 
3.7-15 through Internal Structure - Seismic Results 
 
3.7-19
 
3.7-20 Containment and Internal Structure Mathematical Model 
 
3.7-21 Containment and Internal Structure Mathematical Model 
 
3.7-22 Polar Crane Bracket and Seismic Retainer 
 
3.7-23 Auxiliary Building Mathematical Model 
 
3.7-24 Auxiliary Building Mathematical Model 
 
FNP-FSAR-3
 
3-xiii REV 21  5/08 LIST OF FIGURES
 
3.7-25 through Auxiliary Building Seismic Results 
 
3.7-30
 
3.7-31 through Diesel Generator Building Seismic Results 
 
3.7-36
 
3.7-37 through River Intake Structure Seismic Results 
 
3.7-42
 
3.7-43 through Intake Structure at Storage Pond Seismic Results 
 
3.7-48
 
3.7-49 Vent Stack Seismic Results 
 
3.7-50 Vent Stack Seismic Results 
 
3.7-51 through Pond Spillway Structure Seismic Results 
 
3.7-56
 
3.7-57 Containment and Internal Structure Frequencies and Mode Shapes 
 
3.7-58 Containment and Internal Structure Frequencies and Mode Shapes 
 
3.7-59 Modal Inertia Forces for Containment 
 
3.7-60 Modal Inertia Forces for Containment 
 
3.7-61 Response Spectrum at Reactor Support Elevation 
 
3.7-62 Seismic Instrumentation 
 
3.7-63 Earthquake Elevation Procedure for Category 1 Structures 
 
3.7-64 Mathematical Model of Reactor Internals 
 
3.7-65 First Mode of Vibration of Reactor Internals 
 
3.8-1 Containment Typical Sections and Details 
 
3.8-2 Containment Plans and Sections FNP-FSAR-3
 
3-xiv REV 21  5/08 LIST OF FIGURES
 
3.8.3 Containment Details of Equipment Hatch (Deleted)
 
3.8-4 Containment Details of Personnel Lock (Deleted)
 
3.8-5 Sheathing and Trumpet Detail 
 
3.8-6 Base Details for Steam Generator and Reactor Coolant Pump Foundations 
 
3.8-7 Base Detail for Secondary Shield Wall 
 
3.8-9 Secondary Shield Walls Below El 129'-0" 
 
3.8-10 Secondary Shield Wall El 129'-0" to 166'-6" 
 
3.8-11 Primary Shield Wall 
 
3.8-12 Detail for Base Slab to Cylinder Liner Juncture 
 
3.8-13 Typical Plans Containment 
 
3.8-14 Typical Sections Containment 
 
3.8-15 Thermal Gradient Across Containment Wall 
 
3.8-16 Finite Element Mesh Bottom Half Containment for Axisymmetric Loads 
 
3.8-17 Finite Element Mesh Top Half Containment for Axisymmetric Loads 
 
3.8-18 Model of Containment for Finite El ement Analysis Non-Axisymmetric Loads 
 
3.8-19 Containment Base Slab Finite Element Mesh Non-Axisymmetric Loads 
 
3.8-23 Auxiliary Building Control Room and Spent Fuel Pool Plans at El 155'-0" 
 
3.8-24 Auxiliary Building Section "A-A" 
 
3.8-25 Auxiliary Building Section "B-B" 
 
3.8-26 Diesel Generator Building Plan and Section 
 
3.8-27 River Intake Structure Plan and Section 
 
3.8-28 Intake Structure at Storage Pond Plan and Section 
 
FNP-FSAR-3
 
3-xv REV 21  5/08 LIST OF FIGURES
 
3.8-29 Pond Spillway Structure Plan and Sections 
 
3.8-30 Equipment Hatch Boundary Lines for the SAP Analysis 
 
3.8-31 SAP Finite Element Mesh for the Equipment Hatch 
 
3.8-32 through SAP Analysis of Equipment Hatch 
 
3.8-35
 
3.8-36 Location Plan - Foundations for Category 1 Structures 
 
3.8-37 Containment Base Slab Details 
 
3.8-38 Auxiliary Building Base Slab Details 
 
3.8-39 Foundation Details for Diesel Generator Building, River Intake Structure, and Intake Structure at Storage Pond 
 
3.8-40 Geometry of Personnel Lock and Auxiliary Access Lock 
 
3.8-41 Auxiliary Building Cask Wash and Cask Storage Areas Plan and Section 
 
3.8-42 Auxiliary Building Cask Wash and Cask Storage Areas Section 
 
3.9-1 Vibration Checkout Functional Test Inspection Points 
 
3.9-2 Time-History Dynamic Solution for LOCA Loading 
 
3.11-1 FNP Composite LOCA/MSLB Containment Pressure Profiles
 
3.11-2 FNP Composite LOCA/MSLB Containment Temperature Profiles
 
FNP-FSAR-3
 
3.1-1 REV 21  5/08 3.0 - DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.1  CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA 3.1.1  CRITERION 1 - QUALITY STANDARDS AND RECORDS Structures, systems and components important to safety are designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be
 
performed. Where generally recognized codes and standards are used, they are identified and
 
evaluated to determine their applicability, adequacy, and sufficiency and are supplemented or
 
modified as necessary to assure a quality product in keeping with the required safety function. 
 
A quality assurance program has been established and implemented in order to provide
 
adequate assurance that these structures, system s, and components will satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection, and testing of
 
structures, systems, and components important to safety are maintained by or under the control of the nuclear power unit licensee throughout the life of the unit. 
 
CONFORMANCE 
 
Discussion of the quality standards for those systems which are essential to the prevention of
 
accidents which could affect the public health and safety or lead to a mitigation of their
 
consequences are presented in appropriate sections of the Safety Analysis Report.  (See
 
chapters 3.0, 4.0, 5.0, 6.0, 7.0, 8.0, 9.0, 10.0, 11.0, and 12.0.) 
 
For example, components of the engineered saf eguards systems are designed and fabricated in accordance with established codes and/or standards as required to assure that their quality is in
 
keeping with the safety function of the component. 
 
The mechanical components of the facility have been classified according to their importance in
 
the prevention and mitigation of accidents which could cause undue risk to the health and safety
 
of the public. These classifications are described in section 3.2, along with the codes to which
 
each component was designed. The design criteria for mechanical components are listed in
 
table 3.2-1. Those components listed as ANS Safety Class 1, 2a, 2b, or 3 are designed and
 
manufactured to the quality standards required by Criterion 1. 
 
The tests and inspections applied to the systems are included in the appropriate sections of the
 
FSAR. The quality assurance program followed during plant design and construction was
 
implemented in accordance with section 17.1. The quality assurance program used for plant
 
operations is described in the SNC Quality Assurance Topical Report (QATR). 
 
Drawings, records, correspondence, field books, forms, film, and temperature charts required to
 
show compliance with the codes, tests, and quality control standards as outlined in section 17.1
 
are in the possession of or available to Southern Nuclear Operating Company throughout the
 
life of the plant. This is discussed further in section 13.5. 
 
FNP-FSAR-3
 
3.1-2 REV 21  5/08 3.1.2  CRITERION 2 - DESIGN BASES FOR PROTECTION AGAINST NATURAL PHENOMENA Structures, systems, and components important to safety are designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and
 
seiches without loss of capability to perform their safety functions. The design bases for these
 
structures, systems, and components reflect:
* Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for
 
the limited accuracy, quantity, and period of time in which the historical data have
 
been accumulated.
* Appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena.
* The importance of the safety functions to be performed. 
 
CONFORMANCE 
 
The natural phenomena and their magnitude have been selected in accordance with the
 
probability of their occurrence at the Farley plant site. The designs are based upon the most
 
severe of the natural phenomena recorded for the site, with an appropriate margin to account
 
for uncertainties in historical data. The natural phenomena postulated in the design are
 
presented in sections 2.3, 2.4, and 2.5. The design criteria for the structures, systems, and
 
components affected by each natural phenomenon are presented in sections 3.2, 3.3, 3.4, 3.5, 3.7, and 3.8. Those combinations of natural phenomena and plant originated accidents that are
 
considered in the design are identified in sections 3.8, 3.9, and 3.10. The importance of the
 
safety functions is identified with the classi fication system developed by the American Nuclear Society. Such identification is included in section 3.2. 
 
3.1.3  CRITERION 3 - FIRE PROTECTION Structures, systems, and components important to safety are designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. 
 
Noncombustible and heat resistant materials are used wherever practical throughout the unit, particularly in locations such as the containment and control room. Fire detection and fighting
 
systems of appropriate capacity and capability are provided and designed to minimize the
 
adverse effects of fires on structures, sy stems, and components important to safety.
Firefighting systems are designed to assure that their rupture or inadvertent operation does not
 
significantly impair the safety capability of these structures, systems, and components. 
 
CONFORMANCE 
 
The reactor facility is designed to minimize the probability and effect of fires and explosions. 
 
Noncombustible and heat resistant materials are used in the containment, in the control room, in rooms where safety-related equipment is located, and wherever required throughout the plant. Appropriate equipment and facilities for fire protection, including detection, alarm, and FNP-FSAR-3
 
3.1-3 REV 21  5/08 extinguishing of fire, are provided to protect both plant equipment and personnel from fire, explosion, and the resultant release of toxic vapors. Both wet and dry types of firefighting
 
equipment are provided. Fire protection is pr ovided by deluge sprays, sprinkler systems, piped carbon dioxide systems, hose stream, and port able extinguishers of the type appropriate for each area. 
 
Firefighting systems are designed to assure that their rupture or inadvertent operation will not
 
significantly impair safety-related systems. The fi re protection system consists of a reliable, partially automatic unit engineered and installed in accordance with the requirements of the
 
National Fire Protection Association (NFPA) or American National Standards Institute (ANSI),
the Nuclear Energy Property Insurance Asso ciation (NEPIA), the Occupational Safety and Health Act (OSHA), Nuclear Regulatory Commission guidelines, and the Nuclear Electric
 
Insurance Limited (NEIL) Property Loss and Prevention Standards, in addition to the applicable
 
local codes and regulations. 
 
The fire protection system is provided with test hos e valves for periodic testing. All equipment is accessible for periodic inspection. The fire protection system is described further in subsection
 
9.5.1 and appendix 9B.
 
3.1.4  CRITERION 4 - ENVIRONMENTAL AND MISSILE DESIGN BASES Structures, systems, and components important to safety are designed to accommodate the effects of and be compatible with the environmental conditions associated with normal
 
operation, maintenance, testing, and postulated accidents, including LOCAs. These structures, systems, and components shall be appropriately pr otected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluid, that may result from equipment
 
failures and from events and conditions outside the nuclear power unit. However, dynamic
 
effects associated with postulated pipe ruptures in nuclear power units may be excluded from
 
the design basis when analyses reviewed and approved by the Commission demonstrate that
 
the probability of fluid system piping rupture is extremely low under conditions consistent with
 
the design basis for the piping. 
 
CONFORMANCE 
 
Structures, systems, and components important to safety are designed to accommodate the effects of any environmental conditions at the Farley site that are associated with normal
 
operation, maintenance, testing, and postulated accidents, including LOCAs, and are
 
compatible with these conditions. Design criteria and implementation are presented in sections
 
3.5 and 3.6, and environmental factors are described in sections 3.11, 6.3, and 6.4. These 
 
structures, systems, and components are appropria tely protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from any
 
postulated failure of equipment or structural design or from environmental conditions and
 
accidents outside the nuclear power unit. Postulated breaks in the reactor coolant loop, except
 
for branch line connections, have been eliminated from the structural design basis for both Unit
 
1 and Unit 2, as allowed by the revised General Design Criterion 4. The elimination of these
 
breaks is the result of the application of leak-before-break technology as described in section
 
3.6. Chapter 7.0 lists the motors, instrumentation, and associated cables of protection and
 
safety features systems located inside the cont ainment structure. It also gives the design FNP-FSAR-3
 
3.1-4 REV 21  5/08 requirements in terms of the time that each mu st survive the extreme environmental conditions following a LOCA. 
 
Details of the design, environmental testing, and construction of these systems, structures, and components are included in other sections of chapters 3.0, 5.0, 6.0, 7.0, and 9.0. Evaluation of
 
the performance of safety features and analyses of possible accidents are contained in chapter
 
15.0. 
 
3.1.5  CRITERION 5 - SHARING OF STRUCTURES, SYSTEMS, AND COMPONENTS Structures, systems, and components important to safety are not shared among nuclear power units unless it can be shown that such sharing does not significantly impair their ability to
 
perform their safety functions, including, in the event of an accident in one unit, an orderly
 
shutdown and cooldown of the remaining units. 
 
CONFORMANCE 
 
The Farley Plant is a two-unit plant with the following shared structures, systems, and
 
components that provide features and functions important to safety. Such sharing has been
 
evaluated to ensure that there are no adverse impacts on safety functions.
 
Structures
 
Control room
 
Diesel generator building
 
Auxiliary building HVAC penthouse
 
Ultimate heat sink storage pond and service water intake structure
 
Systems Control room HVAC system
 
Diesel generator fuel storage/transfer system
 
Diesel generator HVAC systems for shar ed diesel generators and switchgear rooms Service water intake structure HVAC
 
SWIS nitrogen backup system
 
Components
 
Three of five diesel generators and auxiliary support equipment
 
Diesel generator 600-V switchgear and MCCs
 
Service water intake structure 600-V switchgear, batteries, and battery chargers
 
Load centers K, L, R, and S
 
Service water discharge pipes to UHS pond
 
Carbon dioxide storage
 
Fire protection water tanksFire protection pumps
 
Spent-fuel cask crane
 
Pressurizer heater service transformer
 
FNP-FSAR-3
 
3.1-5 REV 21  5/08 Functional evaluation of structures, systems, and components shared by the two Farley units are addressed in corresponding sections of the FSAR. 
 
3.1.6  CRITERION 10 - REACTOR DESIGN The reactor core and associated coolant, control, and protection systems are designed with
 
appropriate margin to assure that specified acceptable fuel design limits are not exceeded
 
during any condition of normal operation, including the effects of anticipated operational
 
occurrences. 
 
CONFORMANCE 
 
Each reactor core and associated coolant, control, and protection system is designed with
 
adequate margins to: 
 
A. Preclude any fuel damage during normal operation and operational transients (Condition I) or any transient conditions arising from occurrences of moderate
 
frequency (Condition II).
 
B. Ensure return of the reactor to a safe state following a Condition III fault with only a small fraction of fuel rods damaged, although sufficient fuel damage might occur to
 
preclude resumption of operation without considerable outage time. 
 
C. Assure that the core is intact with acceptable heat transfer geometry following transients arising from occurrences of limiting faults (Condition IV). 
 
Chapter 4.0 discusses the design bases and design evaluation of reactor components including
 
the fuel, reactor vessel internals, and reactivity control systems. Details of the control and
 
protection systems instrumentation design and logic are discussed in chapter 7.0. This
 
information supports the accident analyses in chapter 15.0 which show that acceptable fuel
 
design limits are not exceeded for Condition I and II occurrences. 
 
3.1.7  CRITERION 11 - REACTOR INHERENT PROTECTION The reactor core and associated coolant systems are designed so that in the power operating
 
range the net effect of the prompt inherent nuclear feedback characteristics tends to
 
compensate for a rapid increase in reactivity. 
 
CONFORMANCE 
 
Prompt compensatory reactivity feedback effects are assured, when the reactor is critical, by the negative fuel temperature effect (Doppler effe ct) and by the operational limit on moderator temperature coefficient of reactivity at full pow er. The negative Doppler coefficient of reactivity is assured by the inherent design using low-enrichment fuel. The reload core is designed to
 
have an overall moderator temperature coefficient of reactivity which is less than or equal to
+7.0 pcm/°F for power levels up to 70% with a linear ramp to 0.0 pcm/
°F at 100% power by the FNP-FSAR-3
 
3.1-6 REV 21  5/08 use of fixed burnable absorber (BA) rods, and/or integral fuel burnable absorbers (IFBAs) and/or control rods by limiting the reactivity held down by soluble boron.
 
The core inherent reactivity feedback characteristics are described in section 4.3, (Nuclear
 
Design). Reactivity control by chemical inject ion is discussed in subsection 4.2.3, (Reactivity Control Systems), and subsection 9.3.4, (Chem ical and Volume Control System). Plant Technical Specifications define allowable absorber concentrations. 
 
3.1.8  CRITERION 12 - SUPPRESSION OF REACTOR POWER OSCILLATIONS The reactor core and associated coolant, control, and protection systems are designed to
 
assure that power oscillations which can result in conditions exceeding specified acceptable fuel
 
design limits are not possible or can be reliably and readily detected and suppressed. 
 
CONFORMANCE 
 
Power oscillations of the fundamental mode are inherently eliminated by the negative Doppler and nonpositive moderator temperature coefficient of reactivity. 
 
Oscillations due to xenon spatial effects in the radial, diametral, and azimuthal overtone modes are heavily damped because of the inherent design and the negative Doppler and
 
nonpositive moderator temperature coefficients of reactivity. 
 
Oscillations due to xenon spatial effects in the ax ial first overtone mode may occur. Assurance that fuel design limits are not exceeded by xenon ax ial oscillations is provided as a result of reactor trip functions using the measured axial power imbalance as an input. 
 
The stability of the core against xenon induced power oscillations and the functional
 
requirements of instrumentation for monitoring and measuring core power distribution are
 
discussed in section 4.3 (Nuclear Design). Details of the instrumentation design and logic are
 
discussed in chapter 7.0. 
 
3.1.9  CRITERION 13 - INSTRUMENTATION AND CONTROL Instrumentation is provided to monitor variabl es and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as
 
appropriate to assure adequate safety, including those variables and systems that can affect the
 
fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the
 
containment and its associated systems. Appropr iate controls shall be provided to maintain these variables and systems within prescribed operating ranges. 
 
CONFORMANCE 
 
Instrumentation and control systems are provided to monitor and maintain plant variables
 
including those variables that affect the fission process, integrity of the reactor core, the reactor
 
coolant pressure boundary, and the containment over their expected range for normal FNP-FSAR-3
 
3.1-7 REV 21  5/08 operation, for anticipated operational occurrences, and under accident conditions. These systems comply with the intent of Criterion 13. 
 
The following processes are controlled to maintain key variables within their normal ranges: 
 
A. Reactor power level (manual or automatic by control of thermal load). 
 
B. Reactor coolant temperature (manual or aut omatic by control of rod position).
C. Reactor coolant pressure (manual or automatic by control of heaters and spray in the pressurizer).
D. Reactor coolant water inventory, as indicated by pressurizer water level (manual or automatic control of charging flow).
E. Reactor coolant system boron concentration (manual or automatic control of makeup charging flow).
F. Steam generator inventory on secondary side (manual or automatic control of feedwater flow). 
 
The reactor control system is designed to automatically maintain programmed average temperature in the reactor coolant during steady-state operation and to ensure that plant
 
conditions do not reach reactor trip settings, as the result of design load change. Overall
 
reactivity control is achieved by the combination of soluble boron and control rods. Long term
 
regulation of the core reactivity is obtained by adjustment of the concentration of boron in the
 
reactor coolant. Short-term reactivity contro l for power changes is performed by the reactor control system which automatically positions t he rod cluster control assemblies. The measured inputs to this system are neutron flux, coolant temperature, and turbine load. 
 
The instrumentation functional requirements are given in chapter 4.0 and the instrumentation
 
and control systems are discussed in more detail in chapter 7.0. Operation of the reactor within
 
prescribed limits is enforced by the reactor pr otection system. This system is discussed in section 7.2.
 
A wide spectrum of measurements is displayed fo r operator information and/or is processed to provide alarms. These measurements provide notification and allow correction of conditions
 
having the potential of leading to abnormal or accident conditions. Typical indication (or alarm)
 
measurements are rod position, rod deviation, in sertion limit, rod bottom, rod control system failure, rod control system urgent failure, incore flux and temperature, protection system faults, and protection system test mode. 
 
3.1.10  CRITERION 14 - REACTOR COOLANT PRESSURE BOUNDARY The reactor coolant pressure boundary is designed, fabricated, erected, and tested so as to
 
have an extremely low probability of abnormal leakage, or rapidly propagating failure, and of
 
gross rupture. 
 
FNP-FSAR-3
 
3.1-8 REV 21  5/08 CONFORMANCE 
 
The reactor coolant pressure boundary is designed to accommodate the system pressures and
 
temperatures attained under all expected modes of unit operation, including all anticipated
 
transients, and maintain the stresses within applicable stress limits. The design criteria
 
methods and procedures applied to components of the reactor coolant pressure boundary are
 
discussed in subsection 5.2.1. Reactor coolant pressure boundary materials selection and
 
fabrication techniques ensure a low probability of gross rupture or significant leakage. 
 
In addition to the loads imposed on the system under normal operating conditions, consideration is also given to abnormal loading conditions such as pipe rupture and seismic, as
 
discussed in sections 3.6 and 3.7, respectively. Fracture prevention measures are included to
 
prevent brittle fracture. Refer to the discussion under Criterion 31 for additional information. 
 
The system is protected from overpressure by means of the pressurizer high-pressure reactor trip (section 7.2) and by pressure relieving devices according to the provisions of subsection
 
5.2.2. 
 
The materials of construction of the pressure retaining boundary of the reactor coolant system
 
are protected by control of coolant chemistry from corrosion that might otherwise reduce the
 
system structural integrity during its service lifetime. 
 
The pressure boundary has provisions for inspection, testing and surveillance of critical areas to
 
assess the structural and leaktight integrity of the boundary. The reactor coolant pressure
 
boundary leakage detection systems and inservice inspection program are discussed in
 
subsections 5.2.7 and 5.2.8, respectively. 
 
3.1.11  CRITERION 15 - REACTOR COOLANT SYSTEM DESIGN The reactor coolant system and associated aux iliary, control, and protection systems are designed with sufficient margin to assure that the design conditions of the reactor coolant
 
pressure boundary are not exceeded during any condition of normal operation, including
 
anticipated operational occurrences. 
 
CONFORMANCE 
 
The reactor coolant system and associated aux iliary, control, and protection systems are designed to ensure the integrity of the reactor coolant pressure boundary with adequate
 
margins during normal operation and during Condition I and Condition II transients. The system
 
boundary accommodates loads due to the 1/2 safe shutdown earthquake during normal
 
operation including normal operational transients (Condition II) within upset condition code
 
stress limits. The system boundary accommodates loads due to the safe shutdown earthquake
 
combined with loads due to piping failures such as circumferential pipe ruptures of reactor
 
coolant pipes at junctures with equipment nozzles and connecting pipes at junctures to reactor
 
coolant piping (Condition IV), without propagation of failure to remaining reactor coolant system
 
loops, steam power conversion system, or other piping or equipment needed for emergency cooling. The components of the reactor cool ant system and associated fluid systems are designed in accordance with appropriate ASME and ANSI codes. These codes are identified in FNP-FSAR-3
 
3.1-9 REV 21  5/08 chapter 5.0. The protection system is designed in accordance with IEEE-Std 279. The protection system analyses are given in subsecti on 7.2.2. Overpressure protection is provided by automatic controls, pressure relief valves, and code safety valves. 
 
The selected design margins include operating transient changes due to thermal lag, coolant
 
transport times, pressure drops, system relief valve characteristics, and instrumentation and
 
control response characteristics. 
 
3.1.12  CRITERION 16 - CONTAINMENT DESIGN Reactor containment and associated systems are pr ovided to establish an essentially leaktight barrier against the uncontrolled release of radioactivity to the environment and to assure that
 
the containment design conditions important to safety are not exceeded for as long as
 
postulated accident conditions require. 
 
CONFORMANCE 
 
The containment, penetration rooms, and the engineered safeguards systems are designed to
 
provide protection for the public from the consequences of an unlikely event of the LOCA, which
 
is based on a postulated break of the reactor coolant piping up to and including a double-ended
 
break of the largest reactor coolant pipe. 
 
The containment, penetration rooms, and the engineered safeguards system are designed to
 
safely sustain all internal and external loading conditions resulting from postulated transients
 
and accidents as described in chapter 15.0. Functional capability of the containment, penetration rooms, and engineered safeguards will be maintained for as long as required to
 
protect the public. Due consideration is given to site factors and local environments as they relate to public health and safety. 
 
Refer to section 3.8 and chapters 6.0 and 15.0. 
 
3.1.13  CRITERION 17 - ELECTRIC POWER SYSTEMS An onsite electric power system and an offsite electric power system are provided to permit functioning of structures, systems, and components important to safety. The safety function for each system (assuming the other system is not functioning) provides sufficient capacity and capability to assure that specified acceptable fuel design limits and design conditions of the
 
reactor coolant pressure boundary are not exceeded as a result of anticipated operational
 
occurrences, and the core is cooled and containment integrity and other vital functions are
 
maintained in the event of postulated accidents. 
 
The onsite electric power supplies, including the batteries, and the onsite electric distribution
 
system, have sufficient independence, redundancy, and testability to perform their safety
 
functions assuming a single failure. 
 
FNP-FSAR-3
 
3.1-10 REV 21  5/08 Electric power from the transmission network to the onsite electric distribution system is supplied by two physically independent circuits (not necessarily on separate rights of way) designed and located so as to minimize, to the extent practical, the likelihood of their
 
simultaneous failure under operating and postulated accident and environmental conditions. A
 
switchyard common to both circuits is acceptable. Each of these circuits is designed to be
 
available in sufficient time following a loss of all onsite alternating current power supplies and
 
the other offsite electric power circuit. This assures that specified acceptable fuel design limits
 
and design conditions of the reactor coolant pressure boundary are not exceeded. One of
 
these circuits is designed to be available within a few seconds following a LOCA to assure that
 
core cooling, containment integrity, and other vital safety functions are maintained. 
 
Provisions are included to minimize the probability of losing electric power from any of the
 
remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear
 
power unit, the loss of power from the transmission network, or the loss of power from the onsite electric power supplies. 
 
CONFORMANCE 
 
An onsite electric power system and an offsite electric power system are provided to permit functioning of systems and components important to safety. As discussed in sections 8.2 and 8.3, sufficient capacity and capability have been provided in each system to assure that
 
specified acceptable fuel design limits and design conditions of the reactor coolant pressure
 
boundary are not exceeded as a result of anticipated operational occurrences, and the core is
 
cooled and containment integrity and other vital functions are maintained in the event of
 
postulated accidents. 
 
The onsite electric ac power supply is furnished by five diesel generators. The engineered
 
safeguards loads are divided between the emergency buses of each unit in a balanced, redundant load grouping so that the failure of one diesel generator or one emergency bus will
 
not prevent the safe shutdown of both units. 
 
The onsite electric dc power supply for each unit consists of two redundant battery systems, either of which is adequate to supply the dc power required for the engineered safeguards.
 
Failure of a single component in this system will not impair control of the minimum engineered
 
safeguards required to maintain each unit in a safe condition. 
 
Electric power from the transmission network to the onsite electric distribution system is
 
provided for Units 1 and 2 by six 230- and 500-kV transmission lines. Four of these lines are
 
connected to the 230-kV switchyard and two of these lines are connected to the 500-kV
 
switchyard. The 230- and 500-Kv switchyards are interconnected at the plant by two 230/500-
 
kV autotransformers. Power from the switchyard is provided for the engineered safeguards for
 
each unit through two startup auxiliary transfo rmers via two physically independent circuits which serve alternately as the redundant source for each emergency bus. A failure of a single
 
active component in this power system w ill not prevent its required functioning. 
 
In case of loss of all onsite power and of one offsite power circuit, power requirements for
 
ensuring the specified fuel design limits and the design conditions of the reactor coolant
 
pressure boundary are not exceeded since they are met from the other redundant offsite power
 
source.
FNP-FSAR-3
 
3.1-11 REV 21  5/08 In the event of a LOCA, all the emergency supplies are available to supply safety loads to assure that core cooling, containment integrity, and other vital safety functions are maintained. 
 
The electric power supplies have been designed so as to minimize the probability of losing
 
electric power from any of the remaining supplies as a result of, or coincident with, the loss of
 
power generated by the nuclear power unit, the loss of offsite power, or the loss of onsite
 
power. 
 
3.1.14  CRITERION 18 - INSPECTION AND TESTING OF ELECTRIC POWER SYSTEMS Electric power systems important to safe ty are designed to permit appropriate periodic inspection and testing of important areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the condition of their
 
components. The systems are designed with a capability to test periodically the operability and
 
functional performance of the components of the systems, such as onsite power sources, relays, switches, and buses, and the operability of the systems as a whole and, under
 
conditions as close to design as practical, the full operation sequence that brings the systems
 
into operation, including operation of applicable portions of the protection system, and the
 
transfer of power among the nuclear power unit, the offsite power system, and the onsite power system. 
 
CONFORMANCE 
 
The inspection and testing of components of the electric power systems are divided into two
 
types: 
 
I. Inspection and testing performed during periods of unit outage. 
 
II. Inspection and testing performed with the unit in service. 
 
Type I covers checking for circuit integrity of wiring insulation and connections, of 4160-V
 
switchgear, 600-V load centers, motor control centers, and the interphase circuitry between
 
components falling under Type II. Type II covers the periodic testing of components of the
 
electrical system such as diesel generators, relays, breakers, valves, and motors to assess their
 
operability and functional performance. Testing of systems as a whole for the full operation
 
sequence that brings the systems into operation, including operation of applicable portions of
 
the protection system and the transfer of power fr om the nuclear power unit to the offsite power source, will be done on a periodic basis. The scope of testing involved for these types is
 
covered in subsections 8.3.1 and 8.3.2. 
 
3.1.15  CRITERION 19 - CONTROL ROOM A control room is provided from which actions can be taken to operate the nuclear power unit
 
safely under normal conditions and to maintain it in a safe condition under accident conditions, including LOCAs. Adequate radiation protection is provided to permit access and occupancy of
 
the control room under accident conditions without personnel receiving radiation exposures in FNP-FSAR-3
 
3.1-12 REV 21  5/08 excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. 
 
Equipment at appropriate locations outside the control room is provided with a design capability
 
for prompt hot shutdown of the reactor, including necessary instrumentation and controls to
 
maintain the unit in a safe condition during hot shutdown, and with a potential capability for
 
subsequent cold shutdown of the reactor through the use of suitable procedures. 
 
CONFORMANCE 
 
Following proven power plant design philosophy, vital control stations, switches, controllers, and
 
indicators necessary to start up, operate, and shut down each nuclear unit and maintain safe
 
control of the facility are located in one control room. The control room and the associated post
 
accident ventilation systems are designed in accordance with Category I requirements, as
 
discussed in subsection 3.2.1. 
 
The design of the control room will permit access and occupancy during a LOCA. Sufficient
 
shielding and ventilation are provided to permit occupancy of the control room for a period of 30
 
days following the LOCA, without receiving more than 5 rem integrated whole body dose, or its
 
equivalent, to any part of the body. The shielding is discussed in subsection 12.1.2, ventilation
 
is discussed in subsection 9.4.1, and the control room habitability systems are discussed in
 
section 6.4. 
 
As discussed in subsection 7.4.1, it is possible to bring each unit to the hot shutdown condition, and maintain it in that condition, from outside the control room if access to the control room is
 
prohibited during normal operation. Through the use of suitable procedures, the plant design
 
also provides the potential capability of attaining cold shutdown from outside the control room. 
 
3.1.16  CRITERION 20 - PROTECTION SYSTEM FUNCTIONS The protection system is designed to initiate autom atically the operation of appropriate systems, including the reactivity control systems, to assure that specified acceptable fuel design limits are
 
not exceeded as a result of anticipated operational occurrences, and to sense accident
 
conditions and initiate the operation of syst ems and components important to safety. 
 
CONFORMANCE 
 
The operational limits for the reactor protection system are defined by analyses of plant
 
operating and fault conditions requiring rapid rod insertion to prevent or limit core damage. With
 
respect to acceptable fuel design limits, the system design bases for anticipated operational
 
occurrences are as follows:
 
A. Minimum departure from nucleate boiling ratio (DNBR) will not be less than the safety analysis limit.
 
B. Clad strain on the fuel element shall not exceed 1 percent. 
 
C. No center melt shall occur in the fuel elements.
FNP-FSAR-3
 
3.1-13 REV 21  5/08 A region of permissible core operation is defined in terms of power, axial power distribution, and coolant flow and temperature. The protection syst em monitors these process variables (as well as other process variables and plant conditions). If the region limits are approached during
 
operation, the protection system will automatic ally actuate alarms, prevent control rod withdrawal, or trip the reactor depending on the severity of the condition. 
 
Operation within the permissible region and complete core protection is assured by the overtemperature T and overpower T reactor trips in the system pressure range defined by the pressurizer high pressure and pressurizer low pressure reactor trips, in the event of a transient
 
that is slow with respect to piping delays from t he core to the temperature sensors. In the event that a transient faster than the T responses occurs, high nuclear flux and low coolant flow reactor trips provide core protection. Finally, the thermal transients are anticipated and avoided
 
by reactor trips initiated by turbine trip and primary coolant pump circuit breaker position. 
 
The protection system operates by interrupting power to the rod control power supply. As a
 
result, all full length control and shutdown rods insert by gravity. The Westinghouse protection
 
system design meets the requirements of IEEE-279-1971, "Criteria for Protection Systems for Nuclear Power Generating Stations." 
 
The protection system measures a wide spectrum of process variables and plant conditions. All analog channels that actuate reactor trip, rod stop, and the permissive functions are indicated or
 
recorded. In addition, visual and/or audible alarms are actuated for reactor trip; partial reactor
 
trip, any input channel; and any control variable exceeding its setpoint, any input channel. 
 
These measurements and indications provide the bases for corrective action to prevent the development of accident conditions. In the event of accident conditions, however, the reactor protection system senses the condition, processes the signals used for engineered safety
 
features actuation, and generates the actuation demand. 
 
The reactor trip system is discussed in section 7.2 and the engineered safety features are
 
discussed in section 7.3. 
 
3.1.17  CRITERION 21 - PROTECTION SYSTEM RELIABILITY AND TESTABILITY The protection system is designed for high functional reliability and inservice testability
 
commensurate with the safety functions to be performed. Redundancy and independence
 
designed into the protection system are sufficient to assure that no single failure results in loss
 
of the protection function and to assure that removal from service of any component or channel
 
does not result in loss of the required minimum redundancy unless the acceptable reliability of
 
operation of the protection system can be other wise demonstrated. The protection system is designed to permit periodic testing of its functioning when the reactor is in operation, including a
 
capability to test channels independently to determine failures and losses of redundancy that
 
may have occurred. 
 
CONFORMANCE 
 
The protection system is designed for the high functional reliability and inservice testability
 
commensurate with the safety functions to be performed.
 
FNP-FSAR-3
 
3.1-14 REV 21  5/08 The system consists of a large number of i nput measurement channels, redundant logic trains, redundant reactor trip breakers, and redundant engineered safety features actuation devices. It
 
performs indication and alarm functions in addition to its reactor trip and engineered safety
 
features actuation functions. The design meets the requirements of IEEE Standard 279-1971, "Criteria for Nuclear Power Generating Station Protection Systems." The redundant logic trains, reactor trip breakers, and engineered safety features actuation relays are electrically isolated
 
and physically separated. Further, physical separation of the channels is maintained within the
 
separated trains to the point of logical processing. 
 
Either of the two redundant logic trains performs the required protection function. All channels
 
employed in power operation are sufficiently redundant so that individual testing and calibration, without degradation of the protection function or violation of IEEE Standard 279-1971, can be
 
performed with the reactor at power. Such testing discloses failures or reduction in redundancy
 
that may have occurred. Removal from serv ice of any single channel or component employed during power operation does not result in loss of minimum required redundancy. For example, a two-of-three logic function is placed in the one-of-two mode when one channel is removed
 
from service. 
 
Semiautomatic testers are built into each of the two logic trains. These testers have the
 
capability of testing the major part of the protection system rapidly with the reactor at power. 
 
Between tests, the testers continuously monitor a number of internal protection system points, including train power supply voltages and fuses. The outputs of these monitor circuits are
 
logically processed to provide an alarm in the event of a single failure in either train, and
 
automatic reactor trip in the event of one or more for failures in both trains. Self testing
 
provision is designed into each tester. 
 
The protection system is discussed in section 7.2. 
 
3.1.18  CRITERION 22 - PROTECTION SYSTEM INDEPENDENCE The protection system is designed to assure that the effects of natural phenomena, and of
 
normal operating, maintenance, testing, and postulated accident conditions on redundant
 
channels do not result in loss of the protection function, or are demonstrated to be acceptable
 
on some other defined basis. Design techniques, such as functional diversity or diversity in
 
component design and principles of operation, are used to the extent practical to prevent loss of
 
the protection function. 
 
CONFORMANCE 
 
The protection system has been designed to perform its intended protective functions under the
 
effects of accident conditions or postulated events. The design features that limit the effects of
 
natural phenomena such as tornado, flood, earthquake, and fire, are physical separation and
 
electrical isolation of redundant channels and subs ystems, functional diversity of subsystems, and safe component and subsystem failure modes. The redundant logic trains, reactor trip
 
breakers, and safety features actuation devic es are physically separated and electrically isolated. Physically separate channel cable trays, conduit, and penetrations are provided to
 
ensure independence of redundant elements of each train. Functional diversity and location
 
diversity are designed into the system. For example, the loss of one feedwater pump would FNP-FSAR-3
 
3.1-15 REV 21  5/08 actuate one pressure trip, one high-level trip, one low-level trip, two temperature trips, and one flow trip. The system logic is designed so that, wi th exception of the safety features actuation devices, a zero input represents a trip demand. Hence, severed or shorted channel wiring, loss
 
of power, and the majority of channel component failures are seen by the system as trip demands. 
 
The protection system is tested and qualified under extreme environmental conditions. These tests ensure that the equipment will perform the required functions under accident conditions. 
 
Loss of the protection function through improper testing or failure of the test equipment is
 
guarded against by interlocks that enable the test of only one of the two trains at the same time, bypass trip breakers to maintain the protection function during test, annunciation of the test
 
mode, unambiguous tester readout, and the indication, alarm, and status systems. 
 
Functional and locative diversity designed into the system are defenses against loss of the
 
protection function through postulated accident conditions. For the postulated loss of coolant
 
accident, at least five diverse reactor trip demands and at least two diverse engineered safety
 
features actuation demands would be generated. In addition, manual reactor trip and manual
 
engineered safety features actuation means are provided. 
 
The protection system has been quantitatively evaluated with respect to functional diversity and
 
qualitatively evaluated with respect to common mode susceptibility. These studies indicate that
 
the system is designed to have a very high probability of performing its function in any postulated occurrence. 
 
The reactor protection system and the engineered safety features actuation system are discussed in sections 7.2 and 7.3, respectively. 
 
3.1.19  CRITERION 23 - PROTECTION SYSTEM FAILURE MODES The protection system is designed to fail into a safe state or into a state demonstrated to be
 
acceptable on some other defined basis if conditions such as disconnection of the system, loss
 
of energy (e.g., electric power, instrument air), or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced. 
 
CONFORMANCE 
 
The protection system is designed with due consideration of the most probable failure modes of
 
the components under various perturbations of energy sources and the environment. Each
 
reactor trip channel is designed on the 'deenergize to trip' principle so that a loss of power or
 
disconnection or shorting of a channel causes that channel to go into its tripped mode. 
 
Likewise, loss of voltage to either of the two protection system output devices will trip the
 
reactor. In a two-out-of-three logic circuit, the three channels are equipped with separate
 
primary sensors and each channel is energized from independent electrical buses. A single
 
failure can, in the worst case, cause a single channel to fail to deenergize. The trip signal
 
furnished by the two remaining channels is unimpaired by this event. In addition, 15 internal
 
points in each train are continuously monitored by the semiautomatic testers. Faults involving one logic train are annunciated; multiple faults involving both trains automatically trip the FNP-FSAR-3
 
3.1-16 REV 21  5/08 reactor, although such faults would not necessarily defeat the trip function. All full-length control and shutdown rods will insert by gravity if the rod power supply is lost. 
 
The protection system components have been test ed and qualified for the extremes of the normal environment to which they are subjected. In addition, components are tested and
 
qualified according to individual requirements for the adverse environment specific to their
 
location which might result from postulated accident conditions. 
 
In the event of a loss of the preferred offsite power source, onsite diesel generators provide
 
power to emergency loads. Station batteries are provided to power the vital instrumentation
 
loads. The diesels are capable of supplying power required to operate engineered safeguards
 
pumps and associated valves. A loss of power to one train of emergency core cooling
 
equipment will not affect the ability of the other train to perform its function. 
 
The rod control system, reactor trip system, and engineered safety features actuation systems are discussed in sections 4.2., 7.2, and 7.3, respectively. 
 
3.1.20  CRITERION 24 - SEPARATION OF PROTECTION AND CONTROL SYSTEMS The protection system is separated from control syst ems to the extent that failure of any single control system component or channel, or failure or removal from service of any single protection system component or channel which is common to the control and protection systems, leaves intact a system satisfying all reliability, redundancy, and independence requirements of the
 
protection system. Interconnection of the protec tion and control systems is limited so as to assure that safety is not significantly impaired. 
 
CONFORMANCE 
 
The failure of a single control system component or channel, or the failure or removal from
 
service of any protection system component or channel, which is common to the control and protection systems, leaves intact a syst em satisfying all reliability, redundancy, and independence requirements of the protection system. Interconnection of the protection and control systems is limited so as to assu re that safety is not impaired. 
 
Most functions performed by the reactor protection and the reactor control systems require the
 
same process information. The design philosophy for these systems is to make maximum use
 
of a wide spectrum of diverse and redundant proc ess measurements. The protection system is separate and distinct from the control sy stem. The control system is dependent on the protection system in that control input signals are derived from protection system measurements, where applicable. These control signals are transferred to the control system by isolation amplifiers which are classified as protection system components. No credible failure at the output of an isolation amplifier will prevent the corresponding protection channel
 
from performing its protection function. Such failures include short circuits, open circuits, grounds, and the application of the maximum credible ac and dc voltages. The adequacy of
 
system isolation has been verified by testing under these fault conditions. The design meets all
 
requirements of IEEE 279-1971, "Criteria for Protection Systems for Nuclear Power Generating
 
Stations." 
 
FNP-FSAR-3
 
3.1-17 REV 21  5/08 The reactor protection system and the reactor cont rol system are discussed in sections 7.2 and 7.7, respectively. 
 
3.1.21  CRITERION 25 - PROTECTION SYSTEM REQUIREMENTS FOR REACTIVITY CONTROL MALFUNCTIONS The protection system is designed to assure that specified acceptable fuel design limits are not
 
exceeded for any single malfunction of the reacti vity control systems, such as accidental withdrawal (not ejection or dropout) of control rods. 
 
CONFORMANCE 
 
The protection system design assures that accept able fuel design limits are not exceeded in the event of single reactivity control malfunctions including accidental withdrawal of control rod
 
groups. Analyses of postulated accidents are given in chapter 15.0. 
 
Reactor shutdown with control rods is completely independent of the control functions. The trip
 
breakers will interrupt power to the full-length rod drive mechanisms to trip the reactor
 
regardless of the status of existing control signals. 
 
The reactor control system provides visual displays of the control rod assembly positions and
 
actuates an alarm in the event that deviation of rods occurs within their banks. 
 
Additional information is given by the response to Criterion 10. The reactivity control systems are discussed in subsection 4.2.3; the protection system is discussed in section 7.2; and the
 
electrical control systems are discussed in section 7.7. 
 
3.1.22  CRITERION 26 - REACTIVITY CONTROL SYSTEM REDUNDANCY AND CAPABILITY Two independent reactivity control systems of differ ent design principles are provided. One of the systems uses control rods, preferably including a positive means for inserting the rods, and
 
is capable of reliably controlling reactivity changes to assure that under conditions of normal
 
operation, including anticipated operational occurrences, and with appropriate margin for
 
malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The
 
second reactivity control system is capable of re liably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable
 
fuel design limits are not exceeded. One of the systems is capable of holding the reactor core subcritical under cold conditions. 
 
CONFORMANCE 
 
Two independent reactivity control systems of differ ent design principles are provided. One of the systems uses control rods; the second syst em employs dissolved boron (chemical shim). 
 
The rods are assembled in clusters and are manipulated as groups (a) of clusters. Two functional categories of rods are employed. These categories are full-length shutdown and full-FNP-FSAR-3
 
3.1-18 REV 21  5/08 length control. During operation the shutdown rod banks are fully withdrawn. The full-length control rod system automatically maintain s a programmed average reactor temperature compensating for reactivity effects associated with scheduled and transient load changes. The
 
shutdown rod banks, along with the full-length control banks, are designed to shut down the
 
reactor with adequate margin under conditions of normal operation and anticipated operational
 
occurrences, thereby ensuring that specified fuel design limits are not exceeded. The most
 
restrictive period in core life is assumed in all analyses and the most reactive rod cluster is
 
assumed to stick in/out of core position. The r eactor protection system initiates reactor trip by interrupting power to the rod control power supply. This releases the magnetic latches, and the
 
full-length control and shutdown rods are inserted. The design is inherently fail safe. 
 
The boron system is capable of controlling the rate of reactivity change resulting from planned
 
normal power changes, including xenon burnout, to assure fuel design limits are not exceeded. 
 
This system is capable of maintaining the reactor core subcritical under cold conditions with all
 
rods withdrawn. 
 
The reactivity control systems are discussed in subsection 4.2.3. 
 
3.1.23  CRITERION 27 - COMBINED REACTIVITY CONTROL SYSTEMS CAPABILITY The reactivity control systems are designed to have a combined capability, in conjunction with
 
poison addition by the emergency core cooling syst em, of reliably controlling reactivity changes to assure that under postulated accident conditions, and with appropriate margin for stuck rods, the capability to cool the core is maintained. 
 
CONFORMANCE 
 
The facility steam supply system is provided with the means of making and holding the core
 
subcritical under any anticipated condition with appropriate margin for contingencies. 
 
Combined use of rod control and chemical shim control permit the necessary shutdown margin
 
to be maintained during long term xenon decay and plant cooldown. The single highest worth
 
control rod cluster is assumed to be stuck in its fully withdrawn position in postulated accident
 
analyses. These means are discussed in detail in chapter 4 and subsection 9.3.4. 
 
Under accident conditions when the safety injecti on system is actuated, concentrated boric acid is injected into the reactor coolant system. In case of a LOCA the accumulators will passively
 
inject borated water. Reactivity effects of safety injection are discussed in section 6.3 and
 
evaluated for accident conditions in chapter 15. 
 
3.1.24  CRITERION 28 - REACTIVITY LIMITS The reactivity control systems are designed wi th appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither
 
result in damage to the reactor coolant pressure boundary greater than limited local yielding nor
 
sufficiently disturb the core, its support structures, or other reactor pressure vessel internals to
 
impair significantly the capability to cool the core. These postulated reactivity accidents shall FNP-FSAR-3
 
3.1-19 REV 21  5/08 include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition. 
 
CONFORMANCE 
 
Core reactivity is controlled by a chemical poison dissolved in the coolant, rod cluster control
 
assemblies, and burnable poison rods. The maximum reactivity insertion rates due to
 
withdrawal of a bank of rod cluster control assemblies or by boron dilution are limited. These
 
limits are set so that peak heat generation rate and the departure from nucleate boiling ratio (DNBR) do not exceed the specified limits at overpower conditions. The maximum worth of
 
control rods and the maximum rates of reactivity insertion employing control rods are limited to
 
values which prevent rupture of the coolant pressure boundary or disruption of the core
 
internals to a degree which would impair core cooling capacity. The reactor can be brought to
 
the shutdown condition and the core will maintain acceptable heat transfer geometry following
 
any Condition IV event, such as rod ejection or steam line break. Specifically, faulted condition stress limits are used to contain, within tolerable lim its, the effects of extremely unlikely events. 
 
3.1.25  CRITERION 29 - PROTECTION AGAINST ANTICIPATED OPERATIONAL OCCURRENCES The protection and reactivity control system s are designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational
 
occurrences. 
 
CONFORMANCE 
 
The protection and reactivity control systems ar e designed to assure extremely high probability of performing their required safety functions in the event of anticipated operational occurrences.
 
Redundancy, functional and locative diversity, testability, use of safe failure modes, and
 
analyses are design measures which are employed to assure performance of the required
 
safety functions. Detailed probabilistic analyses of the systems verify this high reliability. The
 
protection system is further discussed under Criteria 20 through 25 and in section 7.2. The
 
reactivity control systems are discussed in sections 4.2 and 7.7. 
 
3.1.26  CRITERION 30 - QUALITY OF REACTOR COOLANT PRESSURE BOUNDARY Components which are part of the reactor coolant pressure boundary are designed, fabricated, erected, and tested to the highest quality standards practical. Means are provided for detecting
 
and, to the extent practical, identifying the location of the source of reactor coolant leakage. 
 
_________________
: a. Two rod groups are considered a rod bank. The groups in a bank are electrically interlocked 
 
to move alternately and to remain within one step of each other.
FNP-FSAR-3
 
3.1-20 REV 21  5/08 The reactivity control systems are discussed in subsection 4.2.3 and section 4.3. 
 
CONFORMANCE 
 
Reactor coolant pressure boundary components are designed, fabricated, inspected, and tested
 
in conformance with ASME Nuclear Power Plant Components Code, Section III. The design
 
bases and evaluations of reactor coolant pressure boundary components are discussed in
 
section 5.2. 
 
Major components are classified as ANS N18.2 Safety Class 1 and are accorded the quality
 
measures appropriate to this classification. 
 
Leakage is detected by an increase in the amount of makeup water required to maintain a
 
normal level in the pressurizer. The reactor vessel closure joint is provided with a temperature
 
monitored leakoff between double gaskets. Leakage inside the containment is drained to the
 
containment sump.
 
Leakage is also detected by measuring the airborne activity and the rate of condensate drained
 
from the containment air recirculation units. Monitoring the inventory of reactor coolant in the
 
system at the pressurizer, volume control tank, and coolant drain collection tanks makes
 
available an accurate indication of integrated leakage. 
 
The reactor coolant pressure boundary leakage detection system is discussed in subsection
 
5.2.7. 
 
3.1.27  CRITERION 31 - FRACTURE PREVENTION OF REACTOR COOLANT PRESSURE BOUNDARY The reactor coolant pressure boundary is designed with sufficient margin to assure that when
 
stressed under operating, maintenance, testing, and postulated accident conditions the
 
boundary behaves in a nonbrittle manner and the probability of rapidly propagating fracture is
 
minimized. The design reflects consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident
 
conditions and the uncertainties in determining material properties, the effects of irradiation on
 
material properties, residual, steady-state and transient stresses, and size of flaws. 
 
CONFORMANCE 
 
The Joseph M. Farley Units No. 1 and 2 are designed to conform to the intent of Criterion 31. 
 
The reactor coolant pressure boundary is designed so that, for all transients, normal, upset, and
 
faulted, the reactor coolant pressure boundary behaves in a nonbrittle manner. The units were
 
designed for 650° and 2500 psia. The normal service temperature and pressure are 550°F and
 
2250 psia. The reactor pressure vessels were designed in accordance with Section III of the
 
ASME Boiler and Pressure Vessel Code, which considers cyclic loading, defect
 
characterization, minimum material toughness, and maximum allowable stresses. Material
 
selection and testing were in accordance with the Summer 1970 Addenda of the ASME Code. 
 
FNP-FSAR-3
 
3.1-21 REV 21  5/08 Normal operating limits are calculated based on the toughness properties of the ferritic components of the pressure boundary in accordance with nonmandatory Appendix G, Section
 
III of the ASME Boiler and Pressure Vessel Code.  [For details, see heatup and cooldown limit
 
curves for normal operation, which are provided in the Pressure Temperature Limits Reports (PTLR)]. 
 
Postulated accident conditions are analyzed using the concepts of fracture mechanics
 
technology with which one quantitatively considers defect size and geometry, applied stresses (residual, thermal, and membrane), and the material properties. 
 
The effects of irradiation are considered in the generation of the heatup and cooldown limit
 
curves, analysis of post operational tests, and for analyses of all other transients and postulated
 
accidents. A reactor vessel materials radiation surveillance program is performed in
 
accordance with ASTM E-185.  (For details see paragraph 5.2.4.4.) 
 
3.1.28  CRITERION 32 - INSPECTION OF REACTOR COOLANT PRESSURE BOUNDARY Components which are part of the reactor coolant pressure boundary are designed to permit
 
periodic inspection and testing of important areas and features to assess their structural and
 
leaktight integrity and to provide an appropriate material surveillance program for the reactor pressure vessel. 
 
CONFORMANCE 
 
The Joseph M. Farley Units' design conforms with the intent of Criterion 32. The Units' reactor
 
coolant pressure boundary design meets the requirements of the ASME Boiler and Pressure Vessel Code, Section XI, which requires access for all required inspections. The design also
 
permits the conduct of a material surveillance program for the reactor pressure vessel. 
 
Additional details of these features can be found in subsections 5.2.8, Inservice Inspection
 
Program, and 5.2.4, Fracture Toughness.
 
3.1.29  CRITERION 33 - REACTOR COOLANT MAKEUP A system to supply reactor coolant makeup for protection against small breaks in the reactor
 
coolant pressure boundary is provided. The system safety function is to assure that specified
 
acceptable fuel design limits are not exceeded as a result of reactor coolant loss due to leakage
 
from the reactor coolant pressure boundary and rupture of small piping or other small
 
components which are part of the boundary. The system is designed to assure that for onsite
 
electric power system operation (assuming offsite pow er is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished using the piping, pumps, and valves used to maintain coolant inventory during
 
normal reactor operation. 
 
FNP-FSAR-3
 
3.1-22 REV 21  5/08 CONFORMANCE 
 
The chemical and volume control system provides a means of reactor coolant makeup and
 
adjustment of the boric acid concentration. Makeup is added automatically if the level in the
 
volume control tank falls below a preset level. High pressure centrifugal charging pumps are
 
provided which are capable of supplying the required makeup and reactor coolant seal injection
 
flow with power available from either onsite or offsite electric power systems. These pumps also serve as high head safety injection pumps. In the event of a loss of coolant larger than the
 
capacity of the normal makeup path, these pumps discharge into the larger safety injection
 
piping. A high degree of functional reliability is assured by provision of standby components and
 
assuring safe response to probable modes of failure. Details of system design are included in
 
subsection 9.3.4 and details of the electric power systems are given in chapter 8.0. 
 
3.1.30  CRITERION 34 - RESIDUAL HEAT REMOVAL A system to remove residual heat is provided.
The system safety function is to transfer fission product decay heat and other residual heat from the reactor core at the rate such that specified
 
acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary
 
are not exceeded. 
 
Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities are provided to assure t hat for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure. 
 
CONFORMANCE 
 
The residual heat removal system (RHRS), in conjunction with the steam and power conversion
 
system, is designed to transfer the fission product decay heat and other residual heat from the
 
reactor core within acceptable limits. The cro ssover from the steam power conversion system to the RHRS occurs at approximately 350°F. 
 
Suitable redundancy is accomplished with the two residual heat removal pumps (located in
 
separate compartments with means available for draining and monitoring leakage), two heat exchangers, and the associated piping and cabling. The RHRS is able to operate on either
 
onsite or offsite electrical power. 
 
Suitable redundancy at temperatures above approx imately 350°F is provided by the steam generators and attendant piping. Details of the system design are given in subsection 5.5.7 and
 
section 6.3. 
 
3.1.31  CRITERION 35 - EMERGENCY CORE COOLING A system to provide abundant emergency core cooli ng is provided. The system safety function is to transfer heat from the reactor core following any loss of reactor coolant at a rate such that FNP-FSAR-3
 
3.1-23 REV 21  5/08 fuel and clad damage that could interfere with continued effective core cooling is prevented and clad metal water reaction is limited to negligible amounts. 
 
Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities are provided to assure that for onsite electric power
 
system operation (assuming offsite power is not av ailable) and for offsite electric power system operation (assuming onsite power is not avail able) the system safety function can be accomplished, assuming a single failure. 
 
CONFORMANCE 
 
By combining the use of passive accumulators with two safety injection pumps and two residual
 
heat removal pumps, emergency core cooling is provided even if there should be a failure of
 
any component in any system. The emergency core cooling system (ECCS) employs a passive system of accumulators, which do not require any external signals or source of power for their operation, to cope with the short-term cooling requirements of large reactor coolant pipe breaks.
 
Two independent and redundant high-pressure and high-flow pumping systems, each capable
 
of the required emergency cooling, are provided for small break protection and to keep the core
 
in a coolable geometry after the accumulators have discharged, following a large break.
 
Adequate design provisions assure the performance of the required safety functions even with the loss of a single component, assuming the electric power is available from either the offsite
 
or onsite electric power sources. Borated water is injected into the reactor coolant system by
 
accumulators, high-head safety injection pumps (charging pumps), and low-head safety
 
injection pumps (RHR pumps). The design meets the intent of the "Interim Policy Statement
 
Criteria for Emergency Core Cooling Systems for Light Water Power Reactors." 
 
The primary function of the ECCS is to deliver borated cooling water to the reactor core in the
 
event of a LOCA. This limits the fuel clad tem perature and thereby ensures that the core will remain intact and in place, with its essential heat transfer geometry preserved. This protection
 
is afforded for: 
 
A. All pipe break sizes up to and including the hypothetical circumferential rupture of a reactor coolant loop. 
 
B. A loss of coolant associated with a rod ejection accident. 
 
The basic criteria for LOCA evaluations are no clad melting, zircaloy water reactions will be
 
limited to an insignificant amount, and the core geometry is to remain essentially in place and
 
intact so that effective cooling of the core will not be impaired. 
 
The zircaloy water reactions will be limited to an insignificant amount so that the accident: 
 
A. Does not interfere with the emergency core cooling function to limit clad temperatures. 
 
B. Does not produce H 2 in an amount that when burned would cause the containment pressure to exceed the design value. 
 
FNP-FSAR-3
 
3.1-24 REV 21  5/08 For any rupture of a steam pipe and the associated uncontrolled heat removal from the core, the ECCS adds shutdown reactivity so that with a stuck rod, no offsite power, and minimum
 
engineered safety features, there is no cons equential damage to the primary system and the core remains in place and intact. With no stuck rod, offsite power, and all equipment operating
 
at design capacity, there is insignificant cladding damage. The ECCS is described in section
 
6.3. 
 
Chapter 15.0 contains an analysis of LOCAs. 
 
3.1.32  CRITERION 36 - INSPECTION OF EMERGENCY CORE COOLING SYSTEM The emergency core cooling system is designed to permit appropriate periodic inspection of
 
important components, such as spray rings in the reactor pressure vessel, water injection
 
nozzles, and piping to assure the integrity and capability of the system. 
 
CONFORMANCE 
 
Design provisions are made for inspection to the extent practical of all components of the
 
emergency core cooling system. An inspection is performed periodically to demonstrate system readiness. 
 
The pressure-containing systems are inspected for leaks from pump seals, valve packing, flanged joints, and safety valves during system testing. 
 
In addition, to the extent that is practical, the critical parts of the reactor vessel internals, injection nozzles, pipes, valves, and safety inje ction pumps are inspected visually or by boroscopic examination for erosion, corrosion, and vibration wear evidence, and by
 
nondestructive inspection when such techniques are desirable and appropriate. 
 
Details of the inspection program for the reactor vessel internals are included in section 5.4. 
 
Inspection of the emergency core cooling system is discussed in subsection 6.3.4. 
 
3.1.33  CRITERION 37 - TESTING OF EMERGENCY CORE COOLING SYSTEM The emergency core cooling system is designed to permit appropriate periodic pressure and
 
functional testing to assure the structural and leaktight integrity of its components, the
 
operability and performance of the active components of the system, and the operability of the
 
system as a whole. Also tested, under conditions as close to design as practical, is the
 
performance of the full operational sequence that brings the system into operation, including
 
operation of applicable portions of the protection system, the transfer between normal and
 
emergency power sources, and the operation of the associated cooling water system. 
 
CONFORMANCE 
 
The components of the system located outside the containment will be accessible for
 
leaktightness inspection during appropriate periodic tests. Each active component of the
 
emergency core cooling system may be individually actuated on the normal power source or FNP-FSAR-3
 
3.1-25 REV 21  5/08 transferred to emergency power sources at any time during plant operation to demonstrate operability. The centrifugal charging/safety inje ction pumps are part of the charging system, and this system is in continuous operation during plant operation. The test of the safety
 
injection pumps employs the minimum flow recirculation test line which connects back to the
 
refueling water storage tank. Remotely operated valves are exercised and actuation circuits
 
tested. The automatic actuation circuitry, valves, and pump breakers also may be checked
 
during integrated system tests performed during a planned cooldown of the reactor coolant system. 
 
Design provisions also include special instrumentation, testing, and sampling lines to perform
 
the tests during plant shutdown to demonstrate proper automatic operation of the ECCS. A test
 
signal is applied to initiate automatic action, and verification made that the safety injection
 
pumps attain required discharge heads. The test demonstrates the operation of the valves, pump circuit breakers, and automatic circuitry. 
 
These tests are described in subsection 6.3.4. 
 
3.1.34  CRITERION 38 - CONTAINMENT HEAT REMOVAL A system to remove heat from the reactor contai nment is provided. The system safety function is to reduce rapidly, consistent with the functioning of other associated systems, the
 
containment pressure and temperature following any LOCA and maintain them at acceptably
 
low levels. 
 
Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities are provided to assure that for onsite electrical power
 
system operation (assuming offsite power is not av ailable) and for offsite electric power system operation (assuming onsite power is not avail able) the system safety function can be accomplished, assuming a single failure. 
 
CONFORMANCE
 
Three systems are provided to reduce contai nment atmosphere temperature and pressure and/or to remove heat from the containment under post accident conditions. These are the low-
 
head safety injection/residual heat removal sy stem, the containment spray system, and the containment cooling system. The design, operation, and reliability of the low-head safety
 
injection/residual heat removal system are discussed in section 6.3, Emergency Core Cooling
 
System.
 
The containment spray system has been designed to spray water into the containment
 
atmosphere, when appropriate, in the event of a MSLB or LOCA, to ensure the containment
 
peak pressure is below its design value. This is accomplished by one of the two trains of
 
containment spray. A detailed description of the containment spray system is provided in
 
subsection 6.2.2.
 
The containment cooling system has been designed to remove heat which would be released to
 
the containment atmosphere during any MSLB or LOCA up to and including the double-ended
 
rupture of the largest system pipe. This is accomplished by one of the four containment air FNP-FSAR-3
 
3.1-26 REV 21  5/08 coolers. A detailed description of the containment cooling system is provided in subsection 6.2.2. 
 
The containment heat removal systems are designed to ensure that the failure of any single
 
active component, assuming the availability of either onsite or offsite power exclusively, does
 
not prevent the systems from accomp lishing their design safety functions.
 
3.1.35  CRITERION 39 - INSPECTION OF CONTAINMENT HEAT REMOVAL SYSTEM The containment heat removal system is designed to permit appropriate periodic inspection of
 
important components, such as the torus, sumps, spray nozzles, and piping to assure the
 
integrity and capability of the system. 
 
CONFORMANCE
 
The containment air coolers and associated service water system piping (subsection 6.2.2)
 
inside containment can be inspected during shut down. The service water system is outside containment and, with the exception of buried piping, can also be inspected periodically. 
 
The containment spray system (subsection 6.2.2) essential equipment, except risers, distribution header piping, spray nozzles, and the containment sump are located outside 
 
containment. The containment sump and the spray pipe and nozzles can be inspected during
 
shutdowns. Those portions of the containment spray suction piping from the containment
 
sump and refueling water storage tank either embedded in concrete or buried are not
 
accessible for inspections. Associated equipment outside the containment can be visually
 
inspected. 
 
3.1.36  CRITERION 40 - TESTING OF CONTAINMENT HEAT REMOVAL SYSTEM The containment heat removal system is designed to permit appropriate periodic pressure and
 
functional testing to assure the structural and leaktight integrity of its components; the
 
operability and performance of the active components of the system; the operability of the
 
system as a whole; and, under conditions as close to the design as practical, the performance
 
of the full operational sequence that brings the system into operation, including operation of
 
applicable portions of the protection system, the transfer between normal and emergency power
 
sources, and the operation of the associated cooling water system. 
 
CONFORMANCE 
 
The normal operation of the four containment coolers will verify structural and leaktight integrity
 
of the system components, as well as demonstrate the operation of the associated cooling
 
system. System piping, valving, pumps, fans (at low speed), coolers, and other components of this system are arranged so that each component can be tested periodically for operability (see
 
subsection 6.2.2). The delivery capability of the containment spray system can be tested
 
periodically to the extent practical up to the last valve before the spray nozzles. These tests can
 
also verify the structural and leaktight integrity of the spray system components (see subsection
 
6.2.2).
FNP-FSAR-3
 
3.1-27 REV 21  5/08 The containment air cooling and containment spray systems are designed to provide the capability for testing the full operational sequence as close to design as practical, including
 
transfer to alternate power sources, during shutdown or refueling (see paragraphs 6.2.2.4 and
 
6.2.3.4). 
 
3.1.37  CRITERION 41 - CONTAINMENT ATMOSPHERE CLEANUP Systems to control fission products, hydrogen, oxygen, and other substances which may be
 
released into the reactor containment shall be provided as necessary to reduce, consistent with
 
the functioning of other associated systems, the concentration and quality of fission products
 
released to the environment following postulated accidents, and to control the concentration of
 
hydrogen or oxygen and other substances in the containment atmosphere following postulated
 
accidents to assure that containment integrity is maintained. 
 
Each system shall have suitable redundancy in components and features, and suitable
 
interconnections, leak detection, isolation, and containment capabilities to assure that for onsite
 
electric power system operation (assuming offsite pow er is not available) and for offsite electric power system operation (assuming onsite power is not available) its safety function can be accomplished, assuming a single failure. 
 
CONFORMANCE 
 
The ECCS recirculation fluid pH is controlled to reduce airborne iodine activity levels inside
 
containment and to retain the removed iodine in solution in the core sump. For a complete
 
description of the design of the containment spray systems, see subsections 6.2.2 and 6.2.3. 
 
A penetration room filtration system is provi ded for control of contaminants from ECCS recirculation leakage following a LOCA. This system is also used to mitigate the
 
consequences of a fuel handling accident in the spent-fuel pool. For a complete description of
 
the design of the penetration room filtration system, see subsection 6.2.3. 
 
A hydrogen control system is provided to cont rol hydrogen concentrations inside containment following an accident. This system consists of redundant hydrogen-oxygen recombiners and
 
appropriate containment atmosphere mixing and sampling systems. Backup protection against
 
excessive containment atmosphere hydrogen conc entrations is provided by a post-accident containment purging system. For a complete description of the design of the containment
 
combustible gas control system see subsection 6.2.5
 
Each of the containment atmosphere cleanup systems has suitable redundancy in components
 
and features, and suitable interconnections, leak detection, isolation, and containment
 
capabilities to assure that for onsite electrical power system operation (assuming offsite power
 
is not available) and for offsite electrical pow er system operation (assuming onsite power is not available), its safety function can be accomplished assuming a single failure. 
 
FNP-FSAR-3
 
3.1-28 REV 21  5/08 3.1.38  CRITERION 42 - INSPECTION OF CONTAINMENT ATMOSPHERE CLEANUP SYSTEMS  The containment atmosphere cleanup systems are designed to permit appropriate periodic
 
inspection of important components such as filter frames, ducts, and piping to assure the
 
integrity and capability of the systems. 
 
CONFORMANCE 
 
Trisodium phosphate is added to the recirculation sump to help reduce airborne iodine activity
 
levels inside containment and to retain the removed iodine in solution in the sump. Inspection
 
of the remaining components of the containment spray system is discussed in the response to
 
Criterion 39. 
 
Components of the penetration room filtration syst em, including ducts, filters, fans, and dampers are located in the auxiliary building and are accessible for physical inspection (see
 
paragraph 6.2.3.4). 
 
The post-accident containment hydrogen-oxygen recombiners and atmosphere mixing system, and selective containment isolation valves associated with the post-accident containment
 
sampling and purging systems are located inside containment and are available for inspection
 
during shutdown or refueling. The trisodium phosphate baskets are also located inside
 
containment and are available for inspection during shutdown and refueling. The remaining
 
components of the containment atmosphere cl eanup systems are located outside containment and are available for periodic inspection (see paragraphs 6.2.3.4 and 6.2.5.4). 
 
3.1.39  CRITERION 43 - TESTING OF CONTAINMENT ATMOSPHERE CLEANUP SYSTEMS The containment atmosphere cleanup systems are designed to permit appropriate periodic
 
pressure and functional testing to assure the structural and leaktight integrity of its components;
 
the operability and performance of the active components of the systems such as fans, filters, dampers, pumps, and valves; the operability of t he systems as a whole; and, under conditions as close to design as practical, the performance of the full operational sequence that brings the
 
systems into operation, including operation of app licable portions of the protection system, the transfer between normal and emergency power sources, and the operation of associated systems. 
 
CONFORMANCE 
 
The capability for testing of the containment spray system has been discussed in the response
 
to Criterion 40. 
 
The penetration room filtration system is designed to provide the capability for testing the
 
penetration room filtration system through the full operational sequence and as close to design
 
conditions as practical, including transfer to alternate power sources, during shutdown or
 
refueling. Such tests assure the operability and performance of the active system components
 
and demonstrate the structural and leaktight integrity of the system components. A capability is
 
also provided for periodic testing and surveillance of the penetration room filtration system filters FNP-FSAR-3
 
3.1-29 REV 21  5/08 to ensure that filter bypass paths have not developed and filter and adsorber materials have not deteriorated beyond acceptable limits. 
 
The post-accident containment combustible gas control system is designed to allow appropriate
 
periodic testing of the system through the full operational sequence and as close to design
 
conditions as practical, including transfer to al ternate power sources, to demonstrate system integrity as well as operability and performance of system active components. For further
 
discussion of the testing capability of this system see subsection 6.2.5. 
 
3.1.40  CRITERION 44 - COOLING WATER A system to transfer heat from structures, sy stems, and components important to safety to an ultimate heat sink is provided. The system safety function is to transfer the combined heat load
 
of these structures, systems, and components under normal operating and accident conditions.
 
Suitable redundancy in components and features and suitable interconnections, leak detection, and isolation capabilities are provided to assure that for onsite electrical power system
 
operation (assuming offsite power is not avail able) and for offsite electric power system operation (assuming onsite power is not avail able) the system safety function can be accomplished, assuming a single failure. 
 
CONFORMANCE 
 
Component cooling water, service water and spent-fuel pool cooling systems are provided to
 
transfer heat from the plant to an ultimate heat sink. These systems have been designed to
 
transfer their respective heat loads under all anticipated operating and accident conditions. 
 
Suitable redundancy, leak detection, and system interconnection and isolation capabilities have
 
been incorporated into the design of these systems to assure that the systems can accomplish
 
all required safety functions, assuming a single failure concurrent with either onsite or offsite
 
power exclusively. 
 
A more complete description of the spent-fuel pool cooling system design is presented in
 
subsection 9.1.3. Descriptions of the design of the other cooling water systems are presented
 
in section 9.2. 
 
3.1.41  CRITERION 45 - INSPECTION OF COOLING WATER SYSTEM The cooling water system is designed to permit appropriate periodic inspection of important
 
components, such as heat exchangers and piping, to assure the integrity and capability of the system. 
 
CONFORMANCE 
 
All safety-related components of the cooling water systems, with the exceptions of buried piping to the river and to pond intake structures and service water piping to the containment coolers
 
that are located within the containment are accessible for periodic inspection. The service FNP-FSAR-3
 
3.1-30 REV 21  5/08 water piping inside the containment is accessible for inspection during reactor shutdown and refueling periods. The integrity of the buried pipe is demonstrated by pressure and functional
 
tests. 
 
Refer to section 9.2 for further discussion of these systems. 
 
3.1.42  CRITERION 46 - TESTING OF COOLING WATER SYSTEM The cooling water system is designed to permit appropriate periodic pressure and functional
 
testing to assure the structural and leaktight integrity of its components; the operability and the
 
performance of the active components of the syst em; the operability of the system as a whole; and, under conditions as close to design as practical, the performance of the full operational
 
sequence that brings the system into operation for reactor shutdown and for loss of coolant
 
accidents, including operation of applicable portions of the protection system and the transfer
 
between normal and emergency power sources. 
 
CONFORMANCE 
 
All cooling water systems operate either conti nuously or intermittently during normal plant operation. This functional operation serves to demonstrate the operability, performance, and
 
structural and leaktight integrity of the syst em components. The cooling water systems are designed to include the capability for testing through the full operational sequence, as close to
 
accident design conditions as practical, including transfer to alternate power sources (see
 
section 9.2). 
 
3.1.43  CRITERION 50 - CONTAINMENT DESIGN BASIS The reactor containment structure, including access openings, penetrations, and the
 
containment heat removal system are designed so that the containment structure and its
 
internal compartments can accommodate, without exceeding the design leakage rate and with
 
sufficient margin, the calculated pressure and temperature conditions resulting from any LOCA.
 
This margin reflects consideration of the effects of potential energy sources which have not
 
been included in the determination of the peak conditions, such as energy in steam generators
 
and energy from metal-water and other chemical reactions that may result from degraded
 
emergency core cooling functioning; the limited experience and experimental data available for defining accident phenomena and containment responses; and the conservatism of the
 
calculational model and input parameters. 
 
CONFORMANCE 
 
The design of the containment is based on the LOCA, coupled with the partial loss of the
 
redundant engineered safety features, which produces the post-LOCA temperature and
 
pressure conditions described in subsection 6.2.1. 
 
A minimum safety margin of 10 percent between the peak calculated post-LOCA containment
 
pressure and the design pressure has been provided. The containment structural design is
 
described in section 3.8.
FNP-FSAR-3
 
3.1-31 REV 21  5/08 3.1.44  CRITERION 51 - FRACTURE PREVENTION OF CONTAINMENT PRESSURE BOUNDARY The reactor containment boundary is designed with sufficient margin to assure that, under
 
operating, maintenance, testing, and postulated accident conditions, its ferritic materials behave
 
in a nonbrittle manner and to assure that the probability of rapidly propagating fracture is
 
minimized. The design reflects consideration of service temperatures and other conditions of the containment boundary material during operation, maintenance, testing, and postulated
 
accident conditions, and the uncertainties in determining material properties; residual, steady-
 
state, and transient stresses; and size of flaws. 
 
CONFORMANCE 
 
Principal load carrying components of ferritic materials used in the containment will not be
 
exposed to the external environment. The ferritic material of the containment liner plate is designed to function as a leaktight membrane only. 
 
In addition, nil ductility transition temperature (NDTT) requirements are not considered relevant
 
for the design of the containment since this is a ligament type of structure wherein the brittle
 
fracture of a ligament could not propagate to adjacent ligaments. 
 
In all areas where the liner plate is the pressure resisting structural element without backup from
 
the concrete, a minimum NDTT of 0°F has been specified, based on a minimum service
 
temperature of 30°F. These areas are the containment access openings which are enclosed on
 
the outside by heated buildings. The equipment hat ch not protected by the auxiliary building may experience service temperatures as low as 0°F; this material has been specified to an NDTT of -30°F. 
 
In all other areas, the liner plate has no structural function since it is backed up by concrete. 
 
Except for small locally thickened areas (up to 2 in.) in the floor and walls for crane brackets, anchorages for main steam pipe ruptures, frames, and similar items, the containment liner plate is 1/4 in. thick. No NDTT has been specified since the rules for NDTT from Section III of the
 
ASME Code for Class B vessels apply to pressure vessels only. In addition, these rules are
 
based on "Fracture Analysis Diagram Procedures for the Fracture-Safe Engineering Design of
 
Steel Structures" by Pellini and Puzak and, as stated by the authors, are not applicable to plates
 
less than 5/8 in. thick. See section 3.8 for details. 
 
Although not classified as a part of the containment for the Farley Nuclear Plant, piping and
 
penetrations through the containment pressure boundary out to the penetration isolation valves
 
have been designed and fabricated with due consideration for brittle fracture prevention. 
 
A selective review of the design and materials used for the Farley Nuclear Plant was performed
 
by the NRC staff, which confirmed compliance with General Design Criterion 51. Documentation
 
to support this conclusion was submitted to the NRC by Alabama Power Company letters dated
 
August 15, 1980; September 22, 1980; and January 30, 1981. 
 
FNP-FSAR-3
 
3.1-32 REV 21  5/08 3.1.45  CRITERION 52 - CAPABILITY FOR CONTAINMENT LEAKAGE RATE TESTING 
 
The reactor containment and other equipment which may be subjected to containment test
 
conditions are designed so that periodic integrated leakage rate testing can be conducted at
 
containment design pressure. 
 
CONFORMANCE 
 
The reactor containment and other equipment which may be subjected to containment test
 
conditions are designed so that periodic integrated leakage rate tests can be conducted. For a
 
complete description of the planned leak rate tests, including system design provisions to
 
accommodate the attendant containment pressure refer to paragraphs 3.8.1.7, 6.2.1.4, and the
 
plant Technical Specifications. 
 
3.1.46  CRITERION 53 - PROVISIONS FOR CONTAINMENT TESTING AND INSPECTION 
 
The reactor containment is designed to permit appropriate periodic inspection of all important
 
areas, such as penetrations, an appropriate surveillance program, and periodic testing at
 
containment design pressure of the leaktightness of penetrations which have resilient seals and
 
expansion bellows. 
 
CONFORMANCE 
 
There are special provisions for conducting individual leakage rate tests on applicable
 
penetrations. Penetrations are visually inspected and pressure tested for leaktightness at
 
periodic intervals. Other inspections are conducted as required by Appendix J of 10 CFR 50. 
 
Inservice inspection of the metallic liner and the pressure retaining concrete structure of the
 
containments of both units, meeting the requirements of Subsections IWE and IWL of the ASME Section XI Code and applicable edition and addenda as required by 10 CFR 50.55a, except where an alternative, exemption, or relief has been authorized by the NRC, will be performed.
Refer to paragraph 3.8.1.7, subsection 6.2.1, and the plant Technical Specifications, subsections 5.5.6 and 5.5.17. 
 
3.1.47  CRITERION 54 - PIPING SYSTEMS PENETRATING CONTAINMENT 
 
Piping systems penetrating primary reactor cont ainment are provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance
 
capabilities which reflect the importance to safety of isolating these piping systems. Such piping
 
systems are designed with a capability to test periodically the operability of the isolation valves
 
and associated apparatus and to determine if valve leakage is within acceptable limits. 
 
CONFORMANCE 
 
Piping systems penetrating the containment hav e been provided with appropriate isolation capabilities. The design of the isolation system incorporates the capability to test the operability
 
of the isolation valves periodically and to determine that valve leakage is within specified limits. 
 
Leak detection capability for the isolation valves is provided as described in paragraph 6.2.1.4. 
 
The containment isolation system is discussed in subsection 6.2.4. 
 
FNP-FSAR-3
 
3.1-33 REV 21  5/08 3.1.48  CRITERION 55 - REACTOR COOLANT PRESSURE BOUNDARY PENETRATING CONTAINMENT Each line that is part of the reactor coolant pressure boundary and that penetrates primary
 
reactor containment is provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as
 
instrument lines, are acceptable on some other defined basis, namely: 
 
(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment. 
(2) One automatic isolation valve inside and one locked closed isolation valve outside containment. 
(3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve
 
outside containment. 
(4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve
 
outside containment. 
 
Isolation valves outside containment are located as close to the containment as practical (and
 
upon loss of actuating power), automatic isolation valves are designed to take the position that
 
provides greater safety. 
 
Other appropriate requirements to minimize the probability or consequences of an accidental
 
rupture of these lines or of lines connected to them are provided, as necessary, to assure
 
adequate safety. Determination of the appropriateness of these requirements, such as higher
 
quality in design, fabrication, and testing, additional provisions for inservice inspection, protection against more severe natural phenomena, and additional isolation valves and
 
containment includes consideration of the population density, use characteristics, and physical
 
characteristics of the site environs. 
 
CONFORMANCE 
 
Each line that is part of the reactor coolant pressure boundary and that penetrates the reactor
 
containment is provided with containment isolation valves, in accordance with this criterion, as
 
discussed in subsection 6.2.4. 
 
Isolation valves outside the containment are located as close to the containment as practical, and automatic isolation valves are designed to take the position that provides greater safety
 
upon loss of actuating power. 
 
Other appropriate requirements to minimize the probability or consequences of an accidental
 
rupture of these lines or of lines connected to them are provided, as necessary, to assure
 
adequate safety. Determination of the appropriateness of these requirements, such as higher
 
quality in design, fabrication, and testing, additional provisions for inservice inspection, protection against more severe natural phenomena, and additional isolation valves and FNP-FSAR-3
 
3.1-34 REV 21  5/08 containment, includes consideration of the population density, use characteristics, and physical characteristics of the site environs. 
 
3.1.49  CRITERION 56 - PRIMARY CONTAINMENT ISOLATION Each line that connects directly to the containment atmosphere and penetrates primary reactor
 
containment is provided with containment isolation valves as follows, unless it can be
 
demonstrated that the containment isolation provisions for a specific class of lines, such as
 
instrument lines, are acceptable on some other defined basis, namely: 
 
(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment. 
(2) One automatic isolation valve inside and one locked closed isolation valve outside containment. 
(3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve
 
outside containment. 
(4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve
 
outside containment. 
 
Isolation valves outside containment are located as close to the containment as practical and, upon loss of actuating power, automatic isolation valves shall be designed to take the position
 
that provides greater safety. 
 
CONFORMANCE 
 
Each line that connects directly to the containment atmosphere and penetrates the primary
 
reactor containment is provided with containment isolation valves in accordance with this
 
criterion, as discussed in subsection 6.2.4. Isolation valves outside the containment are located
 
as close to the containment as practical. Upon loss of actuating power, automatic isolation
 
valves are designed to take the position that provides greater safety. 
 
3.1.50  CRITERION 57 - CLOSED SYSTEM ISOLATION VALVES Each line that penetrates primary reactor containment and is neither part of the reactor coolant
 
pressure boundary nor connected directly to the containment atmosphere has at least one
 
containment isolation valve which is either automatic, or locked closed, or capable of remote
 
manual operation. This valve is outside containment and located as close to the containment
 
as practical. A simple check valve may not be used as the automatic isolation valve. 
 
FNP-FSAR-3
 
3.1-35 REV 21  5/08 CONFORMANCE 
 
Each line that penetrates containment and is neither part of the reactor coolant pressure
 
boundary nor connected directly to the containment atmosphere is provided with an appropriate
 
containment isolation valve arrangement. Refer to subsection 6.2.4 for a complete description
 
of the design and operation of the containment isolation system. 
 
3.1.51  CRITERION 60 - CONTROL OF RELEASE OF RADIOACTIVE MATERIALS TO THE ENVIRONMENT The nuclear power unit design includes means to control suitably the release of radioactive
 
materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during
 
normal reactor operation, including anticipated operational occurrences. Sufficient holdup
 
capacity is provided for retention of gaseous and liquid effluents containing radioactive
 
materials, particularly where unfavorable site environmental conditions can be expected to
 
impose unusual operational limitations upon the release of such effluents to the environment. 
 
CONFORMANCE 
 
Control of waste gas effluents is accomplished by holdup of waste gases in decay tanks until
 
the activity of tank contents and existing env ironmental conditions permit discharges within Technical Specification requirements. Waste gas effluents are monitored at the point of
 
discharge for radioactivity and rate of flow. Sufficient waste gas holdup capacity is provided, as
 
discussed in section 11.3, to cope with all anticipated operational occurrences and site
 
environmental conditions. A decay tank burst would not result in an activity release greater than
 
10 CFR 100 limits, based on 1-percent failed fuel. 
 
Control of liquid waste effluents is accomplished by holdup of waste liquids in storage tanks, batch processing of all liquids, and sampling before controlled rate discharge. Liquid effluents
 
are monitored for radioactivity and rate of flow. The liquid waste disposal system tankage and
 
processing capacity, as described in section 11.2, is sufficient to cope with all anticipated
 
operational occurrences and unfavorable site environmental conditions. 
 
Station solid wastes are prepared in batches for offsite disposal by approved contractors in
 
shielded and reinforced containers which meet Federal Regulation requirements. Sufficient
 
handling capacity is provided, as discussed in section 11.5, to cope with all anticipated
 
operational occurrences. 
 
Chapters 11.0 and 15.0 provide additional information. 
 
3.1.52  CRITERION 61 - FUEL STORAGE AND HANDLING AND RADIOACTIVITY CONTROL  The fuel storage and handling, radioactive waste, and other systems which may contain
 
radioactivity are designed to assure adequate safety under normal and postulated accident
 
conditions. These systems are designed with a capability to permit appropriate periodic
 
inspection and testing of components important to safety; with suitable shielding for radiation FNP-FSAR-3
 
3.1-36 REV 21  5/08 protection; with appropriate containment, confinement, and filtering systems; with a residual heat removal capability having reliability and testability that reflects the importance to safety of
 
decay heat and other residual heat removal; and to prevent significant reduction in fuel storage
 
coolant inventory under accident conditions. 
 
CONFORMANCE 
 
With the exception of spent fuel stored in the ISFSI in accordance with the general license provisions of 10 CFR 72, fuel storage and most waste handling facilities are contained in the auxiliary building; equipment is designed to prevent accidental releases of radioactive material
 
directly to the environment. Components of these systems which are important to safety can be periodically inspected and tested. 
 
The spent-fuel storage pool is designed for the underwater storage of spent-fuel assemblies
 
and control rods after their removal from the reactor. Each unit is designed to accommodate a
 
total of approximately 9 cores. The spent-fuel cask is accommodated in a separate pool adjoining the spent-fuel pool. 
 
The spent-fuel storage racks are located to provide sufficient shielding water over stored fuel
 
assemblies to limit radiation at the surface of the water to no more than 2.5 mrem/h during the
 
storage period. The exposure time during refueling will be limited so that the integrated dose to
 
operating personnel does not exceed the limits of 10 CFR 20. 
 
The waste disposal system is designed to permit controlled handling and disposal of liquid, gaseous, and solid wastes generated during plant operation. The principal design criterion is to
 
ensure that plant personnel are protected against exposure to radiation from wastes in
 
accordance with limits defined in 10 CFR 20. During plant operations members of the public will
 
be protected in accordance with limits defined in technical specifications. 
 
The spent-fuel pool is located within the auxiliary building. The liquid waste processing
 
equipment and the gaseous waste storage and disposal equipment are located within a
 
separate area of the auxiliary building. Both of these areas provide confinement capability in
 
the event of an accidental release of radioactive materials, and both are ventilated with
 
discharges to the vent stack which is monitored. In the event of a fuel handling accident, the
 
ventilation exhaust line will be automatically isol ated and the fuel handling area ventilation fans will be automatically secured. The fuel handling area may then be remotely connected with the
 
penetration room filtration system so that the air is processed by the particulate, absolute, and
 
charcoal filters prior to being released through the vent stack. 
 
Radioactive liquid discharged into the cooling tower blowdown is monitored prior to discharge. 
 
Any accidental leakage from liquid waste processing equipment is collected and transferred to
 
other tanks to prevent uncontrolled releases to the environment. 
 
The spent-fuel pool cooling system removes the residual heat from the spent-fuel pool. The
 
system is required to handle the heat load from typical core discharges from the reactor as described in table 9.1-1, but it can safely accommodate the heat load from the spent-fuel pool
 
while completely loaded with spent-fuel assemblies. 
 
FNP-FSAR-3
 
3.1-37 REV 21  5/08 The spent-fuel pool cooling system design in corporates redundant heat exchangers and pumps, each with 100-percent capability. Normally, one pump draws water from the pool, circulates it
 
through a heat exchanger, and returns it to the pool. Component cooling water cools these heat
 
exchangers. 
 
The spent-fuel storage pool is constructed of reinforced concrete and lined with stainless steel
 
plate. The physical location of the suction and return lines to the storage pool is such that
 
inadvertent loss of coolant from the pool is prevented. 
 
For a more complete description of the design of the fuel storage and handling and radioactive
 
waste systems, see section 9.1 and chapter 11.0. 
 
3.1.53  CRITERION 62 - PREVENTION OF CRITICALITY IN FUEL STORAGE AND HANDLING Criticality in the fuel storage and handling sy stem is prevented by physical systems or processes, preferably by use of geometrically safe configurations. 
 
CONFORMANCE 
 
Criticality in new and spent-fuel storage areas is prevented by physical separation and
 
administrative controls on placement of fuel assemblies. In addition, the presence of borated
 
water controls the criticality in the spent-fuel storage pool. The separation and administrative
 
controls on placement are provided so that criticality is precluded even if pure water replaces
 
the air gap in the new fuel storage or the borated water in the spent-fuel storage area. Criticality
 
prevention is discussed in paragraph 4.3.2.7. 
 
3.1.54  CRITERION 63 - MONITORING FUEL AND WASTE STORAGE Appropriate systems are provided in fuel st orage and radioactive waste systems and associated handling areas to detect conditions that may result in loss of residual heat removal capability
 
and excessive radiation levels and to initiate appropriate safety actions. 
 
CONFORMANCE 
 
Monitoring systems are provided to alarm on exce ssive temperature or low water level in the spent-fuel pool. In the event of a high temperat ure alarm, administrative procedures will provide for checking the cooling water flow to the fuel pool coolers, the operating status of the fuel pool
 
cooling pumps, and the integrity of the fuel pool cooling and purification system. In the event of
 
high airborne gaseous or particulate activity, automatic diversion of the normal fuel building or
 
waste disposal building ventilation system ex haust to the supplementary leak collection and release system will occur.
 
A radiation monitor and an alarm are provided, as required, to warn personnel of an increase in
 
the level of radiation or airborne activity. The radiation monitoring system is described in
 
chapters 11.0 and 12.0. 
 
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3.1-38 REV 21  5/08 3.1.55  CRITERION 64 - MONITORING RADIOACTIVITY RELEASES Means are provided for monitoring the reactor containment atmosphere, spaces containing
 
components for recirculation of LOCA fluids, effluent discharge paths, and the plant environs for
 
radioactivity that may be released from normal operations, including anticipated operational
 
occurrences, and from postulated accidents. 
 
CONFORMANCE 
 
The containment atmosphere is continuously monitored during normal operation by the
 
containment particulate and gas radiation monitor located outside the containment and by the
 
area radiation monitors located within. In the event of accident conditions, samples of the
 
containment atmosphere will be obtained from the containment post accident sample system for laboratory analysis of particulates, iodine, and noble gases that may be present. During normal
 
operation ventilation systems within areas contiguous to the containment structure are
 
monitored by particulate and/or gas radiation moni tors. In addition, the service water outlet from each containment cooler is monitored to ensure that any leakage of radioactive fluids into the
 
service water system will be detected. Radioacti vity levels contained in the effluent discharge paths and in the environs are continuously monitored during normal and accident conditions by
 
the plant radiation monitoring system and by the health physics program. Indications and alarms from the plant radiation monitoring system instruments are provided in the control room.
The radiation monitoring system is described in chapters 11.0 and 12.0. 
 
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3.2-1 REV 21  5/08 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS AND SYSTEMS 3.2.1 SEISMIC CLASSIFICATION A two-level system is used for the Seismic Classification of the structures, components, and
 
systems of the facility. 
: 1. Category I structures, components, and systems. 
: 2. Category II structures, components, and systems. 
 
3.2.1.1  Definitions Structures, components, and systems required for safe shutdown, for immediate or long term core cooling, or for radioactive material confinement following a loss-of-coolant accident (LOCA)
 
to ensure that the public is protected in accordance with 10 CFR 100 guidelines are designed
 
Category I. 
 
Category I structures, components, and systems are designed to withstand the effects of the
 
safe shutdown earthquake (SSE) and 1/2 safe shutdown earthquake (1/2 SSE) as discussed in
 
section 3.7. 
 
When a system as a whole is referred to as Category I, portions not associated with loss of
 
function of the system may be designated as Category II. 
 
Category II structures, components, and systems are those whose failure would not result in the
 
release of significant radioactive material and would not prevent reactor shutdown. All
 
equipment not specifically listed as Category I is included as Category II. 
 
The failure of Category II structures, components, and systems may interrupt power generation. 
 
All Category II structures are designed to conform to Section 2.3.1.4 of the 1970 edition of the
 
Uniform Building Code. 
 
Seismic Classification of structures, systems, and components is in accordance with Regulatory
 
Guide 1.29. 
 
3.2.1.2  Category I Structures
: 1. Containment. 
: 2. Auxiliary building, including all fuel handling  equipment storage areas. 
: 3. Diesel generator building. 
 
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3.2-2 REV 21  5/08  4. River intake structure.(a)   
: 5. Intake structure at storage pond. 
: 6. Storage pond dam and dike. 
: 7. Vent stack.(a) 
: 8. Pond spillway structure. 
: 9. Electrical cable tunnel structure. 
: 10. Category I outdoor tanks. 
: 11. Trisodium phosphate baskets in containment.
 
3.2.1.2.1 Category I Structures with Seismic Restraint Exclusionary Zones Auxiliary building: Zones or areas, defined on drawings D506531 and D356597, do not have any safety-related systems or com ponents. Within these zones push carts, hand tools, and other such devices may be placed without seismically restraining such devices. No safety-related systems or
 
components may be mounted in or run through these areas.
 
3.2.1.3  Category I Mechanical Components and Systems Refer to table 3.2-1 for Category I seismic mechanical components and systems. 
 
3.2.1.4  Category I Electrical Equipment
: 1. 4160-V switchgear (engineered safeguard buses). 
: 2. 4160-V to 600-V transformers (associated with engineered safeguard systems).
: 3. 600-V load centers (engineered safeguard buses). 
: 4. 600-V and 208-V motor-control centers (associated with engineered safeguard systems). 
 
_____________
: a. Not required for safe shutdown of the plant. The original design (Category I) requirements
 
for the river intake structure are no longer required.
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3.2-3 REV 21  5/08  5. Direct-current electrical distribution system (auxiliary building and service water building): 
: a. 125-V dc station batteries. 
: b. Inverters, 125-V dc to 120-V ac (vital ac instrumentation distribution panels).
: c. 125-V dc distribution panels. 
: d. 125-v dc switchgear. 
: e. 125-v dc battery chargers. 
: 6. Vital ac instrumentation and regulated ac distribution panels. 
: 7. Control panels and control boards: 
: a. Auxiliary relay racks. 
: b. Solid-state protection system cabinets. 
: c. Nuclear instrumentation system cabinets. 
: d. Process protection and control system cabinets. 
: e. Emergency power board. 
: 8. Cable tray and conduit supports (associated with engineered safeguard systems). 
: 9. Containment penetration assemblies. 
: 10. Direct-current emergency lighting (except for an 8-hour rated DC battery pack, emergency lighting is installed per Appendix R to 10 CFR 50
 
referenced in FSAR paragraph 9B.4.1.19).
: 11. Diesel generators. 
: 12. Diesel generator control panels. 
: 13. Diesel generator sequencers. 
: 14. Boric acid heat-tracing equipment (functionally nonsafety related).
: 15. Turbine-driven auxiliary feedwater pump uninterruptable power supply.
 
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3.2-4 REV 21  5/08 3.2.1.5  Category I Instrumentation and Control Systems Equipment
: 1. Penetration room filtration system. 
: 2. Radiation monitors for containment purge exhaust lines.
: 3. Radiation monitors for fuel handling area ventilation exhaust line. 
: 4. Post-accident containment combustible gas control system.
: 5. Component cooling water system. 
: 6. Service water system. 
: 7. Auxiliary feedwater system. 
: 8. Power supply inverters for balance of plant instrument panels. 
: 9. Balance of plant instrument panels. 
: 10. Portions of the sampling system which provide containment isolation and which interface with other Category I systems.  (See drawings D-175009, sheet 1, D-175009, sheet 2, D-175009, sheet 3, D-205009, sheet 1, D-205009, sheet 2 and D-205009, sheet 3.)
: 11. Diesel generator control equipment. 
 
3.2.1.6  Structures and Systems of Mixed Category None of the structures in the Farley Nuclear Plant have classifications that are partially Category
 
I and partially Category II. The boundaries of the nuclear classes of piping systems are shown
 
on the Piping and Instrumentation Diagrams (P&ID) in sections as listed in table 3.2-3. 
 
Nuclear Safety Classes 1, 2a, 2b, and 3 are des igned as Seismic Category I systems. Systems that are Nuclear Safety Class NNS, and all piping systems not otherwise indicated as
 
Category I, are non-seismic. The P&ID legend is shown on drawings D-175016, sheet 1, D-175016, sheet 2, D-175016, sheet 3 and figure 1.7-1. 
 
3.2.2 SYSTEM QUALITY GROUP CLASSIFICATION The design criteria are tabulated in table 3.2-1 for all mechanical system components. The
 
design in general complies with the intent of Regulatory Guide 1.26. The actual design
 
standards, however, conform to the standards of the American Nuclear Society, "Nuclear Safety
 
Criteria for the Design of Stationary Pressurized Water Reactor Plants," August 1970 draft. 
 
Regulatory Guide 1.26 was not available at the time of initial equipment design and purchase. 
 
Whenever practicable, equipment has been purchased to meet ASME Section III standards.
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3.2-5 REV 21  5/08 When equipment was purchased before ASME Section III became effective, other design codes, as indicated in table 3.2-1, were used. 
 
The relationship between Safety Class and the ASME Section III Nuclear Class is indicated
 
below.
ANS SAFETY CLASS ASME SECTION III    NUCLEAR CLASS 1 1 2a 2 2b 3 3 3 NNS (Non-nuclear safety) -
 
The system quality group classifications are delineated on the piping and instrumentation
 
diagrams in chapters 5.0, 6.0, 9.0, 10.0, and 11.0. 
 
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TABLE 3.2-1 (SHEET 1 OF 22)
 
==SUMMARY==
OF CRITERIA - MECHANICAL SYSTEM COMPONENTS Component Design Responsibility (1) ANS Safety Class  (2) Code (3) Location (4) Rad Source (5) Rad Seismic (6) Tornado  (7)
REV 21  5/08 REACTOR COOLANT SYSTEM Reactor vessel W 1 III A C S X X
 
Full length CRDM housing W 1 III A C S X X
 
Reactor coolant pump assembly (8) W 1  C S X X
 
Reactor coolant pump casing W 1 P&V I C S X X
 
Reactor coolant pump internals W 1 P&V I C S X X
 
Steam generator (tube side) W 1 III A C S X X (shell side including integral steam flow restrictor) W 2a III A C S X X Pressurizer W 1 III A C S X X
 
Reactor coolant piping to pressure boundary W 1 III 1 C S X X
 
RC system supports W - - C N X X
 
Surge pipe and fittings W 1 III 1 C S X X
 
RC thermowells W 1 III 1 C S X X
 
Safety valves (16) W 1 III A C S X X
 
Relief valves W 1 III A C S X X
 
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TABLE 3.2-1 (SHEET 2 OF 22)
Component Design Responsibility (1) ANS Safety Class  (2) Code (3) Location (4) Rad Source (5) Rad Seismic (6) Tornado  (7)
REV 21  5/08 Valves to RC system boundary W 1 P&V I C S X X Pressurizer relief tank (11) W NSS VIII C S - X
 
CRDM head adapter plugs W 1  B31.7 I C S X X
 
CHEMICAL & VOLUME CONTROL SYSTEM Regenerative HX W 2a III C C S X X
 
Letdown HX  (tube side) W 2a III C AB S X X (shell side)  2b VIII  P X X
 
Mixed bed demineralizer (11) W 3 VIII AB S X X
 
Cation bed demineralizer (11) W 3 VIII AB S X X
 
Reactor coolant filter W 2a III C AB S X X
 
Volume control tank W 2a III C AB S X X
 
Charging/high head safety injection pump (8) W 2a P&V II(30) AB S X X
 
Seal water injection filter W 2a III C AB S X X
 
Letdown orifices W 2a III 2 C S - X
 
Excess letdown HX (tube side) W 2a III C C S X X (shell side)  2b VIII  P X X
 
Seal water return filter W 2a III C AB S X X
 
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TABLE 3.2-1 (SHEET 3 OF 22)
Component Design Responsibility (1) ANS Safety Class  (2) Code (3) Location (4) Rad Source (5) Rad Seismic (6) Tornado  (7)
REV 21  5/08 Seal water HX  (tube side) W 2a III C AB P X X    (shell side)  2b VIII  X X
 
Boric acid tanks (19) A 2b API 650 AB P X X
 
Boric acid filter (11) W 2b III C AB P X X
 
Boric acid transfer pump W 2b P&V III AB P X X
 
Boric acid blender W 2b III 3 AB P X X
 
Resin fill tank (12) (13) A NNS VIII AB N - X
 
Boric acid batching tank (13) W NNS VIII AB N - X
 
Chemical mixing tank(12) (13) W NNS VIII AB N - X
 
Chemical mixing tank orifice A NNS - AB N - X
 
RCP No. 1 seal bypass orifice W 1 III 1 C S - X
 
Reactor makeup water storage tank A 2b III 3 O - X X(23)
 
Reactor makeup water pump A 2b III 3 O - X  X(23)(26)
 
Demineralized water storage tank A NNS - O - - -
 
EMERGENCY CORE COOLING SYSTEM Accumulators  W 2a  III C  C  P X  X
 
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TABLE 3.2-1 (SHEET 4 OF 22)
Component Design Responsibility (1) ANS Safety Class  (2) Code (3) Location (4) Rad Source (5) Rad Seismic (6) Tornado  (7)
REV 21  5/08 RESIDUAL HEAT REMOVAL SYSTEM Residual heat removal/low head safety injection pump (8) W 2a P&V II  AB S X X
 
Residual heat exchanger (tube side) W 2a III C  AB S X X (shell side)  2b VIII  AB P X X
 
CONTAINMENT SPRAY SYSTEM Containment spray pump W 2a P&V II  AB P X X
 
Eductor W 2a III 2  AB P X(a) X
 
CONTAINMENT ISOLATION SYSTEM Valves A 2a P&V II C,AB S,P X X
 
CONTAINMENT COOLING SYSTEM Fans A 2b AMCA (14) C N X X
 
Heat exchanger A 2b VIII C N X X
 
COMPONENT COOLING SYSTEM Pumps A 2b P&V III  AB P X X
 
Unit 1 Heat exchangers (tube side) A 2b VIII  AB N X X (shell side) A 2b VIII  AB P X X
 
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TABLE 3.2-1 (SHEET 5 OF 22)
Component Design Responsibility (1) ANS Safety Class  (2) Code (3) Location (4) Rad Source (5) Rad Seismic (6) Tornado  (7)
REV 21  5/08 Unit 2 Heat exchangers (tube side) A 2b III  AB N X X    (shell side) A 2b III  AB P X X
 
Surge tank (21) A 2b API 620  AB P X X
 
SPENT FUEL POOL COOLING SYSTEM Spent fuel pool heat exchanger (tube side) (8) W 2b III C  AB S X X (shell side) W 2b VIII  AB P X X
 
RESIDUAL HEAT REMOVAL SYSTEM Residual heat removal/low head safety injection pump (8) W 2a P&V II  AB S X X
 
Residual heat exchanger (tube side) W 2a III C  AB S X X (shell side)  2b VIII  AB P X X
 
CONTAINMENT SPRAY SYSTEM Containment spray pump W 2a P&V II  AB P X X
 
Eductor W 2a III 2  AB P X(a) X CONTAINMENT ISOLATION SYSTEM Valves A 2a P&V II C,AB S,P X X
_____________________ a. The components are included as part of the respective piping system model and seismic analysis.
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TABLE 3.2-1 (SHEET 6 OF 22)
Component Design Responsibility (1) ANS Safety Class  (2) Code (3) Location (4) Rad Source (5) Rad Seismic (6) Tornado  (7)
REV 21  5/08 CONTAINMENT COOLING SYSTEM Fans A 2b AMCA (14) C N X X
 
Heat exchanger A 2b VIII C N X X
 
COMPONENT COOLING SYSTEM Pumps A 2b P&V III  AB P X X
 
Unit 1 Heat exchangers (tube side) A 2b VIII  AB N X X (shell side) A 2b VIII  AB P X X
 
Unit 2 Heat exchangers (tube side) A 2b III  AB N X X (shell side) A 2b III  AB P X X
 
Surge tank (21) A 2b API 620  AB P X X
 
SPENT FUEL POOL COOLING SYSTEM Spent fuel pool heat exchanger (tube side) (8) W 2b III C  AB S X X (shell side) W 2b VIII  AB P X X
 
Spent fuel pool pump W 2b P&V III AB S X X
 
Spent fuel pool strainers W NNS  AB S - X
 
Skimmer pump W NNS  AB P - X
 
Spent fuel pool filter (10)(11) W NNS III C AB S - X
 
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TABLE 3.2-1 (SHEET 7 OF 22)
Component Design Responsibility (1) ANS Safety Class  (2) Code (3) Location (4) Rad Source (5) Rad Seismic (6) Tornado  (7)
REV 21  5/08 Spent fuel pool demineralizer (10)(11) W NNS III C AB S - X BORON THERMAL REGENERATION SUBSYSTEM Moderating HX  (tube side) W 3 VIII AB S X X (shell side)  3 VIII AB S X X
 
Letdown chiller HX  (tube side) W 3 VIII AB S X X (shell side) (10)  NNS VIII AB P X X
 
Letdown reheat HX  (tube side) W 2a III C AB S X X (shell side)  3 VIII AB S X X
 
Thermal regeneration demineralizer W 3 III 3 AB S X X
 
Chiller (8) W NNS VIII AB N - X
 
Chiller surge tank (10) W NNS VIII AB N - X
 
Chiller pumps W NNS P&V III AB N - X
 
LIQUID RECYCLE AND WASTE SUBSYSTEM Recycle holdup tank (19) A NNS API 650 AB S - X
 
Recycle evap. feed pump W NNS MS AB S - X
 
Recycle evap. feed demineralizer W NNS VIII AB S - X
 
Recycle evap. feed filter W NNS VIII AB S - X
 
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TABLE 3.2-1 (SHEET 8 OF 22)
Component Design Responsibility (1) ANS Safety Class  (2) Code (3) Location (4) Rad Source (5) Rad Seismic (6) Tornado  (7)
REV 21  5/08 Recycle evaporator W NNS VIII AB S -  X Recycle evap. condensate demineralizer (10) W NNS VIII AB P - X
 
Recycle evap. condensate filter (10) W NNS VIII AB P - X
 
Recycle evap. concentrate filter (10) W NNS VIII AB S - X
 
Recycle evap. reagent tank (13) W NNS VIII AB  - X
 
R.C. drain tank W NNS VIII C S - X
 
R.C. drain tank pump W NNS MS C S - X    (N2G21P001A-N)        API-610        (N2G21P001B-N)
R.C. drain tank HX (tube side) (10) W NNS VIII C S - X (shell side)  2b III C C P X X
 
Waste holdup tank (12)(13) W NNS VIII AB  S - X
 
Waste evap. feed pump W NNS MS AB S - X
 
Waste evap. reagent tank W NNS VIII AB  - X
 
Waste evap. feed filter W NNS VIII AB S - X
 
Waste evaporator W NNS VIII AB S -  X
 
Waste evap. condensate demin. (10) W NNS VIII AB P - X FNP-FSAR-3
 
TABLE 3.2-1 (SHEET 9 OF 22)
Component Design Responsibility (1) ANS Safety Class  (2) Code (3) Location (4) Rad Source (5) Rad Seismic (6) Tornado  (7)
REV 21  5/08 Waste evap. condensate filter (10) W NNS VIII AB P - X Waste evap. condensate tank (10)(12) W NNS VIII AB P - X
 
Waste evap. condensate tank pump W NNS MS AB P - X Chemical drain tank (12)(13) W NNS VIII AB S - X
 
Chemical drain tank pump W NNS MS AB S - X
 
Spent resin storage tank W NNS VIII AB S -  X
 
Spent resin sluice pump W NNS MS AB S - X
 
Spent resin sluice filter W NNS VIII AB S - X
 
Laundry and hot shower tank (10)(12)(13) W NNS VIII AB P - X
 
Laundry and hot shower tank pump W NNS MS AB P - X
 
Laundry and hot shower strainer (10) W NNS  AB P - X
 
Laundry and hot shower filter (10) W NNS VIII AB P - X
 
Floor drain tank (10)(12)(13) W NNS VIII AB P - X
 
Floor drain tank pump W NNS MS AB P - X
 
Waste monitor tank (10)(12)(13) W NNS VIII AB P - X
 
Waste monitor tank pump W NNS MS AB P - X
 
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TABLE 3.2-1 (SHEET 10 OF 22)
Component Design Responsibility (1) ANS Safety Class  (2) Code (3) Location (4) Rad Source (5) Rad Seismic (6) Tornado  (7)
REV 21  5/08 Waste monitor tank demineralizer (10) W NNS VIII AB P - X Waste monitor tank filter (10) W NNS VIII AB P - X
 
Containment sump pump A NNS MS C P - X
 
ES room sump pump A NNS MS AB P - X
 
Drumming header strainer (10) W NNS  AB S - X
 
Floor drain tank filter (10) W NNS VIII AB P - X
 
Floor drain tank strainer (10) W NNS  AB P - X
 
Disposable demineralizers A NNS MS AB S - X
 
Disposable demineralizer pumps A NNS MS AB S - X
 
GAS HANDLING SUBSYSTEM Gas compressor W NNS VIII/MS AB S - X
 
Gas decay tanks W NNS VIII AB S X(D) X
 
Hydrogen recombiner W NNS VIII AB S - X
 
EMERGENCY DIESEL FUEL OIL SYSTEM Transfer pumps A 2b P&V III DB N X X
 
Fuel oil tanks A 2b API 620 B N X X (29)
 
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TABLE 3.2-1 (SHEET 11 OF 22)
Component Design Responsibility (1) ANS Safety Class  (2) Code (3) Location (4) Rad Source (5) Rad Seismic (6) Tornado  (7)
REV 21  5/08 SERVICE WATER SYSTEM Pumps A 2b P&V III S N X X
 
Strainers A 2b VIII S N X X
 
Recirc pipe to wetpit A 2b VIII O N X (26)
 
RIVER WATER SYSTEM Pumps A NNS P&V III R N - -
 
FUEL HANDLING SYSTEM Fuel manipulator crane W 3 - AB N X X
 
Fuel transfer tube (17) W 2a - C/AB N X X
 
Underwater fuel conveyor car and rail system (18) W 3 - AB N X X
 
Fuel pool bridge crane W 3 - AB N X X
 
Polar crane A NNS (9) C N X X
 
Crane supports A NNS - C N X X
 
SAMPLING SYSTEM Sampler cooler A NNS (28) AB S - X
 
Sampler vessel A NNS VIII AB S - X
 
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TABLE 3.2-1 (SHEET 12 OF 22)
Component Design Responsibility (1) ANS Safety Class  (2) Code (3) Location (4) Rad Source (5) Rad Seismic (6) Tornado  (7)
REV 21  5/08 Delay coil A 2a  B31.7 II C S X X REFUELING WATER SYSTEM Pump A NNS - AB P - X
 
Storage tank (20) A 2a III 2 O P X X(23)(29)
 
FIRE PROTECTION SYSTEM Fire pumps A NNS (15) O N - -
 
CONTAINMENT PURGE SYSTEM Fans A NNS - AB N - X
 
Filters A NNS - AB P - X
 
REACTOR VESSEL SUPPORT COOLING SYSTEM Fans A NNS - C N - X
 
CONTROL ROD DRIVE MECHANISM COOLING SYSTEM Fans A NNS - C N - X
 
AUXILIARY BUILDING VENTILATION SYSTEM Fan/coil units A NNS - AB N - X
 
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TABLE 3.2-1 (SHEET 13 OF 22)
Component Design Responsibility (1) ANS Safety Class  (2) Code (3) Location (4) Rad Source (5) Rad Seismic (6) Tornado  (7)
REV 21  5/08 Filters A NNS - AB P - X Pump room air cooling units A 2b - AB P X X Battery room exhaust fans A NNS AMCA AB N X X
 
Battery charger room air cooling units A 2b III-3, AMCA  AB N X X
 
Motor control center and 600 V load center air cooling units A 2b III-3, AMCA  AB N X X
 
600-V load center cooling system fire damper A NNS UL AB N X X Battery room motor operated dampers A NNS AMCA AB N X X
 
PENETRATION ROOM FILTRATION SYSTEM Fans A 2b AMCA(14) AB N X X
 
Filters (HEPA and charcoal) A 2b ORNL-NSIC AB P X X
 
Backdraft dampers A NNS AMCA AB N X X
 
Spent Fuel Pool Area Duct A 2b AMCA O P X (26)
 
Suction dampers backup air supply piping A 3 B31.1 (25) AB N X X Suction dampers air supply accumulators Pressure relief valves A 3 VIII (25) AB N X X
 
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TABLE 3.2-1 (SHEET 14 OF 22)
Component Design Responsibility (1) ANS Safety Class  (2) Code (3) Location (4) Rad Source (5) Rad Seismic (6) Tornado  (7)
REV 21  5/08 CONTROL ROOM VENTILATION SYSTEM Fans A 2b AMCA(14) AB N X X
 
Filters A 2b ORNL-NSIC AB P X X
 
Air conditioning unit A 2b AMCA, III-3 AB N X X
 
Motor-operated dampers/valves A 2b AMCA, III-3 AB N X X
 
Balancing dampers A NNS AMCA AB N X X
 
Fire dampers A NNS UL AB N X X
 
DIESEL BUILDING VENTILATION SYSTEM Fans A 2b AMCA(14) DB N X X
 
Filters A 2b -- DB N X X
 
MAIN STEAM SYSTEM Isolation valves A 2a P&V II AB N X X
 
FEEDWATER SYSTEM Isolation valves A 2a P&V II AB N X X
 
AUXILIARY FEEDWATER SYSTEM Auxiliary feedwater pumps FNP-FSAR-3
 
TABLE 3.2-1 (SHEET 15 OF 22)
Component Design Responsibility (1) ANS Safety Class  (2) Code (3) Location (4) Rad Source (5) Rad Seismic (6) Tornado  (7)
REV 21  5/08 Motor driven A 2b P&V III AB N X X Steam turbine driven A 2b P&V III AB N X (26)
 
Condensate storage tank A 2b III 3 O P X (24)(29)
 
STEAM DUMP SYSTEMS Turbine bypass A NNS - TB N - -
 
Relief valves A 2a P&V II AB N X  X(29)
 
Safety valves (16) A 2a P&V II AB N X  X(29)
 
STEAM GENERATOR BLOWDOWN TREATMENT SYSTEM Blowdown surge tank W NNS VIII AB P - X
 
Blowdown inlet filters W NNS VIII AB P - X
 
Blowdown outlet filter W NNS VIII AB P - X
 
Blowdown discharge-recycle pumps W NNS VIII AB P - X
 
Blowdown heat exchangers W NNS VIII AB P - X
 
Blowdown cation demineralizers W NNS VIII AB P - X
 
Blowdown mixed bed demineralizers W NNS VIII AB P - X
 
Spent resin storage tank W NNS VIII AB P - X
 
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TABLE 3.2-1 (SHEET 16 OF 22)
Component Design Responsibility (1) ANS Safety Class  (2) Code (3) Location (4) Rad Source (5) Rad Seismic (6) Tornado  (7)
REV 21  5/08 Spent resin sluice pump W NNS VIII AB P - X Spent resin sluice filter W NNS VIII AB P - X
 
CONDENSER CIRCULATING WATER SYSTEM Circulating water pumps A NNS - O N - -
 
COMPRESSED AIR SYSTEM(22)
Compressors A NNS - TB N - -
 
After coolers A NNS - TB N - -
 
Air tanks A NNS VIII TB N - -
 
Air dryers A NNS - TB N - -
 
Piping to safety grade component safety boundary A NNS B31.1 TB/C/AB/O N - -        Air filter A NNS - TB N - -
 
Piping to within safety grade component safety boundary W/A  (SAME AS COMPONENT - SEE APPROPRIATE COMPONENT DESCRIPTION)        HYDROGEN SYSTEM Hydrogen vessels A NNS VIII O N - -
 
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TABLE 3.2-1 (SHEET 17 OF 22)
Component Design Responsibility (1) ANS Safety Class  (2) Code (3) Location (4) Rad Source (5) Rad Seismic (6) Tornado  (7)
REV 21  5/08 NITROGEN SYSTEM Nitrogen vessels A NNS VIII O N - -
 
POST-LOCA HYDROGEN CONTROL SYSTEM Post-LOCA hydrogen recombiners W 2b - C N X X
 
Containment post-LOCA hydrogen mixing system W 2b AMCA(14)  C N X X        Post-LOCA, containment hydrogen
 
monitoring W 2b III-2 C P X X  equipment MISCELLANEOUS COMPONENTS Diesel generators A 3 - DB N X X(29)
 
Spent fuel pool A NNS - AB S X X
 
Vent stack A NNS - O N X X (D)
Spent fuel pool H & V system isolation
 
dampers A NNS AMCA AB P X X        Containment venting filter units A 2b III-3, ORNL AB P X X
 
FNP-FSAR-3
 
TABLE 3.2-1 (SHEET 18 OF 22)
 
REV 21  5/08  NOTES    (1) A Alabama Power Company W Westinghouse (2) 1 Safety Class 1 (ANS) 2a Safety Class 2a (ANS)
 
2b Safety Class 2b (ANS)
 
3 Safety Class 3 (ANS)
 
NNS Non Nuclear Safety (ANS)
  (3) III A ASME Boiler and Pressure Vessel Code - Section III, Class A III C ASME Boiler and Pressure Vessel Code - Section III, Class C
 
VIII ASME Boiler and Pressure Vessel Code - Section VIII
 
P&V I ASME Code for Pumps and Valves for Nuclear Power, Class I
 
P&V II ASME Code for Pumps and Valves for Nuclear Power, Class II
 
P&V III ASME Code for Pumps and Valves for Nuclear Power, Class III
 
III 1 ASME Boiler and Pressure Vessel Code - Section III, Class 1
 
III 2 ASME Boiler and Pressure Vessel Code - Section III, Class 2
 
III 3 ASME Boiler and Pressure Vessel Code - Section III, Class 3
 
B31.1 ANSI B31.1 - Power Piping 
 
B31.7 I ANSI B31.7 - Nuclear Power Piping, Class I 
 
B31.7 II ANSI B31.7 - Nuclear Power Piping, Class II 
 
B31.7 III ANSI B31.7 - Nuclear Power Piping, Class III 
 
D100 American Waterworks Association, Standard for Steel Tanks, Standpipes, Reservoirs, and Elevated Tanks for Water Storage, AWWA, D100 FNP-FSAR-3
 
TABLE 3.2-1 (SHEET 19 OF 22)
 
REV 21  5/08  API 610 American Petroleum Institute, Centrifugal Pumps for General Refinery Services   
 
API 620 American Petroleum Institute Recommended Rules for Design and Construction of Large Welded Low Pressure Storage Tanks API 650 American Petroleum Institute, Welded Steel Tanks for Oil
 
AMCA Air Moving and Conditioning Association 
 
MS Manufacturer's Standard 
 
ORNL-NSICOak Ridge National Laboratory, Nuclear Information Center - Design, Construction, and Testing of High Efficiency Air Filtration Systems for Nuclear
 
Application.
UL Underwriters Laboratory (4) C Containment AB Auxiliary Building 
 
TB Turbine Building 
 
B Buried in Ground 
 
DB Diesel Generator Building 
 
R River Water Intake Structure 
 
S Service Water Intake Structure 
 
O Outside (5) S Source of radiation N No source of radiation 
 
P Possible source of radiation (6) X Category I, (Methods used for seismic analysis of Category I systems and components are presented in table 3.7-4).
FNP-FSAR-3
 
TABLE 3.2-1 (SHEET 20 OF 22)
 
REV 21  5/08 X(D) Designed and constructed to the seismic requirements given in Regulatory Guide 1.143, Revision 1 with the exception of the seismic design criteria given in
 
Regulatory Position C.5. The components, systems, and structures are
 
designed to the seismic design criteria given in FSAR section 3.7.   
- Category II (7) X Protected by virtue of location in a structure designed for tornado wind.
X(D) Designed for tornado wind loads.
 
- No protection required (8)  Portions of equipment containing component cooling water will be analyzed for seismic requirements.      (9)  Crane Manufacturers Association of America Specification No. 70 of 1971.      (10)  National Fire Underwriters and Underwriters Laboratory Certification.      (11)  Designed and fabricated to ASME III C, radiographed and so stamped; however, compliance with ASME VIII would be sufficient for this use.      (12)  Outside jurisdiction of ASME, but designed, fabricated, examined, and tested according to ASME Boiler and Pressure Vessel, Section VIII.      (13)  Built to code but not tested    (14)  Performance test required    (15)  National Fire Protection Association Standard No. 20.    (16)  Will meet pressure-relieving requirements of ASME Section III, Article 9.    (17)  That part which is part of the containment pressure boundary.    (18)  Protect against seismic overturning and possible impaling of fuel.    (19)  Quality control requirements include sidewall and nozzles to tank welds examined by magnetic-particle or liquid-penetrant methods; roof, roof-to-
 
sidewall, and bottom welds visually examined; bottom and bottom-to-sidewall
 
welds vacuum box tested FNP-FSAR-3
 
TABLE 3.2-1 (SHEET 21 OF 22)
 
REV 21  5/08 (20)  Quality control requirements include 100-percent radiograph of sidewall welds; roof, roof-to-sidewall, and bottom welds visually examined;
 
bottom and bottom-to-sidewall welds vacuum box tested; bottom-to-
 
sidewall and nozzles-to-tank welds examined by magnetic-particle or liquid penetrant methods    (21)  Quality control requirements include sidewall welds 3/16 in. or under examined by magnetic-particle or liquid-penetrant methods; 100-percent
 
radiograph of sidewall welds over 3/16 in.; roof, roof-to-sidewall, bottom, bottom-to-sidewall, and nozzles-to-tank welds examined by magnetic-
 
particle or liquid-penetrant methods; roof and roof-to-sidewall welds
 
soap tested; bottom and bottom-to-sidewall welds vacuum box tested.      (22)  The compressed air system includes the instrument air system.      (23)  This equipment is surrounded by a concrete wall to protect it from tornado missiles.    (24)  In order to ensure the 150,000-gallon reserve required by Technical Specifications, the lower12 ft of the tanks are designed to withstand ruptures caused by missiles generated by tornadoes. Certain
 
connections to the Unit 1 and Unit 2 CSTs, within the lower twelve feet
 
of the tank, are protected by structures. Reference paragraph 9.2.6.3
 
for a discussion of these connections.    (25)  Deleted    (26)  This component or portions of this system are not required to protected from tornado generated missiles per paragraph 3.5.1.2.2.1.    (27)  Shall meet 10 CFR 50 Appendix B Requirements    (28)  Sample coolers for the RCS sample stream were originally procured in accordance with ASME Code, Section III; however, compliance with
 
ANSI B31.1 is sufficient for this use. The remaining sample coolers are
 
designed in accordance with ANSI B31.1.    (29)  This component or portions of this system have been analyzed for vulnerability to tornado generated missiles and found to have an
 
acceptable probability of survival per paragraph 3.5.1.2.2.2.
FNP-FSAR-3
 
TABLE 3.2-1 (SHEET 22 OF 22)
 
REV 21  5/08 (30)  The 2A charging pump (Q2E21P002A) and 2B charging pump (Q2E21P002B)casing and discharge head have been replaced. The replacement casing and discharge head were designed to meet the requirements of ASME III 1971 edition with summer 1972 addenda.      GENERAL NOTES: 
: 1. The safety-related systems outside the reactor coolant pressure boundary may have more than one quality class of piping and valves. Individual valves and sections of
 
piping are assigned quality classes and codes appropriate to their locations and
 
functions and consistent with the assignments of quality classes of system components
 
in table 3.2-1. 
: 2. All pressure-retaining cast parts of Safety Class 1a and 2a pumps and valves are radiographed (or ultrasonically tested to equivalent standards). Where size or
 
configuration does not permit effective volumetric examination, magnetic-particle or
 
liquid-penetrant examination is substituted. Examination procedures and acceptance
 
standards are at least equivalent to those specified in the applicable class in the code. 
: 3. The reactor coolant system code requirements, including the applicable addenda, are presented in table 3.2-4. 
 
FNP-FSAR-3
 
TABLE 3.2-2
 
==SUMMARY==
OF QUALITY CLASS REQUIREMENTS -
MECHANICAL SYSTEM COMPONENTS
 
REV 21  5/08 Equipment Category/ Analysis Limits    1  SSE+NORMAL + LOCA NLSF, permanent deformation permitted (faulted condition)
SSE+NORMAL 
 
1/2SSE+NORMAL Applicable code stresses (upset condition)    2a and 2b  SSE+NORMAL NLSF, permanent deformation permitted(faulted condition) 1/2SSE+NORMAL Applicable code stresses (upset condition)    3  SSE+NORMAL NLSF, permanent deformation permitted(faulted condition)      _________________
 
SSE Safe-shutdown earthquake 
 
1/2SSE 1/2 Safe-shutdown earthquake 
 
NORMAL Those normal operation occurrences wh ich are expected frequently and regularly in the course of power operation, refueling, maintenance, or maneuvering of the plant.
 
LOCA Loss-of-coolant accident 
 
NLSF No loss of safety function. Permanent deformation permitted to the extent that there is no loss of safety function.     
 
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TABLE 3.2-3 (SHEET 1 OF 6)
LISTING OF P&IDs
 
System Drawing Number
 
REV 21  5/08 Reactor coolant system D-175037 sheet 1  D-175037 sheet 2 D-175037 sheet 3 D-205037 sheet 1 D-205037 sheet 2 D-205037 sheet 3
 
Residual heat removal system D-175041 D-205041 
 
Containment cooling and purge system D-175010 sheet 1 D-175010 sheet 2
 
Penetration filtration system D-175022 
 
Post-accident containment combustible gas control system D-175019 D-205019    Safety injection system D-175038 sheet 1 D-175038 sheet 2 D-175038 sheet 3 D-205038 sheet 1 D-205038 sheet 2 D-205038 sheet 3
 
Auxiliary feedwater system D-175007 
 
Spent fuel pool cooling system D-205043 
 
River water system D-170119 sheet 6 D-170119 sheet 7
 
Service water system D-170119 sheet 1 D-170119 sheet 2 D-175003 sheet 1 D-175003 sheet 2 D-175003 sheet 3 D-175003 sheet 4 D-205003 sheet 1 D-205003 sheet 2 D-205003 sheet 3 FNP-FSAR-3
 
TABLE 3.2-3 (SHEET 2 OF 6)
 
System Drawing Number
 
REV 21  5/08 Component cooling water system D-175002 sheet 1  D-175002 sheet 2 D-175002 sheet 3 D-205002 sheet 1 D-205002 sheet 2 D-205002 sheet 3
 
Demineralized makeup water system D-175047 sheet 1 D-175047 sheet 2 D-205047 
 
Potable and sanitary water system D-170127 
 
Reactor makeup water system D-175036 D-205036 
 
Plant water treatment system figure 9.2-11 
 
Well water system D-170110 sheet 1
 
Compressed air system D-170131 sheet 1 D-170131 sheet 2 D-170131 sheet 3 D-200019 sheet 1 D-200019 sheet 2
 
Service air system D-175035 sheet 1 D-205035 
 
Instrument air system D-175034 sheet 1 D-175034 sheet 2 D-175034 sheet 3 D-205034 sheet 1 D-205034 sheet 2 D-205034 sheet 3 D-205034 sheet 4
 
Sampling system D-175009 sheet 1 D-175009 sheet 2 D-175009 sheet 3 D-205009 sheet 1 D-205009 sheet 2 D-205009 sheet 3 FNP-FSAR-3
 
TABLE 3.2-3 (SHEET 3 OF 6)
 
System Drawing Number
 
REV 21  5/08    Nonradioactive drains and vents D-175005 
 
Radioactive drains and vents D-175004 sheet 1 D-175004 sheet 2
 
Chemical and volume control system D-175039 sheet 1 D-175039 sheet 2 D-175039 sheet 3 D-175039 sheet 4 D-175039 sheet 5  D-175039 sheet 6 D-175039 sheet 7 D-205039 sheet 1 D-205039 sheet 2 D-205039 sheet 3 D-205039 sheet 4 D-205039 sheet 5
 
Boron thermal regeneration system D-175040 sheet 1 D-205040 
 
HVAC and filtration system (control room and computer room)
D-175012 D-205012    Nonradioactive area heating, ventilation system D-175014 sheet 1 D-175014 sheet 2 D-205014 sheet 1 D-205014 sheet 2
 
Spent fuel pool ventilation D-175045 D-205045 
 
Access control area heating, ventilating, and air
 
conditioning D-175001    Engineered safety feature pump rooms ventilating and
 
filtration system D-175001    Radwaste area heating, ventilating and filtration system D-175011 D-175011 sheet 1 sheet 2  D-175011 sheet 3 FNP-FSAR-3
 
TABLE 3.2-3 (SHEET 4 OF 6)
 
System Drawing Number
 
REV 21  5/08  D-205011 sheet 1  D-205011 sheet 2 D-205011 sheet 3 D-205011 sheet 4
 
Turbine building chilled water system D-175031 sheet 1 D-175031 sheet 2 D-175031 sheet 3 D-205031 sheet 1 D-205031 sheet 2 D-205031 sheet 3
 
Turbine building heating, ventilating, air conditioning, and filtration system D-175027    Communication system D-177331 D-177334 sheet 1 D-177334 sheet 2 D-177334 sheet 3 D-175335 D-177336 D-177337 sheet 1 D-177337 sheet 2 D-177337 sheet 3 D-177338 D-177339 D-207331 D-207334 sheet 1 D-207334 sheet 2 D-207336 D-207337 sheet 1 D-207337 sheet 2 D-207339 
 
Diesel generator fuel oil system D-170060 
 
Diesel generator cooling water system D-170119 sheet 3 D-200013 sheet 3
 
Diesel generator starting air system D-170806 sheet 1
 
Diesel generator starting air and control air systems D-170807 sheet 1
 
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TABLE 3.2-3 (SHEET 5 OF 6)
 
System Drawing Number
 
REV 21  5/08 Extraction steam system D-200009 Main stream system D-175033 sheet 1 D-175033 sheet 2 D-170114 sheet 1 D-170114 sheet 2 D-205033 sheet 1 D-205033 sheet 2 D-200007 
 
Chemical injection system D-175000 sheet 1 D-175000 sheet 2
 
Main condenser vacuum system D-170064 
 
Circulation water system D-170119 sheet 9 D-170119 sheet 10 D-200013 sheet 6
 
Condensate and feedwater system D-170117 sheet 1 D-170117 sheet 2 D-170117 sheet 3 D-170117 sheet 4 D-175073 D-200011 sheet 1 D-200011 sheet 2 D-200011 sheet 3 D-205073 
 
Steam generator blowdown processing system D-175071 sheet 1 D-175071 sheet 2 D-175071 sheet 3 D-205071 sheet 1 D-205071 sheet 2 D-205071 sheet 3
 
Waste processing system D-175042 sheet 1 D-175042 sheet 2 D-175042 sheet 3 D-175042 sheet 4 D-175042 sheet 5 D-175042 sheet 6 FNP-FSAR-3
 
TABLE 3.2-3 (SHEET 6 OF 6)
 
System Drawing Number
 
REV 21  5/08  D-175042 sheet 7  D-175042 sheet 11 D-175042 sheet 12 D-205042 sheet 1 D-205042 sheet 2 D-205042 sheet 3 D-205042 sheet 4 D-205042 sheet 5 D-205042 sheet 6 D-205042 sheet 7 D-205042 sheet 9 D-205042 sheet 10
 
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REV 21  5/08 TABLE 3.2-4 COMPONENT CODING
 
Component Code Edition Applicable/Addenda
 
Reactor vessel ASME III (a) 1968 through Summer 1970  Class A 
 
Full length ASME III 1968 through Winter 1969 CRD mechanisms Class A 
 
Steam Generators Tube side ASME III, Class A 1989 No Addenda Shell side ASME III, Class A 
 
Pressurizer ASME III, Class A 1968 through Summer 1970
 
Reactor coolant Nuclear Pump and Valve Code (b) 1968 through March 1970 pump casing Piping and  ASME III, Class 1 (c) 1971 through Summer 1971 fittings   
: a. ASME Boiler and Pressure Vessel Code Section III, Nuclear Vessels, including applicable
 
mandatory Addenda. 
 
Code Cases are not mandatory until included in a mandatory Addendum to the Code. The
 
designer does not require that all Code Cases be applied. Where a specific Code Case is
 
required by the designer it will be identified in the technical requirements. Where a supplier
 
presents justification for applying a specific Code Case, the designer will review the justification
 
and approve or disapprove the request.
 
Hardship exceptions to 10 CFR Part 50.55a are presented in Table 5.2-1. 
: b. ASME Code of Pumps and Valves for Nuclear Power. Reactor coolant pump casing in Unit
 
2 is to ASME III, 1971 ed. 
: c. Three 31-in., 90-degree vane elbows manufactured by Mitsubishi Steel Manufacturing
 
Company, Nagasaki, Japan, meet requirements of ASME III, Class 1, through Summer 1971
 
addenda, except for "N" Stamp (Unit 1 only). At time of procurement of these fittings, Mitsubishi
 
did not have "N" Stamp due to ASME not recognizing foreign manufacturers. 
 
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REV 21  5/08 TABLE 3.2-5 (SHEET 1 OF 4)
ASME CODE CASES FOR CLASS 1 COMPONENTS
 
Code/Case Title 1141 Foreign Produced Steel 
 
1332 Requirements for Steel Forgings 
 
1334 Requirements for Corrosion Resistant Steel Bars 
 
1335 Requirements for Bolting Material 
 
1337 Requirements for Special Type 403 Modified Forgings or Bars (Section III) 
 
1344 Requirements for Nickel-Chromium Age-Hardenable Alloys
 
1345 Requirements for Nickel-Molybdenum-Chromium-Iron Alloys
 
1355 Electroslag Welding 
 
1361 Socket Welds 
 
1364 Ultrasonic Transducers SA-435 (Section II) 
 
1384 Requirements for Precipitation Hardening Alloy Bars and Forgings 
 
1388 Requirements for Stainless Steel-Precipitation Hardening
 
1390 Requirements for Nickel-Chromium, Age-Hardenable Alloys for Bolting 
 
1395 SA-508, Class 2 Forgings-Modified Manganese Content
 
1401 Welding Repair, to Cladding 
 
1407 Time of Examination 
 
1423 Plate, Wrought Type 304 and 316 with Nitrogen Added
 
1433 Forgings, SA-387 
 
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REV 21  5/08 TABLE 3.2-5 (SHEET 2 OF 4)
 
Code/Case Title 1434 Class 8N Steel Casting (Postweld Heat Treatment for SA-487)
 
1448 Use of Case Interpretations of ANSI B31 Code for Pressure Piping
 
1456 Substitution of Ultrasonic Examination 
 
1459 Welding Repairs to Base Metal 
 
1461 Electron Beam Welding 
 
1470 External Pressure Charts for Low Alloy Steel 
 
1471 Vacuum Electron Beam Welding of Tube Sheet Joints 
 
1474 Integrally Finned Tubes (Section III) 
 
1477 B-31.7, ANSI 1970 Addenda 
 
1484 SB-163 Nickel-Chromium-Iron Tubing at a Specified Minimum Yield Strength of 40,000 psi 
 
1487 Evaluation of Nuclear Piping for Faulted Conditions
 
1492 Postweld Heat Treatment 
 
1493 Postweld Heat Treatment 
 
1494 Weld Procedure Qualification Test 
 
1495 Stress Indices in Table NB-3681.2-1 
 
1498 SA-508, Class 2, Minimum Tempering Temperature 
 
1501 Use of SA-453 Bolts in Service Below 800 F without Stress Rupture Tests
 
1504 Electrical and Mechanical Penetration Assemblies 
 
1508 Allowable Stresses, Design Stress Intensity and/or Yield Strength Values
 
1514 Fracture Toughness Requirements FNP-FSAR-3
 
REV 21  5/08 TABLE 3.2-5 (SHEET 3 OF 4)
 
Code/Case Title 1515 Ultrasonic Examination of Ring Forgings for Shell Section of Section III-Class 1 Vessels 
 
1516 Welding of Non-Integral Seats in Valves for Section III Application
 
1517 Material Used in Pipe Fittings 
 
1519 Use of A-105-71 in lieu of SA-105 
 
1521 Use of H. Grades SA-240, SA-479, SA-336 and SA-358 
 
1522 ASTM Material Specifications 
 
1524 Piping 2-in. NPS and Smaller 
 
1525 Pipe Descaled by Other Than Pickling 
 
1526 Elimination of Surface Defects 
 
1527 Integrally Finned Tubes 
 
1528 High Strength SA-508 Class 2 and SA-541 Class 2 Forgings for Section III Construction of Class 1 Components
 
1529 Material for Instrument Line Fittings 
 
1531 Electrical Penetrations, Special Alloys for Electrical Penetration Seals
 
1534 Overpressurization of Valves 
 
1535 Hydrostatic Test of Class 1 Nuclear Valves 
 
1539 Metal Bellows and Metal Diaphragm Steam 
 
1542 Requirements for Type 403 Modified Forgings or Bars for Bolting Material
 
1544 Radiographic Acceptance Standards for Repair Welds 
 
1545 Test Specimens from Separate Forgings for Class 1, 2, 3 and MC
 
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.2-5 (SHEET 4 OF 4)
 
Code/Case Title 1546 Fracture Toughness Test for Weld Metal Section 
 
1547 Weld Procedure Qualification Tests; Impact Testing Requirements, Class 1 
 
1552 Design by Analysis of Section III Class 1 Valves 
 
1557 Plate Steel Refined by Electroslag Remelting 
 
1567 Test Lots for Low Alloy Steel Electrodes 
 
1568 Test Lots for Low Alloy Steel Electrodes 
 
1571 Materials for Instrument Line Fittings, for SA-234 Carbon Steel Fittings 
 
1573 Vacuum Relief Valves 
 
1574 Hydrostatic Test Pressure for Safety Relief Valves 1621 Line Valve Internal and External Items Section III, Class 1, 2, and 3 
 
1690 Stock Materials for Section III Construction 
 
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3.3-1 REV 21  5/08 3.3 WIND AND TORNADO LOADINGS 3.3.1 WIND LOADINGS Wind loadings for Category I structures have been selected on the basis of ASCE Paper No.
 
3269, "Wind Forces on Structures" (1) or as provided in "TORMIS Missile Risk Analysis for Farley Nuclear Plant Units 1 and 2." (3) 3.3.1.1  Design Wind Velocity Category I structures are designed to withstand a basic wind velocity of 115 mph. The
 
recurrence interval of this wind velocity is estimated to be at least 100 years.
(1)  The variation of wind velocity with height is shown in table 3.3-1. 
 
3.3.1.2  Basis for Wind Velocity Selection The "fastest mile of wind" at the Farley Plant site is shown, according to Figure 1 (b) and the
 
ASCE paper, (1) to be 90 mph. As a result of recent hurricane experiences on the Gulf Coast, a design velocity of 105 mph at ground level was selected. For additional conservatism, to
 
account for uncertainties in historical data, this margin of safety has been increased and a
 
ground level wind of 115 mph has been used as the basic design wind. 
 
3.3.1.3  Vertical Velocity Distribution and Gust Factor The wind pressures resulting from the wind velocities shown in table 3.3-1 incorporate the
 
shape factors in both horizontal and vertical directions. A gust factor of 1.1 has been selected
 
for the design and has been incorporated into the wind pressures shown in table 3.3-1. 
 
The gust factor of 1.1 is selected on the basis of ASCE paper No. 3269, "Wind Forces on
 
Structures." (1)  This paper recommends that appropriate gust factors be used for structures that are small enough to be responsive to gusts involving less than 1 mile of passing wind, and that
 
the gust factors bear some relation to the minimum size of gust necessary to envelop the
 
structure and its accompanying pattern of flow. A gust factor of 1.1 will allow for gust of
 
approximately 10-second duration which, in a 115-mph basic wind, would have a length
 
downwind of about 1,700 ft; this factor is adequate for structures having a horizontal dimension, transverse to the wind, of 125 ft and larger. 
 
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3.3-2 REV 21  5/08 3.3.1.4  Determination of Applied Forces The design wind dynamic pressure is calculated by 
 
q  =  0.002558 V 2
where  q  =  pressure in psf
 
V  =  velocity in mph
 
A shape coefficient of 1.3 is applied for building wind loads. Of the total of 1.3 q , 0.9 q is
 
applied as positive pressure to the windward walls, and 0.4 q is applied as negative pressure
 
on the leeward walls, where applicable. A shape factor of 0.6 is applied for plant vent stack
 
wind loads.
 
Wind loads are applied to the structures as uniform static loads on the surface area normal to
 
the wind. 
 
The applied force magnitude and distribution calculated for Category I structures are shown on
 
figure 3.3-1. 
 
3.3.2 TORNADO LOADINGS All above ground Category I structures required to ensure the integrity of the reactor coolant
 
pressure boundary, safe shutdown of the plant, long-term core cooling, or to prevent radioactive
 
releases resulting in offsite exposures comparable to 10 CFR 100 guidelines are also designed
 
to withstand tornado loadings and tornado generated missiles (2) or have been analyzed as discussed in paragraph 3.5.1.2.
 
3.3.2.1  Applicable Design Parameters For Category I structures designed to withstand tornadoes and tornado generated missiles, the
 
three following parameters are applied concurrently, in combinations producing the most critical
 
conditions: 
 
a) Dynamic Wind Pressure 
 
The dynamic wind pressure is caused by a tornado funnel having a peripheral tangential velocity of 300 mph and a forward progression of 60
 
mph. The applicable portions of wind design methods described in ASCE
 
Paper No. 3269 are used, particularly for shape factors. The provisions
 
for gust factors and variation of wind velocity with height are not applied. 
 
The average tornado design dynamic wind pressure is q = 230 psf based
 
on an average wind velocity of 300 mph. 
 
FNP-FSAR-3
 
3.3-3 REV 21  5/08  b) Pressure Differential 
 
The structure interior bursting pressure is taken as rising 1 psi/s for 3 seconds, followed by a 2-second calm, them decreasing at 1 psi/s for 3
 
seconds. This cycle accounts for reduced pressure in the eye of a
 
passing tornado. All fully enclosed Category I structures are designed to
 
withstand the full 3 psi pressure differential. 
 
c) Missile Impingement 
 
A tornado missile is defined as any object set in motion and propelled by a tornado. Three types of tornado missiles are considered; each type is
 
assumed to act independently and only one type may be generated at
 
any one time. It is also assumed that the missiles do not tumble while in
 
flight, and are at any time oriented to have the maximum value: 
 
C d A W
 
where
 
C d = Drag coefficient
 
A  = Projected area of missile exposed to wind
 
W = Weight of missile
 
The three types of missiles are as follows: 
: 1. A 12-ft-long piece of wood 8 in. in diameter (114 lb) traveling end-on at a speed of 300 mph and striking the structure at any elevation. 
: 2. A 10-ft-long steel pipe, schedule 40, 3 in. in diameter (75.8 lb), traveling end-on at a speed of 100 mph and striking the structure at any elevation. 
: 3. A 4,000-lb automobile, traveling end-on at a speed of 50 mph and striking the structure on an impact area of 20 sq ft, with any portion of the impact
 
area being not more than 25 ft above grade. 
 
3.3.2.2  Determination of Forces on Structures Tornado loads are applied to the Category I structures in the same manner as the wind loads
 
described in subsection 3.3.1.4 with the exception that gust factor and variation of wind velocity
 
with height do not apply. The load combinations involving tornadoes are given in subsections
 
3.8.1.3, 3.8.4.3, and 3.8.5.3. 
 
FNP-FSAR-3
 
3.3-4 REV 21  5/08 The load factor selected for tornado loadings is 1.0, based on the short duration of the loading condition, the low probability of a tornado striking a specific geographic point, and the degree of
 
conservatism in the selection of design tornado velocity. This subject is discussed in B-TOP-3. 
 
3.3.2.3 Ability of Category I Structures to Perform Despite Failure of Structures Not Designed for Tornado Loads Failure of Category II structures not designed for tornado loads will not affect the ability of
 
Category I structures to perform their functions for the following reasons: 
: a. Tornado missiles that may be formed by the failure of Category II structures will not exceed the force of those postulated and described in
 
subsection 3.3.2.1, against which Category I structures are designed. 
: b. The structural frame of the Category II turbine building in the vicinity of the auxiliary building has been designed against collapse when subjected
 
to tornado loadings. 
 
FNP-FSAR-3
 
3.3-5 REV 21  5/08 REFERENCES 
: 1. "Wind Forces on Structures," Transactions of the ASCE, Paper No. 3269, 1961. 
: 2. "Design Criteria for Nuclear Power Plants Against Tornadoes," Bechtel Topical Report, B-TOP-3, March 1970. 
: 3. "TORMIS Missile Risk Analysis for Farley Nuclear Plant Units 1 and 2," ARA Report 4733 , March 1999.
 
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.3-1 WIND LOADS WITH GUST FACTOR Dynamic Wall Load (psf) Roof Load (psf)
Height Velocity Pressure Pressure Suction Suction (ft) (mph) q (psf) 0.9q 0.4q 0.7q      0-50 115 42 38 17 30      50-150 140 62 56 25 44      150-400 170 91 82 37 64
 
REV 21  5/08 WIND FORCES AND DISTRIBUTION ON CATEGORY I STRUCTURES JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.3-1
 
FNP-FSAR-3
 
3.4-1 REV 21  5/08 3.4 WATER LEVEL (FLOOD) DESIGN All Category I structures are designed to protect the safety related systems, equipment, and
 
components from the respective probable maximum flood and/or the highest groundwater
 
levels. 
 
Three probable maximum flood levels are used. The design bases for the PMF elevation of 127
 
ft are presented in subsections 2.4.2.1 and 2.4.2.2. The design bases for the PMF elevations of
 
144.2 ft and 192.2 ft are given in subsections 2.4.2.2 and 2.4.8.1, respectively. 
 
3.4.1 FLOOD PROTECTION The design maximum flood elevations for Category I structures are as follows: 
: 1. Containment structure - elevation 144.2 ft 
: 2. Auxiliary building - elevation 144.2 ft 
: 3. Diesel generator building - elevation 144.2 ft 
: 4. Electrical cable tunnel structure - elevation 144.2 ft 
: 5. Category I outdoor tanks - elevation 144.2 ft 
: 6. River intake structure - elevation 127 ft (a) 
: 7. Intake structure at storage pond - elevation 192.2 ft 
: 8. Pond spillway structure - elevation 192.2 ft 
: 9. Storage pond dam and dike - elevation 192.2 ft 
 
The safety-related systems, equipment, and com ponents are protected against floods by virtue of their being located in flood protected structures.  (See table 3.2-1.)  These systems, equipment, and components, except those located in the river intake structure and the intake
 
structure at the storage pond, are located on, above, or flood protected to the plant grade
 
elevation of 154.5 ft. The systems, equipment, and components located in the river intake
 
structure and intake structure at the storage pond are flood protected to elevation 127.0 ft and
 
elevation 195.0 ft, respectively. 
 
Descriptions of the Category I structures which house the safety-related systems, equipment, and components are given in subsections 3.8.1.1 and 3.8.4.1. 
: a. Original design (Category I) requirements are no longer required.
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3.4-2 REV 21  5/08 All exterior or access openings and penetrations are flood protected by watertight concrete walls to grade elevations which are 154.5 ft for the plant area and 195.0 ft for the storage pond
 
area, respectively. Access to all Category I structures is possible only from above grade levels. 
 
3.4.2 ANALYSIS PROCEDURES The foundation slabs and exterior walls of the structures are designed to resist the upward and
 
the lateral pressures caused by the maximum flood levels given in Section 3.4.1. 
 
The hydrostatic pressure acting uniformly at the bottom of the structures is the product of
 
the height to the design flood level and the weight of water which is taken as 63 lb/ft 3 (See figure 3.4-1.) 
 
The horizontal pressure acting on the exterior walls varies with height, from the maximum at
 
the bottom of the wall to zero at the design flood level.  (See figure 3.4-1). 
 
Dynamic water forces associated with phenomena such as flood currents, wind waves, hurricanes, and tsunamis are not considered in the design of the Category I structures.
 
Justifications for their omission are given in subsections 2.4.3.6, 2.4.5 and 2.4.6. 
 
REV 21  5/08 WATER PRESSURE ON STRUCTURES JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.4-1
 
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3.5-1 REV 21  5/08 3.5 MISSILE PROTECTION Category I structures are designed to protect safety-related equipment and components from
 
being damaged by internal and external missiles. 
 
3.5.1 MISSILE BARRIERS AND LOADINGS The missile barriers are designed to resist the missiles selected in subsection 3.5.2. 
 
3.5.1.1  Accident/Incident Generated Missiles Inside Containment A tabulation of barriers and the missiles they have been designed to contain is given in table 3.5-
: 1. The postulated missile loadings are derived from the physical characteristics of the
 
components involved and their respective kinetic ener gy levels. They are given in tables 3.5-2 through 3.5-5. The analytical method used to convert energies into forces and depths of
 
penetration necessary to barrier design is described in subsection 3.5.4. 
 
3.5.1.2  Environmental Load Generated Missiles 3.5.1.2.1 Missile Protection Methods Those systems or components listed in table 3.2-1 that are required for safe shutdown, for
 
immediate or long term core cooling, or to prevent a radioactive release resulting in offsite
 
exposures comparable to 10 CFR 100 guidelines are provided with tornado missile protection by
 
location within Category I structures, burial underground, or missile barriers/shielding or have
 
been analyzed as discussed in paragraph 3.5.1.2.2.
 
Category I structures housing equipment and components vital to a safe shutdown have been
 
designed against penetration by the tornado missiles described in paragraph 3.3.2.1 (c). These
 
structures, having at least 2-ft-thick concrete exterior walls and roof slabs, constitute barriers
 
against missile penetration. Calculations show that the deepest missile penetration of the
 
concrete barriers would be 10 in. Therefore, the 2-ft-thick slabs provide ample protection. 
 
Where concrete spalling due to missile impact is considered, the inside surfaces of the following
 
areas have been protected with corrugated sheet metal:
* Control room.
* HVAC equipment room for the control room.
* Component cooling water surge tank room.
* Spent-fuel pool area. 
 
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3.5-2 REV 21  5/08 3.5.1.2.2 Components Not Requiring Unique Missile Protection Certain Seismic Category I systems and components located outside of Seismic Category I
 
structures are evaluated as not requiring unique tornado missile protection by burial or barriers. 
 
The following two approaches are used in the evaluation of these systems and components relative to a tornado event.
 
3.5.1.2.2.1  Components Not Required for a Tornado Event. The probability of occurrence of a tornado event coincident with another low probability design basis event is so small that no
 
protection from tornado missiles is required for certain Seismic Category I structures, systems, and components which are not otherwise needed for safe shutdown; for immediate or long-term
 
core cooling; to prevent a radioactive release resulting in offsite exposures comparable to 10
 
CFR 100 guidelines; or to support other systems or components which are required for one of
 
those functions.
 
3.5.1.2.2.2  Components with Acceptable Probability of Survival
. Safety-related systems and components required for safe shutdown, for immediate or long-term core cooling, or to
 
prevent a radioactive release resulting in offsite exposures comparable to 10 CFR 100 guidelines
 
required for a tornado event are generally protect ed. A limited amount of unprotected portions of these systems and components is analyzed using probabilistic missile damage analysis as
 
permitted in Standard Review Plan 3.5.1.4, "Missiles Generated By Natural Phenomena."  This
 
analysis is conducted to determine the probability per year of missiles generated by postulated
 
tornadoes striking and damaging these systems and components beyond their failure point. For FNP, the specific acceptance criterion for tornado damage for the unprotected systems and
 
components required for a tornado event is that the cumulative sum of the mean failure
 
probabilities for these systems and components be less than 10
-6 per year per unit. The allowable level of less than 10
-6 per year per unit for the cumulative probability of failure of such systems and components is acceptable if, when combined with reasonable qualitative
 
arguments, the realistic probability can be shown to be lower.
 
The analysis used for FNP is the computer program TORMIS (3)(5)(6), developed by the Electric Power Research Institute (EPRI)
(4) and accepted by the NRC.
 
Systems and components whose analysis using t he TORMIS methodology yield results that cause the less than 10
-6 per year per unit acceptance criterion to be exceeded will be provided with unique barriers to reduce the total failure probability value to below the acceptance criterion.
 
3.5.1.2.3 TORMIS Methodology TORMIS (3)(4)(5)(6) is a methodology developed to predict the probability of damage to nuclear power plant structures and components from to rnadoes. There are four fundamental models in the TORMIS analysis:  wind hazard, site facility, load effects, and system models. Monte Carlo simulation is used to produce numerical estimates of hit and damage probabilities based on the
 
site-specific models.
 
The wind hazard analysis for FNP Units 1 and 2 uses a site-specific analysis to generate a
 
tornado hazard curve specifically for Farley.
FNP-FSAR-3
 
3.5-3 REV 21  5/08 The site facility model was conservatively developed based on a site area walkdown and the specific characteristics, materials, and failure points for Farley structures and components.
 
Load effects are determined based on the TORMIS model missiles, missile transport model, and
 
component characteristics. The missiles utiliz ed in the TORMIS model encompass the three design basis missiles described in paragraph 3.3.2.1.
 
TORMIS implements a methodology developed by EPRI. TORMIS determines the probability of striking walls and roofs of buildings on which penetrations or exposed portions of systems/
 
components are located. The probability is calculated by simulating a large number of tornado
 
strike events at the site for each tornado wind speed intensity scale. After the probability of
 
striking the walls or roof is calculated, the exposed surface area of the particular components is
 
factored in to compute the probability of striking and consequently damaging a particular item.
 
The following provisions apply to the TORMIS analysis for FNP:
: 1. FSAR paragraph 2.3.1.3 estimates an occurrence rate of 3 tornadoes per year per 1-degree square (approximately 4000 square miles), or 7.5 E-04 tornadoes per square
 
mile. This occurrence rate was based on conservative treatment of data from 1955
 
through 1967. As part of the FNP TORMIS analysis, the annual probability of a tornado
 
was determined for the Fujita F-scale wind speeds using regional data available in
 
TORMIS for NRC Region II. A site-specific analysis was performed to generate a
 
tornado data set for the TORMIS analysis of Farley.
 
The National Climatic Data Center (NCDC) files for the years 1950 through 1996 were used as the basic source of data for this investigation. These data were screened to
 
eliminate coding errors in the record fields. In addition, corrections were introduced to
 
account for reporting efficiency and time series, or other potential errors resulting from the
 
indirect characteristics of the available data. The overall tornado occurrence rate
 
computed from the updated regional data in TORMIS is 6.0 E-04 tornadoes per square
 
mile per year. While this rate is slightly lower than the rate in FSAR paragraph 2.3.1.3, it
 
is a more accurate figure based on conservative treatment of the best available data and
 
will therefore be used in lieu of the data cited in FSAR paragraph 2.3.1.3 for those
 
components or portions of systems analyzed in TORMIS.
: 2. The Fujita scale (F-scale) wind speeds will be used in lieu of the TORMIS wind speeds (F-scale) for the F0 through F5 intensities.
: 3. The tornado windfield parameters in the FNP TORMIS analysis were adjusted to increase the wind profile in the lowest 10 m over the original profile in TORMIS. This adjustment
 
applied the ratio of V 0/V 33 in a conservative manner in accordance with the NRC's October 26, 1983, TORMIS SER.
: 4. Detailed surveys of the plant site were performed to characterize and quantify potential missiles for use in the FNP TORMIS analysis. To ensure conservatism, these surveys were performed during a refueling outage when large amounts of material were
 
temporarily stored in outside laydown ar eas around the site. Additionally, ground and aerial photographs were reviewed to estimate the number and type of missiles which FNP-FSAR-3
 
3.5-4 REV 21  5/08 could originate from remote areas of the site. The total number of missiles used in the FNP TORMIS analysis was 51,864.
: 5. The FNP analysis will not deviate from the TORMIS program as described in reference 4 of FSAR section 3.5, except as noted in items 1 through 4 above.
 
3.5.1.3  Site Proximity Missiles There are no guided missile installations in the vicinity of the Farley Nuclear Plant. 
 
At the time of construction of Farley Nuclear Plant the only landing strip within a radius of 5 miles
 
from the Farley site was a 3000-ft landing strip for the paper company at Cedar Springs, Georgia, approximately 3.5 miles south of the plant site. 
 
Aircraft using the strip were light, twin-engined business planes comparable to a Cessna 401-A, which has a gross weight of 6300 lb. The orientation of this landing strip was N 30 degrees E;
 
therefore, takeoffs and landing approaches were not in the direction of the Farley Nuclear Plant
 
site. The landing strip is now abandoned. 
 
A new 5400-ft landing strip, capable of handling jet engined aircraft, has been constructed by the
 
paper company at Cedar Springs, Georgia. The new strip is located approximately 4 to 5 miles south of the old landing strip and 7 to 8 miles from the Farley site. The strip has approaches
 
oriented NW and SE and is used by jet aircraft as well as conventional aircraft. The jet aircraft
 
are six-to-eight-passenger business jets comparable to a Lear Jet Model 23, which has a gross
 
weight of 12,500 lb. The paper company at Cedar Springs, Georgia, has indicated that pilots will
 
be instructed to avoid the Farley Nuclear Plant site area during both takeoffs and landing
 
operations. 
 
For these reasons, aircraft generated missiles are not considered. 
 
3.5.1.4  Accident/Incident Generated Missiles Inside Category I Structures Other than Containment A tabulation of barriers and the missiles they have been designed to contain is given in table 3.5-
: 6. The postulated missile loadings for the rod drive motor generator sets are derived from the
 
physical characteristics of these components and their respective kinetic energy levels, as given
 
in table 3.5-7. 
 
3.5.2 MISSILE SELECTION 3.5.2.1  Missile Selection Within the Containment The systems located inside the containment hav e been examined to identify and select potential missiles. The basic approach was to ensure design adequacy against generation of missiles, rather than allow missile formation and then contain their effects.
FNP-FSAR-3
 
3.5-5 REV 21  5/08 The following components have been considered to have a potential for missile generation: 
: a. Control rod drive mechanism, driv e shaft, and the drive shaft and drive mechanism latched together. 
: b. Certain valves defined below. 
: c. Temperature and pressure element assemblies. 
 
The limiting case considered for design is a drive shaft ejected from the reactor through the top
 
of the rod travel housing. The following sequence of events is assumed:  The drive shaft and
 
control rod cluster are forced out of the core by the differential pressure of 2500 psi across the
 
drive shaft.  (The drive shaft and control rod clus ter, latched together, are assumed fully inserted when the accident starts.)  After approximately 12 ft of travel, the rod cluster control spider hits
 
the underside of the upper support plate. Upon impact the flexure arms in the coupling joining
 
the drive shaft and control cluster fracture, completely freeing the drive shaft from the control rod
 
cluster. The control cluster would be completely stopped by the upper support plate; however, the drive shaft would continue to be accelerated upward to hit the missile shield structure
 
provided. 
 
The valves considered for missile potential are those in the region where the pressurizer extends
 
above the operating deck, such as the pressurizer safety valves, the motor-operated isolation
 
valves in the relief line, the air-operated relief valves, and the air-operated spray valves. 
 
Although failure of these valves is considered improbable, failure of the valve bonnet body bolts, nevertheless, has been considered and provisions made to ensure integrity of the containment liner from the resultant bonnet missile. 
 
The only probable source of jet-propelled missiles from the reactor coolant piping and piping
 
systems connected to the reactor coolant system is the type represented by the temperature and pressure element assemblies. The resistance temperature element assemblies can be of two
 
types:  "with well" and "without well". Two rupt ure locations have been assumed for each type of temperature element assembly:  one around the weld between the boss and the pipe wall for
 
each assembly, and another at the weld (or thread) between the temperature element assembly
 
and the boss for the "without well" element or the weld (or thread) between the well and the boss
 
for the "with well" element. 
 
A temperature element is installed on the reactor coolant pumps close to the radial bearing
 
assembly. A hole is drilled in the gasket and sealed on the internal end by a steel plate. In
 
evaluating missile potential, it is assumed that this plate could break and the pipe plug on the
 
external end of the hole could become a missile. 
 
In addition, it is assumed that the welding between the instrumentation well and the pressurizer
 
wall could fail and the well and sensor assembly could become a jet propelled missile. 
 
Finally, it is assumed that the pressurizer heaters could become loose and become jet propelled missiles. 
 
FNP-FSAR-3
 
3.5-6 REV 21  5/08 3.5.2.2  Missiles Selected Outside the Containment The tornado generated missiles selected for the design of the Farley Plant structures are
 
described in subsection 3.3.2. 
 
3.5.2.3  Missile Selection Within Category I Structures Other Than Containment The systems located inside Category I structures other than the containment have been
 
examined to identify potential missiles. The following components are considered to have a
 
potential for missile generation: 
: a. Flywheels of two rod drive power supply motor generator sets. 
 
The electric motors of the rod drive power supply motor generator sets are designed to operate
 
at 1800 rpm. In the unlikely event of an overspeed generated flywheel missile, the steel
 
protective shield which closely encircles the flywheel would contain the missile and prevent it
 
from impacting any safety-related components. 
 
The steel protective shields are designed to contain a spectrum of probable flywheel fragment
 
missiles generated at an overspeed of 150 percent of the operating speed as indicated in table
 
3.5-7. For conservation, the initial translational energy of the governing missile is increased by
 
10 percent for the design of the steel protective shield. 
 
3.5.3 SELECTED MISSILES The missiles selected inside the containment are given in tables 3.5-1 through 3.5-5. 
 
The origin, weight, impact velocity, impact area, and all other parameters necessary to determine the missile penetration are listed in these tables. The calculated depth of penetration into a 2-ft-
 
thick concrete slab is also given. 
 
The missiles selected outside the containment are given in paragraph 3.3.2.1(c). 
 
3.5.4 BARRIER DESIGN PROCEDURES The internal and external missile barriers have been designed to resist missile penetration in
 
order to protect systems and components so that the failure of one system or component cannot cause the failure of another system or component. 
 
Missile barriers are constructed of concrete, steel or a combination of concrete and steel in order
 
to provide protection from the effects of missiles.
 
Barriers are designed based on the pertinent characteristics of the potential targets, postulated
 
missiles, and barrier materials including the materials ability to provide protection from
 
penetration, perforation, and spalling. The methods and procedures used to evaluate missile
 
impact on structures and barriers and the analytical methods used to convert energies into forces FNP-FSAR-3
 
3.5-7 REV 21  5/08 and depths of penetration necessary for barrier design are described in NAVDOCKS P-51 (1) and Bechtel Topical Report BC-TOP-9A (2).
The analysis for the depth of missile penetration in reinforced concrete was carried out using the
 
following modified Petry formula as presented in NAV DOCKS P-51. 
 
  (3.5-1)  D  = KA p V' (3.5-2)  V' = log 10[1 + V 2/215000]    (3.5-3)  D' = D  [1 + e
-4 (a'-2)] 
 
a' = T/D
 
where 
 
D  = depth of penetration of an infinitely thick slab (inches) 
 
K  = an experimentally obtained materials coefficient for penetration (k = 0.0022 for 5000 psi reinforced concrete)
A p  = sectional pressure, obtained by dividing the weight of the missile by the maximum cross sectional area (expressed as pounds per square foot) 
 
V'  = velocity factor 
 
V  = terminal or striking velocity in feet per second 
 
D'  = actual depth of penetration in a slab of finite thickness (inches)
 
T  = thickness of resisting slab (inches) 
 
The design basis for concrete barrier thickness within the reactor containment is planned to
 
provide a barrier approximately three times thicker than the depth of missile penetration. As a
 
result, 2 ft of concrete was chosen to satisfy the above criterion. Substituting the value of 2 ft for
 
T in equation 3.5-3, the actual depth of penetration, D', was calculated as shown in tables 3.5-2
 
through 3.5-5. 
 
For the external missiles, a minimum of 2 ft of concrete has also been used in the plant design, providing protection against penetration. A summary of Category I structures utilizing concrete
 
designed against missile penetration and the thickness provided is given below: 
 
Thickness (in.)
Auxiliary building, exterior walls and roof slabs (see note 1) 24
 
Containment dome 39
 
FNP-FSAR-3
 
3.5-8 REV 21  5/08 Containment wall 45 Diesel generator building 24
 
River intake structure 24
 
Intake structure at storage pond 24
 
RWST & RMWST shield walls 24
 
Note 1: The walls of the main steam room venting structure are heavy welded steel grating which provides protection against penetration of tornado-generated missiles.
 
Equipment and piping located outside the containment which are required for safe shutdown, long-term core cooling, or to prevent a radioactive release resulting in offsite exposures
 
comparable to 10 CFR 100 guidelines are provided with tornado missile protection either by
 
location within Category I structures, burial under- ground, designed missile barriers/shielding, or
 
have been analyzed as discussed as in paragraph 3.5.1.2.2.
 
3.5.5 MISSILE BARRIER FEATURES Figure 3.8-2, drawing D-176151, figures 3.8-9, 3.8-10, 3.8-11, 3.8-13, 3.8-14, drawings D-
 
205205, D-205206, D-205207, and figures 3.8-23, 3.8-24, 3.8-25, 3.8-26, 3.8-27, 3.8-28, and
 
3.8-29 show the layout and principal design features of the barriers and structures designed to resist missiles. 
 
FNP-FSAR-3
 
3.5-9 REV 21  5/08 REFERENCES
: 1. NAVDOCKS P-51 , "Design of Protective Structures," Bureau of Yards and Docks,  Dept. of the Navy, August 1950.
: 2. BC-TOP-9A , "Design of Structures for Missile Protection," Revision 2, September 1974.
: 3. TORMIS Missile Risk Analysis for Farley Nuclear Plant Units 1 and 2, ARA Report 4733 , March 1999.
: 4. EPRI NP-2005 , Tornado Missile Simulation and Methodology, Volumes I and II, Final Report, August 1981.
: 5. REA 97-1409 response, SCS to SNC letter FP 99-0429, "Tornado Missile Broadness Review and PRA Analysis," August 5, 1999.
: 6. REA FS040491501-01 response, SNC letter PS-04-2337, "Re-evaluate TORMIS Analysis of EDG Silencers," December 3, 2004.
 
FNP FSAR-3
 
REV 21  5/08 TABLE 3.5-1 MISSILE BARRIERS INSIDE CONTAINMENT Missiles Barriers    1. Control rod drive mechanism Integral control rod drive missile shield.
See figure 3.8-13.
: 2. Deleted    3. Drive shaft See figure 3.8-13. 4. Drive shaft latched to drive See figure 3.8-13. mechanism 
: 5. Valve bonnets in the area where 2-ft-thick concrete shield wall. pressurizer extends above the See figures 3.8-13 and 3.8-14.
operating deck (el 155 ft) 
 
(Pressurizer safety valves, motor-operated isolation valves, air-operated relief valves, and air-operated spray valves.) 
: 6. Instrumentation assembly Reactor primary shield walls. See figures 3.8-13 and 3.8-14.
: 7. Pressurizer heater 2-ft concrete shield walls and pressurizer  heater missile shield. See figures 3.8-13  and 3.8-14.
 
FNP FSAR-3
 
REV 21  5/08 TABLE 3.5-2 CRDM - MISSILE CHARACTERISTICS
 
Weight to        Impact Area    Missile Weight O.D. Travel Outside Ratio A  Velocity Kinetic Energy Description (lb) (in.) Housing (ft) (psf) (ft/s) (ft-lb) Energy Ratio(1)
Drive shaft    120 1.75 3    7,200 130 32,000 0.38 Drive shaft        latched to 1,500 3.75 3 19,565 (2) (2) (2) drive mechanism       
: 1. The missile barrier is a 1 1/2-in.-thick steel plate. T he penetration evaluation is performed using the Stanford Research In stitute formul ae (equation 6.203  in U.S. Reactor Containment Technology, Vol. 1). The energy ratio is the ratio of missile energy to the energy required for the missile to completely  penetrate the barrier.
: 2. The critical missile is the drive shaft alone. It is t he limiting case and envelops the dr ive shaft latched to drive mecha nism case.
FNP FSAR-3
 
REV 21  5/08 TABLE 3.5-3 VALVE - MISSILE CHARACTERISTICS
 
Depth of    Flow  Weight-  Penetration in Discharge Thrust Impact to-Impact-  a 2-ft thick Weight  Area  Area  Area Area Ratio, Velocity    Barrier Missile Description  (lb)  (in.2)  (in.2) (in.2) Ab (psf)      ft/s        (in.)   
 
Safety relief valve bonnet 350  2.86  80 24 2,100 110 1.31 (3 x 6 in. or 6 x 6 in.)       
 
3-in. motor-operated 400 5.5 113 28 2,057 135 1.92 isolation valve bonnet (plus motor and stem)       
 
2-in. air-operated relief  75 1.8  20 20 540 115 0.37 valve bonnet (plus stem)       
 
3-in. air-operated spray 120 5.5  50 50 345 190 0.61 valve bonnet (plus stem)       
 
4-in. air-operated spray 200 9.3  50 50 576 190 1.06 valve       
 
FNP FSAR-3
 
REV 21  5/08 TABLE 3.5-4 PIPING TEMPERATURE ELEMENT ASSEMBLY - MISSILE CHARACTERISTICS
: 1. For a tear around the weld between the boss and the pipe:
Characteristics "without well" "with well"
 
Flow discharge area 0.11 in.
2 0.60 in.2  Thrust area 7.1 in.
2 9.6 in.2  Missile weight 11.0 lb 15.2 lb Area of impact 3.14 in.
2 3.14 in.2 A p  = Missile weight 504 psf 697 psf        Impact Area 
 
Velocity 20 ft/s 120 ft/s
 
Depth of penetration in a 2-ft thick barrier 0.012 in. 0.518 in.
: 2. For a tear at the junction between the tem perature element assembly and the boss for the  "without well" element and at the junction bet ween the boss and the well for the "with well"  element. 
 
Characteristics "without well" "with well"
 
Flow discharge area 0.11 in.
2 0.60 in.2  Thrust area 3.14 in.
2 3.14 in.2  Missile weight 4.0 lb 6.1 lb Area of impact 3.14 in.
2 3.14 in.2 A p  = Missile weight 183 psf 279 psf        Impact Area 
 
Velocity 75 ft/s 120 ft/s
 
Depth of penetration in a 2-ft thickness barrier 0.006 in. 0.205 in.
 
FNP FSAR-3
 
REV 21  5/08 TABLE 3.5-5 CHARACTERISTICS OF OTHER MISSILES POSTULATED WITHIN CONTAINMENT
 
Reactor Coolant Pump Instrument Temperature  Well of Pressurizer
 
Characteristics  Element    Pressurizer  Heaters 
 
Weight 0.25 lb 5.5 lb 15 lb
 
Discharge area 0.50 in.
2 0.442 in.
2 0.80 in.2 Thrust area 0.50 in.
2 1.35 in.2 2.4 in.2 Impact area 0.50 in.
2 1.35 in.2 2.4 in.2 A p = Missile weight 72 psf 587 psf 900 psf      Impact Area   
 
Velocity 260 ft/s 100 ft/s 55 ft/s
 
Depth of penetration in a 2-ft thick 0.23 in. 0.30 in. 0.14 in.
barrier   
 
FNP FSAR-3
 
REV 21  5/08 TABLE 3.5-6 MISSILE BARRIERS AWAY FROM CONTAINMENT
 
Missiles Barriers Rod drive power supply motor- Steel protective shield over generator set flywheel flywheel
 
FNP FSAR-3
 
REV 21  5/08 TABLE 3.5-7 ROD DRIVE POWER SUPPLY MOTOR-GENERATOR SET FLYWHEEL MISSILE CHARACTERISTICS
: a. Governing the design of steel protective shield. Missile Missile Missile Missile Missile Missile Missile  1  2  3  4  5  6  7          Flywheel fragment angle (degrees) 90 120 133 134 (a) 135 150 180 Flywheel fragment weight (lb) 329 438 486 489 493 548 657
 
Rotational speed at failure (rpm) 2,700 2,700 2,700 2,700 2,700 2,700 2,700 Rotation speed at failure, percent  150 150 150 150 150 150 150 of operating speed       
 
Initial velocity (ft/s) 249 229 219 218 217 204 176
 
Initial translational energy 0.317 0.357 0.361 0.361 0.360 0.355 0.317 (ft-lb x 10
: 6)
FNP-FSAR-3 3.6-1 REV 21  5/08 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING This section describes the design bases and protective measures which are used to ensure that
 
the containment, vital equipment, and other vital structures are adequately protected from
 
dynamic effects associated with the postulated rupture of piping, including the reactor coolant system. 
 
Relative to interfaces for the piping systems described in this section, Bechtel designed and
 
performed layouts for all auxiliary piping systems except the reactor coolant loop (RCL). The RCL is a generic layout except for the latitude of relocating, along the piping length, the various
 
branch nozzles. This was jointly designed and approved by Westinghouse and Bechtel. Bechtel
 
provided Westinghouse with the appropriate layout information to permit Westinghouse to
 
analyze the RCL piping and other Class I branch piping for which Westinghouse is responsible. 
 
Postulated breaks in the RCL, except for Accumulator and Residual Heat Removal branch line
 
connections, have been eliminated from the structural design basis for both Unit 1 and Unit 2, as
 
allowed by the revised GDC-4.
(3)  The elimination of these breaks is the result of the application of leak-before-break technology as presented in reference 4 and 5. Leak-before-break
 
technology was evaluated in NRC miscellaneous letters dated 1/15/92 and 8/12/91 and satisfies
 
the NRC acceptance criteria contained in NUREG 1061, Volume 3, dated November 1984, and
 
GDC 4.
 
3.6.1 SYSTEMS IN WHICH DESIGN BASIS PIPING BREAKS ARE POSTULATED TO OCCUR  Design basis piping breaks and piping cracks are postulated to occur in the RCLs and in all lines
 
outside the reactor coolant piping system that have a normal operating temperature above 200°F and a normal operating pressure above 275 psig. 
 
PIPING SYSTEMS INSIDE CONTAINMENT
 
Piping systems inside containment in which piping breaks and cracks are postulated to occur are
 
as follows: 
 
A. RCLs (branch connections only, as discussed above). B. ASME III Class 1 branch lines from the reactor coolant system.
C. Following ASME III Class 2 and 3 lines.
* Main steam lines (3).
* Main feedwater lines (3).
* Steam generator blowdown lines (3).
* Normal charging line (CVCS).
FNP-FSAR-3 3.6-2 REV 21  5/08
* Alternate charging lines (CVCS).
* Charging line to pressurizer spray lines (CVCS).
* Letdown line (CVCS).
* Reactor coolant pump seal water (CVCS). 
 
PIPING SYSTEMS OUTSIDE CONTAINMENT
 
The piping systems outside containment in which piping breaks are postulated to occur are
 
discussed and outlined in appendix 3K. 
 
3.6.2 DESIGN BASIS METHODS AND PIPING BREAK CRITERIA 3.6.2.1  Criteria The design basis for the postulated pipe rupture includes not only the break criteria, but also the
 
criteria to protect other piping and vital systems from the effects of the postulated rupture. 
 
A loss of reactor coolant accident is assumed to occur for a pipe break down to the restraint of
 
the second normally open automatic isolation valve (Case II in figure 3.6-1) on outgoing lines, (a) and down to and including the second check valve (Case III in figure 3.6-1) on incoming lines
 
normally with flow. A pipe break beyond the restraint or second check valve will not result in an
 
uncontrolled loss of reactor coolant if either of the two valves in the line are closed. 
 
Accordingly, both of the automatic isolation valves are suitably protected and restraints
 
positioned as close to the valves as possible so that a pipe break beyond the restraint will not
 
jeopardize the integrity and operability of the valves. This criterion takes credit for only one of
 
the two valves performing its intended function. For normally closed isolation or incoming check
 
valves (Case I and IV in figure 3.6-1) a loss of reactor coolant accident is assumed to occur for
 
pipe breaks on the reactor side of the valve. 
 
Branch lines connected to the reactor coolant system are defined as "large" for the purpose of
 
these criteria as having an inside diameter greater than 4 in. up to the largest connecting line, generally the pressurizer surge line. Rupture of these lines results in a rapid blowdown from the
 
reactor coolant system and protection is basica lly provided by the accumulators and the low head safety injection pumps (residual heat removal pumps). 
: a. It is assumed that motion of the unsupported line containing the isolation valves could cause
 
failure of the operators of both valves to function.
FNP-FSAR-3 3.6-3 REV 21  5/08 Branch lines connected to the reactor coolant system are defined as "small" if they have an inside diameter equal to or less than 4 in. This size is such that emergency core cooling
 
system analyses have shown acceptable peak clad temperature results for a break area of up to 12.5 in.
2 corresponding to 4-in. inside diameter piping. 
 
Engineered safety features are provided for core cooling and boration, pressure reduction, and
 
activity confinement in the event of a loss of reactor coolant or steam or feedwater line break incident, to ensure that the public is protected in accordance with 10 CFR 100 guidelines. 
 
These safety systems have been designed to provi de protection for a reactor coolant system pipe rupture of a size up to and including a double ended severance of the reactor coolant
 
system main loop. 
 
In order to assure the continued integrity of the vital components and the engineered safety
 
systems, consideration is given to the consequential effects of the pipe break itself to the
 
extent that: 
 
A. The minimum performance capabilities of the engineered safety systems are not reduced below that required to protect against the postulated break. 
 
B. The containment leaktightness is not decreased below the design value. 
 
C. Propagation of damage is limited in type and/or degree to the extent that: 
: 1. A pipe break which is not a loss of reactor coolant will not cause a loss of reactor coolant or steam or feedwater line break. 
: 2. A reactor coolant system pipe break will not cause a steam feedwater system pipe break and vice versa. 
 
In the unlikely event that one of the small pressurized lines should fail and result in a loss-of-
 
coolant accident, the piping must be restrained or arranged to meet the following criteria in
 
addition to A through C above: 
 
A. Break propagation must be limited to the affected leg; i.e., propagation to the other leg of the affected loop and to other 
 
loops will be prevented. 
 
B. Propagation of the break in the affected leg is permitted but must be limited to a total break area of 12.5 in.
2 (4-in. inside diameter). The exception to this case is when the initiating small break is the high head 
 
safety injection line. Further propagation is not permitted for this case. 
 
C. Damage to the high head safety injection lines connected to the other leg of the affected loop or to the other loops must be prevented. 
 
D. Propagation of the break to high head safety injection line connected to the affected leg must be prevented if the line break results in a loss of 
 
core cooling capability due to a spilling injection line.
FNP-FSAR-3 3.6-4 REV 21  5/08 3.6.2.2  Reactor Coolant Loops Pipe break locations are postulated in the reactor coolant loop using the methods and criteria in
 
WCAP 8082 (2). The applicability of WCAP 8082 to the Farley Plant has been verified by reactor coolant loop analysis. The results of the analysis indicate the following: 
 
A. All locations of the reactor coolant loop piping are below 2.4 S m in stress intensities and have fatigue usage factors that are less than 0.2 (the
 
fatigue usage factors are, in fact, less than 0.1), except for the locations
 
identified in WCAP 8082. The report thus applies to this plant.
Consequently, no break locations other than those identified in WCAP 8082 need to be postulated. 
 
B. The component displacements at support interfaces that are listed in WCAP 8082 are typical displacements and were included in the report to indicate the relative magnitude of displacements at the interface. The
 
displacements at the interfaces for the Farley plant are of the same
 
relative magnitude as the displacements provided in WCAP 8082.
With respect to the component displacements at the design basis break locations, WCAP 8082 assumes, for purposes of analysis, double-ended area breaks for all points where circumferential breaks are postulated
 
except at the reactor vessel nozzles. At the reactor vessel nozzles, circumferential breaks with limited break area are postulated since the
 
concrete shield wall prevents the development of double-ended area
 
breaks. The displacements at these two design basis break locations are
 
100 square inches. However, displacements at all other points are not
 
required for review since the break areas employed at these points are
 
absolute maximum.  (See paragraph 6.2.1.3.10.B for further justification.) 
 
Additional information for the reactor coolant loop analysis relative to methods and analytical
 
procedures, jet impingement forcing functions, discharge flow areas, and break opening
 
areas/displacements is contained in attachment F to appendix 3K and paragraph 6.2.1.3.10.
 
According to WCAP-8082 (Reference 2), eleven break locations were postulated in the RCS primary loop piping including breaks at the nozzle welds of three large branch lines (Accumulator, RHR and Surge lines). Nine of these break locations including the one at the
 
surge line branch connection have subsequently been e liminated from the structural design basis through the application of leak-before-break (LBB) technology. Two postulated break locations
 
at the nozzle welds of two branch lines (Accumulator and Residual Heat Removal) still exist. 
 
The detailed fracture mechanics techniques used in this evaluation are discussed in references 4
 
and 5.
 
3.6.2.3  Class 1 Branch Lines Pipe break locations postulated for ASME III Class 1 Branch Lines meet the intent of Regulatory
 
Guide 1.46. The specific criteria are as follows: 
 
FNP-FSAR-3 3.6-5 REV 21  5/08  1. ASME Section III Code Class 1 piping (a) breaks should be postulated to occur at the following locations in each piping run (b) or branch run: 
: a. The terminal ends. 
: a. Piping is pressure retaining components consisting of straight or curved pipe and pipe fittings (e.g., elbows, tees, and reducers).
: b. A piping run interconnects components such as pressure vessels, pumps, and valves that act
 
to restrain pipe movement beyond that required for design thermal displacement. A branch run
 
differs from a piping run only in that it originates at a piping intersection as a branch of the main
 
pipe run.
FNP-FSAR-3 3.6-6 REV 21  5/08  b. At intermediate locations between terminal ends where the primary plus secondary stress intensities (circumferential or longitudinal) derived on an
 
elastically calculated basis under the loadings associated with specified
 
seismic events (a) and operational plant conditions (b) exceed 2.4 S m for austenitic steel. 
: c. At intermediate locations between terminal ends where the cumulative usage factor U (c) derived from the piping fatigue analysis under the loadings associated with specified seismic events and operational plant
 
conditions exceeds 0.1.
 
Postulated breaks in the pressurizer surge line have been eliminated from the structural design
 
basis through the application of leak-before-break (LBB) technology. The detailed fracture
 
mechanics techniques used in this evaluation are discussed in reference 5. Application of LBB
 
allows the elimination of the dynamic effects of pipe rupture for these break locations.
 
The requirement to postulate arbitrary intermediate breaks has been eliminated from the
 
structural design basis (including resultant dy namic and environmental effects) as allowed by NRC Generic Letter 87-11, "Relaxation in Arbitrary Intermediate Pipe Rupture Requirements."
 
3.6.2.4  Class 2 and 3 Lines Methods and criteria for ASME III Class 2 and 3 piping lines for piping systems inside and
 
outside containment are outlined in appendix 3K. 
 
Specific location criteria for break points are as follows: 
 
ASME Section III Code Class 2 and 3 piping breaks will be postulated to occur at the following
 
locations in each piping run or branch run: 
: a. Specified seismic events are earthquakes that produce at least 50 percent of the vibratory
 
motion of the safe shutdown earthquake. 
: b. Operational plant conditions include normal reactor operation, upset conditions (e.g.,
anticipated operational occurrences), and testing conditions.
: c. U is the cumulative usage factor as specified in Section III of the ASME Boiler and Pressure
 
Vessel Code. 
 
FNP-FSAR-3 3.6-7 REV 21  5/08  a. The terminal ends. 
: b. At intermediate locations between terminal ends where either the circumferential or longitudinal stresses derived
 
on an elastically calculated basis under the loadings
 
associated with specified seismic events and operational
 
plant conditions exceed 0.8(S h + S A).(a)
The requirement to postulate arbitrary intermediate breaks has been eliminated from the
 
structural design basis (including resultant dy namic and environmental effects) as allowed by NRC Generic Letter 87-11, "Relaxation in Arbitrary Intermediate Pipe Rupture Requirements."
 
3.6.2.5  Break Types The following types of breaks will be postulated at the locations identified in subsections 3.6.2.3
 
and 3.6.2.4. 
: a. Longitudinal breaks will be considered only in piping runs and branch runs 4 in. nominal pipe size and larger. Circumferential
 
breaks will be considered only in piping runs and branch runs
 
exceeding 1 in. nominal pipe size. 
: b. The local stress field at the break location in the pipe will determine whether a circumferential  or longitudinal break or
 
both will be postulated. 
: c. Longitudinal breaks are parallel to the pipe axis and oriented at any point around the pipe circumference unless a
 
preferential direction can be justified by analysis. The break
 
area is equal to the sum of the effective cross sectional flow
 
area upstream of the break location and downstream of the
 
break location or is equal to a break area determined by test
 
data which defines the break geometry. Dynamic forces
 
resulting from such breaks will be normal to the pipe axis. 
: a. S h is the stress calculated by the rules of NC-3600 and ND-3600 for Class 2 and 3 components, respectively, of the ASME Code Section III Winter 1972 Addenda.
 
S A is the allowable stress range for expansion stress calculated by the rules of NC-3600 of the ASME Code, Section III, or the USA Standard Code for Pressure Piping, ANSI B31.1.0-1967.
FNP-FSAR-3 3.6-8 REV 21  5/08  d. Circumferential breaks are perpendicular to the pipe axis, and the break area is equivalent to the internal cross sectional
 
area of the ruptured pipe. Reduced cross sectional opening
 
areas can be used if the pipe motion is physically restricted by
 
pipe restraints or other restraining structures. Dynamic forces
 
resulting from such breaks are assumed to separate the piping
 
axially and cause whipping in any direction normal to the pipe
 
axis.
3.6.3 DESIGN LOADING COMBINATIONS In determining pipe break locations, a worst case combination of the following load conditions is
 
considered--thermal expansion, deadweight, seismic, seismic anchor movement, internal
 
pressure, and load due to steam hammer, relief thrust, etc., where applicable. 
 
3.6.3.1  Reactor Coolant Piping As described in section 5.2, the forces associated with rupture of reactor piping systems are
 
considered in the design of supports and restraints in order to ensure continued integrity of vital
 
components and engineered safety features. 
 
3.6.3.2  Class 1 Branch Lines For Class 1 piping, the design loading combinations and design stress limits are given in section
 
5.2. 
 
3.6.3.3  Class 2 and 3 Lines Stress analysis results used in determining pipe break locations for Class 2 and 3 lines outside
 
containment are given in the applicable piping stress calculations. 
 
Stress analysis results used in determining pipe break locations for Class 2 and 3 lines inside
 
containment are given in the applicable piping stress calculations for the following systems: main
 
steam, main feedwater, CVCS normal and alternate charging lines, CVCS letdown lines and
 
steam generator blowdown lines. Thrust loads for Class 2 and 3 lines inside containment are
 
given in Table 3.6-2.
 
The piping in the charging line to the pressurizer spray and the reactor coolant pump seal water
 
lines are field run. The stress analyses of these lines are found in the applicable piping stress
 
calculation for the subject piping. 
 
FNP-FSAR-3 3.6-9 REV 21  5/08 3.6.4 DYNAMIC ANALYSIS The dynamic analyses and orientations applicable to the main reactor coolant loop piping system
 
are presented in WCAP 8082.
The dynamic analyses, postulated pipe break location, and orientations for lines outside the
 
reactor coolant pressure boundary and for Class 2 and 3 lines inside containment are discussed
 
in appendix 3K, High Energy Line Pipe Break (Outside Containment). 
 
Class 1 system branch lines and Class 2 and 3 lines inside containment are analyzed using the
 
methods outlined in appendix 3K. This analysis takes into account the movement of supports, forcing functionings, dynamics, and design criteria. 
 
3.6.4.1  Postulated Break Locations Reactor Coolant Loops 
 
The breaks postulated in the reactor coolant loop for the Farley Nuclear Plant are identical to
 
those postulated in WCAP 8082 (2), except as explained in paragraph 3.6.2.2. 
 
Class 1 Branch Lines 
 
The analysis methods used by Westinghouse for Class 1 branch lines are described below: 
: 1. Deadweight 
 
The deadweight loading is defined as consisting of the dry weight of the piping and the weight of the water contained in piping during normal
 
operation. 
: 2. Thermal Expansion 
 
The thermal movements of the terminal ends are considered in addition to the thermal expansion of the branch piping. 
 
The cold and hot moduli of elasticity, the coefficient of thermal expansion at the metal temperature, external movements transmitted to the piping as
 
described above, and the temperature ri se above the ambient temperature define the required input data to perform the flexibility analysis for thermal
 
expansion. 
: 3. Earthquake Loads 
 
The intensity and character of the earthquake motion that produces forced vibration of equipment mounted within the containment building are
 
specified in terms of the floor response spectrum curves at various
 
elevations within the containment building. The 1/2 SSE and SSE floor
 
response spectrum curves for earthquake motions are given in reference 9, FNP-FSAR-3 3.6-10 REV 21  5/08 section 5.2. Code Case N-411 Damped Response Spectra may be used for piping as referenced in Section 3.7.
: 4. Pressure 
 
The design and steady state operating pressures are evaluated in accordance with the requirements of the ASME Section III code. 
: 5. Transients 
 
In addition to the deadweight, thermal expansion, and seismic loads, the ASME code requires that the through-wall temperature distribution be
 
evaluated for Class 1 piping. A heat transfer analysis is performed using
 
the anticipated transients to determine this temperature distribution. 
: 6. Analytical Methods 
 
The static and dynamic structural analyses assume linear elastic behavior and employ the displacement (stiffness) matrix method and the normal
 
mode theory for lumped-parameter, multi-mass structural representation to
 
formulate the solution. The complexity of the physical system to be
 
analyzed requires the use of a computer for solution based on an idealized
 
mathematical model. 
: 7. Effect of Design Basis Accident (DBA) 
 
The motions induced at the reactor coolant loop branch piping nozzle interface as a result of a rupture in the primary coolant are applied as
 
terminal displacements at the nozzle connections. 
: 8. Static Load Solutions 
 
The static solutions for deadweight, thermal expansion conditions are obtained by using the WESTBYN computer program. The computer input
 
consists of the piping model, stiffness matrices representing various
 
supports for static behavior, and the appropriate load condition. Coordinate
 
transformations for rotation from the local or support coordinate system to
 
the global system are applied to the stiffness matrices prior to their input. 
: 9. Normal Mode Response Spectral Seismic Load Solution 
 
The stiffness matrices representing various supports for dynamic behavior are incorporated into the model after transformations for rotation from local
 
to the RCL global system. The response spectra for the 1/2 SSE or SSE load case are applied along the X, or Z, and Y axes simultaneously. From
 
the input data, the overall stiffness ma trix of the three-dimensional system is generated. The stiffness matrix is manipulated to obtain a reduced
 
stiffness matrix, associated with the mass points only. The reduced matrix
 
is inverted to give the flexibility matrix of the system. A product matrix (also FNP-FSAR-3 3.6-11 REV 21  5/08 known as the dynamic matrix) formed by the multiplication of the flexibility and mass matrices is used to solve for the natural frequencies and normal
 
modes by the modified Jacobi method. The modal participation factor
 
matrix is computed and combined with the appropriate seismic response
 
spectra values to give the amplitude of the modal coordinate for each
 
mode. Then the forces, moments, deflections, rotations, support structure
 
reactions, and stresses are calculated for each significant mode. The total
 
seismic response is computed by combining the contributions of the
 
significant modes by the square-root-of-the-sum-of-the-squares method. 
 
The method of analysis is presented in appendix 3L. 
 
Postulated break locations for Class 1 branch lines are developed using the criteria outlined in
 
subsection 3.6.2.3. The stress analysis methodology for Class 1 branch lines is presented in
 
appendix 3L. Stress analysis results are found in the applicable stress calculations for the
 
following systems:  reactor coolant system drain, pressurizer auxiliary spray, and reactor coolant pump seal water injection. These stress results are utilized by the criteria of subsection 3.6.2.3
 
in determining pipe break locations.
 
Class 2 and 3 Lines 
 
Postulated break locations for the following Class 2 and 3 piping systems are developed using
 
the criteria outlined in subsection 3.6.2.4:  main steam, main feedwater, CVCS normal and
 
alternate charging, and CVCS letdown line.
 
Stress analysis results of the subject systems are utilized by the criteria of subsection 3.6.2.4 in
 
determining pipe break locations.
 
3.6.5 PROTECTIVE MEASURES The fluid discharged from the ruptured piping will produce reaction and thrust forces in the RCL
 
system. These effects are considered in ensuring the continued integrity of the vital components
 
and the engineered safety features. 
 
To accomplish this in the design, a combination of component restraints, barriers, and layout are
 
utilized to ensure that for a loss-of-coolant or steam feedwater line break, propagation of damage
 
from the original event is limited, and the co mponents as needed are protected and available. 
 
Protective measures for high energy lines outside the reactor coolant pressure boundary are
 
discussed in appendix 3K, High Energy Line Pipe Break (Outside Containment). 
 
3.6.5.1  Pipe Whip Restraints Reactor Coolant Loop
 
Large branch lines attached to the reactor coolant loop piping are restrained to meet the
 
following criteria:
FNP-FSAR-3 3.6-12 REV 21  5/08  a. Propagation of the break to the unaffected loops must be prevented to ensure the delivery capacity of the accumulators and low head pumps. 
: b. Propagation of the break in the affected loop is permitted to occur but must not exceed 20 percent of the area of the line which initially ruptured. This
 
criterion has been voluntarily applied so as not to increase substantially the
 
severity of the loss-of-coolant. c. Where restraints on the lines are necessary in order to prevent impact on and subsequent damage to the neighboring equipment or piping, restraint
 
type and spacing will be chosen so that a plastic hinge on the pipe at the
 
two support points closest to the break is not formed. 
 
Additional discussion of pipe restraint design criteria is found in reference 2.
 
Class 1 Branch Lines
 
Pipe whip restraints for Class 1 branch lines are shown on applicable civil design drawings. 
 
Class 2 and 3 Lines
 
Pipe whip restraint design criteria along with the description of a typical pipe whip restraint for
 
Class 2 and 3 lines are given in appendix 3K. 
 
Pipe whip restraint locations for Class 2 and 3 lines are shown on applicable civil design
 
drawings.
 
3.6.5.2  Jet Impingement In addition to pipe restraints, barriers and layout are used to provide protection from blowdown
 
jet and reactive forces on cabling, instrumentation, and equipment necessary for safe shutdown
 
of the reactor. 
 
In addition, the refueling cavity walls, various structural beams, and the operating floor enclose
 
each reactor coolant loop into a separate compartment, thereby preventing an accident, which
 
may occur in any loop, from affecting another loop or the containment liner. The portion of the
 
steam and feedwater lines within the containment have been routed behind barriers which
 
separate these lines from all reactor coolant piping. 
 
In reviewing the mechanical aspects of these lines, it has been demonstrated by Westinghouse
 
Nuclear Energy System tests that lines hitting equal or larger size lines of same schedule will not cause failure of the equal or larger line; e.g., a 1-inch line, should it fail, will not cause
 
subsequent failure of a 1 in. or larger size line. The reverse, however, is assumed to be
 
probable; i.e., a 4-in. line, should it fail and whip as a result of the fluid discharged through the
 
line, could break smaller size lines such as neighboring 3 in. or 2 in. lines.
 
Bending of a broken stainless steel pipe section such as that used in the reactor coolant system
 
branch lines does not cause this section to become a missile. This design basis has been FNP-FSAR-3 3.6-13 REV 21  5/08 demonstrated by performing bending tests on large and small diameter, heavy and thin walled stainless steel pipes. 
 
RC Loop Jet Impingement
 
Jet impingement loads on the primary equipment and supports due to the breaks postulated in
 
RCL are based upon the dynamic piping displacem ent response determined from the loop LOCA analyses. The jet impingement loads on the adjacent structures are evaluated using the
 
methods outlined in appendix 3K. 
 
Class 1, 2 and 3 lines Jet Impingement
 
Jet impingement methods and analyses for Class 1, 2, and 3 lines for both inside and outside
 
containment are outlined in appendix 3K. 
 
3.6.5.3  Separation and Redundancy The separation and redundancy of equipment and safety features that have been designed for
 
protection against the effects of pipe break are outlined in appendix 3K. 
 
3.6.6 STRUCTURAL ANALYSIS 3.6.6.1  Outside Containment The structural analysis for high energy pipe breaks outside containment is discussed in
 
attachment G, appendix 3K. 
 
3.6.6.2  Inside Containment The containment internal structures, which have the most critical loading conditions during a pipe
 
break event, consist
 
of: 
: 1. The primary shield wall. 
: 2. The secondary shield wall, which encloses the steam generator and pressurizer compartments. 
: 3. The floor slab at elevation 129 ft 0 in. 
 
The geometry of these structures is described in subsection 3.8.3.1. 
 
The structural loads and loading combinations for each postulated break are in accordance with
 
Sections B and C of "Structural Design Criteria for Evaluating the Effects of High Energy Pipe
 
Breaks on Category I Structures Outside the Containment," Document (B) of the NRC.
FNP-FSAR-3 3.6-14 REV 21  5/08 The finite element method was employed for the analysis and Bechtel's SAP1.8 computer program was utilized for performing the finite element analysis. A description of the computer
 
program is in attachment G, appendix 3K. 
 
3.6.6.2.1 Finite Element Model Three finite element models were developed for the analysis of the structures under
 
investigation. Due to the symmetry or similari ty of the geometry and the loads, only parts of each of these structures were modeled. The results and conclusions obtained from the analysis of
 
these models will also apply to the other parts of the structures not included in the models. 
 
Three-dimensional brick elements were used for the primary shield wall. Plate elements were
 
used for the secondary shield walls (steam generator compartment walls), slab, and the
 
pressurizer compartment walls. The boundary conditions include partial or complete fixity
 
against displacement and rotation, depending upon the structure boundary restraint conditions
 
under thermal and other loadings. 
 
3.6.6.2.2 Results of Analysis Table 3.6-7, summarizing the results of this linear, elastic, finite element analysis, indicates that
 
the walls and slab are sufficiently strong to resist the various combinations of loads associated
 
with a high energy pipe break inside the containment, with a margin of safety provided by the
 
load increases and load factors used in the analysis.(a) 3.6.6.3  Pipe Whip Restraint Design The analytical approach and design of the pipe whip restraints are described in attachment B, appendix 3K. 
: a. Reference section 6.2.
FNP-FSAR-3 3.6-15 REV 21  5/08 REFERENCES 
: 1. Szyslowski, J. J. and Salvatori, R., "Determination of Design Pipe Breaks for the Westinghouse Reactor Coolant System," WCAP-7503 , Rev 1, February 1972. 
: 2. PWR Staff, "Westinghouse Technical Position on Discrete  Break Locations and Types for the LOCA Analysis of the  Primary Coolant Loop," WCAP-8082 , May 1973. 
: 3. Modification of GDC-4, Final Rule, 52 FR 41288, October 27, 1987.
: 4. "Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the Joseph M. Farley Units 1 and 2 Nuclear Power Plants,"
(Proprietary Class 2) WCAP-12825 , January 1991.
: 5. "Technical Justification for Eliminating Pressurizer Surge Line Rupture from the Structural Design Basis for Farley Units 1 and 2," (Proprietary Class 2) WCAP-12835 , April 1991.
 
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.6-2 THRUST LOADS DUE TO A FULL AREA PIPE RUPTURE (CLASS 2 AND 3 PIPING)
Temperature Pressure Thrust Force System Line Size    (F)      (PSIG)      (lbs)
Main steam 32 in. 547 1005 285,000
 
Main feedwater 14 in. 442 1055 109,400
 
CVCS normal and 3 in. 485 2350 12,000 alternate charging   
 
lines          CVCS letdown line 3 in. 550 2350 12,000 from Class 1 interface to regenerative heat exchanger   
 
CVCS letdown line 2 in. 380 2250 4700 from regenerative heat exchanger to containment penetration   
 
    (before flow orifice) 3 in. 380 2250 12,000
 
CVCS letdown line 2 in. 380 550 3150 from regenerative heat exchanger to containment penetration   
 
    (after flow orifice) 3 in. 380 550 6900
 
Steam generator blowdown 2 in. 547 1055 4730
 
line     
 
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.6-7 ANALYSIS RESULTS Critical  Critical  Maximum  Allow  Thickness Postulated  Load Pressure Pressure P allow Structure  (in.)  Pipe Break Combination    P      P allow    P          Primary 108 Hot Leg D + L + Ta 124 150 1.21 Shield  Rupture + Ra + 1.5P 615 667 1.08 Wall     
 
Steam 42 Cold Leg D + L + Ta 57 60 1.05 Generator  Rupture + Ra + 1.5P Compartment Wall     
 
Pressurizer 24 Spray Line D + L + Ta 20 22 1.10 Compartment  Rupture + Ra + 1.5P Wall       
 
Slab 36 Cold Leg D + L + Ta 57 58 1.02 El 129'-0" 36 Rupture + Ra + 1.5P Notes:  1. Loads and load combinations are in accordance with Sections B and C of "Structural Design Crit eria for Evaluating the Effects of High-Energy Pipe Breaks on Category I Structures Outside t he Containment," Document (B) of the NRC.
: 2. Two pressure values are given for the primary shield wall. The first one is the differential pressure uniformly applied inside the reactor cavity. The second one is the possible localized pressure with in the inspection chamber at the reactor nozzles.
: 3. Maximum pressure values are the same as those used in the critical load combinations, and were obtained by multiplying the c alculated peak  pressure by a factor 1.4 x 1.2 = 1.68 to account for uncertainty and the dynamic factors, respectively. 
 
REV 21  5/08 LOSS OF REACTOR COOLANT ACCIDENT BOUNDARY LIMITS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.6-1
 
FNP-FSAR-3
 
3.7-1 REV 21  5/08 3.7 SEISMIC DESIGN The criteria for determining the adequacy of Seismic Category I mechanical and electrical
 
equipment for the Farley Nuclear Plant are described in various areas of the FSAR. In some cases, the criteria are specified in general terms to require verification by tests or analyses. In other cases, more specific criteria are specified such as verification in accordance with IEEE Standard 344-1971.
 
Historically, it should be noted that the FNP Unit 2 seismic qualification program, i.e., IEEE 344-71
 
type qualification, was previously audited by the NRC's Seismic Qualification Review Team (SQRT). It was concluded in NUREG-0117 Supplement No. 5 (dated March, 1981) Safety
 
Evaluation Report related to the operation of Unit 2 that "the licensee's seismic qualification
 
program provides reasonable assurance that the seismic category I mechanical and electrical
 
equipment is adequately qualified, meets the applicable requirements of General Design Criterion
 
2, and is, therefore, acceptable for full-power operation."
 
By letter dated February 19, 1987, the NRC issued Generic Letter (GL) 87-02, "Verification of
 
Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved
 
Safety Issue (USI) A-46."  On May 22, 1992, the NRC issued GL 87-02 Supplement 1. As
 
documented in NUREG-1211, "Regulatory Analysis for Resolution of Unresolved Safety Issue A-46, Seismic Qualification of Equipment in Operating Plants," GL 87-02 is applicable to Farley Nuclear
 
Plant (FNP) Unit 1, since Unit 1 had not previously been audited by the SQRT. Southern Nuclear
 
Operating Company (SNC) replied to GL 87-02 by letter dated September 10, 1992. The SNC letter included a commitment to use the Seismic Qualification Utility Group (SQUG) methodology as
 
documented in the Generic Implementation Procedur e (GIP) for resolution of seismic issues identified in GL 87-02 for FNP Unit 1. The SQUG methodology is based on application of
 
earthquake experience data to verify the seismic adequacy of equipment. The seismic evaluation
 
for FNP Unit 1 was completed, and the results were documented in a document entitled
 
"Unresolved Safety Issue A-46 Summary Report."
This document was submitted to the NRC by letter dated May 18, 1995 as a 10 CFR 50.54(f) response. SNC received an SER dated July 9, 1998, concerning FNP Unit 1 USI A-46 resolution and it stated that SNC's USI A-46 program
 
implementation resulted in safety enhancement s beyond the original licensing basis and SNC actions provide sufficient basis to close the USI A-46 review at the facility.
 
3.7.1 SEISMIC INPUT Geologic and seismologic surveys of the si te have been made to establish two "design earthquakes" with different intensities of ground motion. These are the 50 percent safe shutdown
 
earthquakes (1/2 SSE) and safe shutdown earthquakes (SSE) with different intensities of ground
 
motion. The 1/2 SSE, previously called operating basis earthquake (OBE) in the Preliminary Safety
 
Analysis Report, is postulated to be the earthquake that could be expected to occur at the site
 
during the operating life of the plant. The SSE represents the strongest earthquake that is
 
hypothetically postulated to occur during an infinite period. 
 
The plant site geologic and seismologic investigations and recommendations are discussed in
 
section 2.5. As specified in the following paragraphs, the intensity postulated to occur at the site for
 
both the 1/2 SSE and SSE is defined from the history of seismic activity in the area around the site.
 
FNP-FSAR-3
 
3.7-2 REV 21  5/08 3.7.1.1  Design Response Spectra The safe shutdown earthquake and 50 percent safe shutdown earthquake are specified in terms of
 
a set of idealized, smooth curves, called the design spectra because they specify a range of values
 
for two of the important properties of an earthquake ground motion, i.e., the maximum ground
 
acceleration and the frequency distribution.
 
The SSE is that earthquake which produces the vibratory ground motion for which Category I
 
structures, systems and components are designed to remain functional. Category I structures, systems, and components will also be designed to withstand the effects of vibratory motion of at
 
least 50 percent of the safe shutdown earthquakes in combination with other appropriate loads. 
 
Figure 3.7-1 shows the 1/2 SSE spectra for 0, 0.5, 1.0, 2.0, 3.0, and 5.0 percent of critical damping, with a horizontal ground peak acceleration of 0.05 g and vertical ground acceleration of 0.033 g. 
 
Figure 3.7-2 shows the SSE spectra for 0, 0.5, 1.0, 2.0, 3.0 and 5.0 percent of critical damping, with a
 
horizontal ground peak acceleration of 0.10 g and vertical ground acceleration of 0.067 g. 
 
The bases for the selection of 1/2 SSE and SSE ground accelerations are presented in section 2.5. 
 
These design spectra are obtained by modifying Newmark's curves.
(1) To obtain these curves, variations in site conditions, foundation properties, and amplification factors of previous distant and
 
nearby earthquakes (where the location of the origin is known), were taken into account. One of the
 
parameters defining the design spectra is the spectrum amplification ratio, which is the ratio of the
 
peak spectrum acceleration to the ground acceleration for a particular magnitude of damping. For
 
this site, a ratio of 3.5 is used for the period range of 0.15 to 0.50 second of the 2 percent critical
 
damping design spectrum. 
 
Section 2.5.1(6) of BC-TOP-4 (Rev. 1) discusses the derivation of the shape of the design spectra. 
 
These design spectra are based on the existing strong motion earthquake ground records of
 
various durations, and are recorded at sites' having different geologic conditions, epicentral
 
distances, and their associated spectral amplification factors. 
 
A discussion of the effects of historical seismic events on the site is given in section 2.5. Because
 
the modified design spectra are based on the properties of several strong motion records of the
 
earthquakes recorded at sites of various geologic conditions and epicentral distances, the effects of
 
duration, distance, and depth are automatically taken into account. 
 
3.7.1.2  Design Response Spectra Derivation The synthesized time history accelerogram, normalized to 0.10 g, is shown in figure 3.7-3 (SSE
 
synthetic time history). The same synthesized time history accelerogram, normalized to 0.05 g, is
 
used for 1/2 SSE analysis. In the vertical direction, the same accelerogram is normalized to 0.067 g
 
for SSE and 0.033 g for 1/2 SSE. 
 
The synthesized time history is obtained through modification of a time history selected from
 
simulated motion, for a total duration of 24 seconds with a uniform time increment of 0.01 second. 
 
FNP-FSAR-3
 
3.7-3 REV 21  5/08 The spectral values of the synthesized time history for the 1/2 SSE are equal to or greater than those on the 1/2 SSE ground response spectrum for 2 percent critical damping, as shown in figure
 
3.7-4. The spectral values of the synthesized time history for the SSE are equal to or greater than
 
those on the SSE ground response spectrum for 5 percent critical damping as shown in figure 3.7-
: 5. The damping values that are used for the generation of instructure response spectra are 2 and 5
 
percent for 1/2 SSE and SSE respectively for the prestressed concrete structures and reinforced
 
concrete structures as presented in table 3.7-1. Because of this fact, 2 and 5 percent critical
 
damping response spectra envelop the corresponding 1/2 SSE and SSE response spectra for the
 
range of 115 frequencies tabulated in table 3.7-2. These 115 frequencies are sufficient to describe
 
a response spectrum accurately for engineering purposes. 
 
3.7.1.3  Critical Damping Values The specific percentage of critical damping values used for Category I structures, systems, components, and soil are provided in table 3.7-1. 
 
In lieu of damping values given in Table 3.7-1, ASME Code Case N-411 damping values may
 
be used in piping analysis. Use of N-411 damping values will adhere to the conditions and
 
limitations contained in the Code Case and Regulatory Guide 1.84.
 
Energy dissipation in structures is generally represented by equivalent viscous dampers.
Evaluation of the damping coefficients is based on material, stress level, and the type of
 
connections used in the structural system. The damping values used in the response spectrum
 
design approach are those in table 3.7-1 which are based on a paper by N. M. Newmark and W. J.
 
Hall, "Seismic Design Criteria for Nuclear Reactor Facilities," and another paper by N. M. Newmark, "Design Criteria for Nuclear Reactors Subjected to Earthquake Hazards."  These values are used in
 
conjunction with the modal representation of the structure and are expressed as a percentage of
 
critical damping. 
 
The allowable stress levels for 1/2 SSE and working load combinations have been established as
 
normal code allowables. The allowable stress levels for SSE and yield load combinations have
 
been established at 85 percent of the compressive strength for concrete and 90 percent of yield for
 
steel which will maintain the materials of construction within the elastic range. 
 
3.7.1.4  Bases for Site Dependent Analysis Site dependent analysis is not used to develop the shape of the design response spectra. 
 
3.7.1.5  Soil Supported Category I Structures Outdoor tanks are the only major Category I structures founded on soil. The depth of soil over
 
bedrock (Lisbon formation) is about 55 ft. 
 
FNP-FSAR-3
 
3.7-4 REV 21  5/08 3.7.1.6  Soil Structure Interaction Soil structure interaction is taken into account in the dynamic analysis of the containment and other
 
Category I structures. For the lumped mass model, the soil stiffness of the foundation is
 
represented by introducing equivalent springs for the foundation medium, whereas the base mat is
 
assumed to be relatively rigid. Horizontal, vertical, and rocking spring constants are obtained from
 
the theory of a rigid base resting on an elastic half-space. 
 
A lumped mass model of a structure and the foundation is shown in figure 3.7-6. The constants k x and k are the equivalent spring stiffnesses for horizontal translation and rocking, respectively, and k z is the spring stiffness for vertical translation. The formulas for computing the equivalent spring stiffness for the cases of circular base mat and rectangular base mat are as follows: 
: a. Circular Base
 
Motion    Spring Constant
 
Horizontal    k X = 32(1-)GR        7-8    Rocking k = 8GR 3                3(1-)    Vertical k z = 4GR              1-  in which    = Poisson's ratio of foundation medium
 
G = shear modulus of foundation medium
 
R = radius of the circular base mat
 
FNP-FSAR-3
 
3.7-5 REV 21  5/08  b. Rectangular Base Motion    Spring Constant Horizontal   
()kGBL xx=+21    Rocking k G BL=1 2    Vertical k G BL zz=1  in which  and G are as defined previously, and
 
B = width of the base mat perpendicular to the direction of horizontal excitation
 
L = length of the base mat in the direction of horizontal excitation
 
x , , z = constants that are functions of the dimensional ratio,      L/B. See figure 3.7-7.
 
The shear modulus is obtained from the shear wave velocity and mass density of the soil using the
 
following relationship: 
 
G = (V s)2      144g
 
G = shear modulus, lb/in.
2      = density, lb/ft 3
Vs = shear wave velocity, ft/s
 
g = 32.174, ft/s 2
These springs are entered at the base of the model. 
 
3.7.2 SEISMIC SYSTEM ANALYSIS This subsection describes the seismic analysis performed for Category I structures. Category I
 
structures were designed using a dynamic analysis. 
 
FNP-FSAR-3
 
3.7-6 REV 21  5/08 3.7.2.1  Seismic Analysis Methods The seismic analysis methods applied to all Category I structures,  systems, and components are
 
identified in tables 3.7-3 and 3.7-4. 
 
Analysis of Category I structures, systems, and components is accomplished, where applicable, using the response spectra or time history approach, which utilizes the natural period, mode
 
shapes, and appropriate damping factors of the particular system. Where analytical methods of
 
analysis do not produce results of a significant confidence level or where analysis appears
 
undesirable, dynamic testing of equipment is used to ensure functional integrity. 
 
An important step in the seismic analysis of Category I systems or structures is the procedure used
 
for modeling. The system is represented by lum ped masses and a set of springs idealizing both the inertia and stiffness properties of the system. 
 
A complete dynamic analysis including soil structure interaction is performed on the containment
 
and all Category I structures to determine their behavior during an earthquake. The modal
 
response spectrum technique is used in the seismic design of all Category I structures. The
 
analysis is accomplished in the following 5 steps: 
: 1. Reduce the structure into a mathematical model in terms of lumped masses and stiffness coefficients. 
: 2. Obtain the natural frequencies and mode shapes of the model. 
: 3. Evaluate and determine the proper damping values. 
: 4. Determine the resulting internal forces on the structure, using the appropriate earthquake response spectra. 
: 5. Determine the spectrum response curves to be used in the analysis of the equipment located at all levels. The spectrum response curves are
 
generated at the modal mass points. 
 
In building the mathematical model, the locations for lumped masses are chosen at floor levels and
 
points considered of critical interest. Between mass points the structural properties are reduced to
 
uniform segments of cross-sectional area, effe ctive shear area and moments of inertia. 
 
The analysis utilizes the values from the ground response spectra for this site. Acceleration values
 
are selected for each mode, based on damping and natural frequency. The inertia forces, shears, moments, accelerations, and displacements of a sufficient number of the individual modes are
 
combined by taking the square root of the summation of the squares (SRSS) of the individual modal
 
values. Procedures for combining modal responses are presented in subsection 3.7.3.4. 
 
A separate analysis is made on the model for the horizontal and vertical earthquake accelerations, the vertical being two-thirds of the horizontal spectral values. The results from both analyses are
 
combined to obtain the critical response values. For structures, the model is analyzed separately
 
for both horizontal directions, and the results of each are combined separately with those from the
 
vertical analysis.
FNP-FSAR-3
 
3.7-7 REV 21  5/08 The mathematical model of the structure and results of seismic dynamic analysis for the Category I structures in the north to south, east to west, and vertical directions are shown in figures 3.7-8
 
through 3.7-21 and 3.7-23 through 3.7-56. 
 
The following information is obtained from the preceding analyses: 
: a. Inertial forces. 
: b. Accelerations. 
: c. Structural displacements. 
: d. Horizontal shears. 
: e. Horizontal moments. 
: f. Vertical axial forces. 
 
The mathematical model is analyzed for its frequencies and mode shapes; then the dynamic
 
response at the mass points is obtained by application of the synthesized time history earthquake at
 
the base of the structure. The input time history used is synthesized so that the computed spectral
 
values are greater than or equal to the spectral values of the design spectrum for all periods. The
 
output time history response is obtained for any mass point desired. 
 
From the time history response of a particular mass point, a spectrum response curve is developed
 
and enveloped into a design spectrum. These envel opes are widened by at least 10 percent by period to account for uncertainties in the structural model and input. For all rigid and flexible
 
equipment, the maximum acceleration is obtained by the spectrum response curves developed at
 
various elevations and other points of attachment. These curves are generated using a
 
synthesized time history with horizontal components normalized to 0.05 g and 0.10 g ground
 
accelerations for the 1/2 SSE and SSE, respectively. Both horizontal and vertical excitations are
 
applied to the structure, and curves are generated for each direction. 
 
The stability of structures from the combined horizontal and vertical earthquake excitation is
 
considered by taking modal SRSS moments about the foundation level and comparing them
 
against the resisting moment of the structural deadweight. 
 
The general approach employed in the dynamic anal ysis of Category I equipment and component design is based on the response spectrum techni que where applicable. The time history analysis of Category I structures, as previously explained, generates instructure response spectrum curves and time histories at various support elevations for use in analysis of systems and equipment. 
 
At each level of the structure where vital items are located, horizontal response spectra for each
 
of the two major axes of the structure and a vertical response spectrum are developed. The floor
 
response spectrum is smoothed so that the response curve is an upper bound envelope of all the
 
acceleration points. Whenever the response curve comes to a peak, the curve is made flat in a region +/-10 percent of that peak frequency. When items are supported at two or more elevations, the response spectrum of each elevation is superimposed on each other and the resulting
 
spectrum is the upper bound envelope of all the individual spectrum curves considered.
FNP-FSAR-3
 
3.7-8 REV 21  5/08 Simplified analytical models are used for analys is of systems and equipment; however, where one or two degree of freedom models do not provide a suitable representation of the systems or equipment under consideration, multi-mass models are used in accordance with the lumped
 
parameter modeling techniques and normal mode t heory. Piping analysis is handled in systems using lumped mass models outlined above. Special att ention is given to the flexibility or rigidity characteristics of the piping networks using strategically placed restraints and snubbers to ensure
 
predictability of structural integrity under the specified seismic conditions. 
 
To determine the effect of an earthquake on Westinghouse equipment, a dynamic analysis based
 
on a discrete mass mathematical model is performed. Although a mechanical component may be
 
analyzed using a mathematical model with as much complexity as allowed by the capacity of the computer and the computer code, the analysis is meaningful only when this detailed model also
 
represents the effective utilization of the theory on which the computer code is built. Specifically, there are at least three things that are considered when establishing the mathematical model. They
 
are the limiting values for items such as the degrees of freedom, sections, members, anchors, joints, and bellows, etc; the maximum allowable ratio of member rigidity; and the basic theory limitations. A
 
computer code such as WESTDYN can be used to obtain the natural frequencies, mode shapes, absolute and relative displacements, absolute accelerations, and the stresses. The equipment
 
design is determined to be adequate from the stress margin and by displacements limited to the
 
operating tolerance. 
 
For certain Category I equipment and components wher e dynamic testing becomes a necessity to ensure functional integrity, test performance data and results reflect the following: 
: a. Performance data of equipment which, under the specified conditions, has been subjected to equal or greater dynamic loads than those to be
 
experienced under the specified seismic conditions.
: b. Test data from previously tested comparable equipment which, under similar conditions, has been subjected to equal or greater dynamic loads
 
than those specified. 
: c. Actual testing of equipment in accordance with one of the following methods: 
: 1. The equipment is subjected to a sinusoidal excitation, sweeping through the desired range of significant
 
frequencies, using input acceleration amplitudes for the
 
forcing function which simulates the specified seismic
 
conditions. 
: 2. The equipment is subjected to a transient sinusoidal motion synthesized by pulse exciting a group of approximate octave
 
filters so that the response of the shake table and the
 
duration of load simulates the artificial response spectrum
 
curve at the building floor elevation of interest. 
 
A detailed description of dynamic analysis and testing requirements is given in sections 3.9 and
 
3.10. Table 3.7-4 identifies which qualification method is used.
FNP-FSAR-3
 
3.7-9 REV 21  5/08 The mathematical models used for the dynamic anal ysis of Category I structures, systems, and components, and the results of the analysis are shown in figures 3.7-8 through 3.7-21 and 3.7-23
 
through 3.7-56. Figure 3.7-6 shows a typical lumped mass model for a cantilevered system.
Figures 3.7-64 and 3.7-65 show mathematical model and first mode of vibration of the reactor
 
internals, respectively. 
 
The allowable stresses for the 1/2 SSE and SSE loads in combination with other loads are in
 
accordance with section 3.8.
 
For the 1/2 SSE, the resulting stresses and deflections are limited to those that do not interrupt
 
normal operation of the plant; whereas for the SSE, the resulting stresses and deflections are
 
limited to those that do not prevent a safe and orderly shutdown of the plant. 
 
3.7.2.2  Natural Frequencies and Response Loads A summary of natural frequencies is presented in table 3.7-5. Typical mode shapes for the
 
containment and internal structures are shown in figures 3.7-57 and 3.7-58. Typical response loads (inertia forces) for each mode of the containment are shown in figures 3.7-59 and 3.7-60. The
 
response spectrum at the reactor support elevation is shown in figure 3.7-61. 
 
The natural frequencies of Westinghouse supplied components are considered in the system
 
seismic analysis. The natural frequencies of the components themselves are above the seismic
 
cutoff frequency. 
 
The natural frequencies are listed in the component stress reports filed with the NRC. 
 
3.7.2.3  Procedures Used to Lump Masses The regular lumping techniques, which consist of lumping the continuous mass distribution at
 
discrete joints referred to in section 3.7 as mass points are used in constructing some of the
 
mathenatical models. The location of the lumped masses are chosen at floor levels and points
 
considered of critical interest, such as equipment. The lumped masses are computed from tributary
 
structure dead loads and fixed equipment loads. The model used to generate response spectra for
 
the containment structure utilizes a consistent mass matrix. The term "consistent" describes both
 
the inertial and deformation shapes of the structure that are consistent with stiffness formulation. In
 
the matrix formulation for consistent masses, the floor masses and tributary fixed equipment
 
masses are added to the diagonal mass matrix. 
 
3.7.2.4  Rocking and Translational Response Summary A fixed base mathematical model for the dynamic system analyses is not assumed. As described in subsection 3.7.1.6, a simplified lumped mass and soil spring approach has been used to
 
characterize soil structure interaction. For more details, refer to BC-TOP-4 (Rev. 1), Section 3.3. 
 
FNP-FSAR-3
 
3.7-10 REV 21  5/08 3.7.2.5  Methods Used to Couple Soil with Seismic System Structures Methods used to couple soil with seismic system structures are discussed in subsection 3.7.1.6. A
 
finite element analysis for the layered site has not been used to couple the soil and the
 
Seismic System structures and components. 
 
3.7.2.6  Development of Floor Response Spectra A modal synthesis method is used to develop the response spectra as described in BC-TOP-4 (Rev. 1), Sections 4.2 and 5.2. The modal response spectra multi-mass method was not used to
 
develop floor response spectra. 
 
3.7.2.7  Differential Seismic Movement of Interconnected Components Differential seismic movement of interconnected components has been considered. The stress and
 
deformation criteria for structures are provided in section 3.8. The stress and deformation criteria
 
for piping are described in BP-TOP-1. 
 
The effect of differential seismic movement of interconnected components between floors is
 
considered in the analysis when it is within Westinghouse scope of responsibility. The
 
interconnected components subjected to differential movement will be within the applicable stress
 
and deformation limits. 
 
3.7.2.8  Effects of Variations on Floor Response Spectra The instructure response spectra computed from the time history instructure acceleration response
 
generally reflect two parameters, the amplificati on of the free field input produced by the soil and structural system, and the frequency content associated with these amplification regions. 
 
The criteria selected for enveloping these computed curves with the smooth design spectra are: 
: a. The instructure design spectrum envelopes the computed spectra at all points. 
: b. The minimum frequency shift is either computed as above or
+/-10 percent, whichever is larger. 
 
3.7.2.9  Use of Constant Vertical Load Factors A vertical, seismic-system, multi-mass, dynamic analysis method has been used for seismic
 
analysis of all Category I structures. The mathematical models are discussed in subsection 3.7.2.1. 
 
Constant vertical load factors are not used as the vertical floor response load for the seismic design
 
of safety-related systems and components within Westinghouse scope of responsibility. 
 
FNP-FSAR-3
 
3.7-11 REV 21  5/08 3.7.2.10 Methods Used to Account for Torsional Effects The dynamic analysis of structures is covered in subsection 3.7.2.1. The method presented, in
 
general, describes lumped mass modeling techniques used to analyze Category I structures. The
 
effects of shear stress due to torsion are considered in the analysis. 
 
A lumped mass mathematical model of the auxiliary building incorporating the eccentricity of the masses was generated. Calculated torsional frequencies are much higher than the translational
 
frequencies for the auxiliary building and internal structures, as shown in table 3.7-6. Therefore, the
 
torsional coupling has been neglected in the mathematical model of these structures. However, to
 
assure the adequacy of the design, the effect of the torsional moment has been taken into account.
 
The torsional moment is determined by the product of the shear force and eccentricity between the
 
center of mass and the center of rigidity. 
 
3.7.2.11 Comparison of Responses Table 3.7-7 gives a comparison of results for the acceleration of the containment shell, based on
 
the response spectrum and time history methods. 
 
3.7.2.12 Methods for Seismic Analysis of Dams The analytical methods and procedures that have been used for the seismic system analysis of the
 
storage pond dam and dikes is described in Appendix 2B.
 
3.7.2.13 Methods to Determine Category I Structure Overturning Moment The overturning moments of the Category I structures were calculated by the response spectrum
 
method. The stability of the structures is checked by combining the overturning moment, dead load
 
of the structure, and vertical acceleration. The soil reaction under the containment is obtained by
 
considering the linear stress distribution under a rigid base mat subjected to the worst combined
 
effects of overturning moment, dead load, and vertical acceleration. 
 
3.7.2.14 Analysis Procedure for Dampings In general, the models employed in the seismic analysis represent more than one material and the
 
characteristic mode shapes have component deflections due to translation and rotation of the soil
 
and structural deformation of steel and concrete. The mode shapes are broken down to
 
component deflections due to the various material deformations. Damping values applied to each
 
mode are computed as the summation of the absolute deflection multiplied by the associated
 
damping for that material divided by the absolute summation of all component deflections. 
 
As an example of the technique, consider a st ructure whose motion is primarily composed of flexural displacement and foundation rotation. The mode shape must be broken down into its rotational and flexural components, denoted as R and F , respectively. Since the rotation is due to the fact that the structure is supported on a flexible foundation, the foundation damping, denoted as FNP-FSAR-3
 
3.7-12 REV 21  5/08 R , will influence the total damping. Denoting the structure's flexural or material damping by F , the composite damping is determined using the following equation: 
 
C = RR + FF    R + F The above formula may be regarded as an approximat e technique to determine a composite damping value when the structural motion consists of both a rocking effect with the soil interaction and flexure of
 
the building. If rocking is predominant, then soil damping alone is assigned to that mode. The
 
converse holds true when flexure is predominant. 
 
3.7.3 SEISMIC SUBSYSTEM ANALYSIS 3.7.3.1  Determination of Number of Earthquake Cycles 3.7.3.1.1 Category I Systems and Components Other Than NSSS Procedures to determine the number of earthquake cycles for piping during one seismic event are
 
discussed in BP-TOP-1, (Rev. 1) Section 6.0. For equipment designed on the basis of analytical
 
results, the design criteria used assumed elastic behav ior. Therefore, the number of loading cycles is of no concern. Neither is it of any concern for Category I structures, since the calculated stresses and
 
strains are below yield. 
 
3.7.3.1.2 NSS System Where fatigue analyses of mechanical systems and components are required, Westinghouse specifies
 
in the equipment specification that five occurrences of 1/2 SSE and SSE, each having ten cycles of
 
maximum response for each occurrence, be analyzed. The fatigue analyses are performed as part of
 
the stress report. 
 
3.7.3.2  Basis for Selection of Forcing Frequencies Forcing frequencies are not selected but are calculated in accordance with BC-TOP-4, Rev. 1, Section
 
5.3-2. 
 
3.7.3.3  Root Mean Square Basis The term used to describe the procedure for combination of modal responses is "square root of the
 
sum of squares" (SRSS). 
 
FNP-FSAR-3
 
3.7-13 REV 21  5/08 3.7.3.4  Procedure for Combining Modal Responses The criteria for combining modal responses (shears, moments, stresses, deflections, and/or
 
accelerations) for the response spectrum modal analysis are as follows: 
: a. The SRSS method of combining all modal responses is used. 
: b. All modes up to a frequency of 30 hz are used in the analysis. 
: c. When closely spaced frequencies of two or more modes occur, only those modes' responses are combined in an absolute manner; the
 
resulting total is treated as that of a pseudo-mode and then combined
 
with the rest of the modes in an SRSS manner. 
: d. The criterion used to determine whether mode frequencies are closely spaced is whether the frequencies differ from each other by less than
 
20% of the lower frequency. Also, multiples of lower mode frequencies
 
are compared with higher mode frequencies, and the same 20 percent
 
comparison is made. 
: e. For analyses for which Westinghouse has the responsibility, the total seismic response may be obtained by combining the individual modal seismic response
 
and the individual modal responses, using the SRSS method. For systems
 
having modes with closely spaced frequencies, this method is modified to include
 
the possible effect of these modes. The groups of closely spaced modes are
 
chosen so that the difference between the frequencies of the first mode and the
 
last mode in the group does not exceed 10 percent of the lower frequency. The
 
combined total response for systems which have such closely spaced modal
 
frequencies is obtained by adding to the square root of the sum of the squares of
 
all modes, the product of the responses of the modes in each group of closely spaced modes and a coupling factor, . This can be represented mathematically as:      K K N 1 K 1 N M K N 1 i S 1 j 2 i 2 T R R 2 R R j j j+=+====  where    T R = total response i R = absolute value of response of mode i N = total number of modes considered S = number of groups of closely spaced modes
 
FNP-FSAR-3
 
3.7-14 REV 21  5/08 j M = lowest modal number associated with group j      of closely spaced modes j N = highest modal number associated with group j      of closely spaced modes K = coupling factor with 1 2 K K K}1{K++=  and
  ()[]2/1 2 K K K 1=    d k K K t 2+=        K = frequency of closely spaced mode K (rad/sec)
K = fraction of critical damping in closely spaced      mode K
 
t d = duration of the earthquake (sec.)
 
In addition to the above methods, any of the methods described in USNRC Regulatory Guide 1.92, Revision 1 may be used for modal combination in the analysis of replacement components.
: f. Modal response combination for piping analysis is described in appendix 3L. 
 
3.7.3.5  Significant Dynamic Response Modes BC-TOP-4 (Rev. 1), Appendices F and H, describe the analysis techniques used when the peak of the
 
spectra method is employed by the Category I equipment suppliers. 
 
For piping, this is covered in BP-TOP-1, Rev. 1, Section 2.0 and Appendix D.
FNP-FSAR-3
 
3.7-15 REV 21  5/08 The static load equivalent or static analysis method involves the multiplication of the total weight of the equipment or component member by the specified se ismic acceleration coefficient. The magnitude of the seismic acceleration coefficient is established on the basis of the expected dynamic response
 
characteristics of the component. Components which can be adequately characterized as a single-
 
degree-of-freedom system are considered to have a modal participation factor of one. Seismic acceleration coefficients for multi-degree-of-freedom systems, which may be in the resonance region
 
of the amplified response spectra curves, are increased by 50 percent to account conservatively for
 
the increased modal participation. 
 
3.7.3.6  Design Criteria and Analytical Procedures For Piping The relative seismic movements between buildings, between floors in buildings, and between major
 
components and buildings are applied to the pipe anchors and restraints in a rational or conservative
 
manner. Movements between buildings and between buildings and components are always
 
considered to be out of phase in such a way that their relative movements are maximum. The
 
resulting stresses are classed as secondary and are combined with thermal expansion stresses. 
 
These stresses are held below the appropriate code allowable limits. 
 
3.7.3.7  Basis for Computing Combined Response The bases for the methods used to determine the combined horizontal and vertical amplified response
 
loadings for the seismic design of the piping and equipment are discussed in BP-TOP-1, Revision 1, and BC-TOP-4, Revision 1, except that in all cases the maximum horizontal response in one direction
 
is combined with the vertical response by the "SRSS" approach. The combined response is then used
 
in the stress analyses. 
 
3.7.3.8  Amplified Seismic Responses A constant vertical load factor is not used for the seismic design of Category I structures, components, and equipment. 
 
3.7.3.9  Use of Simplified Dynamic Analysis The simplified seismic analysis methods and procedures are used only for the design of 2 in. and
 
under piping that is field routed. The design requires the piping system to be supported by means of
 
hangers, restraints, and anchors in a continuous run of simple shapes, for which the natural
 
frequencies have been established by previous analyses and found to fall within the rigid frequency
 
range or acceptable stress limits. 
 
A summary of typical results is given in table 3.7-8. 
 
FNP-FSAR-3
 
3.7-16 REV 21  5/08 3.7.3.10 Modal Period Variation The procedures used to account for modal period variation in the mathematical models for Category I
 
structures due to variation in material properties are discussed in subsection 3.7.2.8. 
 
The materials employed in safety-related system s under Westinghouse scope of supply are standard.
The material properties that can affect a variation in modal period are well known, and the known
 
variation in these properties does not account for any measurable or significant shift in period or
 
increase in seismic loads. 
 
3.7.3.11 Torsional Effects of Eccentric Masses The seismic mass model accounts for the effect of masses that are offset from the pipe centerline. 
 
Components with eccentric masses are modeled by placing the component's mass at its calculated
 
center of gravity and connecting this mass to the pipe centerline with a rigid connection. The inertia
 
forces calculated from the response spectra curves are applied at this lumped mass point. Therefore, any forces or moments, including torsion, resulting from eccentric masses are accounted for in the
 
seismic analysis. 
 
3.7.3.12 Piping Outside Containment The differential movement of all Category I piping located outside containment is included in the stress
 
analysis. 
 
Movements are always considered to be out of phase in such a manner that the relative movements
 
are maximum. These movements are then imposed on all anchors and restraints. The resulting
 
stresses are classed as secondary and combined with thermal expansion stresses. These stresses
 
are held below the appropriate code allowable limits. 
 
BC-TOP-4 (Rev. 1), Section 6, discusses the techniques used to predict structural stresses in buried
 
Category I piping for seismic loadings. The criteria require piping to remain functional when exposed
 
to loadings predicted by use of the site design spectra. This is assured by limiting the strains to 40
 
percent of the ultimate strain of the pipe material.
 
3.7.3.13 Interaction of Other Piping With Category I Piping The interface between Category I piping and non-Category I piping is always an anchor. The anchor
 
prevents seismic motion on the non-Category I side from affecting the Category I side. The anchor is
 
designed so that under the most conservative combination of thermal, weight, and seismic loads from
 
both sides of the anchor, the anchor can maintain separation of motions. Seismic loads from the non-
 
Category I side are estimated using a simplified dynamic analysis. 
 
FNP-FSAR-3
 
3.7-17 REV 21  5/08 3.7.3.14 Field Location of Supports and Restraints Seismic supports and restraints for seismic Category I piping are located so that the stresses, as
 
determined by the dynamic analysis, are less than t he appropriate code allowable limits. When rigid seismic supports result in excessive thermal loads on piping or equipment, snubbers or dampers are
 
used. 
 
The pipe support contractors' pipe restraint locations and detailed support drawings are reviewed by
 
pipe stress engineers to ensure that they conform to requirements. In addition, a field inspection of the
 
pipe supports is made by stress engineers to ensur e that supports have been installed properly and meet design requirements. 
 
For 2 in. and under Category I piping, a Bechtel field installation manual is provided so that field
 
engineers can properly design and locate pipe supports and restraints. When the field engineers have
 
completed their designs, they are reviewed by pipe stress engineers. 
 
3.7.3.15 Seismic Analyses for Fuel Elements, Control Assemblies, and Control Rod Drives Fuel assembly responses resulting from a safe shutdown earthquake were analyzed using time
 
history integration techniques. The time history motions of the core plates and the core barrel
 
used as the seismic input were obtained from the reactor vessel and internals system model. 
 
The acceleration spectra of time histories at the reactor vessel support elevation encompass the
 
corresponding design spectra for the plant site.
 
The seismic response of the fuel assemblies is analyzed to determine structural design
 
adequacy. Component stresses were obtained through the use of finite element computer
 
modeling. Detailed discussions of the analysis methodology for evaluating the faulted condition
 
loads on the fuel assembly design are contained in references 2, 4, and 5. The resulting
 
combined seismic and LOCA loads are given in paragraph 4.2.1.1.2.
 
The control rod drive mechanisms (CRDMs) are se ismically analyzed to confirm that system stresses under seismic conditions do not exceed allowable levels as defined by the ASME
 
Boiler and Pressure Vessel Code Section III for "upset" and "faulted" conditions. Based on
 
these stress criteria, the allowable seismic stresses in terms of bending moments in the
 
structure are determined. The CRDM is mathem atically modeled as a system of lumped and distributed masses. The model is analyzed under appropriate seismic excitation, and the
 
resultant seismic bending moments along the length of the CRDM are calculated. These values
 
are then compared to the allowable seismic bending moments for the equipment to ensure
 
adequacy of the design. 
 
3.7.4 SEISMIC INSTRUMENTATION PROGRAM 3.7.4.1  Comparison with NRC Regulatory Guide 1.12 The original seismic instrumentation for the FNP was installed to meet the guide lines of NRC
 
Regulatory Guide 1.12. The seismic instrumentation provides data to determine if the plant can FNP-FSAR-3
 
3.7-18 REV 21  5/08 continue to operate safely after an earthquake. New advances in seismic instrumentation technology have made available instruments that can analyze seismic data on site, faster and more accurately than the original instrumentation installed at FNP. The NRC has accepted the
 
following EPRI reports as an acceptable approach to redefine the seismic monitoring
 
requirements and determine plant action following an earthquake:
 
Seismic Instrumentation in Nuclear Power Plants for Response to OBE Exceedance: 
 
Guideline for Implementation, EPRI TR-104239, July 1994
 
A Criterion for Determining Exceedance of the Operating Basis Earthquake, EPRI
 
NP-5930, July 1988 Guidelines for Nuclear Plant Response to an Earthquake, EPRI NP-6695, December
 
1988
 
Standardization of the Cumulative Absolute Velocity (CAV), EPRI TR-100082, December 1991
 
The seismic monitoring system for FNP consists of the following instruments which meet the seismic requirement of EPRI TR-104239:
 
A. Three triaxial strong-motion acceleration sensors (time history) connected to an on-line computer for automatic data retrieval and analysis. System computes
 
OBE exceedance and provides operator alarm.
B. Two self contained strong-motion accelerographs (time history).
 
3.7.4.2  Location and Description of Instrumentation For location and summary of instrumentation see figure 3.7-62. 
: 1. Acceleration Sensors Connected to Seismic Monitoring Panel in Control Room Three triaxial strong-motion acceleration sensors are located in the following plant
 
areas:
: a. Two sensors are located on the exterior surface of the containment. One sensor is rigidly mounted on the containment base slab (elevation 104 ft)
 
and the other is rigidly fastened to the containment wall directly above at
 
elevation 212 ft 7 in.
: b. One sensor is located in the free field south of the containment, to be used as a free field instrument. It is far enough from the containment to avoid any
 
effect of the containment on the sensor.
The three axes of each triaxial strong-motion acceleration sensor have an orientation
 
common with the others to permit accurate phase correlation of all channels. The FNP-FSAR-3
 
3.7-19 REV 21  5/08 sensors are rigidly mounted to the structure so that seismic records can be directly related to any structure movement.
 
The function of each acceleration sensor is to provide data on the frequency, amplitude, and phase relationship of the seismic response of the containment and
 
other seismic Category I structures. These measurements will be used to evaluate
 
the effect of the seismic disturbance on the structures and to assess their post-
 
disturbance integrity. The seismic monitoring instrumentation is not connected to
 
the plant safety systems.
 
The monitoring system is an automatic dat a retrieval and analysis system based on a high speed computer. The system remains on at all times continuously monitoring
 
the signals from the acceleration sensors.
The digital triggering system continuously monitors the signals and when the motion exceeds the adjustable, preset threshold, the system retrieves the time history data from storage and automatically performs the preprogrammed calculations for operator review. The system printer provides
 
print and plot copies of the results for record and review. The system provides
 
alarms for event, OBE, and loss of power. The system is powered by internal rechargeable batteries, which provide sufficient reserve power in the event of ac
 
power failure.
: 2. Self-Contained Triaxial Accelerographs (Time History)
Two self-contained strong-motion triaxial accelerographs are installed at elevation
 
155 ft 0 in. in the diesel generator building and at elevation 167 ft 3 in. in the service
 
water intake structure. These instruments will not be connected to the control room.
 
These instruments are actuated by integral, digital triggers which have an
 
adjustable, preset threshold for each channel. When the motion exceeds the
 
threshold setting, the integral digital solid state recorder retrieves the time history
 
data from storage and continues to record data until the unit de-triggers. The
 
recorded data can be retrieved and analyzed to determine plant impact. The unit is
 
powered by internal rechargeable batteries, which provide power in the event of ac
 
power failure. The unit has external indicators for event and loss of ac power.
 
3.7.4.3  Control Room Operator Notification The seismic monitoring panel will annunciate in the control room to alert the operator for an
 
event, OBE exceedance and loss of power.
 
3.7.4.4  Comparison of Measured and Predicted Responses If the seismic monitoring panel in the control room is triggered by an event from one or more of the strong-motion acceleration sensors connected to the panel, the system will automatically retrieve and analyze the data. The data analysis is rapid and automatic so operators can
 
evaluate the event. An outline of the order of actions to be taken is given in figure 3.7-63.
 
FNP-FSAR-3
 
3.7-20 REV 21  5/08 3.7.5 SEISMIC DESIGN CONTROL 3.7.5.1  Seismic Design Control - Construction Phase This section describes the design control measures which were used during the construction
 
phase to ensure that adequate seismic input data (including necessary feedback from structural
 
and system dynamic analysis) were specified to vendors of purchased Category I components and equipment. Three organizations are involved in procuring Category I components. These
 
include Southern Company Services, Inc., Bechtel Power Corporation, and the nuclear steam
 
supplier, Westinghouse. 
 
The primary design organizations involved in the seismic design of the various structures, systems and components for the FNP are Westinghouse Electric Corporation, Bechtel Power
 
Corporation, and Southern Company Services, Inc. 
 
Components designed by others which fall under one of the three primary areas of responsibility
 
are designed to the overall seismic requirements and checked by one of these organizations. 
 
Responsibilities are as follows: 
: a. Westinghouse Electric Corporation is responsible for design of the nuclear steam supply system (NSSS). This includes the reactor vessel, steam generator, pressurizer, NSSS supports, primary coolant piping, and the emergency core
 
cooling systems. 
: b. Bechtel Power Corporation is responsible for the design of the containment, auxiliary building, and all of the safety-rela ted systems in these two buildings not furnished by Westinghouse. In addition, Bechtel has responsibility for reviewing
 
seismic designs originated by Southern Company Services. 
: c. Southern Company Services, Inc., has responsibility for designing those structures and systems not contained in either the cont ainment or the auxiliary building. These seismic designs are reviewed by Bechtel Power Corporation. 
 
Westinghouse Supplied Equipment and Components
: 1. To ensure that Westinghouse supplied NSSS Category I mechanical components meet the seismic design criteria, the following procedures are implemented: 
: a. Equivalent static acceleration factors are determined for each Category I component based upon the amplified ground acceleration response
 
spectrum curves and the location of the component within the structure. 
 
This acceleration factor is included in the equipment specification, and the
 
vendor must certify the adequacy of the component to meet this seismic
 
requirement. Equipment specifications to vendors require that
 
Westinghouse supplied Category I auxiliary pumps are designed by the
 
vendor to operate during horizontal and vertical accelerations of 1.0 g and
 
0.6 g respectively and simultaneously. The sum of the primary stresses FNP-FSAR-3
 
3.7-21 REV 21  5/08 does not exceed Section III of the ASME Code for pressure-containing members. If qualification is by test results or by response analysis, the input
 
frequencies for referenced "g" loadings is 5 to 15 hz. 
 
Category I tanks are designed by Westinghouse to withstand the simultaneous horizontal and vertical forces resulting from the SSE. The
 
vendor is also required to perform a static analysis and to comply with ASME
 
Section III. 
 
Category I valves are designed by the vendor to withstand seismic loadings equivalent to 3.0 g in the horizontal direction and to 2.0 g in the vertical
 
direction and perform all functions within the specification. 
: b. The vendor's drawings and calculations are reviewed to determine whether the component meets all specification requirements. 
: c. Based upon engineering judgment and detailed analyses on similar equipment, the cognizant engineer will: 
: i. Accept the component. 
 
ii. Reject the component as inadequate or recommend modifications. 
 
iii. Require that the engineering analysis section review the drawing details and perform a detailed analysis, if deemed necessary, using
 
one of the methods described in the following paragraph. 
: d. To conform to the above criteria, seismic analysis of selected NSSS Category 1 components, including heat exchangers, pumps, tanks, and
 
valves, is performed using one of three methods depending on the relative
 
rigidity of the equipment being analyzed: 
: i. Equipment that is rigid and rigidly attached to the supporting structure is analyzed for a g-loading equal to the
 
acceleration of the supporting structure at the appropriate
 
elevation 
 
ii. Equipment that is not rigid, and therefore a potential for response to the support motion exists, is analyzed for the
 
peak of the floor response curve with appropriate damping
 
values iii. In some instances, nonrigid equipment is analyzed using a multiple degree of freedom modal analysis, including the
 
effect of modal participation factors and mode shapes, together with the spectral motions of the floor response
 
spectrum defined at the support of the equipment. The
 
inertial forces, moments, and stresses are determined in FNP-FSAR-3
 
3.7-22 REV 21  5/08 each mode. They are then summed using the square-root-of-the-sum-of-the-squares method. 
 
The analyses described above are performed on mechanical equipment selected on a generic and size basis to verify that the
 
equipment meets the seismic criteria listed in the equipment
 
specification. Westinghouse has established the following criteria
 
for protection and engineered safety system equipment: 
 
For the SSE, the equipment is analyzed to ensure that it does not lose capability to perform its function; i.e., shut the plant down and
 
maintain it in a safe shutdown condition.
To ensure that the equipment will perform its intended function during a safe shutdown earthquake, the deflections and stresses
 
obtained by the seismic analysis are added to those associated
 
with the operational mode of the equipment to verify that
 
clearances are not exceeded and stresses are within allowable
 
limits. 
: 2. For protection grade instrumentation and control equipment: 
 
For either earthquake (1/2 SSE or SSE), the equipment is designed to ensure that it does not lose its capability to perform its function; i.e., shut the plant down and
 
maintain it in a safe shutdown condition. 
 
For the SSE there may be permanent deformation of the equipment provided that the capability to perform its function is maintained. 
 
Typical protection system equipment is subjected to type tests under simulated seismic accelerations to demonstrate its ability to perform its functions. 
 
Type testing is being done on equipment by Westinghouse using conservatively large accelerations and applicable frequencies. Analyses such as those done for
 
structures are not done for the reactor protection system equipment. However, the
 
peak accelerations and frequencies used are checked against those derived by
 
structural analyses of 1/2 SSE and SSE loadings. 
 
A Westinghouse topical report, WCAP-7397-L , (and supplement), provides the seismic evaluation of safety related equipment. The type tests covered by this
 
report are applicable to the Farley Nuclear Plant. 
 
The control board is not considered to be protection equipment. Typical switches and indicators for safeguards components have been tested to determine their
 
ability to withstand seismic forces without malfunction which would defeat automatic
 
operation of the required component. 
 
The control boards are stiff and past experience indicates that the amplification due to the board structure is sufficiently low so that the acceleration seen by the device FNP-FSAR-3
 
3.7-23 REV 21  5/08 is considerably less than the acceleration that the device was shown to withstand in testing. 
 
Bechtel and Southern Company Services Specified Equipment and Components
 
Bechtel and Southern Company Services Specifications for Category I equipment incorporate a
 
section on seismic design criteria. Category I valves and dampers are designed by the vendor to
 
withstand seismic loadings vertical direction and to perform all functions within the specification. 
 
For other Category I equipment, the vendor was provided, as a part of the design specifications, the seismic response spectra, generated by a time history, which have been developed for the
 
particular equipment location, and a list of damping factors. The specification requires the vendor
 
to do one of the following: 
: 1. Perform a seismic analysis based on the appropriate damping factor and response spectrum as well as the natural frequency of his equipment. 
: 2. If it is not practical to calculate the natural frequency of the equipment, use the maximum acceleration of the spectrum curve for the seismic analysis. 
: 3. Subject prototype equipment to a test demonstrating its ability to perform its intended function during and after seismic disturbance. 
 
The current applicable seismic design data are provided to each vendor and certification is
 
required from each vendor that his equipment will function during the SSE. This certification may
 
consist of calculations checked by an engineer knowledgeable in the design of such equipment or
 
of a written certification that the equipment has successfully passed tests of forces equal to or
 
higher than those stated in the seismic requirement and has been exposed to these severe
 
vibration requirements. The method of analysis, calculation, or testing is reviewed and approved
 
by the responsible engineer. 
 
3.7.5.2  Seismic Design Control - Operational Phase During the operational phase, FNP will exercise the same controls as described in paragraph
 
3.7.5.1, except the responsibilities  shall be as directed by Southern Nuclear Operating Company. 
 
FNP-FSAR-3
 
3.7-24 REV 21  5/08 REFERENCES 
: 1. Newmark, N. M., "Design Criteria for Nuclear Reactors Subjected to Earthquake Hazards," Proceedings, IAEA Panel on Aseismic Design and Testing of Nuclear Facilities , Japan Earthquake Engineering Promotion Society, Tokyo, May 1967. 
: 2. Gesinski, T. L., "Fuel Assembly Safety Analysis For Combined Seismic and Loss of Coolant Accident," WCAP-7950 , July 1972. 
: 3. Vogeding, E. L., "Seismic Testing of Electrical and Control  Equipment," WCAP-7817 and Supplement I, December 1971. 
: 4. Davidson, S. L. and Iorii, J. A., ed., "Verification Testing and Analyses of the 17 x 17 Optimized Fuel Assembly," WCAP-9401-P-A , August 1981.
: 5. Davidson, S. L., ed., "Reference Core Report - VANTAGE 5 Fuel Assembly,"
WCAP-10444-P-A , September 1985.
 
FNP FSAR-3
 
REV 21  5/08 TABLE 3.7-1 PERCENTAGE OF CRITICAL DAMPING FACTORS
 
1/2 Safe  Shutdown Safe-Shutdown Earthquake Earthquake (E')
0.05 g Ground 0.10 g Ground
 
Type of Structure Acceleration Acceleration
 
Vital piping (a) 0.50 1.00 Welded steel plate 1.00 2.00 assemblies 
 
Welded steel frame 2.00 5.00 structures 
 
Bolted and riveted 3.00 5.00
 
steel (b)
Reinforced concrete 2.00 5.00 structures and equipment supports 
 
Prestressed concrete 2.00 5.00 structures 
 
Soil damping 4.00 7.00
: a. ASME Code Case N-411 damping values may be used as stated in paragraph 3.7.1.3.
: b. Regulatory Guide 1.61 damping values are used in the analysis of the reactor vessel head assembly structure.
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.7-2 SYSTEM PERIOD INTERVAL Fre-  Fre-  Fre-  Fre-quency  quency  quency  quency
 
No. (Hz) No. (Hz) No. (Hz) No. (Hz) 1 0.10 31 0.43 61 1.87  91  8.07 2 0.105 32 0.45 62 1.96  92  8.48 3 0.11 33 0.48 63 2.06  93  8.90 4 0.115 34 0.50 64 2.16  94  9.35 5 0.12 35 0.53 65 2.27  95  9.81 6 0.125 36 0.55 66 2.38  96 10.30 7 0.13 37 0.58 67 2.50  97 10.82 8 0.14 38 0.61 68 2.63  98 11.36 9 0.15 39 0.64 69 2.76  99 11.93 10 0.155 40 0.67 70 2.90 100 12.52 11 0.16 41 0.70 71 3.04 101 13.15 12 0.17 42 0.74 72 3.19 102 13.81 13 0.18 43 0.78 73 3.35 103 14.50 14 0.19 44 0.81 74 3.52 104 15.22 15 0.20 45 0.86 75 3.70 105 15.98 16 0.21 46 0.90 76 3.88 106 16.78 17 0.22 47 0.94 77 4.08 107 17.62 18 0.23 48 0.99 78 4.28 108 18.50 19 0.24 49 1.04 79 4.50 109 19.43 20 0.25 50 1.09 80 4.72 110 20.40 21 0.27 51 1.15 81 4.96 111 21.42 22 0.28 52 1.20 82 5.20 112 22.49 23 0.29 53 1.26 83 5.46 113 23.62 24 0.31 54 1.33 84 5.74 114 24.80 25 0.32 55 1.39 85 6.02 115 26.04 26 0.34 56 1.46 86 6.33 27 0.36 57 1.54 87 6.64 28 0.37 58 1.61 88 6.97 29 0.39 59 1.69 89 7.32 30 0.41 60 1.78 90 7.69 
 
FNP FSAR-3
 
REV 21  5/08 TABLE 3.7-3 METHODS USED FOR SEISMIC ANALYSES OF CATEGORY I STRUCTURES
 
Method of Analysis Response  Applicable Stress  Spectra Time-History or Deformations 
 
Category I Structures Analysis Analysis Criteria Remarks Containment X X Refer to Section 3.8.1.5 
 
Auxiliary building X X Refer to Section 3.8.4.5 
 
Diesel generator building X X " 
 
River intake structure (a) X X "
Intake structure at X X "
storage pond   
 
Storage pond dam and - - - See Section 2.5 dike (earth fill)   
 
Vent stack X - Refer to Section 3.8.4.5 
 
Pond spillway structure X - " 
 
Electrical cable tunnel X - "
structure   
 
Category I outdoor tanks X - "       
 
__________
: a. Original design (Category I) r equirements are no longer required.
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.7-4 (SHEET 1 OF 10)
METHODS USED FOR SEISMIC ANALYSES OF CATEGORY I SYSTEMS AND COMPONENTS Method of Analysis        Applicable Response  Stress  Category I Equivalent Spectra Time-History  or Deformation Systems and Components Static Load Analysis    Analysis    Tests    Criteria      Remarks        REACTOR COOLANT SYSTEM     
 
Reactor Vessel X    See section 5.2 
 
Full-length CRDM housing  X  " 
 
Part-length CRDM housing  X  " 
 
Reactor coolant pump  X  " 
 
Steam generator X    " 
 
Pressurizer  X  "        Reactor coolant loop piping (1)  X(2) X(1)  "  and piping to pressure boundary (2)              RC system supports  X  " 
 
Surge pipe and fittings  X  " 
 
RC Thermowells    X " 
 
Safety valves X    " 
 
Relief valves X    " 
 
Valves to RC system boundary X    " 
 
CRDM head adapter plugs  X  " 
 
CHEMICAL AND VOLUME CONTROL SYSTEM     
 
Generative HX  X  See section 3.9 
 
Letdown HX  X  "        Mixed-bed demineralizer  X  " 
 
Cation bed demineralizer  X  " 
 
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.7-4 (SHEET 2 OF 10)
Method of Analysis Applicable    Response        Stress Category I Equivalent Spectra Time-History  or Deformation Systems and Components Static Load Analysis    Analysis    Tests      Criteria      Remarks Reactor coolant filter X    " 
 
Volume control tank  X  " 
 
Charging/high head X  X " Tests were run to safety injection pump      determine natural frequency of the foundation system to meet seismic criteria.
 
Seal water injection X filter    " 
 
Excess letdown HX  X   
    "
Seal water return X filter    " 
 
Seal water HX  X   
    "
Boric acid tanks  X    Per API 650
 
Boric acid filter X     
    "
Boric acid transfer pump X     
 
Boric acid blender  X   
    "
Reactor makeup water  X  Wt/% exceeding storage tank    90% of yield stresses and/or loss of function 
 
EMERGENCY CORE COOLING SYSTEM     
 
Accumulators  X   
    "
Boron injection tank  X   
    "
BIT recirculation pump X     
 
Boron injection surge  X  See section 3.9 tank FNP-FSAR-3
 
REV 21  5/08 TABLE 3.7-4 (SHEET 3 OF 10)
Method of Analysis Applicable Response        Stress Category I Equivalent Spectra Time-History  or Deformation Systems and Components Static Load Analysis    Analysis    Tests      Criteria      Remarks        RESIDUAL HEAT REMOVAL SYSTEM              Residual heat X removal/low head safety injection pump            "  Residual heat  X exchanger              CONTAINMENT SPRAY SYSTEM    "        spray additive tank  X            Containment spray pump X            CONTAINMENT ISOLATION    "
SYSTEM            "  Valves X            CONTAINMENT COOLING SYSTEM            "  Fans  X            Heat exchanger  X            COMPONENT COOLING    "
SYSTEM            "  Pumps X            Heat exchangers  X            Surge tank  X  "        SPENT FUEL POOL COOLING    " Per API 650 SYSTEM            "  Spent fuel pool heat  X exchanger              Spent fuel pool pump X FNP-FSAR-3
 
REV 21  5/08 TABLE 3.7-4 (SHEET 4 OF 10)
Method of Analysis          Applicable Response        Stress Category I Equivalent Spectra Time-History  or Deformation Systems and Components Static Load Analysis    Analysis    Tests      Criteria      Remarks        BORON THERMAL REGENERATION SUBSYSTEM              Moderating HX  X            Letdown chiller HX  X          "  Letdown reheat HX  X          "  Thermal regeneration  X demineralizer    "        LIQUID RECYCLE AND WASTE    "
SUBSYSTEM              Recycle holdup tank  X    Per API 650
 
Recycle evaporator feed X pump    "        Recycle evaporator feed  X  "
demineralizer              Recycle evaporator feed filter X    "        Recycle evaporator X    "        LIQUID RECYCLE AND WASTE      SUBSYSTEM    "        R.C. drain tank HX  X            Waste holdup tank  X          "  Waste evaporator feed X pump    "        Waste evaporator feed X    "
filter              Waste evaporator X    "        Spent resin storage  X tank    "        Spent resin sluice pump X    "
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.7-4 (SHEET 5 OF 10)
Method of Analysis              Applicable    Response  Stress Category I Equivalent Spectra Time-History  or Deformation  Systems and Components Static Load Analysis    Analysis    Tests      Criteria      Remarks Spent resin sluice filter X    " 
 
Floor drain tank  X  "
ES room sump pump (a) X    "        GAS HANDLING SUBSYSTEM Gas compressor X  X " Vibration tests were conducted        To determine seismic capability Gas decay tanks  X  " 
 
Hydrogen recombiner X    " 
 
EMERGENCY DIESEL FUEL OIL SYSTEM Transfer pumps  X  "
Fuel oil tanks X    " 
 
SERVICE WATER SYSTEM Pumps  X  "
Strainers  X  " 
 
RIVER WATER SYSTEM Pumps(a)  X  "        FUEL HANDLING SYSTEM Fuel manipulator crane  X  "
Fuel transfer tube X    " 
 
Underwater fuel conveyor X    "
car and rail system     
 
Fuel pool bridge crane  X  " 
 
Polar crane X    " 
: a. Pumps originally seismically analyzed as Seismic Category I, but have been downgraded to Seismic Category II.
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.7-4 (SHEET 6 OF 10)
Method of Analysis Applicable Response        Stress  Category I Equivalent Spectra Time-History  or Deformation  Systems and Components Static Load Analysis    Analysis    Tests      Criteria      Remarks        Crane supports X    "        REFUELING WATER SYSTEM Storage tank  X  Wt/% exceeding 90% of yield stresses and/or loss of function AUXILIARY BUILDING      VENTILATION SYSTEM ES AIR COOLING UNITS Heat exchanger  X  "        Fan  X  "        PENETRATION ROOM      FILTRATION SYSTEM Fans  X  "        Filters (HEPA and  X  "  charcoal)              CONTROL ROOM VENTILATION SYSTEM Fans  X  "        Filters  X  "        Air handling unit  X  "        Condensing unit    X "        DIESEL BUILDING      VENTILATION SYSTEM Fans    X "        Filters    X "        MAIN STEAM SYSTEM Isolation valves X    "
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.7-4 (SHEET 7 OF 10)
Method of Analysis Applicable Response        Stress Category I Equivalent Spectra Time-History  or Deformation Systems and Components Static Load Analysis    Analysis    Tests      Criteria      Remarks FEEDWATER SYSTEM     
 
Isolation valves X    " 
 
AUXILIARY FEEDWATER SYSTEM     
 
Auxiliary feedwater pumps  X  "
motordriven, steam turbine driven     
 
Condensate storage tank  X  " 
 
STEAM DUMP SYSTEMS     
 
Relief valves X    " 
 
Safety valves X    " 
 
ELECTRICAL COMPONENTS AND SYSTEMS     
 
4160-v switchgear    X " 
(engineered safe-guard buses)     
 
4160-v to 600-v trans-  X  "
formers (associated with engineered safe-guard systems)     
 
600-v load centers    X " Test on prototype (engineered safe-guard buses)     
 
600-v and 208-v motor-    X " Test on prototype control centers (associated with engineered safeguard systems)     
 
125-v dc station    X " Test on three cells batteries     
 
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.7-4 (SHEET 8 OF 10)
Method of Analysis Applicable Response  Stress Category I Equivalent Spectra Time-History  or Deformation Systems and Components Static Load Analysis    Analysis    Tests      Criteria      Remarks Inverters, 125-v dc to 120-v ac    X See section 7.1  (vital ac instrumentation distribution panels)
ELECTRICAL COMPONENTS AND SYSTEMS     
 
125-v dc distribution    X " Tests on two panels panels      selected at random
 
120-v vital ac instrumentation    X "
and regulated ac distribution panels     
 
125-v dc switchgear    X " Tests on prototype 125-v dc battery    X " Test on one charger chargers     
 
Solid-state protection    X "
system cabinets     
 
Reactor trip switchgear    X " 
 
Nuclear instrumentation    X "
system cabinets     
 
Process protection and    X "
control system cabinets     
 
Cable tray supports (associated with  X  "
engineered safeguard system)     
 
Auxiliary relay    X "
racks     
 
Containment penetration    X Wt/% exceeding Test on one medium assemblies    90% of yield voltage penetration stresses and/or assembly plus test on loss of function a composite assembly comprised of 1000-V dc power and 600-V control and instrument cables
 
Turbine driven auxiliary feedwater pump    X See section 7.1  uninterruptable power supply FNP-FSAR-3
 
REV 21  5/08 TABLE 3.7-4 (SHEET 9 OF 10)
Method of Analysis Applicable Response  Stress Category I Equivalent Spectra Time-History  or Deformation Systems and Components Static Load Analysis    Analysis    Tests      Criteria      Remarks Emergency power board  X  X " Instruments and        switches are tested
 
Direct-current emergency    X " Test on prototype lighting     
 
Diesel generators  X  " 
 
Diesel generator control panels    X " 
 
Diesel generator    X " Test on one panel sequencers     
 
Boric acid heat-tracing  X  "
equipment     
 
Balance of plant    X Wt/% loss of instrument cabinets    function and equipment contained therein     
 
Equipment contained within  X  X "
balance of plant instrument cabinets     
 
Containment purge  X  X "
radiation monitors     
 
Fuel handling area  X  X "
radiation monitors     
 
SAMPLING SYSTEM     
: 1. Cabinet  X  Wt/% exceeding 90% of yield stresses and w/o loss of function 
: 2. Tubing, valves, X    Wt/% loss of coolers, sample    function vessels     
 
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REV 21  5/08 TABLE 3.7-4 (SHEET 10 OF 10)
Method of Analysis                Applicable Response  Stress  Category I Equivalent Spectra Time-History  or Deformation  Systems and Components Static Load Analysis    Analysis    Tests      Criteria      Remarks        ELECTRICAL COMPONENTS AND SYSTEMS Balance of plant field X  X "
mounted instruments     
 
Instrument valves for X    "
field mounted instruments     
 
Instrument lines for X    Wt/% exceeding 
 
field mounted      code allowable  instruments    stresses Isolation devices    X  for output of AMSAC     
 
FNP FSAR-3
 
REV 21  5/08 TABLE 3.7-5 NATURAL FREQUENCIES FOR CATEGORY I STRUCTURES Direction North-South    East-West  Vertical Structure  Mode    Mode    Mode    1st 2nd 3rd 4th 5th 1st 2nd 3rd 4th 5th 1st 2nd 3rd 4th 5th      Frequency (Hz)      Frequency (Hz)      Frequency (Hz)
 
Containment
 
&  4.20 12.62 16.87 25.00 33.41  4.20 13.19 17.54 27.01 33.76 10.77 21.17 43.47  -  - internal                structure               
 
Auxiliary  8.89 25.79 35.79 41.76  -  8.20 23.10 29.66 38.82  -  9.85 49.0  -  -  - building               
 
Diesel  1.44 31.54 54.69  -  -  1.44 34.44 53.35  -  - 14.01 74.09  -  -  - generator building               
 
River intake (b)  .23  6.60 16.33 28.69 41.86  .26  7.66 21.68 44.28  - 10.22 41.65  -  -  - structure               
 
44.29(a)
Intake structure  .15  .27  1.19 11.28 12.47  .24  1.14  9.71 12.51 36.51  5.14 42.72  -  -  - storage pond               
: a. 6th mode
: b. The river intake structure was originally designed as a Category I structure, but has since been downgraded to non-seismic.
 
FNP-FSAR-3
 
REV 31  5/08 TABLE 3.7-6 COMPARISON OF TRANSLATIONAL AND TORSIONAL FREQUENCIES
 
Horizontal Translational Torsional Mode Frequency Frequency Structure  NR      (Hz)        (Hz)(a)        Internal structures 1 16.21  26.72 2 41.51  63.31 Auxiliary building 1  8.89  34.92 2 25.77 120.35
 
3 35.97  -
 
4 41.76  -
: a. Frequencies greater than 25 Hz.
 
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.7-7 CONTAINMENT SHELL COMPARISON OF RESPONSE SPECTRUM AND TIME HISTORY ANALYSIS, SAFE SHUTDOWN EARTHQUAKE (EAST - WEST DIRECTION)
Absolute Acceleration, g
 
Response Spectrum Time History Elevation, ft  Analysis  Analysis 99.5 .0630 (a) .1206    116.50 .0782 (a) .1293    129.00 .0933 (a) .1341    142.00 .1077 .1384    155.00 .1215 .1564    174.00 .1451 .1742    193.00 .1705 .2008    212.58 .2011 .2380    228.12 .2255 .2706    243.67 .2519 .3049    251.42 .2659 .3216    269.17 .2927 .3514    282.17 .3101 .3705
: a. A minimum of 0.10 g has been used.
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.7-8 MAXIMUM ALLOWABLE SPAN BETWEEN SEISMIC RESTRAINTS FOR PIPES 2 IN. AND UNDER Max. Allowable Span of Pipe + Natural Pipe Diam. Water + Insulation Frequency fn (a)    (in.)            (ft)            (Hz) 3/8 3.5 26.6    1/2 4.0 27.1    3/4 5.0 22.6    1 6.0 20.2    1-1/2 7.0 22.1    2 8.0 21.4
: a. Figures based on a limiting natural frequency of 20 Hz, derived from the following equation:
 
fn = natural frequency - Hz
 
    = 0.743 fn EI W L=2  where: L = span ft. E = Young's modulus of elasticity - psi    I = Moment of inertia (in.
: 4)
W = Pipe weight/ft
 
REV 21  5/08 1/2 SAFE SHUTDOWN EARTHQUAKE GROUND SPECTRA 0.05 g (HORIZONTAL & VERTICAL)
JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-1
 
REV 21  5/08 SAFE SHUTDOWN EARTHQU AKE GROUND SPECTRA 0.10 g (HORIZONTAL & VERTICAL)
JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-2
 
REV 21  5/08 SYNTHESIZED TIME HISTORY (1/2 SSE & SSE)
JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-3
 
REV 21  5/08 TIME HISTORY SPECTRUM ENVELOPE ON RESPONSE SPECTRUM (1/2 SSE)
JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-4
 
REV 21  5/08 TIME HISTORY SPECTRUM ENVELOPE ON RESPONSE SPECTRUM (SSE)
JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-5
 
REV 21  5/08 A LUMPED-MASS MODEL OF STRUCTURE FOUNDATION SYSTEM JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-6
 
REV 21  5/08 CONSTANTS x,  AND z FOR RECTANGULAR BASES JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-7
 
REV 21  5/08 CONTAINMENT - SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-8
 
REV 21  5/08 CONTAINMENT - SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-9
 
REV 21  5/08 CONTAINMENT - SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-10
 
REV 21  5/08 CONTAINMENT - SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-11
 
REV 21  5/08 CONTAINMENT - SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-12
 
REV 21  5/08 CONTAINMENT - SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-13
 
REV 21  5/08 INTERNAL STRUCTURE - SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-14
 
REV 21  5/08 INTERNAL STRUCTURE - SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-15
 
REV 21  5/08 INTERNAL STRUCTURE - SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-16
 
REV 21  5/08 INTERNAL STRUCTURE - SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-17
 
REV 21  5/08 INTERNAL STRUCTURE - SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-18
 
REV 21  5/08 INTERNAL STRUCTURE - SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-19
 
REV 21  5/08 CONTAINMENT AND INTERNAL STRUCTURE MATHEMATICAL MODEL JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-20
 
REV 21  5/08 CONTAINMENT AND INTERNAL STRUCTURE MATHEMATICAL MODEL JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-21
 
REV 21  5/08 POLAR CRANE BRACKET AND SEISMIC RETAINER JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-22
 
REV 21  5/08 AUXILIARY BUILDING MATHEMATICAL MODEL JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-23
 
REV 21  5/08 AUXILIARY BUILDING MATHEMATICAL MODEL JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-24
 
REV 21  5/08 AUXILIARY BUILDING SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-25
 
REV 21  5/08 AUXILIARY BUILDING SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-26
 
REV 21  5/08 AUXILIARY BUILDING SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-27
 
REV 21  5/08 AUXILIARY BUILDING SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-28
 
REV 21  5/08 AUXILIARY BUILDING SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-29
 
REV 21  5/08 AUXILIARY BUILDING SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-30
 
REV 21  5/08 DIESEL GENERATOR BUILDING SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-31
 
REV 21  5/08 DIESEL GENERATOR BUILDING SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-32
 
REV 21  5/08 DIESEL GENERATOR BUILDING SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-33
 
REV 21  5/08 DIESEL GENERATOR BUILDING SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-34
 
REV 21  5/08 DIESEL GENERATOR BUILDING SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-35
 
REV 21  5/08 DIESEL GENERATOR BUILDING SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-36
 
REV 21  5/08 RIVER INTAKE STRUCTURE SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-37
 
REV 21  5/08 RIVER INTAKE STRUCTURE SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-38
 
REV 21  5/08 RIVER INTAKE STRUCTURE SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-39
 
REV 21  5/08 RIVER INTAKE STRUCTURE SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-40
 
REV 21  5/08 RIVER INTAKE STRUCTURE SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-41
 
REV 21  5/08 RIVER INTAKE STRUCTURE SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-42
 
REV 21  5/08 INTAKE STRUCTURE AT STORAGE POND SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-43
 
REV 21  5/08 INTAKE STRUCTURE AT STORAGE POND SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-44
 
REV 21  5/08 INTAKE STRUCTURE AT STORAGE POND SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-45
 
REV 21  5/08 INTAKE STRUCTURE AT STORAGE POND SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-46
 
REV 21  5/08 INTAKE STRUCTURE AT STORAGE POND SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-47
 
REV 21  5/08 INTAKE STRUCTURE AT STORAGE POND SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-48
 
REV 21  5/08 VENT STACK SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-49
 
REV 21  5/08 VENT STACK SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-50
 
REV 21  5/08 POND SPILLWAY STRUCTURE SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-51
 
REV 21  5/08 POND SPILLWAY STRUCTURE SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-52
 
REV 21  5/08 POND SPILLWAY STRUCTURE SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-53
 
REV 21  5/08 POND SPILLWAY STRUCTURE SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-54
 
REV 21  5/08 POND SPILLWAY STRUCTURE SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-55
 
REV 21  5/08 POND SPILLWAY STRUCTURE SEISMIC RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-56
 
REV 21  5/08 CONTAINMENT AND INTERNAL STRUCTURE FREQUENCIES AND MODE SHAPES JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-57
 
REV 21  5/08 CONTAINMENT AND INTERNAL STRUCTURE FREQUENCIES AND MODE SHAPES JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-58
 
REV 21  5/08 MODAL INERTIA FORCES FOR CONTAINMENT JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-59
 
REV 21  5/08 MODAL INERTIA FORCES FOR CONTAINMENT JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-60
 
REV 21  5/08 RESPONSE SPECTRUM AT REACTOR SUPPORT ELEVATION JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-61
 
REV 21  5/08 SEISMIC INSTRUMENTATION JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-62
 
REV 21  5/08 EARTHQUAKE EVALUATION PROCEDURE FOR CATEGORY 1 STRUCTURES JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-63
 
REV 21  5/08 MATHEMATICAL MODEL OF REACTOR INTERNALS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-64
 
REV 21  5/08 FIRST MODE VIBRATION OF REACTOR INTERNALS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.7-65
 
FNP-FSAR-3
 
3.8-1 REV 21  5/08 3.8 DESIGN OF CATEGORY I STRUCTURES
 
3.8.1 CONCRETE CONTAINMENT The containment completely encloses the reacto r, the reactor coolant systems, the steam generators, and portions of the auxiliary and engineered safeguards systems. It ensures that
 
an acceptable upper limit for leakage of radioactive materials to the environment will not be exceeded even if gross failure of the reactor coolant system occurs. The structure provides
 
biological shielding during normal operation and following a loss-of-coolant accident (LOCA). It
 
also provides a vapor containment following an accident inside the containment. Further
 
information relative to the containment is covered in topical BC-TOP-5, which provides the
 
bases for design, construction, testing, and surveillance for the prestressed concrete
 
containment. 
 
3.8.1.1  Description of the Containment
: a. Physical Description 
 
The containment is a prestressed, reinforced concrete cylindrical structure with a
 
shallow domed roof and a reinforced concrete foundation slab with provision for a
 
reactor cavity at the center. The cylindrical portion of the containment is
 
prestressed by a post-tensioning system composed of horizontal and vertical
 
tendons. The horizontal tendons are placed in three 240-degree segments using
 
three buttresses spaced 120 degrees apart as supports for the anchorages. The
 
dome has a three-way tendon pattern in which groups of tendons intersect at 120
 
degrees. The concrete foundation is a conventionally reinforced mat. A
 
continuous access gallery is provided beneath the base slab for installation and
 
inspection of the vertical tendons.
A 1/4-in.-thick welded steel liner is attached to the inside face of the concrete. 
 
The floor liner is installed on top of the foundation slab and is then covered with
 
concrete. 
 
Principal nominal dimensions of the containment are as follows: 
 
Interior diameter (ft)    130 Interior height (ft)    183 Cylindrical wall thickness (ft)    3 3/4 Dome thickness (ft)    3 1/4 Foundation slab thickness (ft)  9  Liner plate thickness (in.)    1/4 Internal free volume (ft
: 3)    2.0 x 10 6  The geometry and typical details of the containment and liner plate are shown in
 
figures 3.8-1, 3.8-2, and drawing D-176145.
 
FNP-FSAR-3
 
3.8-2 REV 21  5/08  b. Description of Post-Tensioning System 
 
The prestressed, post-tensioning system is a low relaxation Inland-Ryerson
 
BBRV buttonhead system using 170 wires of 1/4-in. diameter per tendon. The
 
pertinent features of the post-tensioning system are given in table 3.8-1. The
 
tendons are installed in metal sheaths which form ducts through the concrete
 
between anchorage points. Trumpets, which are enlarged ducts attached to the
 
bearing plate, allow the wires to spread out at the anchorage to suit washer hole
 
spacing and facilitate field buttonheading of wires. Figure 3.8-5 shows details of
 
trumpet and sheathing. Sheaths are provided with a valved vent at the highest
 
points of curvature to permit release of pockets of entrapped air during greasing
 
operations. Drains are provided at the lowest points of curvature to remove
 
accumulated water prior to installing tendons. In the process of greasing
 
operations, the vents and drains are closed and sealed. 
 
The prestressing wire is protected against atmospheric corrosion during its
 
shipment and installation, and during the life of the containment. Prior to
 
shipment, the wire is coated with a thin film of petrolatum containing rust
 
inhibitors. The interior surface of the sheathing is coated with a suitable material
 
during manufacture to minimize removal of the petrolatum from the tendon wires
 
during pulling through the sheathing. The sheathing filler material used for
 
permanent corrosion protection is a modified, refined petroleum oil base product.
 
The material is pumped into the sheathing after stressing. 
 
The vertical tendons are anchored at the top of the ring girder and at the bottom
 
of the foundation slab. The hoop tendons are anchored at buttresses 240
 
degrees apart, bypassing an intermediate buttress. The anchorages of each
 
successive hoop tendon are progressively offset 120 degrees from the one
 
beneath it. The three way dome tendons are anchored at the side of the ring
 
girder.
For the arrangement of the prestressing tendons, especially at penetrations, anchorage zones, connections and joints, see figures 3.8-1, 3.8-2, and drawing
 
D-176145.
: c. Description of the Equipment Hatch, Personnel Locks, and Electrical Penetration.
 
The geometry of the equipment hatch is shown on drawing D-176145. The
 
personnel access lock and the auxiliary access lock are both shown on figure
 
3.8-40. The electrical penetrations are shown on drawing D-176151. 
 
3.8.1.2  Applicable Codes, Standards and Specifications The following codes, standards, specifications, design criteria, NRC Regulatory Guides, and
 
industry standard practices constitute the basis for the design and construction of the
 
containment. Modifications to these codes, st andards, etc. are made when necessary, to meet
 
the specific requirements of the structure. These modifications are indicated in the sections
 
where references to the codes, standards, etc. are made.
FNP-FSAR-3
 
3.8-3 REV 21  5/08 Codes  1. ACI 214-65 "Recommended Practices for Evaluation  of Compression Test Results of Field Concrete".
: 2. ACI 301-66 "Specifications for Structural  Concrete for Buildings".
: 3. ACI 306-66 "Recommended Practice for Cold  Weather Concreting".
: 4. ACI 307-69 "Specification for the Design  and Construction of Reinforced Concrete Chimneys".
: 5. ACI 311-64 "Recommended Practice for Concrete  Inspection".
: 6. ACI 315-65 Manual of Standard Practice for Detailing Reinforced Concrete Structures.
: 7. ACI 318-63 "Building Code Requirements for    Reinforced Concrete".
: 8. ACI 605-59 "Recommended Practice for Hot  Weather Concreting".
: 9. ACI 613-54 "Recommended Practice for Selecting  Proportions for Concrete".
: 10. ACI 614-59 "Recommended Practices for Measuring,  Mixing, and Placing Concrete".
: 11. AISC Manual of Steel Construction,  1969 Edition.
: 12. AWS D2.0-69 "Specifications for Welded  Highway and Railway Bridges".
: 13. ASME "Boiler and Pressure Vessel Code",  Sections III, VIII, and IX -
1968 Edition.
: 14. ICBO "Uniform Building Code" - 1970 Edition. 15. SBCC "Southern Standard Building Code" - 1969 Edition.
FNP-FSAR-3
 
3.8-4 REV 21  5/08  16. CFR  Federal Register , Title 29, Part 1910, Department of Labor, Occupational Safety, and Health Standards.
: 17. CFR Code of Federal Regulations , Title 10, Part 100, Appendix A,    Part 50.
: 18. AWS D1.1-86 "Structural Welding Code - Steel". 19. NCIG-01, Rev. 2 "Visual Weld Acceptance Criteria for Structural Welding at Nuclear Power Plants" - EPRI NP-5380.
Standards  1. ACI Manual of Concrete Inspection - sp 2  Specifications
: 1. CMAA  "Specifications for Electric Overhead Traveling Cranes" -
No. 70 - 1970 Edition. 2. ASTM  The specifications utilized are identified in the applicable 
 
sections.
Design Criteria
: 1. ASCE "Wind Forces on Structures", Paper No. 3269. 2. NRC "Nuclear Reactor and Earthquake", Publication TID 7024.
 
Industry Standard Practices Naval Documents "Design of Protective Structures - A New Concept of 
 
Structural Behavior" - Naval Document P-51.
NRC Regulatory Guides Regulatory Guide No. 1.10 "Mechanical (Cadweld) Splices in Reinforcing Bars of Category I Concrete Structures."
 
Regulatory Guide No. 1.12 "Instrumentation for Earthquakes."
 
Regulatory Guide No. 1.13 "Fuel Storage Facility Design Basis."
 
Regulatory Guide No. 1.15 "Testing of Reinforcing Bars for Category I Concrete 
 
Structures."
 
Regulatory Guide No. 1.18 "Structural Acceptance Test For Concrete Primary Reactor Containments."
 
FNP-FSAR-3
 
3.8-5 REV 21  5/08 Regulatory Guide No. 1.19 "Nondestructive Examination of Primary Containment Liner Welds."
Regulatory Guide No. 1.28 "Quality Assurance Program Requirements - Design and Construction."
Regulatory Guide No. 1.29 "Seismic Design Classification."
Regulatory Guide No. 1.31 "Control of Stainless Steel Welding."
Regulatory Guide No. 1.35 Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures."
Regulatory Guide No. 1.38 "Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water Cooled Nuclear Power Plants,"
Regulatory Guide No. 1.46 "Protection Against Pipe Whip Inside Containment."
Regulatory Guide No. 1.54 "Quality Assurance Requirements for Protective Coatings Applied to Water Cooled Nuclear Power Plants."
 
Regulatory Guide No. 1.55 "Concrete Placement in Category I Structures."
Regulatory Guide No. 1.59 "Design Basis Floods for Nuclear Power Plants."
Regulatory Guide No. 1.64 "Quality Assurance Program Requirements for the Design of Nuclear Power Plants." 
 
Discussions on the compliance with and interpretations of Regulatory Guides are presented in
 
appendix 3A. 
 
FNP-FSAR-3
 
3.8-6 REV 21  5/08 BECHTEL POWER CORPORATION TOPICAL REPORTS
 
B-TOP-3 "Design Criteria for Nuclear Power Plants Against Tornadoes," March 1970.
 
BP-TOP-1 "Seismic Analysis of Piping Systems," February 1974.
 
BC-TOP-4 "Seismic Analysis of Structures and Equipment for Nuclear Power Plants," September 1972.
BC-TOP-5 "Prestressed Concrete Nuclear Reactor Containment Structures," December 1972.
 
BC-TOP-7 "Full Scale Buttress Test for Prestressed Nuclear Containment Structures," Revision 0, September 1972.
 
BC-TOP-8 "Tendon End Anchor Reinforcement Test," Revision 0, September 1972.
 
BC-TOP-1 "Containment Building, Liner Plate Design Report," Revision 1, December 1972. BN-TOP-1 "Testing Criteria for Integrated Leak - Rate Testing of Primary Containment Structures for Nuclear Power Plants," Revision 1, November 1972.
 
BC-TOP-9A "Design of Structures for Missile Protection," Revision 2, September 1974.
 
3.8.1.2.1 Discussions of Codes and Standard Specifications
: a. American Concrete Institute, Building Code Requirements for Reinforced Concrete (ACI-318-71) 
 
The ACI-318-71 code has not been used in the design of the containment. 
: b. American Concrete Institute, Concrete Shell Structures (ACI-334) 
 
In order to ensure consistency of the analysis results, equilibrium checks of internal stresses and external loads were made in a similar manner to that
 
described in Sections 202 (d) and 202 (e) of the ACI Committee 334 Report, "Concrete Shell Structures Practice and Commentary," Journal of the ACI , Title No. 61, Proceedings V 61, No. 9, pp. 1091-1108, 1964. 
: c. American Concrete Institute, Specifications for the Design and Construction of Reinforced Concrete Chimneys (ACI-307-69) 
 
The Specifications for the Design and Construction of Reinforced Concrete Chimneys (ACI-307-69) is not used for consideration of self-relieving effects of thermal stresses. Instead, reference is made to the considerations outlined in
 
Section 2.5.6.3.3 of ACI-349 Committee Report.
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3.8-7 REV 21  5/08  d. American Concrete Institute and American Society of Mechanical Engineers, "Proposed Standard Code for Concrete Reactor Vessels and Containments
."  (ACI-359) 
 
The joint-ACI-ASME (ACI-359) document, Proposed Standard Code for Concrete Reactor Vessels and Containments , has not been used in the design of the containment. 
 
3.8.1.2.2 Structural Specifications Structural specifications are prepared to cover the areas related to design and construction of
 
the containment. These specifications emphasize important points of the industry standards for
 
the design and construction of the containment, and reduce options that otherwise would be
 
permitted by the industry standards. Unless specifically noted otherwise, these specifications
 
do not deviate from the applicable industry standards. They cover the following areas: 
: a. Concrete material properties. 
: b. Placing and curing of concrete. 
: c. Reinforcing steel and splices. 
: d. Post-tensioning system. 
: e. Liner plate and penetration assemblies. 
 
3.8.1.3  Loads and Loading Combinations The containment is designed for all credible conditions of loadings, including normal loads, loads resulting from a loss-of-coolant accident, test loads, and loads due to adverse
 
environmental conditions. 
 
Critical loading combinations are those caused by a postulated loss of reactor coolant, by a
 
postulated earthquake, or by a pipe rupture in the containment. 
 
Wind and tornado loads, flood design bases, and seismic loads are given in sections 3.3, 3.4
 
and 3.7, respectively. Missile effects and the postulated pipe rupture effects are discussed in
 
sections 3.5 and 3.6. Chapter 15.0, "Accident Analyses", provides information on the design
 
pressure load. 
: 1. Loads 
 
The following loads are considered:  dead loads; live loads; prestressing loads; earthquake loads; pipe rupture loads; loss-of-coolant accident loads; operating
 
thermal loads; wind and tornado loads; external pressure loads; hydrostatic
 
loads; and test loads. 
 
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3.8-8 REV 21  5/08  a. Dead Loads: 
 
Structural dead loads consist of the weight of the containment wall, dome, base slab, interior framing and slabs, all internal structures, equipment, and major piping and electrical conductors. 
: b. Live Loads: 
 
Live loads consist of design floor loads, equipment live loads, and all live loads transmitted by the internal structures. 
 
The operating floor slab is designed for either of these live loads: 
 
Floor gratings          450 psf Concrete slabs      1000 psf 
 
Equipment live loads are those specified on drawings supplied by the manufacturer of the equipment. 
: c. Prestressing Loads: 
 
The compressive forces due to the prestressing tendons are taken into consideration. 
: d. Earthquake Loads: 
 
Earthquake loads are predicated on a basis of the 1/2 safe shutdown earthquake (1/2 SSE), having a horizontal ground acceleration of 0.05 g
 
and a vertical ground surface acceleration of 0.033 g. 
 
In addition, a safe shutdown earthquake (SSE), having a horizontal ground acceleration of 0.10 g and a vertical ground surface acceleration
 
of 0.067 g, is used to check the design to ensure that loss of structural
 
functions will not occur.
 
Seismic response spectrum curves are given in section 3.7 for both horizontal and vertical ground motions. A dynamic analysis is used to
 
compute the seismic loads for the design of structural elements. 
: e. Pipe Rupture Loads: 
 
Pipe rupture loads represent the forces or pressure on the structure due to the rupture of any one pipe.
: f. Loss-of-Coolant Accident Loads: 
 
The design pressure and temperature of the containment are greater than the peak pressure and temperature that would result from a postulated
 
complete blowdown of the reactor coolant. This might occur through the FNP-FSAR-3
 
3.8-9 REV 21  5/08 rupture of the reactor coolant system, up to and including the hypothetical
 
double ended severance of the largest reactor coolant pipe. 
 
Pressure transients resulting from the LOCA (see section 6.2) serve as the basis for a containment design pressure of 54 psig. 
 
The design pressure will not be exceeded during any subsequent long term pressure transients caused by the combined effects of heat sources.
 
These effects will be overcome by the combination of safety features and
 
heat sinks. 
 
The temperature gradient through the containment wall during operating conditions and during LOCA is shown in figure 3.8-15. The variation of
 
temperature with time and the expansion of the liner plate with
 
temperature are considered in determining the thermal stresses
 
associated with the LOCA. 
: g. Operating Thermal Loads: 
 
The temperature gradient through the containment wall during normal operating condition is shown in figure 3.8-15. For this condition, a low
 
mean winter temperature of 20°F is assumed at the exterior surface of the
 
concrete, and an operating temperature of 120°F on the interior surface of
 
the concrete. For normal operation or any other long term period, the
 
concrete temperature shall not exceed 150°F, except for local areas, which are allowed to have increased temperatures not to exceed 200°F, such as heat affected zones around penetrations. Thermal loads caused
 
by transient wall temperatures during a prolonged shutdown are also
 
considered in the design. 
 
The thermal loads caused by the expansion and contraction of piping and equipment are considered in the design whenever applicable. 
: h. Wind and Tornado Loads: 
 
The wind loadings and tornado loadings are discussed, respectively, in subsections 3.3.1 and 3.3.2. 
 
The containment is designed to withstand the effects of these wind and tornado loadings, and to provide protection against tornado missiles. 
 
The structure is analyzed for tornado loadings not coincident with the SSE.   
: i. External Pressure Loads: 
 
A pressure of 3 psi from the exterior to the interior of the containment is assumed and applied as an external pressure on the containment. 
 
FNP-FSAR-3
 
3.8-10 REV 21  5/08  j. Hydrostatic Loads: 
 
Lateral hydrostatic pressure resulting from ground or flood water, as well as buoyant forces resulting from the displacement of ground or flood
 
water by the structure, are accounted for in the design. 
 
The water levels considered are: 
 
Ground water, elevation 125 ft.
Flood water, elevation 144.2 ft. 
: k. Test Loads: 
 
Upon completion of construction, the containment and its penetrations are tested at 115 percent of the design pressure. This pressure is considered
 
in the design. 
: 2. Loading Combinations 
 
In general, two types of loading cases are considered in the design of the containment. 
: a. The design loading case for which the working stress method is used. 
: b. The factored loading case for which the ultimate stress method is used. 
 
The following terms are used in the loading combination equations: 
 
C = Required capacity of the containment to resist factored loads. 
 
    = Capacity reduction factor (defined in subsection 3.8.1.3.1)
D = Dead loads of containment, interior structures, and equipment, plus any other permanent contributing loads. 
 
E = 1/2 Safe-shutdown earthquake load. 
 
E'= Safe Shutdown Earthquake load. 
 
F = Prestress load. 
 
H = Force on structure due to thermal expansion of pipes under operating conditions. 
 
L = Appropriate live load. 
 
P = Design accident pressure load. 
 
R = Force or pressure on structure due to rupture of any one pipe.
FNP-FSAR-3
 
3.8-11 REV 21  5/08 T a= Thermal loads due to the accident temperature gradient through the wall, based on a temperature corresponding to the unfactored design accident
 
pressure. 
 
T o= Thermal loads due to the normal operating temperature gradient through the walls. 
 
T s= Thermal loads due to transient wall temperatures over a prolonged shutdown.  (20°F at outside face, 70°F at center, 212°F at inside face.) 
 
W = Wind load. 
 
W t= Tornado load. 
: a. Design Loading Case: 
 
In the basic working stress design, the containment is designed for the following loading combinations: 
: 1) D + F + L (construction case) 
: 2) D + F + L + T o + E (or W) (operating case) 
: 3) D + F + L + P + T a (design accident case)
: 4) D + F + L + T S + E (or W) (prolonged shutdown case) 
: 5) D + F + L + 1.15P (test case) 
: b. Factored Loading Case: 
 
This loading case utilizes the capacity of the structure to verify its ability to withstand loading combinations in excess of the maximum that could be
 
expected under the LOCA conditions. The design of the containment
 
satisfies the following loading combinations: 
: 1) C = 1/ (1.0D + 1.5P + 1.0T a + 1.0F) 
: 2) C = 1/ (1.0D  + 1.25P + 1.0T a  + 1.25H            + 1.25 E (or 1.25W)+ 1.0F)
: 3) C = 1/ (1.0D  + 1.25H + 1.0R + 1.0F            + 1.25E (or 1.25W) +1.0T o)    4) C = 1/ (1.0D  + 1.25H + 1.0F + 1.0W t + 1.0T o) 
: 5) C = 1/ (1.0D  + 1.0P + 1.0T a + 1.0H + 1.0E' + 1.0F)
 
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3.8-12 REV 21  5/08    6) C = 1/ (1.0D  + 1.0H + 1.0R + 1.0E'            1.0F + 1.0T o)
Equation 1 ensures that the containment will have the capacity to withstand pressure loadings at least 50 percent greater than those calculated for the postulated loss-of-
 
coolant accident alone. 
 
Equation 2 ensures that the containment will have the capacity to withstand loadings at least 25 percent greater than those calculated for the postulated loss-of-coolant accident
 
with a coincident 1/2 SSE. 
 
Equation 3 ensures that the containment will have the capacity to withstand earthquake loadings 25 percent greater than those calculated for the 1/2 SSE coincident with the
 
associated rupture of any attached piping. 
 
Equation 4 ensures that the containment will have the capacity to withstand a tornado loading. 
 
Equations 5 and 6 ensure that the containment will have the capacity to withstand either the postulated loss-of-coolant accident or the rupture of any attached piping coincident
 
with the SSE. 
 
3.8.1.3.1 Capacity Reduction Factors The capacities of all load carrying structural elements are reduced by capacity reduction factors
() as given below. These factors provide for the possibility that small adverse variations in material strengths, workmanship, dimensions, control, and degree of supervision, while individually within the required tolerances and the limits of good practice, occasionally may
 
combine to result in undercapacity. 
 
Capacity reduction factors: 
 
  = 0.90 for concrete in flexure. 
 
  = 0.85 for tension, shear, bond, and anchorage in concrete. 
 
  = 0.75 for spirally reinforced concrete compression members. 
 
  = 0.70 for tied compression members. 
 
  = 0.90 for fabricated structural steel. 
 
  = 0.90 for mild reinforcing steel (not prestressed in direct tension excluding splices). 
 
  = 0.90 for mild reinforcing steel with welded or mechanical splices.  (For lap splices,  = 0.85 as above for bond and anchorage.)
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3.8-13 REV 21  5/08
  = 0.95 for prestressed tendons in direct tension. 
 
3.8.1.3.2 Prestress Losses In accordance with ACI 318-63, the design provides for the following prestress losses: 
: a. Slip at anchorage. 
: b. Elastic shortening of concrete. 
: c. Creep of concrete. 
: d. Shrinkage of concrete. 
: e. Relaxation of steel stress. 
: f. Frictional loss due to intended or unintended curvature in the tendons. 
 
The following relationships and assumed values have been used in conjunction with these
 
categories of prestress losses: 
: a. Slip at Anchorage 
 
There is no loss for hoop, dome, or vertical tendons due to slippage at anchorage since the buttonheaded BBRV system has physi cal characteristics that eliminate
 
this loss. Lift off readings will be made to confirm that seating losses are
 
negligible. 
: b. Elastic Shortening of Concrete 
 
csc sE2Ex=    where  s is the change in tendon stress due to elastic shortening, c is the maximum concrete stress for that general area, E S is the modulus of elasticity of the steel, and E c is the modulus of elasticity of the concrete at the time of stressing. 
 
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3.8-14 REV 21  5/08  c. Creep of Concrete Creep loss  =
s'c c 6Exx in in10x317  where c is the maximum concrete stress for that general area, 'c is the concrete stress level at which the value of 317 x 10 6 in/in was determined, and ES is the modulus of elasticity of the steel. 
: d. Shrinkage of Concrete 
 
Shrinkage loss = 170 x 10 6 in/in x E S where E S is the modulus of elasticity of the steel. e. Relaxation of Steel Stress 
 
The relaxation loss was assumed to be 8 1/2 percent of the seating stress of the tendon. 
: f. Frictional Losses 
 
The curvature friction coefficient, µ, and the wobble friction coefficient, K, which were used are 0.08 and 0.0003, respectively. 
 
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3.8-15 REV 21  5/08 The following tabulation shows the magnitude of losses and the final effective prestress at the end of 40 years for a typical dome, hoop, and vertical tendon: (a)    DOME (ksi) HOOP (ksi) VERTICAL (ksi)
 
Average overstress 179.8 192.2 176.7
 
Average friction 11.3 36.0 6.4 loss   
 
Seating loss 0.0 0.0 0.0
 
Average seating 172.4 173.4 172.3 stress   
 
Elastic shortening 5.0 5.0 2.7 loss   
 
Creep loss 7.6 7.6 4.1
 
Shrinkage loss 4.9 4.9 4.9
 
Relaxation loss 14.7 14.7 14.6
 
40-year minimum 136.3 124.0 144.0 effective stress   
 
________________
: a. The operating licenses for both FNP units have been renewed and the original licensed
 
operating terms have been extended by 20 year
: s. Containment tendon prestress was evaluated as a time-limited aging analysis in accordance with 10 CFR 54.21 during the license renewal process. Continued trending of prestress losses for tendons will ensure that minimum effective prestress for typical dome, hoop, and vertical tendons will remain above the required
 
values for the period of extended operation (see chapter 18, subsection 18.4.3).
 
FNP-FSAR-3
 
3.8-16 REV 21  5/08 3.8.1.4  Design and Analysis Procedures
 
The containment is analyzed for various loading combinations, considering the values of individual loads that generate the most significant stress condition for each component and member of the structure. 
 
The critical areas for analysis are as follows: 
: a. The intersection between cylinder wall and base slab. 
: b. Ring girder. 
: c. Behavior of the base slab relative to an elastic foundation. 
: d. Transient temperature gradients in the steel liner plate and concrete. 
: e. Tendon anchorage zones. 
 
Classical theory, empirical equations, and numerical methods are applied as necessary for the
 
analysis of structural elements. They are described in Section 7 of BC-TOP-5. 
 
The design methods incorporate several phases as described in Section 6 of BC-TOP-5. 
 
Improved assumptions as to material properties including the effects of creep, shrinkage and
 
cracking on concrete are used in design. Analysis and design of tendon anchorage zones and
 
reinforcement in buttresses are discussed in BC-TOP-5, BC-TOP-7, and BC-TOP-8. The
 
method of analyzing the effects of penetrations, the thickening of walls, reinforcements and
 
embedments, etc., are discussed in Section 7 of BC-TOP-5. The design of the liner and its
 
anchorage system is covered in BC-TOP-1 and BC-TOP-5. Information on analyses for
 
computation of seismic loads is provided in section 3.7. 
 
3.8.1.4.1 Analytical Techniques 
 
The analysis of the containment consists of two parts:  the axisymmetric analysis and the non-axisymmetric analysis. The axisymmetric analysis is performed by utilizing a finite element computer program for combinations of the individual loading cases of dead, live, temperature, pressures, and prestress loads. The axisymmetric finite element representation of the
 
containment assumes that the structure is axisymmetric. This does not account for the
 
buttresses, penetration, brackets, and anchors. These items, together with the lateral loads due
 
to earthquakes, winds, tornados, and various c oncentrated loads, are considered in the non-axisymmetric analysis. 
: 1. Axisymmetric Analysis 
 
The containment is considered an axisymmetric structure for the overall analysis. Although there are deviations from this ideal shape, the deviations are
 
usually localized and can be handled by s pecial analyses; hence, axisymmetric analyses are considered acceptable.
FNP-FSAR-3
 
3.8-17 REV 21  5/08  The axisymmetric analysis of the cont ainment is performed by Bechtel's "FINEL" computer program (CE 316-4) based on the finite element method. Because of program limitations, the upper and lower portions of the containment are
 
analyzed separately, to permit the use of a greater number of elements for those
 
areas of the structure which are of major concern, i.e., the ring girder and the
 
haunch connecting the cylindrical shell to the base slab. 
 
The entire concrete structure is modeled by continuously interconnected elements. The geometry of the mesh allows for the locations and shapes of
 
narrow elements representing reinforcing steel superimposed on the
 
corresponding concrete elements. 
 
The liner plate is simulated by a layer of elements attached to the interior surfaces of the concrete structure. 
 
The finite element mesh of the structure is extended into the foundation to account for the elastic nature of the foundation materials and its effect on the
 
behavior of the base slab. The tendon access gallery is designed as a separate
 
structure.
The use of the finite element analysis permits accurate determination of the stress pattern at any location on the structure. 
 
The finite element mesh for axisymmetric loads is shown on figures 3.8-16 and 3.8-17. 
: 2. Non-Axisymmetric Analysis 
 
The non-axisymmetric analyses of the containment include the following: 
: a. Seismic loadings. 
: b. Wind and tornado loadings. 
: c. Buttress and tendon anchorage zones. 
: d. Large penetration openings. 
: e. Small penetration openings. 
: f. Non-axisymmetric internal structure and equipment. 
: a. Seismic Loadings 
 
The analysis of the non-axisymmetric seismic loadings is performed by Bechtel's "Axisymmetric Shell and Solid Computer Program" (ASHSD), described in
 
appendix 3F. 
 
Details of the analysis are described in section 3.7.
FNP-FSAR-3
 
3.8-18 REV 21  5/08  b. Wind and Tornado Loadings 
 
Wind and tornado loadings are discussed in section 3.3. 
: c. Buttress and Tendon Anchorage Zones 
 
The containment has three buttresses. At each buttress, two out of any group of three-hoop tendons are spliced by anchoring on the opposite faces of the buttress, with the third tendon continuous through the buttress. 
 
Between the opposite anchorages in the buttress, the compressive forces exerted by the spliced tendons are twice as large as elsewhere on the buttress.
This value, combined with the effect of the tendon which is continuous
 
throughout the buttress, is 1.5 times the prestressing forces acting outside the
 
buttress. The thickness of the buttress is approximately 1.5 times the thickness
 
of the wall. Hence, the hoop stresses and strains, as well as the radial
 
displacements, may be considered as being nearly constant all around the
 
structure.
The vertical stresses and strains, caused by the vertical post-tensioning become constant a short distance away from the anchorages because of the stiffness of
 
the cylindrical walls. The effects of the buttresses on the overall behavior of the
 
containment are negligible under dead and prestressing loads. The stresses and
 
strains remain nearly axisymmetric despite the presence of the buttresses. 
 
The design of the tendon anchorage zones is based on two test programs conducted by Bechtel to demonstrate the adequacy of several reinforcing patterns for use in anchorage zone concrete in the base slab, buttresses and ring
 
girder. These tests have been undertaken to develop a more efficient design to
 
reduce reinforcement congestion and thereby facilitate the placement of high
 
quality concrete around the tendon anchorages. The test programs are as
 
follows: 
: 1. A full scale model of a simulated containment buttress containing several patterns of reinforcememt and types of tendon anchorages was constructed and tested. A detailed description of the test is presented in
 
Bechtel Topical Report BC-TOP-7. 
: 2. Two large concrete test blocks containing two patterns of reinforcement with different proportions of reinforcing bars were constructed and tested.
A detailed description of the test is presented in Bechtel Topical Report
 
BC-TOP-8.
The test results demonstrated satisfactory performance of the test anchorages.
The design of the tendon anchorage zones is based on the results and
 
recommendations of these tests. 
 
In addition, the local stress distribution in the immediate vicinities of the bearing plates has been investigated using the following methods.
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3.8-19 REV 21  5/08  1. The Guyon Equivalent Prism Method:  This method is based on the experimental photoelastic results and the equilibrium considerations of homogeneous and continuous media. It also considers the relative
 
bearing plate dimensions of the anchorages. 
: 2. The experimental test data presented by S. J. Taylor at the March 1967 London Conference of the Institute of Civil Engineers:  These data are used to evaluate the effect of the biaxial stresses at the anchorages, including the effects of the trumpets welded to the bearing plates. 
: 3. Leonhardt's Formula for determining the bursting forces in the anchorage zone of a prestressed concrete member. 
: d. Large Penetration Openings 
 
Large penetrations are defined as those having an inside diameter equal to or greater than 2.5 times the containment wall thickness. The equipment hatch falls
 
into this category.
The stresses at the opening are predicted by Bechtel's Computer program "SAP" (CE 779), which is capable of performing a static analysis of linear elastic three  dimensional structures utilizing the finite element method. 
 
The points delineating the outermost boundaries of the analytical model are located at two
 
penetration diameters beyond the edges of the opening, so that the behavior of the model along
 
the boundaries is compatible with that of the undisturbed cylindrical wall. 
 
Typical details of the equipment hatch are shown on drawing D-176145. Figure 3.8-30 shows the equipment hatch boundary lines. Figure 3.8-31 shows the finite element mesh used for the analyses of the equipment hatch. Figures 3.8-
 
32 through 3.8-35 show typical iso stress plots for the load combination D + F +
 
1.5P + T a. Typical results of the analyses of the equipment hatch are given in table 3.8-4. 
: e. Small Penetration Openings Small penetration openings are defined as those having an inside diameter less than 2.5 times the containment wall thickness.
The determination of the stresses at the openings due to applied moments and forces are based on a paper by Eringen et al.
(1)    Results of these analyses show the stresses to be well within the allowable limits.
Typical details of small penetrations are shown on drawings D-176145 and
 
D-176151. 
 
FNP-FSAR-3
 
3.8-20 REV 21  5/08 f. Non-Axisymmetric Internal Structure and Equipment The stresses due to the non-axisymmetric portion of the internal structure and equipment are combined with the stresses resulting from the other loading combinations described in subsection 3.8.1.3. 
 
3.8.1.4.2 Steel Liner Plate and Penetrations 
 
The steel liner plate and penetrations are designed to serve as the leakage barrier for the
 
containment. Typical details for the liner plate and penetrations are shown in figures 3.8-1, 3.8-2, and drawings D-176145 and D-176151. 
 
The design of the liner plate considers the composite action of the liner and the concrete
 
structure and includes the transient effects on the liner due to temperature changes during
 
construction, normal operation, and the loss-of-coolant accident. The changes in strains to be
 
experienced by the liner due to these effects, and those at the pressure testing of the
 
containment, are considered. 
 
The stability of the liner is achieved by anchoring it to the concrete structure. At all
 
penetrations, the liner is thickened to reduce stress concentration. The thickened plate is also
 
anchored to the concrete. 
 
For a detailed description of the liner plate stability, see appendix 3I. 
 
Insert plates are provided in the liner to transfer concentrated loads to the wall, slab, and dome
 
of the containment. Examples of these concentrated loads are polar crane brackets and floor
 
beam brackets. A typical bracket detail is shown in figure 3.8-1. 
 
The topical Report BC-TOP-1, "Consumers Power Company Palisades Nuclear Power Plant, Containment Building Liner Plate Design Report," October 1969, prepared by Bechtel Power Corporation, and submitted to the Nuclear Regulatory Commission, constitutes the basic
 
approach used in the design of the liner plate. 
 
There are minor differences in the design of the Farley Nuclear Plant from that presented in the
 
topical report. They are listed below. 
: a. The 1/4-in. liner plate material is ASTM A-285, Grade A, with a specified yield stress of 24,000 psi, instead of ASTM A-442, which has a specified yield stress of 30,000 psi. The lower yield stress would only tend to decrease the loads on
 
the anchors, as indicated in Section 3.4 of the report. 
: b. The welding of the stiffeners is 3/16 - 6 x 12 rather than 3/16 - 4 x 12. This does not invalidate the analysis, since the spring constants used are similar. 
: c. The stiffeners on the thickened plates are not welded with a double fillet weld.
Instead, the 3/16 - 6 x 12 welding is used for all stiffeners. 
 
FNP-FSAR-3
 
3.8-21 REV 21  5/08  d. A self-supporting dome is used rather than the truss supported dome. Details of the dome are shown in figure 3.8-2.
3.8.1.4.3 Description of Computer Programs 
 
Computer programs used in the design and analysis of the containment are described in
 
appendix 3F. 
 
3.8.1.5  Structural Acceptance Criteria
 
The fundamental acceptance criterion for the containment is the successful completion of the
 
structural integrity test, with measured responses within the limits predicted by analyses. The
 
limits are predicted based on test load combinations and code allowable values for stress, strain, or gross deformation for the range of material properties and construction tolerances
 
specified as described in Topical Report BC-TOP-5. In this way the margins of safety
 
associated with the design and construction of the containment are, as a minimum, the accepted margins associated with nationally recognized codes of practice. 
 
The structural integrity test is planned to yield information on both the overall response of the
 
containment and the response of localized areas, such as major penetrations and buttresses, which are important to its design functions. 
 
The design and analysis methods, as well as the type of construction and construction materials, are chosen to allow assessment of the structure's capability throughout its service life. Additionally, surveillance testing provides further assurances of the structure's continuing
 
ability to meet its design functions. 
 
Tables 3.8-2 and 3.8-3 show the calculated stresses and strains, respectively, as well as the
 
allowables, taken from critical sections of the containment structure. The ratios of the allowable
 
stresses and strains to the calculated stresses and the strains yield the margins of safety at
 
selected critical sections. Deviations in allowable stresses for the design loading conditions in
 
the working stress method are permitted if the factored load capacity requirements are fully
 
satisfied. 
 
For the margins of safety related to major local areas of the containment, such as the equipment
 
hatch, refer to table 3.8-4. The ratio of the allowable stress to the calculated stress, yields the
 
margins of safety. This information, together with the test information documented in BC-TOP-7 and BC-TOP-8, permits the assessment of the margins of safety for anchorage zones. 
 
The effect of three dimensional stress/strain fields on the behavior of the structure has been
 
considered in the "FINEL" computer program. 
 
3.8.1.6  Materials, Quality Control, and Special Construction Techniques
 
The following basic materials are used in the construction of the containment structure: 
 
FNP-FSAR-3
 
3.8-22 REV 21  5/08
: a. Concrete For: Tendon access gallery f'c (psi) =  5,000  Base slab f'c (psi) =  5,000  Cylindrical wall and dome f'c (psi) =  5,500    b. Reinforcing steel 
 
Deformed bars ASTM A-615 f y  (psi) = 60,000  Grade 60 Spiral bars ASTM A-82 f y  (psi) = 70,000
: c. Structural and miscellaneous steel
 
Rolled shapes, bars and plates ASTM A-36 f y  (psi) = 36,000 High strength bolts ASTM A-325 or A-490 Stainless steel ASTM A-240 Type 304 
: d. Containment steel liner plate and penetration sleeves
 
1/4-in. liner plates ASTM A-285 f y  (psi) = 24,000  Grade A Insert plates ASTM A-516 f y  (psi) = 38,000  Grade 70 Penetration sleeves pipes ASME SA-333 f y  (psi) = 32,000  Grade 6 Plates ASME SA-516 f y  (psi) = 38,000  Grade 70 
 
FNP-FSAR-3
 
3.8-23 REV 21  5/08 e. Post-tensioning system Prestressing wires ASTM A-421-65 f's  (psi) =240,000 Type BA      Bearing plates ASTM A-36-70a f y    (psi) = 36,000 Anchor heads HR 4142 or 4140 Alloy steel Bushing HFSM 4142 Tubing      Shims ASTM A 36-70a f y (psi) - 36,000    0.40/0.50 carbon    sheet steel Sheathing 22-gauge galvanized corrugated tubing
 
Materials and their quality control requirements are described in the following subsections. 
 
3.8.1.6.1 Reinforced Concrete 
: a. Concrete 
 
All concrete work is done in accordance with ACI 318-63, "Building Code Requirements for Reinforced Concrete", and ACI 301-66, "Specifications for
 
Structural Concrete for Buildings", except as otherwise stated herein, or in the
 
appropriate job specifications or design drawings. 
 
The concrete is a dense, durable mixture of sound coarse aggregates, fine aggregates, cement, and water. In some areas, fly ash is substituted for portions
 
of cement used in the concrete. Admixtures are added to improve the quality and workability of the plastic concrete during placement and to retard the set of
 
concrete. The sizes of aggregates, water reducing additives, and slumps are
 
selected to maintain low limits on shrinkage and creep. 
: b. Aggregates 
 
Aggregates comply with ASTM C-33-69, "Specifications for Concrete Aggregates". Acceptability of the aggregates is based on the initial tests listed in table 3.8-5.
Certain user tests, as indicated in table 3.8-5,  are performed on the aggregates used in every 5,000 cubic yards of concrete produced. 
 
FNP-FSAR-3
 
3.8-24 REV 21  5/08  In addition, a daily inspection control program is carried out during construction to ascertain the consistency in the potentially variable characteristics such as gradation and organic content. 
: c. Cement 
 
Cement is Type II, low-alkali cement in accordance with ASTM C-150-68, "Specification for Portland Cement", and is tested to comply with the
 
requirements of ASTM C-114-67, "Chemical Analysis of Hydraulic Cement". The
 
inspection and testing of cement, in addition to the initial tests performed by the
 
cement manufacturer, are indicated in table 3.8-6.
User tests are performed on the cement used in every 5,000 cubic yards of concrete produced. In addition, ASTM Tests C-191-65 and C-109-64 are performed periodically during construction to check the environmental effects of
 
storage on cement characteristics. 
: d. Fly Ash 
 
Fly ash conforms to ASTM C-618-68T Class F, "Fly Ash and Raw or Calcined Natural Pozzolans for Use in Portland Cement Concrete," and is tested to comply with the requirements of ASTM C-311-68, "Sampling and Testing Fly Ash for Use
 
as an Admixture in Portland Cement Concrete." 
 
The producer is required initially to test and then submit data on each lot of fly ash furnished. User tests, as indicated in table 3.8.7, are performed for each
 
2,500 cubic yards of concrete produced. In addition, periodic tests in accordance
 
with ASTM C-109-64 are performed during construction to check the
 
environmental affects of storage on fly ash. 
: e. Water 
 
Water used in mixing concrete is free from injurious amounts of acid, alkali, organic matters, and other deleterious substances as determined by AASHO-T-
: 26. In addition, user tests are performed quarterly on the mixing water after the
 
initial tests. 
 
The acceptance criteria for the mixing water are as follows: 
 
Criteria Percentage (Max.)
Alkalinity in terms of Calcium carbonate  0.025 Total organic solids  0.025 Total inorganic solids  0.050 Total chlorides  0.025
 
Total sulfates as SO 4  0.025 FNP-FSAR-3
 
3.8-25 REV 21  5/08  f. Admixtures 
 
The selected water reducing agent MBHC, manufactured by the Masters Builders Company, possesses a shrinkage reduction effect similar to the types prescribed by ASTM C-494-68, "Specifications for C hemical Admixtures for Concrete." 
 
An air entraining agent, Vinsol Resin, also manufactured by the Masters Builders Company, is added to the concrete mix to increase workability.
Admixtures containing chlorides are not used. 
: g. Concrete Mix Design 
 
Concrete mixes are designed in accordance with ACI 613-54, "Recommended Practice for Selecting  Proportions for Concrete," using materials qualified and
 
accepted for this work. Only concrete mixes meeting the design requirements
 
specified for the structures are used.
Trial mixes are tested in accordance with the applicable ASTM specifications as indicated below.
ASTM Test C-39-66 Compressive strength of molded concrete
 
cylinders.
C-143-69 Slump of Portland Cement concrete.
C-192-69 Making and curing concrete test specimens, in the laboratory.
C-231-68 Air content of freshly mixed concrete by the pressure method.
C-232-58 Bleeding of concrete.
 
Concrete test cylinders are cast from the mix proportions selected for construction, to determine the following properties: 
: 1. Compressive strength (ASTM C-39-66). 
: 2. Thermal diffusivity (ASTM C-342-67). 
: 3. Autogenous shrinkage (ASTM C-342-67). 
: 4. Thermal coefficient of expansion (ASTM-342-67). 
: 5. Modulus of elasticity and Poisson's ratio (ASTM-469-65). 
: 6. Uniaxial creep (ASTM C-512-69). 
 
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3.8-26 REV 21  5/08  h. Concrete Testing 
 
During construction, concrete is sampled and tested to ascertain conformance to the specifications. Concrete samples are taken from the mix in accordance with ASTM C-172-68, "Sampling Fresh Concrete," and six cylinders, three sets of two
 
cylinders each, are prepared from each sampling and cured in accordance with
 
ASTM C-31-69, "Making and Curing Concrete Compressive and Flexural
 
Strength Test Specimens in the Field."
The tests consist of the following: 
: 1. Determination of air content in accordance with ASTM C-231-68. 
: 2. Determination of unit weight in accordance with ASTM C-138-63. 
: 3. Slump test in accordance with ASTM C-143-69. 
: 4. Compressive strength test in accordance with C-39-66. 
: 5. Determination of temperature. 
 
The frequency and extent of these tests are as follows: 
 
One partial test for every 35 cubic yards mixed at the batch plant, consisting of a determination of slump and air content only.
One complete test for every 100 cubic yards mixed  at the batch plant. 
 
One complete test for every 210 cubic yards discharged from the truck. 
 
One partial test for every 210 cubic yards discharged into the forms, consisting of a determination of temperature and slump only. This partial
 
test is performed at the forms, and is required only when concrete is
 
discharged into the forms using pumps or conveyors.
 
The tests conducted at the truck discharge are performed on the concrete previously tested at the batch plant. The tests conducted at the forms are performed on concrete previously tested at the truck discharge. Hence, the
 
concrete tested at the forms is previously tested at the truck discharge and also
 
at the batch plant.
In addition, all concrete discharged from the truck is visually examined by an experienced inspector during the course of discharge from the truck and samples
 
are obtained and tested whenever the concrete appears to have excessive slump. 
 
The locations at which the sampled concrete is placed are marked. 
 
FNP-FSAR-3
 
3.8-27 REV 21  5/08  i. Concrete Placement 
 
All concrete for the containment base slab, cylindrical wall, dome and all walls exceeding 2 1/2 ft in thickness has a placing temperature of not less than 40°F nor more than 70°F. All other concrete including walls and elevated slabs has a placing
 
temperature of not less than 40°F nor more than 85°F.
If it is necessary to keep the temperat ure of the concrete from exceeding the above maximums, approved measures for reducing the temperature of the concrete are
 
employed, such as: 
: 1. Cooling the mixing water. 
: 2. Cooling the aggregates. 
: 3. Shading the materials and facilities from direct rays of the sun. 
: 4. Insulating water supply lines. 
: 5. Introducing flaked ice into the mix. 
 
In general, all procedures for hot weather concreting are in accordance with ACI-605-59. 
 
During cold weather, frozen materials or materials containing ice are not used. To protect the new poured concrete from freezing, enclosures or coverings are placed
 
over the concrete. Whenever the outdoor temperature is below 40°F, steam or heat
 
is introduced into the enclosures or coverings to maintain a temperature of not less
 
than 50°F for at least 72 hours, or for as much longer as is necessary to insure
 
proper rate of curing the concrete. The enclosures or coverings used in connection
 
with curing remain in place and intact for at least 24 hours after the artificial heating
 
is discontinued.
In general, all procedures for cold weather concreting are in accordance with ACI-306-66. 
: j. Bonding of Concrete Between Lifts 
 
Horizontal construction joints are prepared for receiving the next lift by either wet sandblasting, by cutting with an air water jet, or by bush hammering. 
 
When wet sandblasting is employed, it is continued until all laitance, coatings, stains, and other foreign materials are removed. The surface of the concrete is
 
washed thoroughly to remove all loose materials. 
 
When air water jet cutting is used, it is performed after initial set has taken place but before the concrete has taken its final set. The surface is cut with a high pressure
 
air water jet to remove all laitance and to expose clean, sound aggregates, but not
 
so as to undercut the edges of the larger particles of the aggregates. After cutting, FNP-FSAR-3
 
3.8-28 REV 21  5/08 the surface is washed and rinsed as long as there is any trace of cloudiness of the wash water. When it is necessary to remove accumulated laitance, coatings, stains, and other foreign materials, wet sandblasting is used before placing the next lift, to
 
supplement air water jet cutting.
The horizontal surface is wet immediately before the concrete is placed. 
 
Surface set retardant compounds are not used. 
 
3.8.1.6.2 Reinforcing Steel 
 
All reinforcing steel conforms to ASTM A-615-68, "Deformed Billet-Steel Bars for Concrete
 
Reinforcement," Grade 60. Spiral reinforcing steel conforms to ASTM A-82-66, "Cold Drawn Steel Wire for Concrete Reinforcement," 
 
Mill test reports are obtained from the reinforcing steel supplier for each heat of steel, to ensure
 
that the physical and chemical properties of the steel are in compliance with the ASTM
 
specifications. In addition, user tests consisting of tension and bend tests, in accordance with ASTM A-615-68, are performed to supplement the standard mill tests. One tension test and
 
one bend test are required for each 50 tons of each bar size from each heat of steel, with the
 
exception that bend tests are not performed on No. 14 and No. 18 bars. 
 
All user tests are performed on full size bars. 
 
Bars No. 11 and smaller are generally lap spliced in accordance with ACI 318-63. Bars No. 14
 
and No. 18 are Cadweld spliced exclusively. 
 
Splicing reinforcing bars by welding is not done. 
 
Procedures for splicing reinforcing bars using the Cadweld process are defined in appendix 3C. 
 
3.8.1.6.3 Structural and Miscellaneous Steel 
 
All structural and miscellaneous steel conforms to the following ASTM specifications: 
 
Rolled shapes, bars, and plates  A-36-69 High strength bolts    A-325 or A-490 Stainless steel      A-240, Type 304
 
Mill test reports are obtained for all materials used with the exceptions of hand rails, toe plates, kick plates, stairs, and ladders. 
 
Detailing, fabrication, and erection of the structural and miscellaneous steel are in accordance
 
with the Manual of Steel Construction, 1969 edition. 
 
Welding is done in accordance with AWS D 2.0, "Specification for Welded Highway and Railway
 
Bridges."
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3.8-29 REV 21  5/08 3.8.1.6.4 Containment Steel Liner Plate and Penetration Sleeves 
 
Since the liner plate is not a pressure vessel and its function is to serve as a leaktight
 
membrane, the design, construction, inspection, and testing of the liner plate are not covered by
 
any recognized codes or standards. However, for Unit 1, components of the liner plate that must resist the full containment design pressure, such as the penetrations, personnel locks and
 
equipment hatch, are designed, fabricated, constructed, and tested to meet the requirements of
 
Subsection B of Section III, Nuclear Vessels of the 1968 ASME Code. For Unit 2, components
 
that must resist the full containment design pressure are designed, fabricated, constructed, and
 
tested to meet requirements of Subsection NE of Section III of the 1971 ASME Code. 
 
Since the principal stresses of the liner due to thermal expansion are in compression, and no
 
significant tensile stresses are expected from the internal pressure loadings, special nil ductility transition temperature requirements are not applied to the liner plate materials. However, for Unit 1, all materials for the liner components which must resist tensile stresses resulting from
 
internally applied pressure, such as the penetration sleeves, are impact tested in accordance
 
with the requirements of Article 12 of Section III, Nuclear Vessels, of the 1968 ASME Code. For
 
Unit 2, impact tests are performed to paragraph NE 2300 of Section III of the 1971 ASME Code.
 
All welding procedures, welders, and welding operators used in the fabrication and erection of the steel liner plate and penetration sleeves are qualified by tests in accordance with Section IX
 
of the 1968 ASME Code, for Unit 1, and to the 1971 ASME Code for Unit 2. 
 
The quality control procedures for welding and for non-destructive inspection of welds are
 
defined in Appendix 3-G. 
 
3.8.1.6.5 Post-Tensioning System 
 
The prestressed, post-tensioning system selected for the containment is an Inland-Ryerson
 
BBRV buttonhead system. 
: 1. Tendons 
 
The tendons are composed of stabilized, low relaxation wires of 1/4-in. diameter with a minimum tensile strength of 240,000 psi in accordance with ASTM
 
A-421-65, Type BA. The pertinent features of the tendons are as follows: 
 
Number of wires  170 Ultimate tensile capacity (kips) 2000 End anchorage  Buttonheads
 
Sampling and testing of the tendon material conform to ASTM A-421-65. 
: 2. Anchorages 
 
The basic performance requirements for the end anchors of the tendons are stated qualitatively by the Seismic Committee of the Prestressed Concrete
 
Institute and published in their Journal of June, 1966, as follows:
FNP-FSAR-3
 
3.8-30 REV 21  5/08    "All anchors of unbonded tendons should develop at least 100 percent of the guaranteed ultimate strength of the tendons.
The anchorage gripping should function in such a way that no
 
harmful notching would occur on the tendons. Any such
 
anchorage system used in earthquake areas must be capable of
 
maintaining the prestressing force under sustained and fluctuating
 
loads and the effect of shock. Anchors should also possess
 
adequate reserve strength to withstand any overstress to which
 
they may be subjected during the most severe probable
 
earthquake. Particular care should be directed to accurate
 
positioning and alignment of end anchors." 
 
The end anchors used are capable of developing 100 percent of the minimum tensile strength of the tendons. Furthermore, the end anchors are capable of maintaining integrity for 500 cycles of loads corresponding to an average axial
 
stress variation between 0.7 and 0.75 f' s , at a repetition rate of one cycle in 0.1 second. This requirement sets the minimum acceptable limits on fatigue effects due to notching by the end anchor and tendon performance in response
 
to earthquake loads.
The number of cycles caused by the earthquake loads is predicted as only 30 of a total of 100, resulting from using all the strong ground motions which exceed
 
one half of the peak ground motion of the earthquake. However, it is
 
conservatively set at 500. 
 
The stress variations due to the earthquake motion alone are predicted as being 10 percent of the total of the estimated stress variations of 0.04 f'
: s. The estimated 0.04 f' s stress variations, in turn, result from the combinations of earthquake, wind, and incident loadings. Analyses made during the investigation
 
include consideration of tendon excitation, both parallel and perpendicular to the
 
tendon axis. 
 
The anchorage assemblies, including the bearing plates, are capable of transmitting the ultimate loads of the tendons into the structure without brittle
 
fracture at an anticipated lowest service temperature of 20°F. 
: 3. Sheathing 
 
Sheaths for the tendons are classified as concrete forms and are not subjected to any Standard codes. They provide a void in the concrete in which the tendons are installed, stressed and greased after the concrete is placed.
The sheaths are fabricated from 22 gauge, galvanized and corrugated ferrous metal tubing, which has an internal diameter of 4-3/4 in. clear of corrugations.
Couplers are provided at all field splices and sealed by tapes.
After sheathing installation, and prior to concrete placement, the sheathing is surveyed to assure accurate alignment. An inspection is also performed to
 
ascertain that all sheaths are continuous and unblocked by obstructions.
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3.8-31 REV 21  5/08  Before installation of the tendons, the sheathing is cleaned to remove all water and debris.
Vent tubing and temporary valves are provided to permit drainage at all low points. Splash caps at the ends of all sheaths, to prevent concrete and laitance from entering into the sheaths during construction, are provided. 
: 4. Corrosion Protection 
 
Suitable atmospheric corrosion protection is maintained for the tendons from the point of manufacture to the installed locations. The atmospheric corrosion protection provides assurance that the tendon integrity is not impaired due to
 
exposure to the environment. 
 
Prior to shipment, a thin film of petroleum oil based rust inhibitor, Visconorust 1601 and 1702 Amber, manufactured by the Vi scosity Oil Company, is applied to the tendons in accordance with the manufacturer's instructions. After the tendons are installed and stressed, the interior of the sheathing is pumped full of
 
a modified, thixotropic, refined petroleum oil based product to provide corrosion
 
protection. The tendon and anchors are also encapsulated by gasketed end
 
caps which are filled with the corrosion protection material and sealed against the
 
bearing plates.
Testing of the permanent corrosion protection material indicates that there are no significant amounts of chlorides, sulfides, or nitrates present. However, to further
 
verify the chemical composition of the filler material, test samples are obtained
 
from each shipment and analyzed as follows: 
: a. Water soluble chlorides (Cl) are determined in accordance with ASTM D512-67 with a limit of accuracy of 0.5 ppm. 
: b. Water soluble nitrates (NO3) are determined by the Water and Sewage Analysis Procedure of the Hach Chem ical Company, Ames, Iowa, or by ASTM D-992 Brucine Method. 
: c. Water soluble sulfides (S) are determined in accordance with American Public Health Association (APHA) standards with a limit of accuracy of 1 ppm. The APHA Standard methods (Methylene Blue procedure) or the
 
Hach Chemical Company method are used. 
 
Acceptance criteria of the corrosion protection materials are as follows: 
: 1. As shipped from the point of manufacture: 
 
a) Chlorides - 2 ppm max.
b) Nitrates - 4 ppm max.
c) Sulfides - 2 ppm max.
 
FNP-FSAR-3
 
3.8-32 REV 21  5/08  2. Onsite user test
 
a) Chlorides - 5 ppm max.
b) Nitrates - 5 ppm max.
c) Sulfides - 5 ppm max.
: 5. Prestressing Sequences 
 
The criteria of the prestressing sequence are based on the design requirements to limit the membrane tension in concrete to 1.0 
'cf to minimize unbalanced loads or differential stresses in the structure.
Prestressing begins after the concrete in the wall and the dome has reached the specified f' c (5500 psi). The construction opening will be closed prior to prestressing. All tendons are tensioned from both ends. Each hoop tendon wraps approximately 240 degrees around the containment. Three hoop tendons
 
forming one complete band are tensioned simultaneously and every other band
 
is stressed starting from the top of the base slab to approximately 30 ft below the
 
bottom of the ring girder. Vertical tendons are placed in two layers spaced
 
equally on either side of the containment wall center line. Three vertical tendons, approximately 120 degrees apart, are tensioned simultaneously; then every
 
fourth tendon is stressed, taking three at a time. Three groups of dome tendons
 
are oriented at 60 degrees to each other and are anchored at three different
 
levels in the ring girder. Three dome tendons, one in each group, are tensioned
 
simultaneously. The procedure for prestressing is carefully worked out with the
 
vendor so that all the tendons proceed in such a manner that the containment
 
structure will not be eccentrically loaded at any phase.
Table 3.8-8 summarizes the prestressing sequences.
 
3.8.1.6.6 Containment Interior Coating System 
 
Coating materials specified have been tested by their manufacturers under simulated operating
 
and incident conditions and shipments of the coating materials are accompanied by vendor certification of compliance. Coatings will be selected from inorganic zinc and epoxy compounds
 
that are resistant to radiation and have successfully withstood tests in an environment
 
simulating post design basis accident conditions, in cluding exposure to chemical sprays. Such tests are designed to ensure that the selected coatings will suffer no significant loss of adhesion or deterioration which could contribute particles capable of interfering with free flow of the
 
emergency core cooling system. Current references for the selection of coatings include B. J.
 
Newby, "Applicability of Conventional Protective Coatings to Reactor Containment Building," IN-1169, June 1968.
 
Table 6.2-37 summarizes the coating systems used and surface areas covered inside the
 
containment.
 
FNP-FSAR-3
 
3.8-33 REV 21  5/08  1. Containment Steel Liner Plate Coatings 
 
Surface preparation of the interior (exposed) surfaces of the steel liner plate is done in the shop by abrasive blast cleaning of each plate from edge to edge in accordance with the Steel Structures Painting Council (SSPC) Specification
 
SSPC-SP6-63, "Commercial Blast Cleaning" for inorganic zinc primer or with
 
SSPC-SP10 "Near White Metal Blast Cleaning" for epoxy primer. The plates are
 
then primed with one coat of Ameron Dimetcote No. 6 inorganic zinc coating or
 
Ameron Amercoat 90 to an average dry film thickness of 3.0 mils or 5.0 mils, respectively.
 
Areas damaged by welding, such as strikes, or those damaged by the removal of temporary attachments for erection, are repaired and recoated so that they are equivalent to the original conditions.
: 2. Containment Interior Coatings 
: a. Steel Surfaces 
 
Carbon steel surfaces, including structural and miscellaneous steel, uninsulated piping, and equipment, which are located in areas subject to hard usage or radioactive contamination are blast cleaned in accordance
 
with Specification SSPC-SP6-63, "Commercial Blast Cleaning" for
 
inorganic zinc primer or with SSPC-SP10 "Near White Metal Blast
 
Cleaning" for epoxy primer. Within 8 hours after blast cleaning, the
 
surfaces are primed with one coat of Ameron Dimetcote No. 6 (inorganic
 
zinc primer), or Ameron Amercoat 90 to an average dry film thickness of
 
3.0 mils or 5.0 mils, respectively.
: b. Concrete and Masonry Surfaces
 
All concrete and masonry surfaces, including floors, wainscot, walls, columns, pilasters, and ceilings, are chemically cleaned by either caustic wash or acid etching, or by blast cleaning. A surfacer or epoxy coating is
 
then applied over the entire area.
: 3. Curing
 
Curing of the newly painted surface areas is as recommended by the coating manufacturer. Special precautions are taken during the curing period so that rain, snow, fog, moisture, condensation, or any foreign contaminant does not
 
come into contact with or adhere to the freshly painted surface areas.
: 4. Recoating (One Additional Finish)
 
The recoating of the existing coating systems (except Amercoat 90) in items 1 and 2 above will be performed with Amercoat 90 or 66 finish as per FNP Coating
 
Manual. The maximum dry film thickness of existing finish (Amercoat 66) and FNP-FSAR-3
 
3.8-34 REV 21  5/08 recoat finish (Amercoat 90 or 66) will be in accordance with FSAR tables 6.2-2 and 6B-1.
 
3.8.1.7  Testing and Inservice Surveillance Requirements
 
Testing and inservice surveillance requirements for the containment include the following:
: a. Pre-operational structural acceptance test.
: b. Pre-operational integrated leak rate test.
: c. Post-tensioning system tendon surveillance.
: d. Post-tensioning system tendon end anchorage concrete surveillance.
: e. Containment surfaces and steel liner plate surveillance.
: f. Inservice Inspection of pressure retaining concrete surfaces and metallic liner and attachments.
3.8.1.7.1 Pre-Operational Structural Acceptance Test
 
In accordance with NRC acceptance criteria (NRC SER NUREG 75/034 dated May 2,1975),
prior to initial fuel loading, the containment is subjected to a pressure proof test equivalent to
 
115 percent of the containment design pressure as required by NRC Regulatory Guide 1.18. 
 
This test demonstrates that the containment is capable of resisting the postulated accident
 
pressure. In addition, by measuring the structural response and comparing the results with
 
analytical predictions, the test also serves to verify the anticipated structural behavior.
 
Measuring systems, pressurization procedure, deflection and temperature measurements, crack
 
pattern mapping, and data acquisition schedules for the pre-operational structural acceptance
 
test are described in appendix 3H, "Containment Structural Acceptance Test".
 
3.8.1.7.2 Pre-Operational Integrated Leakage Rate Testing
 
Pre-operational integrated leakage rate testing of the containment is conducted in accordance
 
with the procedures described in the Bechtel Power Corporation Topical Report BN-TOP-1, "Testing Criteria for Integrated Leakage Rate Testing of Primary Containment Structures for
 
Nuclear Power Plants", Revision 1, November 1, 1972.
 
3.8.1.7.3 Post-Tensioning System Tendon Surveillance
 
The objective of the containment tendon surveillance program during the lifetime of the plant is
 
to provide a systematic means of assessing the continued quality of the post tensioning system.
 
The program is intended to furnish sufficient inse rvice historical evidence to provide a measure FNP-FSAR-3
 
3.8-35 REV 21  5/08 of confidence in the condition and the functional capability of the system, as well as an opportunity for timely corrective measures s hould adverse conditions, such as excessive corrosion, be detected.
 
[HISTORICAL] [Conformance to NRC Regulatory Guide 1.35, "I nservice Surveillance of the Ungrouted Tendons in Prestressed Concrete Containment Struct ures," is discussed in appendix 3A.
In general, the containment tendon surveillance program for the Farley Nuclear Plant is modeled after the programs now in progress and is similar to tho se established by the Point Beach Nuclear Power Plant Units 1 and 2 (Docket Nos. 50-266 and 50-310, respectively), Turkey Point Unit 3 (Docket No.
50-250), and the Palisades Plant Unit 1 (Docket No. 50-255), with the exception of the number of surveillance tendons involved.
To implement the surveillance program for each un it, 21 surveillance tendons are provided. Nine of these 21 tendons are redundant and are not required for the design prestress level. All tendons in the containment structure are uniformly spaced. Those to be used for surveillan ce will be selected after installation, tensioning, and greasing, from groups of tendons similar with respect to environmental exposure, structural function, and prestress losses, as follows:
: a. Six dome tendons, randomly but representatively distributed.
: b. Five vertical tendons, randomly but representatively distributed.
: c. Ten hoop tendons, randomly but representatively distributed.
The program also provides for additional testing and inspection should nonconformance or unexpected conditions be detected.
The frequency of the tendon surveillance, after the structural integrity test, will be as follows:
: a. One year after test.
: b. Three years after test.
: c. Five years after test.
: d. 5 years thereafter for the life of the plant.
The containment tendon surveillance program will include the following:
: a. Liftoff Liftoff will be performed by properly calibrated jacks at the stressing ends of all 21 surveillance tendons, with simultaneous meas urements of tendon elongations and jacking forces. Allowable elongations and jacking forces, temperature effects, and allowable tolerances will be established prior to the tests. The liftoff will include an unloading cycle going down to essentially complete detensi oning of the tendon to identify broken or damaged wires.
FNP-FSAR-3
 
3.8-36 REV 21  5/08  b. Tendon Inspection One dome tendon, one vertical tendon, and one hoop tendon will be completely detensioned. One wire from each of these te ndons will be removed as a sample, visually inspected for corrosion and pitting, and tested for physical properties. In addition, both ends of all surveillance tendons will be in spected for any deformed buttonheads and broken wires. At each successive inspection the samples will be selected from different tendons. 
: c. Wire Tests Tensile and elongation tests will be performed on at least three wire samples taken from the removed wire, one at each end and one at the middle of the wire. The test results for each wire will include, as a minimum, the yield and ultimate strength and percent of elongation under load at failure. 
: d. Anchorage Assembly Inspection The tendon anchorage assemblies will be vi sually inspected for any deleterious conditions such as corrosion or cracks. Th e shim thickness will be measured. The surrounding concrete will be visually inspected for indications of abnormal material behavior. 
: e. Sheath Filler Inspection A sample of sheath filler will be taken from each surveillance tendon for visual and laboratory examinations.
: f. Re-Tensioning After wire removal, the tendons will be re-tensioned to the stress level measured at the liftoff reading, and then checked by a final liftoff reading. Any change in total shim thickness will be recorded.
The acceptance criteria for containment tendon surveillance are as follows:
: a. The prestress force for each tendon shall be not less than the allowable lower bound nor greater than the allowable upper bound forces at the time of the test. b. An acceptance limit of not more than one defective tendon out of the total sample population shall be set. If one sample te ndon is defective, an adjacent tendon on each side of the defective tendon shall also be tested. If both these tendons are acceptable, as defined in "a", the surveillance program shall then proceed considering the single deficiency as unique and acceptable. However, if either adjacent tendon is defective, or if more than one tendon out of the original sample population is defective, abnormal degradation of the containment structure is indicated. The commission will be notified in accordance with  Regulatory Guide 1.16, "Reporting of Operating Information," except that the initial report may be made within 30 days of the completion of the tests,  and the detailed report may follow within 90 da ys of the completion of the tests.
 
FNP-FSAR-3
 
3.8-37 REV 21  5/08  c. Failure below the guaranteed ultimate strength of any one of the three tendon material sample tests will be considered an indication of abnormal degradation of the containment, and the commission will be notified as indicated above.
Unless there is evidence of abnormal degradation of th e containment, after the third surveillance the tendon sample population will be reduced from a tota l of 21 tendons to nine tendons. These will include three dome tendons, three vertical tendons, and three horizontal tendons, selected randomly but representatively.
Since the FNP Unit 2 containment is identical to that of Unit 1, and since these two units are located on the same site and are built by the same constructor in the same manner at the same time, the surveillance program for Unit 2 consists of visual examination of the same numbers and types of tendons as those of Unit 1, to the extent practical, without dismantli ng the load-bearing components of the anchorage.
3.8.1.7.4  Tendon End Anchorage Concrete Surveillance Surveillance will be performed on the concrete surfaces surrounding some tendon anchorages. The locations of these concrete surfaces will be designated prior to the pre-operational structural acceptance test, on the following basis: 
: a. Hoop Tendon Anchorage Four locations on one buttress and one location between the faces of the buttress. 
: b. Vertical Tendon Anchorages Two locations on the top of the ring girder. 
: c. Dome Tendon Anchorages Two locations on the side of the ring girder.
The frequency of the tendon anchorage con crete surveillance will be as follows: 
: a. Immediately prior to and after the pre-operational structural integrity test. 
: b. During the first type A containment leakage rate test only.
The surveillance program will include the following: 
: a. Visual inspection of the end anchorage concrete exterior surfaces. 
: b. The mapping of the predominantly visible concrete crack patterns. 
: c. The measurement of crack widths, by th e use of optical comparators or wire feeler gauges, and the notation of the length, orientation and location of the cracks. 
 
FNP-FSAR-3
 
3.8-38 REV 21  5/08 The acceptance criteria for the tendon anchorage concrete surveillance are as follows:
The measurements and observations of the cracks and crack patterns of the concrete surfaces will be compared with tho se taken from other post-tensioned concrete containment structures, and with the previous measurements and observations at the same location on the containment. The cracks and crack patterns will be acceptable if the comparison is in agreement with the predictions.]
3.8.1.7.5 Containment Surfaces and Steel Liner Plate Surveillance 3.8.1.7.5.1 Containment Surfaces Visual Inspection
 
A periodic visual inspection of exposed accessible interior and exterior surfaces of containment is conducted in accordance with the FNP Containment Leakage Rate Testing Program and containment inspection plan.
 
3.8.1.7.5.2 Liner Plate Surveillance
  [HISTORICAL][The steel liner plate will be examined and measured at four easily accessible locations to monitor its performance. The areas used for surveillance will be chosen as follows: 
: a. Two areas where the liner plate has a measurable initial inward curvature.
: b. Two areas where the local liner plate has a measurable initial outward curvature.
The frequency of the liner plate surveillance will be as follows: 
: a. Prior to and shortly after the pre-operational structural integrity test. 
: b. During the shutdown for the first type A containment leakage test only.
The surveillance program will include the following: 
: a. Measurement of the inward/outward curvature relative to a fi xed chord at each location. 
: b. Measurement of temperature at the liner pl ate and exterior concrete at these locations. 
: c. In addition to the four areas profiled, f our more areas will be surveyed for any indication of strain concentrations.
The acceptance criteria for the steel liner plate surveillance are as follows:
Measurements and observations will be compared with those made previously at the same locations. The performance of the liner plat e will be considered satisfactory if there are no significant differences in the measurements and observations.]
FNP-FSAR-3
 
3.8-39 REV 21  5/08 3.8.1.7.6 Inservice Inspection of Pressure Retaining Concrete Surfaces and Metallic Liner and Attachments.
[HISTORICAL][On September 9, 1996, the NRC amended its regulations to incorporate by reference the 1992 edition with 1992 addenda of Subsecti ons IWE and IWL of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code with modifications, in 10 CFR 50.55a. The new rules require certain containment liner and concrete inspections/examinations to be performed prior to September 9, 2001, and to be repeated on a regular basis thereafter.
Containment repair and replacement requirements of the new rules including preservice examinations after repair or replacement were effective on September 9, 1996. Relief from this effective date until March 15, 1997, was requested by FNP in order to revise the plant repair/replacement program. The periodic inservi ce inspections will be incorporated into the individual Inservice Inspection Programs and Plans prior to the required implementation date of September 9, 2001.]
Some information in paragraph 3.8.1.7.3 through 3.8.1.7.5 above was applicable until 10 CFR 50.55a was amended in 1996 (ref. 61 FR 41303, August 8, 1996). 10 CFR 50.55a was amended in 1996 to incorporate Subsections IWE and IWL of the 1992 edition and 1992 Addenda of the ASME Section XI Code. Th is amendment also required an expedited implementation schedule.
The tendon surveillance program will be implemented per ASME Section XI, Subsection IWL.
The main features of this program are mentioned below, as described in the Containment Inspection Plan.
: a. Subsection IWL specifies the required number of tendons to be examined during an inspection. Both FNP units require a minimum and maximum number of each type of tendon as long as other applicable criteria are met.
: b. To develop a history and to correlate the observed data, one tendon from each group should be kept unchanged after the initial selection, and these unchanged tendons should be identified as control tendons.
: c. Tendon forces are acceptable if the average of all measured tendon forces for each type of tendon is equal to or greater than the minimum required prestress specified at the anchorage for that type of tendon.
 
The ASME Section XI Inservice Inspection Progr am's IWE and IWL inspections are credited as license renewal aging management activities (see chapter 18, subsection 18.2.1).
 
3.8.2 STEEL CONTAINMENT SYSTEM (ASME CLASS MC COMPONENTS) 
 
As described in subsection 3.8.1, the containment is a prestressed, reinforced concrete
 
structure; therefore this section does not apply to the design of the Farley Nuclear Plant. 
 
3.8.3 INTERNAL STRUCTURES 
 
The containment internal structures are all Category I, consisting of:
FNP-FSAR-3
 
3.8-40 REV 21  5/08  A. The reactor cavity and primary shield wall. 
 
B. The secondary shield wall, which encloses the steam generator and pressurizer compartments.
C. The refueling canal. 
 
D. The floor slabs. 
 
3.8.3.1  Description of the Internal Structures
 
A. Reactor Cavity and Primary Shield Wall The reactor cavity is a heavily reinforced concrete structure which houses the reactor and provides the primary shielding barrier. 
 
The reactor is supported on six special pads and shoes;  these vessel supports are mounted on structural steel supports embedded in the reactor cavity
 
concrete. The wall of the cavity structure provides missile protection for the
 
containment structure and liner plate in the event of a hypothetical loss-of-coolant
 
accident. The reactor cavity is basically designed to contain the internal
 
pressures resulting from the loss-of-coolant accident. During normal operating
 
and maintenance inspection, the cavity wall provides biological shielding. Finally, the cavity wall structure acts as the support structure for the reactor and
 
transmits loads to the foundation mat. Refer to figure 3.8-11 for typical
 
dimensions, location, and orientation of the cavity.
B. Secondary Shield Walls The secondary shield walls are thick reinforced concrete walls anchored into the base slab to ensure stability and prevent uplift. See figure 3.8-7 for base anchorage detail. Figures 3.8-9 and 3.8-10 show the dimensions and extent of
 
the secondary shield wall. 
 
The steam generator compartment is a reinforced concrete structure housing the steam generator, reactor coolant pumps, and the reactor coolant loops. There
 
are three steam generator compartments with one reactor coolant loop in each. 
 
Refer to figures 3.8-13 and 3.8-14 for the layout and typical dimensions of the
 
compartments. The compartments are fo rmed by the secondary shielding walls on the exterior and by the reactor and the refueling canal walls on the interior. 
 
The steam generator compartment, which extends 11 ft 6 in. above the operating
 
floor slab, and the foundation base mat form the bottom. The pressurizer
 
compartment is built integrally with the secondary shielding wall. 
 
C. Refueling Canal
 
The refueling canal is used during refueling operations to transfer the new and spent fuel elements between the reactor and the fuel handling building. It is also FNP-FSAR-3
 
3.8-41 REV 21  5/08 a laydown area for the reactor upper and lower internals. The canal is lined with stainless steel plate and is filled with borated water to a depth of 34 ft during
 
refueling. The canal is open at the top. A missile protection slab above the
 
reactor head serves also as biological shielding in the vertical direction. Refer to
 
figures 3.8-13 and 3.8-14 for the structural arrangement showing the refueling
 
canal shield. 
 
D. Floor Slabs
 
The operating floor surrounds the refueling cavity wall, the secondary shield walls and the containment wall. The floor slab is supported by the refueling cavity walls and the secondary shielding walls.
The floor gratings and the floor slab supporting the containment coolers in turn are supported by structural steel beams spanning between the operating floor
 
slab and the containment wall. 
 
3.8.3.2  Applicable Codes, Standards and Specifications
 
The applicable codes, standards, specifications , regulatory guides, and other documents used in the structural design of the internal structures are covered in subsection 3.8.1.2. 
 
3.8.3.3  Loads and Loading Combinations
 
The internal structures are designed for all credible conditions of loadings, including normal
 
loads, loads resulting from a loss-of-coolant accident, test loads, and missile generated loads. 
 
Critical loading combinations are those caused by a postulated earthquake, or by a pipe rupture
 
within the containment. 
: 1. Loads 
 
The following loads are considered:  dead loads, live loads, earthquake loads, pipe rupture loads, operating  thermal loads, and test loads. 
: a. Dead Loads  (D) 
 
Structural dead loads consist of the weight of walls, framing and slabs, partitions, platforms, all permanent equipment, and major piping and
 
electrical conductors. 
: b. Live Loads  (L) 
 
Live loads consist of design floor loads, laydown loads, equipment live loads, fuel handling equipment and material loads. 
 
FNP-FSAR-3
 
3.8-42 REV 21  5/08    The operating floor slab is designed for either of these live loads: 
 
Floor gratings  450 psf    Concrete slabs 1000 psf Equipment live loads are those specified on drawings supplied by the manufacturer of the equipment. 
: c. Earthquake Loads (E, E')
 
Earthquake loads are predicated on a basis of (E) the 1/2 safe shutdown earthquake (1/2 SSE), having a horizontal ground surface acceleration of 0.05 g and a vertical ground surface acceleration of 0.033 g.
In addition, a safe shutdown earthquake (SSE), (E') having a horizontal ground surface acceleration of 0.10 g and a vertical ground surface
 
acceleration of 0.067 g, is used to check the design to ensure that loss of
 
structural function will not occur. 
 
Seismic response spectrum curves are given in section 3.7 for both horizontal and vertical ground motions. A dynamic analysis is used to
 
compute the seismic loads for the design of structural elements. 
: d. Pipe Rupture Loads  (R) 
 
Pipe rupture loads are the local jet forces or pressures resulting from a postulated rupture of any one pipe. 
: e. Thermal loads  (T o , T a)
The thermal effects during normal operating condition (T o), as well as the thermal effect due to the accident temperature gradient (T a), are included. 
: 2. Loading Combinations 
 
In general, two types of loading cases are considered in the design of the internal structures. 
: a. The design loading case for which the working stress method is used. 
: b. The factored loading case for which the ultimate stress method is used. 
 
The following terms are used in the loading combination equations: 
 
C = Required capacity of the containment to resist factored loads. 
 
  = Capacity reduction factor. 
 
FNP-FSAR-3
 
3.8-43 REV 21  5/08  D = Dead loads of containment, interior structures, and equipment, plus any other permanent contributing loads.
E = 1/2 safe-shutdown earthquake load. 
 
E' = safe shutdown earthquake load. 
 
H = Force on structure due to thermal expansion of pipes under operating conditions.
L = Appropriate live load. 
 
P = Design accident pressure load. 
 
R = Force or pressure on structure due to rupture of any one pipe. 
 
T a = Thermal loads due to the accident temperature gradient through the wall, based on a temperature corresponding to the unfactored design accident
 
pressure. 
 
T o = Thermal loads due to the normal operating temperature gradient through the walls. 
 
The load combinations, equations and load factors are as follows: 
: a. The design loading case for which the working stress method is used: 
 
D + L    (construction case)
D + L + H + T o + E  (operating case)  D + L + P + T a    (design accident case)
: b. The factored loading case for which the ultimate stress method is used: 
 
C = 1/ (1.0 D + 1.0 R + 1.25 E)  C = 1/ (1.0 D + 1.25 H + 1.25 E)  C = 1/ (1.0 D + 1.0 R + 1.0 E')  C = 1/ (1.0 D + 1.25 H + 1.0 E' + 1.0 R)  C = 1/ (1.0 D + 1.25 H)
The capacities of all load carrying structural elements are reduced by capacity reduction factors
() as given below. These factors provide for the possibility that small adverse variations in material strengths, workmanship, dimensi ons, control, and degree of supervision, which individually are within the required tolerances and the limits of good practice, occasionally may combine to result in undercapacity. 
 
Capacity reduction factors: 
 
  = 0.90 for concrete in flexure.
FNP-FSAR-3
 
3.8-44 REV 21  5/08
  =  0.85 for tension, shear, bond and anchorage in concrete.
  =  0.75 for spirally reinforced concrete compression members.
 
  =  0.70 for tied compression members.
 
  =  0.90 for fabricated structural steel.
The deflections and deformations of the internal structures and supports are checked to ensure
 
that the functions of the equipment and engineered safeguards are not impaired. 
 
3.8.3.4  Design and Analysis Procedures
 
The basic techniques of analyzing the internal structures can be classified into two groups:
 
conventional methods involving simplifying assumptions such as those found in beam theory, and those based on plate and shell theories of different degrees of approximation. The strength
 
methods given in the ACI-318-63 code are used. The internal structures are provided with
 
connections capable of transmitting axial and lateral loads to the containment base slab. 
 
The containment interior structure is designed to provide structural supporting elements for the
 
entire nuclear steam supply system (NSSS) as well as required shielding. Basic supporting
 
components are designed using both reinforced concrete and structural steel as appropriate. 
 
All design aspects are integrated with the design criteria of the nuclear steam system supplier
 
and include particular attention to the combined thermal and dynamic effects particularly evident
 
during earthquake conditions. Thrusts are taken by rigid members and by shock suppressors. 
 
Design loads for the interior structure are listed and described in subsection 3.8.3.3. 
 
Design of the interior structures evolves around four basic systems: the reactor coolant system;
 
the main steam system; the engineered safeguards system; and the fuel handling system.
The structures which house or support the basic systems are designed to sustain the factored loads described in subsection 3.8.3.3. 
 
The design bases to be applied are given as follows: 
: a. All operating loads, thermal loads, seismic loads, and thermal deformations at the levels indicated in subsection 3.8.3.3. 
: b. Loads and deformations resulting from a LOCA and its associated effects. 
: c. Pressure buildup in locally confined areas such as the primary shielding cavity or the secondary shielding room. 
: d. Jet forces resulting from the impingement of the escaping fluid upon adjacent structures. 
: e. Pipe whipping following a break in the reactor coolant system pipe. 
 
FNP-FSAR-3
 
3.8-45 REV 21  5/08  f. Rapid rise in temperature and accompanying rise in pressure. 
: g. Missiles as described in section 3.5. 
 
The magnitude of thrust forces and pressure buildup resulting from a pipe break is determined
 
from appropriate blowdown values. 
 
The interior areas where local pressure buildup is significant are the reactor cavity and the three
 
secondary shield areas. Ultimate strength design is used for these interior areas in accordance
 
with ACI 318-63. A strain limit of 0.003 in./in.
is used for concrete and steel. Sketches showing structural details of the interior structures are shown in figures 3 8-9 to 3 8-14. 
 
Seismic analyses for the interior structures conform to the appropriate procedures outlined in
 
section 3.7. 
 
The mathematical model includes equipment of significant mass values as part of the lumped masses at the appropriate elevation. The seismic loads are determined using the procedures of
 
the design response spectrum technique of anal ysis. Bending moments and shears resulting from appropriate earthquake loads are combined according to the load combinations described in subsection 3.8.3.3. The equipment seismic shear is resisted by the anchorage system, anchor bolts, and by additional shear studs. See figure 3.8-6 for details. 
 
Various structural components of the interior structures are analyzed and designed individually
 
for governing loading conditions. 
 
For the operating condition analysis, the concrete is assumed to be uncracked and the stresses
 
are limited to those specified in ACI 318-63. For the accident condition analysis, ultimate
 
strength approach is employed. In this analysis the results of the unfactored loading cases are
 
multiplied by appropriate load factors as described in subsection 3 8 3 3. The resulting concrete stresses are limited to 0.85 f
'c and the reinforcing steel stresses are limited to 90 percent of the guaranteed minimum yield given in the appropriate ASTM specification. 
 
The main considerations in establishing the structural design criteria for the internal structures
 
are to provide a structure that will withstand the differential pressure within the cavity in the event of an accident, and to minimize the effects of the pipe rupture force utilizing supports and
 
restraints. Loads and deformations resulting from a LOCA and its associated effects on any
 
one of the basic systems are restricted so that pr opagation of the failure to any other system is prevented. In addition, a failure in one loop of the nuclear steam supply system is restricted so
 
that propagation of the failure to the other loop is prevented. Localized concrete yielding is
 
permitted when it is demonstrated that the yield capacity of the component is not affected, and
 
that this small localized yielding does not generate missiles that could damage the structure. 
 
Full recognition is given to the time increments associated with these postulated failure
 
conditions, and yield capacities are appropriately increased when a transient analysis
 
demonstrates that the rapid strain rate justifies this approach. The walls are also designed to
 
provide adequate protection for potential missile generation that could damage the containment
 
liner. 
 
FNP-FSAR-3
 
3.8-46 REV 21  5/08 The effect of radiation generated heat on the internal structures has been considered in the design of the primary and secondary shield walls. The shield wall thicknesses were determined
 
on the basis of the radiation shielding requirements, much higher than those required for
 
structural purposes. This additional thickness provides a reserve strength greater than required
 
to offset minor damages to the structures due to a LOCA. Since high temperatures are
 
damaging to concrete, provisions are made to maintain a constant temperature through
 
ventilation. The ventilation within the containment has been designed to cool the area
 
surrounding the shield walls in order to prevent any appreciable loss of structural strength due
 
to gamma and neutron heating. 
 
The final designs of the interior structure and equipment supports are reviewed to assure that
 
they can withstand applicable pressure loads, jet forces, pipe reactions, and earthquake loads without loss of function. The deflections or deformations of the structures and supports are
 
checked to ensure that the functions of the containment and safety feature systems are not
 
impaired. 
 
3.8.3.4.1 Reactor Cavity and Primary Shield 
 
For the normal operating condition, the reactor cavity is designed to withstand the stresses due
 
to dead loads and thermal loads. Under this condition, the stresses in the concrete and the
 
reinforcing steel are kept below those permissibl e for the working stress design as stipulated in ACI 318-63 Code. In the stress analysis, flexure tensile cracking is permitted but is controlled by the bonded reinforcing steel. 
 
For the hypothetical LOCA condition, the cavity wall is designed to withstand jet forces and internal pressurization without gross damage to the cavity structure. Local damage to the cavity
 
in the immediate vicinity of the NSSS component failu re is inevitable. However, vital parts of the containment are protected from this failure to ensure a post accident leaktight containment
 
structure. 
 
The reactor cavity is designed to withstand a static equivalent internal pressure of 225 psi due
 
to the LOCA. This pressure is assumed to be acting on the entire cavity for a duration of one
 
second. For this loading case the concrete is assumed cracked across the entire section. The
 
reinforcing steel resists all the stresses. The maximum stress level in the rebar under this
 
loading condition is limited to the ultimate c apacity of the rebar as modified by appropriate capacity reduction factor. 
 
3.8.3.4.2 Refueling Canal 
 
For the refueling condition, the walls are designed for the hydrostatic head due to 35 ft of water
 
and are checked for the effect of hydrodynamic loads due to 1/2 SSE and SSE motions. The
 
pressure loads on the side of the steam generat or compartment and hydrostatic loads do not occur simultaneously. 
 
FNP-FSAR-3
 
3.8-47 REV 21  5/08 3.8.3.4.3 Steam Generator Compartments 
 
The compartments are designed for an internal pressure of 30 psi due to a loss-of-coolant
 
accident resulting from a hypothetical double ended rupture of a reactor coolant pipe. 
 
3.8.3.5  Structural Acceptance Criteria
 
The limiting values of stress, strain, and gross deformations are established by the following
 
criteria: 
: a. To maintain the structural integrity when subjected to the worst load combinations. 
: b. To prevent structural deformations from displacing the equipment to the extent that the equipment suffers a loss of safety-related function.
The allowable stresses are those specified in the applicable codes. The stress contributions
 
due to earthquakes are included in the load combinations described in subsection 3.8.3.3. 
 
Structural deformations were found not to be a controlling criterion in the design of the internal
 
structures. 
 
Table 3.8-9 lists the load combinations, the calculated and allowable stresses, as well as the
 
method of analysis used in the design of the main structural components of the internal
 
structures. The ratio of the allowable to the calculated stresses yields the safety margins. 
 
3.8.3.6  Materials, Quality Control and Special Construction Techniques
 
The following basic materials are used in the construction of the internal structures: 
: a. Concrete f'c (psi)=  5,500    b. Reinforcing steel 
 
Deformed bars ASTM A-615 f y  (psi) = 60,000  Grade 60 Spiral bars ASTM A-82 f y  (psi) = 70,000    c. Structural and miscellaneous steel 
 
Rolled shapes, bars and plates ASTM A-36 f y  (psi) = 36,000  Crane rails ASCE High strength bolts ASTM A-325 or A-490 Stainless steel  ASTM A-240 Type 304 FNP-FSAR-3
 
3.8-48 REV 21  5/08
: d. Reactor cavity steel liner plate and
 
penetration sleeves 1/4-in. liner plates ASTM A-285 f y  (psi) = 24,000 Grade A  Insert plates ASTM A-516 f y  (psi) = 38,000 Grade 70  Penetration sleeves pipes ASME SA-333 f y  (psi) = 32,000 Grade 6    plates ASME SA-516 f y  (psi) = 38,000 Grade 70 
: e. Interior coating system Steel liner plate (reactor cavity)
Primer    Finish coat (wainscot)
Carbon steel surface Primer    Finish coat (wainscot)
Finish coat (above wainscot)
Concrete and masonry surfaces Surfacer Primer    The materials and the quality control procedures have been described in subsection 3.8.1.6. 
 
Part of the "C" Secondary Shield Wall for Unit 1 and the "A" Secondary Shield Wall for Unit 2
 
was removed and restored as a part of the steam generator replacement activities. The
 
Cadwelding process used for wall restoration differed in some respects from the one used
 
during initial construction and described in Appendix 3C. The differences are:
: 1. The original Cadwelding process used for construction of the shield walls, as described in Appendix 3C, specified preheating the reinforcing steel to 300
°F to ensure that Cadweld splices were free of moisture. The Cadwelding process used for the
 
restoration of the shield walls did not specify a specific preheat temperature. Vendor
 
practices at the time of shield wall restoration provided detailed instructions for
 
preheating, but did not specify a temperature.
: 2. The original Cadwelding process used both sister and production splice testing as specified in Appendix 3C. The Cadwelding process used for the restoration used sister
 
splice methodology only. The sister splice testing is in conformance with the statement
 
of conformance for Regulatory Guide 1.10 in Appendix 3A.
 
Additionally, the concrete mix used for restoration of the wall meets the later editions of the
 
applicable codes and standards described in Section 3.8.1.6.1. The restored wall meets the FNP-FSAR-3
 
3.8-49 REV 21  5/08 design strength requirements and is adequate to satisfy the other functions of the wall described in the FSAR.
 
The internal structures are built of reinforced concrete and structural steel, using proven
 
methods common to heavy industrial construction.
No special construction techniques have been employed in the construction of the internal structures. 
 
The effect of various amounts of radiation on the internal structures has been considered in the
 
calculations for the primary and secondary shield walls. 
 
The shield wall thicknesses were determined on the basis of the radiation shielding
 
requirements, much higher than those required for structural purposes. This additional
 
thickness provides a reserve strength greater than required to offset minor damages to the
 
structures due to a LOCA. Since high temperat ures are damaging to concrete, provisions are made to maintain a constant temperature through ventilation. The ventilation within the containment has been designed to cool the area surrounding the shield walls in order to prevent
 
any appreciable loss of structural strength due to gamma and neutron heating. 
 
3.8.3.7  Testing and Inservice Surveillance Requirements
 
A formal program of testing and inservice surveillance is not considered for the internal
 
structures. The internal structures are not directly related to the functioning of the containment
 
concept. Therefore, no testing or surveillance is required.
 
For the period of extended operation, periodic inspections of the Category I containment internal
 
structures by the Structural Monitoring Program are required license renewal aging
 
management program activities (see chapter 18, subsection 18.2.10).
 
3.8.4 OTHER CATEGORY I STRUCTURES 
 
Category I structures other than the containment and the internal structures are listed below: 
: a. Auxiliary building. 
: b. Diesel generation building. 
: c. River intake structure.(a)  d. Intake structure at storage pond. 
: e. Storage pond, dam, and dike. 
: f. Pond spillway structure. 
: g. Electrical cable tunnel. 
: h. Outdoor Category I tanks.
FNP-FSAR-3
 
3.8-50 REV 21  5/08  i. Plant vent stack.(a)  3.8.4.1  Description of the Structures
 
A. Auxiliary Building The auxiliary building consists of four floors below grade (el 155 ft) and two floors above. The containment and the auxiliary building are separated by a 3-in.-wide expansion joint. 
 
According to analysis, this gap is adequate to prevent the two structures from coming in contact with one another during an earthquake or design basis accident. Flood protection design of the
 
auxiliary building is described in section 3.4. 
 
The auxiliary building houses the following major plant facilities related to safety: 
: 1. New fuel and spent fuel handling, storage, and shipment facilities. 
: 2. Control room and related facilities (shared by both units). 
: 3. Radwaste disposal system facilities. 
: 4. Chemical and volume control facilities. 
: 5. Engineered safety features system (ESF). 
: 6. Penetration room. 
: 7. Access control area (area is located in unit 1 auxiliary building, but is shared by both units). 
 
The auxiliary building is constructed of reinforced concrete below and above grade. All
 
columns, slabs, and structural walls are of reinforced concrete. The roof is a reinforced
 
concrete slab, with a minimum thickness of 2 ft, designed to prevent penetration of missiles.
Drawings D-176002 through D-176007 for Unit 1 and D-206002 through D-206007 for Unit 2, and figures 3.8-23, 3.8-24, and 3.8-25 show various plans and sections of the auxiliary building.
 
The principal features of the new and spent fuel handling, storage, and shipment facilities are
 
shown in figures 3.8-23 and 3.8-25. 
 
The fuel handling facilities are served by a 125-ton overhead crane capable of handling heavy
 
loads, such as the spent fuel cask, and a spent fuel handling bridge crane which runs on rails
 
mounted on the operating floor. The overhead spent fuel cask crane is prevented from traveling
 
over the spent fuel pool by means of mec hanical stops and administrative controls. 
 
_____________
: a. Not required for safe shutdown of the plant and the river intake structure is no longer required to be maintained as Category I.
FNP-FSAR-3
 
3.8-51 REV 21  5/08 Mechanical anti-derailing devices mounted on the wheel assemblies of the overhead crane bridge and trolley prevent the crane from being dislodged from its rails due to horizontal motion
 
during an earthquake. The vertical acceleration due to an earthquake is not large enough to
 
overcome the crane's downward load due to gravity.
The same devices prevent overturning of the crane gantry in tornado winds. The gantry also has manual locking devices to prevent horizontal movement during design wind conditions and tornado winds. 
 
The spent fuel cask storage area is separated from the fuel transfer canal by a 3 ft 9-in.-thick, reinforced concrete wall and a gate at the fuel transfer slot. The gate will be in a closed position
 
during cask handling, completely isolating the cask storage area from the fuel transfer canal. 
 
The base slab of the cask storage area is composed of 5 ft of reinforced concrete placed over a
 
9 ft 5 in. concrete fill, which in turn rests on a 5 ft-thick reinforced concrete mat bearing directly
 
on the Lisbon formation. 
 
The spent fuel cask wash area is separated from the cask storage area and the fuel transfer canal by 3 ft-thick and 3 ft 9 in.- thick, respectively, reinforced concrete isolation walls. The
 
base slab of the cask wash area is composed of 5 ft reinforced concrete placed over a 30 ft
 
compacted backfill, which in turn rests on a 5 ft-thick reinforced concrete mat bearing directly on
 
the Lisbon formation. 
 
Figures 3.8-41 and 3.8-42 show the details of the spent fuel cask storage and the cask wash
 
areas. 
 
The fuel transfer from and into the containment is accomplished through the fuel transfer tube. 
 
Expansion joint bellows at the fuel transfer tube provide for the relative movement between
 
containment, containment internals, and the spent fuel pool. The bellows allow for all loading
 
conditions including the safe shutdown earthquake and maximum hydraulic pressure. The
 
design of the expansion bellows considers the maximum computed relative axial and lateral
 
displacement of the fuel transfer tube occurring simultaneously. Access for inspection and
 
maintenance of the bellows is provided. See figure 3.8-2 for details of the refueling transfer
 
tube. The spent fuel bundles are stored in stainless steel racks in the spent fuel pool. 
 
The spent fuel pool walls and base slab are constructed of thick (6 ft to 7 ft walls and 5-ft slab)
 
reinforced concrete. The inside face of the walls and base slab are lined with 1/4-in.-thick
 
stainless steel liner plate to provide leaktightness. The reinforced concrete superstructure of
 
the fuel handling area protects the spent fuel pool from the environment. 
 
B. Diesel Generator Building This reinforced concrete building, housing the diesel generators essential to safe plant shutdown, is a one-story, box-type structure. Reinforced concrete interior walls are provided to physically separate the diesel generators from each other. Figure 3.8-26
 
shows the general configuration of the building. 
 
The foundation, which consists of a reinforced concrete mat slab supported by concrete caissons, is anchored to the Lisbon formation, and is structurally separated from the
 
electrical cable tunnels by means of free joints. The 2 ft 6 in. exterior reinforced
 
concrete walls and the 2 ft roof slab provide protection against missiles, as described in
 
section 3.5.
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3.8-52 REV 21  5/08 C. River Intake Structure The river intake structure was originally designed as a Category I structure; however, it is not required to be maintained as Category I. It is a reinforced concrete, box-type structure which houses the river water pumps. It consists basically of two levels. The
 
upper level is a concrete enclosure which houses the river water pumps. The lower level consists of passages to supply the pumps with water. Trash racks and traveling screens
 
keep debris from entering the pump suction lines via the water passages. Figure 3.8-27
 
shows the configuration of the structure. It is placed on a base mat which bears directly
 
on the Lisbon formation. 
 
D. Intake Structure at Storage Pond The intake structure at the storage pond is a redundant Category I, reinforced concrete, box-type structure which houses the pumps, wetwell, and traveling screens. 
 
Figure 3.8-28 shows the configuration of the structure. The structure has a reinforced
 
concrete caisson foundation which is anchored into the Lisbon formation.
E. Storage Pond, Dam, and Dike
 
See appendix 2B to chapter 2.0 for the description of this structure.
 
F. Pond Spillway Structure
 
A description of this structure is given in section 2.4.8.1. In addition, figure 3.8-29 shows its configuration.
G. Electrical Cable Tunnel The electrical cable tunnel is a Category I, reinforced concrete, tubular type structure which encloses the emergency electrical cable between the diesel generator and auxiliary buildings. The emergency electrical cable is required for a safe shutdown of the plant. The reinforced concrete tunnel has a mat foundation which bears for the most
 
part on the Moody's Branch formation.
H. Outdoor Category I Tanks The following tanks are Category I: 
: 1. Refueling water storage tank. 
: 2. Reactor makeup water storage tank. 
: 3. Condensate storage tank. 
 
These tanks are cylindrical in shape and are supported by concrete mats resting on compacted backfill. The tanks are of steel plate construction and are designed to withstand seismic
 
generated loads. Shield and retaining walls are provided for the RWST and RMWST only to
 
safeguard the quantities of water in the tanks required for a safe shutdown of the plant. The
 
CST contains no shield or retaining walls, but has its bottom 12 feet reinforced to withstand FNP-FSAR-3
 
3.8-53 REV 21  5/08 ruptures caused by tornado-generated missiles. A tabulation of the tank capacities and dimensions is given below: 
 
Tank Capacity (gal) I.D. (ft) Height (ft)
Refueling water 500,000 46 41  storage tank Reactor makeup water storage tank 200,000 34 32 Condensate storage 500,000 46 41  tank 3.8.4.2  Applicable Codes, Standards and Specifications The following codes, standards, specifications, design criteria, NRC Regulatory Guides and industry standard practices constitute the basis for the design and construction of all Category I structures other than the containment. Modifications to these codes, standards, etc. are made
 
where necessary, to meet the specific requirement s of the structures. These modifications are indicated in the sections where references to the codes, standards, etc. are made. 
 
Codes  ACI 214-65 "Recommended Practices for Evaluation of Compression Test Results of Field Concrete."
ACI 301-66 "Specifications for Structural Concrete for Buildings."
 
ACI 306-66 "Recommended Practice for Cold Weather Concreting."
 
ACI 311-64 "Recommended Practices for Concrete Inspection."
 
ACI 315-65 "Manual of Standard Practice for Detailing Reinforced Concrete Structures."
 
ACI 318-63 "Building Code Requirements for Reinforced Concrete."
 
ACI 605-59 "Recommended Practice for Hot Weather Concreting."
 
ACI 347-63 "Recommended Practice for Concrete Formwork."
 
ACI 613-54 "Recommended Practice for Selecting Proportions for Concrete."
 
ACI 614-59 "Recommended Practice for Measur ing, Mixing, and Placing Concrete."
AISC  Manual of Steel Construction , 1963 and 1969 Editions.
 
AWS D1.1-86 Structural Welding Code - Steel.
AWS D2.0-69 Specifications for Welded Highway and Railway Bridges.
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3.8-54 REV 21  5/08 NCIG-01, "Visual Weld Acceptance Criteria for Structural Welding at Nuclear Power Plants" -  Rev. 2  EPRI NP-5380.
 
ICBO  Uniform Building Code , 1970 Edition.
SBCC  Southern Standard Building Code , 1969 Edition.
CFR  Code of Federal Regulations, Title 29, Chapter XVII, "Occupational Safety and  Health Standards."
 
Specifications
 
CMAA  Specifications for Electric Overhead Traveling Crane - No. 70, 1970 Edition.
ASTM  The specifications used are identified in the applicable subsections.
 
Design Criteria
 
ASCE  "Wind Forces on Structures," Paper No. 3269.
 
AEC  "Nuclear Reactor and Earthquake" - Publication TID 7024.
 
NRC Regulatory Guides
 
Regulatory Guide No. 1.10 "Mechanical (Cadweld) Splices in Reinforcing Bars of Concrete Containments."  Regulatory Guide No. 1.13 "Fuel Storage Facility Design Basis."
 
Regulatory Guide No. 1.15 "Testing of Reinforcing Bars for Concrete Structures."
 
Regulatory Guide No. 1.28 "Quality Assurance Program Requirements - Design and Construction."
Regulatory Guide No. 1.29 "Seismic Design Classification."
 
Regulatory Guide No. 1.31 "Control of Stainless Steel Welding."
 
Regulatory Guide No. 1.38 "Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage and  Handling of Items for Water Cooled Nuclear
 
Power Plants."
Regulatory Guide No. 1.55 "Concrete Placement in Category I Structures."
 
Regulatory Guide No. 1.64 "Quality Assurance Program Requirements for the Design of Nuclear Power Plants."
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3.8-55 REV 21  5/08 Bechtel Corporation Topical Reports BC-TOP-4  "Seismic Analysis of Structur es and Equipment for Nuclear Power Plants,"  (Rev. 1)  Sept 1, 1972
 
3.8.4.3  Loads and Loading Combinations
 
All Category I structures are designed for all credible conditions of loadings, including normal
 
loads, loads resulting from a pipe rupture where applicable, and loads due to adverse
 
environmental conditions. 
 
3.8.4.3.1 Loads 
 
The following loads are considered in the design: 
: a. Dead loads. 
: b. Live loads. 
: c. Earthquake loads. 
: d. Pipe rupture loads. 
: e. Thermal loads. 
: f. Wind and tornado loads. 
: g. Hydrostatic loads. 
: h. Cask drop loads. 
 
A. Dead loads. 
 
Structural dead loads consist of the weight of framing, roof, floors, walls, partitions, platforms, hangers, cable trays, pipes with fluid, and equipment dead loads, as specified on the drawings supplied by the manufacturers of the
 
equipment installed within the structure.
B. Live Loads 
 
Live loads consist of design floor loads, pool and tank liquid weights, and equipment live loads as specified on the dr awings supplied by the manufacturers of the equipment installed within the structure.
 
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3.8-56 REV 21  5/08  C. Earthquake Loads 
 
Earthquake loads are predicated on a basis of 1/2 of the safe shutdown earthquake (SSE) and having a horizontal ground acceleration of 0.05g, and a vertical ground acceleration of 0.033g. 
 
In addition, a safe shutdown earthquake (SSE) having a horizontal ground acceleration of 0.10g, and a vertical ground acceleration of 0.067g, is used to
 
check the design to ensure that loss of structural functions would not occur. 
 
Seismic response spectrum curves are given in section 3.7 for both horizontal and vertical ground motions. A dynamic analysis is used to compute the seismic loads for the design of structural elements.
D. Pipe Rupture Loads 
 
Pipe rupture loads include the jet-impingement forces from postulated pipe breaks, differential pressures that might build up across compartments, and
 
loads due to pipe whipping or pipe restraint. 
 
E. Thermal Loads 
 
Thermal loads are those induced in the spent fuel pool floor and walls due to the thermal gradients across these elements. Thermal gradients may be
 
caused by an increase in water temperature during operating conditions, or by
 
an accident. The interior temperatures of the pool are assumed to be 180°F for
 
both normal and abnormal conditions. The ambient temperature is assumed to
 
be 50°F for exterior walls and 70°F for the internal walls.
 
F. Wind and Tornado Loads 
 
The wind loadings and tornado loadings are discussed in section 3.3. 
 
All Category I structures are designed to withstand the effects of the wind and tornado loadings, and to provide protection against tornado missiles for all
 
Category I systems and components within the structures.
The structures are analyzed for tornado loadings not coincident with the safe shutdown earthquake. 
 
G. Hydrostatic Loads 
 
Lateral hydrostatic pressure loads and buoyant forces resulting from the displacement of ground and flood waters are applied to the structures, as
 
discussed in section 3.4. 
 
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3.8-57 REV 21  5/08  H. Cask Drop Loads 
 
Special lifting devices, as discussed in section 9.1.4.2.2.5, are provided for cask crane operation to prevent the dropping of the cask. However, a static equivalent load of 1000 Kips (over a 38.5 ft 2 area) has been considered in the design of the spent fuel pool slab.
3.8.4.3.2 Loading Combinations 
 
The load combinations and load factors for Category I structures listed in Section 3.8.4.1 are as
 
follows: 
 
C = 1/ (1.0D + 1.0R + 1.25E (or 1.25W))
 
C = 1/ (1.0D + 1.25H + 1.25E (or 1.25W))
 
C = 1/ (1.0D + 1.0R + 1.0E')
 
C = 1/ (1.0D + 1.25H + 1.0E' + 1.0R)
 
C = 1/ (1.0D + 1.0W t + 1.25H)
Where: 
 
C = required capacity of the structures. 
 
D = dead load of structure and equipment plus any other permanent loads contributing stress, such as soil or hydrostatic loads. In addition, a portion of "live load" will be added when such load is expected to be present during plant
 
operation. An allowance will also be made for future permanent loads. R = force of pressure on structure due to rupture of any one pipe. 
 
H = force on structure due to thermal expansion of pipes under operation conditions.
 
E = 1/2 safe shutdown earthquake resulting from horizontal ground surface acceleration of 0.05g; vertical acceleration is 2/3 horizontal acceleration. 
 
E'= safe shutdown earthquake resulting from horizontal ground surface acceleration of 0.10g; vertical acceleration is 2/3 horizontal acceleration. 
 
W = wind load. 
 
W t= tornado wind load including differential pressure. 
 
The capacity reduction factors that will be used in the design are: 
 
  = 0.90 for reinforced concrete in flexure.
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3.8-58 REV 21  5/08
  = 0.85 for tension, shear, bond, and anchorage in reinforced concrete.
  = 0.75 for spirally reinforced concrete compression members.
 
  = 0.70 for tied compression members.
 
  = 0.90 for fabricated structural steel.
 
  = 0.90 for reinforcing steel in direction tension.
3.8.4.4  Design and Analysis Procedures
 
The analysis procedures for the Category I structures listed in subsection 3.8.4.1 are based on
 
conventional methods, as found in standard textbooks and handbooks used in universities and
 
engineering practice. 
 
The design procedures of these Category I structures are in accordance with design methods of
 
accepted standards and codes where applicable for normal operating loads. 
 
The seismic analysis of these structures is covered in section 3.7. The structures are
 
proportioned to maintain elastic behavior when subjected to various combinations of dead
 
loads, live loads, wind loads, tornado loads, seis mic loads, and LOCA loads. The upper limit of elastic behavior is the yield strength of the effective load carrying structural materials. 
 
3.8.4.5  Structural Acceptance Criteria
 
The limiting values of stress, strain, and gross deformations are established by the following
 
criteria: 
: a. To maintain the structural integrity when subjected to the worst load combinations. 
: b. To prevent structural deformations from displacing the Category I equipment to the extent that it suffers a loss of function.
The allowable stresses are those specified in the applicable codes. The stress contributions
 
due to earthquakes are included in the load combinations described in subsection 3.8.4.3. 
 
Structural deformations were found not to be a controlling criterion in the design of the Category
 
I structures, listed in subsection 3.8.4.1. 
 
Tables 3.8-10 through 3.8-14 give the load combinations, the calculated and allowable stresses, as well as the method of analysis used in the design of the main structural components of the
 
structures. The ratio of the allowable to the calculated stresses yields the safety margins. 
 
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3.8-59 REV 21  5/08 3.8.4.6  Materials, Quality Control, and Special Construction Techniques
 
The following basic materials are used in the construction of the Category I structures listed in
 
subsection 3.8.4.1. 
: a. Concrete 
 
Auxiliary building f'c (psi)  = 5,000 Diesel generator building f'c (psi)  = 3,000 f'c (psi)  = 4,000 All other structures listed in subsection 3.8.4.1 f'y (psi)  = 4,000
: b. Reinforcing steel Deformed bars ASTM A-615 f y  (psi)  = 60,000 Grade 60 Spiral bars ASTM A-82 f y  (psi)  = 70,000
: c. Structural and miscellaneous steel Rolled shapes,    bars and plates ASTM A-36 f y  (psi)  = 36,000  High strength ASTM A-325 bolts A-193 B-7, or A-490 Stainless steel ASTM A-240 Type 304 Insert plates ASTM A-516 f y  (psi)  = 38,000  (Auxiliary Grade 70    building)
 
The materials and the quality control procedures have been described in paragraph 3.8.1.6. 
 
The Category I structures listed in paragraph 3.8.4.1 are built of reinforced concrete and
 
structural steel, using proven methods common to heavy industrial construction. No special construction techniques have been employed in t he construction of these structures. 
 
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3.8-60 REV 21  5/08 3.8.4.7  Testing and Inservice Surveillance Requirements
 
No structural preoperational testing of the Category I structures is planned. During the life of
 
the plant, periodic inspections of the structures will be made to employ visual inspection for apparent structural deterioration such as large cracks and excessive deflection of structural members. 
 
For the Category I structures required to be maintained as Category I, periodic inspections
 
performed under the Structural Monitoring Program and Service Water Pond Dam Inspection
 
Program (as applicable) are required license r enewal aging management program activities for the period of extended operation (see chapter 18, subsections 18.2.10 and 18.2.3). The noncivil features of the outdoor tanks (e.g., fluid-retaining) are age-managed separately from
 
these structural inspection programs as part of the associated fluid systems.
 
All seam and plug welds in the spent fuel pool liner plate were vacuum box tested upon
 
completion of the welding. Where vacuum box testing was not possible, liquid penetrant testing was performed. The service water pond dam and spillway will be inspected during the period of
 
extended operation on a periodic basis in accordance with NRC Regulatory Guide 1.127, Revision 1.
 
The spent fuel pool has a system that provides for leakage to be detected at any time in the life
 
of the plant. This system consists of troughs under the liner plate which lead to a collection
 
system where leakage can be observed. 
 
3.8.5 FOUNDATIONS AND CONCRETE SUPPORTS 
 
All Category I structures are founded, either directly or by means of caissons, on the Lisbon
 
formation. 
 
3.8.5.1  Description of the Foundations and Supports
 
A. Containment
 
The containment foundation is a conventionally reinforced circular concrete mat, 9 ft thick with a diameter of 146 ft 6 in. bearing directly on the Lisbon formation.
The reactor cavity is located in the center of the mat and forms an integral part of
 
the foundation. Figure 3.8-36 shows the relative position of the two containment
 
foundations and the other Category I structure's foundations. 
 
Figure 3.8-37 shows cross-sections of the containment base slab. 
 
The internal structures that support the large equipment, such as steam generators and reactor coolant pumps, are anchored to the base slab in order to transfer the loads. Figure 3.8-6 shows a typical detail of anchorage to the base
 
slab for the steam generator and reactor coolant pumps. 
 
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3.8-61 REV 21  5/08  Figure 3.8-37 shows the reinforcing pattern at the junction of the base slab and containment wall.
B. Auxiliary Building
 
The Auxiliary Building foundation is a reinforced concrete slab 5 ft 0 in. thick, 430 ft long by 300 ft wide, bearing directly on the Lisbon formation. In the eastern section of the Auxiliary Building, specifically east of column line P, the structure is
 
supported on spread footings 9 ft 0 in. x 9 ft 0 in. x 4 ft 0 in. which bear on the
 
Lisbon formation. The loads are transmitted through cast-in place reinforced
 
concrete columns 3 ft 6 in. diameter. 
 
Refer to figure 3.8-36 for location of auxiliary building foundations in relation to other Category I structures. 
 
Figure 3.8-38 shows typical base slab and foundation details. 
 
C. Diesel Generator Building The Diesel Generator Building foundation is a 4 ft 0 in.-thick reinforced concrete slab bearing on 5 ft 0 in.-diameter cast-in place caissons which transfer the loads
 
to the Lisbon formation. Figure 3.8-39 shows details of caissons, caisson to
 
base slab connection, and load transfer mechanism. Figure 3.8-36 shows the
 
relative position of the diesel generator building foundation to the other
 
Category I structures. 
 
The diesel generator foundations are projected 4 1/2 in. above the base slab and form an integral part of it. 
 
D. River Intake Structure The river intake structure foundation is a 5-ft 0-in.- thick (3 ft 0 in. in the bay area) reinforced concrete slab which bears directly on the Lisbon formation. 
 
The load transfer is done through bearing walls and columns which support the superstructure. 
 
Refer to figure 3.8-36 for location of river intake structure foundation relative to other Category I structures.
 
E. Intake Structure at Storage Pond
 
The intake structure at storage pond foundation is a 3-ft-thick reinforced concrete slab bearing on 6 ft, 7 ft, and 10 ft-diameter cast in place caissons which transfer
 
the loads to the Lisbon formation.
Refer to figure 3.8-36 for location of intake structure at the storage pond foundation relative to other Category I structures. 
 
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3.8-62 REV 21  5/08  Figure 3.8-39 shows a typical detail of caisson and base slab. 
 
Bearing walls and columns support the superstructure. 
 
F. Electrical Cable Tunnel The electrical cable tunnel foundation is a reinforced concrete slab bearing directly on the Moodys Branch formation for the most part. In areas where direct
 
bearing was not possible, caissons anchored into the Lisbon formation were
 
used.
G. Category I Outdoor Tanks There are three Category I outdoor tanks: 
: 1. Refueling water storage tank. 
: 2. Reactor makeup water storage tank. 
: 3. Condensate storage tank. 
 
The foundations of these tanks are 4 ft 0 in.-thick reinforced concrete slabs bearing on compacted fill.
The tank foundations are physically separated from each other as shown in figure 3.8-36.
3.8.5.2  Applicable Codes, Standards and Specifications
 
The applicable codes, standards and specifications are discussed in the following subsections: 
 
Containment    -  3.8.1.2 Internal Structures  -  3.8.3.2 Other Category I Structures  -  3.8.4.2 
 
3.8.5.3  Loads and Loading Combinations Containment foundation loads and loading combinations are discussed in subsection 3.8.1.3. 
 
Foundation loads and loading combinations for other Category I structures are discussed in
 
subsection 3.8.4.3. 
 
3.8.5.4  Design and Analysis Procedures
 
Design and Analysis Procedures for the Containment including the base slab are discussed in
 
Topical Report BC-TOP-5 Sections 6.0 and 7.0.
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3.8-63 REV 21  5/08 The basic techniques for analysis and design of the foundations for all other Category I structures are the conventional methods which involve simplifying assumptions such as are
 
found in the theory of concrete structures. Stresses resulting from local moments, torques, concentrated reactions and uniform loadings are computed by these methods. These methods are further discussed in subsections 3.8.3.4 and 3.8.4.4. 
 
3.8.5.5  Structural Acceptance Criteria
 
The foundations of all Category I structures are designed to meet the same structural
 
acceptance criteria as the structures themselves. These criteria are discussed in subsections
 
3.8.1.5, 3.8.3.5 and 3.8.4.5. 
 
The limiting conditions for the foundation medium together with a comparison of actual capacity and estimated structure loads are found in chapter 2.0, appendix 2B. 
 
3.8.5.6  Materials, Quality Control, and Special Construction Techniques The foundations and equipment supports are built of reinforced concrete using proven methods
 
for heavy industrial construction. The description of the materials and the quality control
 
procedures, as well as special construction tec hniques for foundations, are the same as those discussed in paragraphs 3.8.1.6, 3.8.3.6, and 3.8.4.6 and chapter 17.0.
 
3.8.5.7  Testing and Inservice Surveillance Requirements Testing and inservice surveillance are not required and are not planned for foundations of
 
structures or supports. A discussion of the test program which serves as the basis for the soils
 
investigation and foundation evaluation may be found in chapter 2.0, appendix 2B. 
 
For the period of extended operation, periodic inspections of the Category I structures required
 
to be maintained as Category I (see paragraph 3.8.4.7) are credited license renewal aging
 
management program activities for aging managem ent of foundations and concrete supports.
 
3.8.6 MASONRY WALLS
 
As documented in SER NUREG 0117, Supplement 5 to NUREG 75/034 dated March 1981, the
 
NRC acceptance criteria associated with the seismic design of masonry walls is contained in IE
 
Bulletin 80-11 and the applicable requirements of GDC 2 and GDC 4. Additionally, as
 
documented in NRC miscellaneous letter dated Oct. 24, 1984, the following three options are
 
accepted by the NRC to seismically qualify masonry walls:
: 1. Reanalyze walls qualified by the energy-balance technique by linear elastic working stress approach as recommended in the staff acceptance criteria (SRP Section 3.8.4, Appendix A) and implement modifications to walls as needed.
 
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3.8-64 REV 21  5/08 2. Develop rigorous nonlinear time-history analysis techniques capable of capturing the mechanism of the walls under cyclic loads. Different stages of behavior should be accurately modeled; elastic uncracked, elastic cracked, and inelastic cracked with
 
yielding of the central rebars. Then, a lim ited number of dynamic tests (realistic design earthquake motion inputs at top and bottom of the wall) should be conducted to demonstrate the overall conservatism of the analysis results. In this case, "as-built"
 
walls should be constructed to duplicate the construction details of a specific plant.
: 3. For walls qualified by energy-balance technique, conduct a comprehensive test program to establish the basic non-linear behavioral characteristics of masonry walls (i.e., load-
 
deflection hysteretic behavior, ductility rations, energy absorption and post yield envelopes) for material properties and construction details pertaining to masonry walls in question. The behavior revealed from tests should then be compared with that of
 
elastic-perfectly-plastic materials for which the energy balance technique was originally
 
developed. If there are significant differences, then the energy balance technique should
 
be modified to reflect the actual wall behavior.
 
Structural Monitoring Program inspections of seismically qualified masonry walls are credited license renewal aging management program activities for the period of extended operation (see chapter 18, subsection 18.2.10).
 
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3.8-65 REV 21  5/08
 
==REFERENCES:==
: 1. Welding Research Council (WRC) Bulletin No. 102, "State of Stress in a Circular Cylinder Shell with a Circular Hole," By Eringen, Naghdi and Thiel.
FNP-FSAR-3 REV 21  5/08 TABLE 3.8-1 POST-TENSIONING SYSTEM - BBRV (170)
Designation  Wires Ultimate 2000 kips capacity 
 
Design 1200 kips capacity 
 
Minimum 240 ksi tensile strength 
 
Relaxation 8 1/2 percent
@ 0.70 f
 
Ductility 4 percent (a)
End Buttonhead anchorage 
 
Anchor Head H.R. 4140 material or 4142 Alloy Steel
 
Bushing H.F.S.M.
4142 Tubing Shim ASTM A-36-70a
.40/.50 Carbon Sheet Steel
 
Bearing ASTM A-36-70a plate 
: a. When measured in a gauge length of 10 inches (for wire only).
FNP-FSAR-3 TABLE 3.8-2  (SHEET 1 OF 8)
STRESS ANALYSIS RESULTS
 
REV 21  5/08 FNP-FSAR-3 TABLE 3.8-2 (SHEET 2 OF 8)STRESS ANALYSIS RESULTS NOTATION AND NOTES FOR STRESS AND STRAIN TABLES NOTATION: D=DEAD LOAD FF=MINIMUM GUARANTEED PRESTRESS LEVEL FI=INITIAL PRESTRESS LEVEL P=INTERNAL PRESSURE E=1/2 SAFE SHUTDOWN EARTHQUAKE EI=SAFE SHUTDOWN EARTHQUAKE TA=ACCI DENT TEMPERATUR E fl=COMPRESSIVE STRENGTH OF CONCRETE c t=THICKNESS OF SECTION Pm=MERIDIONAL STEEL PERCENTAGE Ph=HOOP STEEL PERCENTAGE NOTES: THE HIGHER ALLOWABLE CONCRETE STRESS IS FOR THE DOME AND WALL.THE LOWER ALLOWABLE CONCRETE STRESS IS FOR THE BASE SLAB ONLY.DEFLECTIONS ARE MEASURED HORIZONTALLY FOR THE WALL ANDCALLY FOR THE BASE SLAB.DEFLECTIONS FOR THE DOME ARE THE RESULANT OF THE VERTICAL AND HORIZONTAL COMPONENTS OF DEFLECTION AT THAT SECTION.THE PREDICTED VALUES OF STRAIN IN CONCRETE AND REINFORCEMENT SHOWN IN THESE TABLES ARE THE VALUES RELATIVE TO THE ASSUMED REFERENCE TEMPERATURE OF 70°F AND ARE COMPUTED BY THE"FINEL" PROGRAM (CE-316-4)
BASED ON THE 3*DIMENSIONAL STRESS-STRAIN*TEMPERATURE RELATIONSHIP.
THE CONCRETE STRAIN VALUES SHOWN IN THESE TABLES ARE THE MAXIMUM VALUES OF STRAIN AT EACH PARTICULAR SECTION;THEREFORE, THE CONCRETE STRAINS MAY NOT NECESSARILY EQUAL THE REINFORCEMENT STRAINS IN MAGNITUDE AT EACH SECTION.THE CONCRETE STRESSES SHOWN IN THESE TABLES ARE THE MAXIMUM VALUES OF STRESS FOR EACH PARTICULAR SECTION.REV 21 5/08 FNP-FSAR-3 TABLE 3.8-2  (SHEET 3 OF 8)
STRESS ANALYSIS RESULTS DEAD LOAD AND INITIAL PRESTRESS (D + F I)      REV 21  5/08 FNP-FSAR-3 TABLE 3.8-2  (SHEET 4 OF 8)
STRESS ANALYSIS RESULTS DEAD LOAD, FINAL PRESTRESS, AND 115% DESIGN PRESSURE (D + F F + 1.15P)
REV 21  5/08 FNP-FSAR-3 TABLE 3.8-2  (SHEET 5 OF 8)
STRESS ANALYSIS RESULTS DEAD LOAD, FINAL PRESTRESS, DESIGN PRESSURE, AND THERMAL ACCIDENT (D + F F + P + T A)      REV 21  5/08 FNP-FSAR-3 TABLE 3.8-2  (SHEET 6 OF 8)
STRESS ANALYSIS RESULTS DEAD LOAD, FINAL PRESTRESS, 150% DESIGN PRESSURE, AND THERMAL ACCIDENT (D + F F + 1.5P + T A)      REV 21  5/08 FNP-FSAR-3 TABLE 3.8-2  (SHEET 7 OF 8)
STRESS ANALYSIS RESULTS DEAD LOAD, FINAL PRESTRESS, 125% DESIGN PRESSURE, 125% OF 1/2 SAFE SHUTDOWN EARTHQUAKE, AND THERMAL ACCIDENT (D + F F + 1.25P + 1.25E + T A)
REV 21  5/08 FNP-FSAR-3 TABLE 3.8-2  (SHEET 8 OF 8)
STRESS ANALYSIS RESULTS DEAD LOAD, FINAL PRESTRESS, DESIGN PRESSURE, SAFE SHUTDOWN EARTHQUAKE, AND THERMAL ACCIDENT (D + F F + P + E' + T A)      REV 21  5/08
 
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TABLE 3.8-3  (SHEET 1 OF 6)
CONTAINMENT STRAINS (X 10
-6) DEAD LOAD AND INITIAL PRESTRESS (D + F I)      REV 21  5/08 FNP-FSAR-3
 
TABLE 3.8-3  (SHEET 2 OF 6)
CONTAINMENT STRAINS (X 10
-6) DEAD LOAD, FINAL PRESTRESS, AND 115% DESIGN PRESSURE (D + F F + 1.15P)
REV 21  5/08 FNP-FSAR-3
 
TABLE 3.8-3  (SHEET 3 OF 6)
CONTAINMENT STRAINS (X 10
-6) DEAD LOAD, FINAL PRESTRESS, DESIGN PRESSURE, AND THERMAL ACCIDENT (D + F F + P + T A)      REV 21  5/08 FNP-FSAR-3
 
TABLE 3.8-3  (SHEET 4 OF 6)
CONTAINMENT STRAINS (X 10
-6) DEAD LOAD, FINAL PRESTRESS, 150% DESIGN PRESSURE, AND THERMAL ACCIDENT (D + F F + 1.5P + T A)      REV 21  5/08 FNP-FSAR-3
 
TABLE 3.8-3  (SHEET 5 OF 6)
CONTAINMENT STRAINS (X 10
-6) DEAD LOAD, FINAL PRESTRESS, 125% DESIGN PRESSURE, 125% OF 1/2 SAFE SHUTDOWN EARHTQUAKE, AND THERMAL ACCIDENT (D + F F + 1.25P + 1.25E + T A)      REV 21  5/08 FNP-FSAR-3
 
TABLE 3.8-3  (SHEET 6 OF 6)
CONTAINMENT STRAINS (X 10
-6) DEAD LOAD, FINAL PRESTRESS, DESIGN PRESSURE, SAFE SHUTDOWN EARHTQUAKE, AND THERMAL ACCIDENT (D + F F + P + E' + T A)      REV 21  5/08
 
FNP-FSAR-3 TABLE 3.8-4  (SHEET 1 OF 8)
CONTAINMENT STRESSES IN EQUIPMENT HATCH AREA
 
REV 21  5/08 FNP-FSAR-3 TABLE 3.8-4 (SHEET 2 OF 8)CONTAINMENT STRESSES IN EQUIPMENT HATCH AREA NOTATION AND NOTES FOR STRESS AND STRAIN TABLES NOTATION: D DEAD LOAD FF=0 MINIMUM GUARANTEED PRESTRESS LEVEL FI..INITIAL PRESTRESS LEVEL P=INTERNAL PRESSURE E..1/2 SAFE SHUTDOWN EARTHQUAKE EI=SAFE SHUTDOWN EARTHQUAKE TA" ACCIDENT TEMPERATURE NOTES: THIRTEEN POINTS, POSITIONED RADIALLY ABOUT THE CENTER OF THE EQUIPMENT HATCH, ARE SELECTED FOR TABULATION.
THIS COVERS THE LOCAL AREA WHERE STRESS CONCENTRATIONS DUE TO THE PRESENCE OF A LARGE OPENING CAN BE FELT.EQUIPMENT HATCH POLAR COORDINATE SYSTEM IS USED FOR LOCATIONS 1,2,&3.RADIAL STRESS (MERIDIONAL IN THE TABLE)IN THESE LOCATIONS IS NEGLIGIBLE, AND THUS IS NOT TABULATED.
CONTAINMENT GLOBAL CYLINDRICAL COORDINATE SYSTEM IS USED FOR OTHER LOCATIONS.
FOR CONCRETE, ONLY COMPRESSive STRESS IS CONSIDERED.
THE LINER PLATE YIELDS IN SOME LOADING COMBINATIONS.
STRAIN WILL GOVERN THE DESIGN.THEREFORE, MAXIMUM STRAIN IS ALSO TABULATED FOR THE LINER PLATE.REV 21 5/08 FNP-FSAR-3 TABLE 3.8-4  (SHEET 3 OF 8)
CONTAINMENT STRESSES IN EQUIPMENT HATCH AREA DEAD LOAD AND INITIAL PRESTRESS (D + F I)      REV 21  5/08 FNP-FSAR-3 TABLE 3.8-4  (SHEET 4 OF 8)
CONTAINMENT STRESSES IN EQUIPMENT HATCH AREA DEAD LOAD, FINAL PRESTRESS, AND 115% DESIGN PRESSURE (D + F F + 1.15P)
REV 21  5/08 FNP-FSAR-3 TABLE 3.8-4  (SHEET 5 OF 8)
CONTAINMENT STRESSES IN EQUIPMENT HATCH AREA DEAD LOAD, FINAL PRESTRESS, DESIGN PRESSURE, AND THERMAL ACCIDENT  (D + F F + P + T A)      REV 21  5/08 FNP-FSAR-3 TABLE 3.8-4  (SHEET 6 OF 8)
CONTAINMENT STRESSES IN EQUIPMENT HATCH AREA DEAD LOAD, FINAL PRESTRESS, 150% DESIGN PRESSURE, AND THERMAL ACCIDENT (D + F F + 1.5P + T A)      REV 21  5/08 FNP-FSAR-3 TABLE 3.8-4  (SHEET 7 OF 8)
CONTAINMENT STRESSES IN EQUIPMENT HATCH AREA DEAD LOAD, FINAL PRESTRESS, 125% DESIGN PRESSURE, 125% OF 1/2 SAFE SHUTDOWN EARTHQUAKE, AND THERMAL ACCIDENT (D + F F + 1.25P + 1.25E + T A)
REV 21  5/08 FNP-FSAR-3 TABLE 3.8-4  (SHEET 8 OF 8)
CONTAINMENT STRESSES IN EQUIPMENT HATCH AREA DEAD LOAD, FINAL PRESTRESS, DESIGN PRESSURE, SAFE SHUTDOWN EARTHQUAKE, AND THERMAL ACCIDENT (D + F F + P + E' + T A)
REV 21  5/08
 
FNP-FSAR-3 REV 21  5/08 TABLE 3.8-5 AGGREGATE TESTS ASTM  Results to Initial User's Daily No. Title Be Achieved Test Test Test C-33 Gradation To conform with spec X  X
 
C-40 Organic impurities To conform with spec X  X
 
C-87 Mortar making properties To conform with spec X 
 
C-88 Soundness To conform with spec X X 
 
C-117 Specific gravity and No. 200 sieve Design mix calculations X 
 
C-127 Specific gravity and absorption (fine aggregates) Design mix calculations X 
 
C-128 Specific gravity and absorption (fine aggregates) Design calculations X 
 
C-131 Los Angeles abrasion To conform with spec X X 
 
C-136 Sieve analysis To conform with spec X 
 
C-142 Clay lumps To conform with spec X 
 
C-227 Potential reactivity (mortar bar) To conform with spec X 
 
C-289 Potential reactivity (chemical) To conform with spec X X 
 
C-295 Petrographic To conform with spec X 
 
FNP-FSAR-3 REV 21  5/08 TABLE 3.8-6 CEMENT TESTS ASTM  Initial  Periodic No. Type of Test Test User's Test Tests C-109 Compressive strength X X X
 
C-114 Chemical analysis X X 
 
C-115 Fineness-turbidimeter X X 
 
C-151 Autoclave expansion (Soundness) X X 
 
FNP-FSAR-3 REV 21  5/08 TABLE 3.8-7 FLY ASH TESTS ASTM  Initial  Periodic
 
No. Type of Test Test User's Test Tests     
 
C-109 Compressive strength X X X
 
C-114 Chemical analysis X X 
 
C-151 Autoclave expansion(Soundness) X X 
 
C-188 Specific gravity X X 
 
C-311 Sampling and testing X X 
 
FNP-FSAR-3 REV 21  5/08 TABLE 3.8-8 PRESTRESSING SEQUENCES Numbers of Tendons Phase Hoop Dome Vertical Total Description 1 57 - - 57 Between the top of base slab and approximately      30 ft below the bottom of the ring grider.
2 6 - - 63 Between approximately 30 ft below the bottom      of the ring girder and the uppermost tendon.
3 - - 33 96 Every fourth tendon of three vertical tendon      groups at 120 degrees apart.
4 - - 33 129 Repeat Phase 3 with the tendons immediately      adjacent to the last tendons.
5 174 Stressing every other tendon on alternate sides      of the center ones moving outward.
6 - - 33 207 Continue stressing every fourth tendon of      three vertical tendon groups at 120 degrees apart.
7 - - 31 238 Stressing the remaining vertical tendons.      8 54 - - 292 The remaining 50 percent of tendons specified in Phase 1.
9 18 - - 310 The remaining 75 percent of the tendons specified in Phase 2.
10 358 Stressing the remaining out-most dome tendons toward the      center of each group.
Total 135 93 130 358 
 
FNP-FSAR-3 REV 21  5/08 TABLE 3.8-9 CALCULATED RESULTS - INTERNAL STRUCTURES TOTAL ALLOWABLE    CALCULATED REINF.
REINF. STEEL STEEL  DESCRIPTION  STRESS STRESS 
 
OF MEMBER  LOCATION OF MEMBER LOAD COMBINATION    KSI        KSI    REMARKS 5 ft 0 in. con- Reactor cavity wall from EL 79 ft 0 in. 1.0 D + 1.0 L + 1.0 To 19.5 20 WSD crete wall to EL 102 ft 0 in. 1.0 D + 1.0 L + 1.0 To + 1.0 P 28.2 54 USD
 
9 ft 8 in. con- Primary shield wall from EL 102 ft 0 in. 1.0 D + 1.0 P + 1.25 E crete wall to EL 129 ft 0 in. 1.0 D + 1.25 To + 1.25E 49.3 54 USD 1.0 D + 1.25 To + 1.0E' + 1.0P   
 
2 ft 6 in. con- Secondary shield wall from EL 104 ft 0 1.0 D + 1.0 R 47.6 54 USD crete wall in. to EL 125 ft 9 in.   
 
3 ft 9 in. and Refueling cavity wall 1.0 D + 1.0 E' 33.9 54 USD 5 ft 0 in. con-crete wall     
 
3 ft 3 in. con- Slab at EL 129 ft 0 in. 1.0 D + 1.0 L + 1.0 E 25.4 32 WSD crete slab  1.0 D + 1.0 L +/- 1.0 E' + 1.0 P 43.0 54 USD
 
3 ft 6 in. con- Secondary shield wall from EL 129 ft 1.0 D + 1.0 L 18.3 24 WSD crete wall 0 in. to EL 152 ft 0 in. 1.0 D + 1.0 L + 1.0 P 44.3 54 USD 1.0 D + 1.0 L + 1.0 R 42.5 54 USD
 
3 ft 0 in. con- Slab at EL 155 ft 0 in. 1.0 D + 1.0 L + 1.0 E 23.0 32 WSD crete slab  1.0 D + 0.1 L + 1.0 E' + 1.0 R 40.4 54 USD
 
2 ft 0 in. and Secondary shield wall above 1.0 D + 1.0 P 44.8 54 USD 3 ft 0 in.
concrete wall     
 
D =  Dead load  L =  Live load T =  Operating temperature load  P =  Pressure load E =  1/2 safe shutdown earthquake load  E'=  Safe shutdown earthquake load R =  Pipe rupture load  WSD = Working stress design USD = Ultimate strength design
 
FNP-FSAR-3 REV 21  5/08 TABLE 3.8-10 (SHEET 1 OF 2)
CALCULATED RESULTS - AUXILIARY BUILDING Total      Calculated Allowable Stress, Or Stress, Or  Description  Load Required Maximum Of Member  Location Of Member Combination Capacity Capacity Remarks 3-ft 6-in. diam. Cols for cask crane along concrete column Col. line V from EL 100-ft D + L + E P= 498 K P m = 1970 K WSD  0-in. to 146-ft 6-in. south-east quadrant   
 
4-ft 6-in. diam. Spent fuel pool support D + L + E P= 3153 K P m = 3730 K WSD concrete column between Col lines O and T, and 2 and 9.8 - EL 100-ft 0-in.   
 
2-ft 6-in. diam. Cols supporting electric D + L + E P= 753 K P m = 947 K WSD concrete column penetration rooms from EL 100-ft 0-in. to 175-ft 0-in.   
 
2-ft 3-in. by Floor in demineralizer D + L + E + Pipe M= 1885 ft-K M m = 2000 ft-K WSD 5-ft 6-in. area at Col. line 14, concrete beam between Col. lines M and N - EL 139-ft 0-in.   
 
2-ft 0-in. by Floor in hot machine D + L + E M= 665.5 ft-K M m = 725 ft-K WSD 4-ft 0-in. shop between Col. Lines concrete beam 18 and 19, and R and U-EL 155-ft 0-in.   
 
D = Dead load M = Maximum moment L = Live load P = Maximum load E = Earthquake load WSD = Working stress design M = Missile load  H = Hydrostatic load S = Surcharge load  W = Tornado load
 
FNP-FSAR-3 REV 21  5/08 TABLE 3.8-10 (SHEET 2 OF 2)
Total      Calculated Allowable Stress, Or Stress, Or  Description  Load Required Maximum Of Member  Location Of Member Combination Capacity Capacity Remarks 2-ft 0-in. Floor in radwaste filter D + L + E + Pipe M= 12.9 ft-K M m = 29.8 ft-K WSD concrete slab room between Col. Lines J and N, and 17 and 18 -
EL 139-ft 0-in.   
 
2-ft 0-in. by Floor in hot machine shop D + L + E M= 665.5 ft-K M m = 725 ft-K WSD 4-ft 0-in. between Col. Lines 18 and concrete beam 19, and R and U - EL 155-ft 0-in.   
 
2-ft 0-in. Floor in radwaste filter room D + L + E + Pipe M= 12.9 ft-K M m = 29.8 ft-K WSD concrete slab between Col. Lines J and N, and 17 and 18 - EL139-ft 0-in.   
 
2-ft 0-in. con- Roof slabs D + L + M M= 43.91 ft-K M m = 59.8 ft-K WSD crete roof slab            5-ft 0-in. Floor in cask wash area D + L + 1000 K M= 80.2 ft-K M m = 143 ft-K WSD concrete slab
 
between Col. lines P and (Impact)
Q, and 7 and 9.3 - EL 139-ft      0-in.   
 
3-ft 6-in. Wall below EL 121-ft 0-in. D + L + H + S M+ 235 ft-K M m = 274 ft-K WSD concrete wall Col. lines P northeast quadrant - east wall   
 
2-ft 0-in. Wall above EL 155-ft 0-in. D + L + W t + M 1 M= 20.5 ft-K M m = 36 ft-K WSD concrete wall southwest quadrant - west wall, typical   
 
D = Dead load  M = Maximum moment L = Live load  P = Maximum load E = Earthquake load  WSD = Working stress design M = Missile load  S = Surcharge load H = Hydrostatic load  W = Tornado load FNP-FSAR-3 REV 21  5/08 TABLE 3.8-11 CALCULATED RESULTS - DIESEL GENERATOR BUILDING Total Calculated Allowable Description  Stress, Or Required Stress, Or Maximum 
 
Of Member Location Of Member Load Combination Capacity Capacity Remarks Caissons (Typ.) Top elevation 151-ft 0-in. 1.0 D + 0.5 L + 1.0 E P u  = 330 K P u  = 450 K USD  5-ft 0-in 54-ft 0-in. long  M u  = 2480 K M u  = 3250 K USD Slab 4-ft 0-in. Ground Slab on caissons 1.0 D + 0.5 L + 1.0 E M u  = 277 K M u  = 345 K USD thick 33-ft 0-in. EL 155-ft 0-in.
Span     
 
Exterior walls 2-ft Exterior walls 1.0 D + 0.5 L + 1.0 W T M u  = 94 ft-K M u = 350 ft-K USD 6-in. thick, 20-ft 0-in. span     
 
Interior walls 1-ft Interior walls 1.0 D + 0.5 L + 1.0 E M u  = 31 ft-K M u  = 40 ft-K USD 6-in. thick, 20-ft 0-in. span     
 
Roof 2-ft 0-in. Roof EL 177-ft 0-in. 1.0 D + 0.5 L + 1.0 W T M    = 78 ft-K M u  = 175 ft-K USD thick 33-ft 0-in. span     
 
D = Dead load  W = Tornado wind load L = Live load strength  P = Ultimate load E = Earthquake load design  M = Ultimate moment USD = Ultimate FNP-FSAR-3 REV 21  5/08 TABLE 3.8-12 CALCULATED RESULTS - RIVER INTAKE STRUCTURE Total      Calculated Allowable Stress, Or Stress, Or Description  Required Ultimate 
 
Of Member Location of Member Load Combination Capacity Capacity Remarks Base slab EL 61 ft 6 in. D + L + H + S M b  = 3859 ft-K M m  = 4107 ft-K WSD 6 ft 0 in. thick M t  = 2538 ft-K M m  = 2590 ft-K Base slab EL 64 ft 0 in. D + L + H + S M b  = 635 ft-K M b  = 676 ft-K WSD Bay area 3 ft 0 in. thick Bay area  M t  = 267 ft-K M t  = 436 ft-K  3 ft 0 in. thick     
 
Walls Bays area D + L + H + S M  = 380 ft-K M m  = 443 ft-K WSD Walls 3 ft 0 in. thick Exterior walls D + L + H + S M  = 119 ft-K M m  - 147 ft-K WSD Roof Slab 2 ft 0 in. thick EL 128 ft 0 in D + L M  = 43 ft-K M m  = 45 ft-K WSD
 
D = Dead load  Mb = Moment at bottom L = Live load  Mt = Moment at top H = Hydrostatic load  Mm = Maximum moment S = Surcharge load WSD = Working stress design
 
FNP-FSAR-3 REV 21  5/08 TABLE 3.8-13 CALCULATED RESULTS - INTAKE STRUCTURE AT STORAGE POND Total      Calculated Allowable Stress, Or Stress, Or Description  Required Ultimate 
 
Of Member  Location of Member Load Combination Capacity Capacity Remarks Base slab EL 151 ft 6 in. D + L + H + S M  = 43 ft-K M m  = 72 ft-K WSD 3 ft 0 in. thick     
 
Exterior walls See figure 3.8-28 D + L + H + S M  = 176 ft-K M m  = 242 ft-K WSD 4 ft 0 in. thick     
 
Columns See figure 3.8-28 D + L M  = 59 ft-K M m  = 381 ft-K WSD Typically 3 ft 0 in. P  = 304 ft-K P m  = 1414 ft-K WSD Floor slab Operating floor D + L M b  = 61 ft-K M m  = 88 ft-K WSD 3 ft 0 in. thick  M t  = 161 ft-K M m  = 176 ft-K Roof slab 2 ft 6 in. thick See figure 3.8-28 D + L + W M  = 59 ft-K M m  = 74 ft-K WSD
 
D = Dead load  M b = Moment at bottom  L = Live load  M t = Moment at top  H = Hydrostatic load  M m = Maximum moment  S = Surcharge load  P m = Maximum load  W = Tornado wind load  WSD = Working stress design FNP-FSAR-3 REV 21  5/08 TABLE 3.8-14 CALCULATED RESULTS - ELECTRICAL CABLE TUNNELS Total      Calculated Allowable Stress, Or Stress, Or Description  Required Ultimate 
 
Of Member  Location of Member Load Combination Capacity Capacity Remarks Base slab Electrical cable tunnels con- (D + L + E) 0.75 M  + 50 ft-K M m  = 60 ft-K WSD 2-ft 0-in. thick necting intake structure with diesel generator building and auxiliary building   
 
Walls Electrical cable tunnels con- (D + L + E) 0.75 M  = 32 ft-K M m  = 34 ft-K WSD 1-ft 6-in. thick necting intake structure with diesel generator building and auxiliary building   
 
Roof slab Electrical cable tunnels con- (D + L + E) 0.75 M  = 36 ft-K M m  = 60 ft-K WSD 2-ft 0-in. thick necting intake structure with diesel generator building and auxiliary building   
 
D = Dead load M = Maximum moment L = Live load WSD = Working stress design E = Earthquake load
 
REV 21  5/08 CONTAINMENT TYPICAL SECTIONS AND DETAILS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-1
 
REV 21  5/08 CONTAINMENT PLANS & SECTION JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-2
 
REV 21  5/08 CONTAINMENT DETAILS OF EQUIPMENT HATCH JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-3
 
REV 21  5/08 CONTAINMENT DETAILS OF PERSONNEL LOCK JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-4
 
REV 21  5/08 SHEATHING AND TRUMPET DETAIL JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-5
 
REV 21  5/08 BASE DETAILS FOR STEAM GENERATOR AND REACTOR COOLANT PUMP FOUNDATIONS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-6
 
REV 21  5/08 BASE DETAIL FOR SECONDARY SHIELD WALL JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-7
 
REV 21  5/08 SECONDARY SHIELD WALLS BELOW EL. 129'-0" JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-9
 
REV 21  5/08 SECONDARY SHIELD WALL EL. 129'-0" TO 166'-6" JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-10
 
REV 21  5/08 PRIMARY SHIELD WALL JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-11
 
REV 21  5/08 DETAIL FOR BASE SLAB TO CYLINDER LINER JUNCTURE JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-12
 
REV 21  5/08 TYPICAL PLANS CONTAINMENT JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-13 (SHEET 1 OF 2)
 
REV 21  5/08 TYPICAL PLANS CONTAINMENT JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-13 (SHEET 2 OF 2)
 
REV 21  5/08 TYPICAL SECTIONS CONTAINMENT JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-14 (SHEET 1 OF 2)
 
REV 21  5/08 TYPICAL SECTIONS CONTAINMENT JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-14 (SHEET 2 OF 2)
 
REV 21  5/08 THERMAL GRADIENT ACROSS CONTAINMENT WALL JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-15
 
REV 21  5/08 FINITE ELEMENT MESH BOTTOM HALF CONTAINMENT FOR AXISYMMETRIC LOADS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-16
 
REV 21  5/08 FINITE ELEMENT MESH TOP HALF CONTAINMENT FOR AXISYMMETRIC LOADS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-17
 
REV 21  5/08 MODEL OF CONTAINMENT FOR FINITE ELEMENT ANALYSIS NON-AXISYMMETRIC LOADS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-18
 
REV 21  5/08 CONTAINMENT BASE SLAB FINITE ELEMENT MESH NON-AXISYMMETRIC LOADS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-19
 
REV 21  5/08 AUXILIARY BUILDING CONTROL ROOM & SPENT FUEL POOL PLANS AT EL. 155'-0" JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-23 (SHEET 1 OF 2)
 
REV 21  5/08 AUXILIARY BUILDING CONTROL ROOM & SPENT FUEL POOL PLANS AT EL. 155'-0" JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-23 (SHEET 2 OF 2)
 
REV 21  5/08 AUXILIARY BUILDING SECTION A-A JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-24
 
REV 21  5/08 AUXILIARY BUILDING SECTION B-B JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-25
 
REV 21  5/08 DIESEL GENERATOR BUILDING PLAN AND SECTION JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-26 (SHEET 1 OF 2)
 
REV 21  5/08 DIESEL GENERATOR BUILDING PLAN AND SECTION JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-26 (SHEET 2 OF 2)
 
REV 21  5/08 RIVER INTAKE STRUCTURE PLAN AND SECTION JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-27 (SHEET 1 OF 2)
 
REV 21  5/08 RIVER INTAKE STRUCTURE PLAN AND SECTION JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-27 (SHEET 2 OF 2)
 
REV 21  5/08 INTAKE STRUCTURE AT STORAGE POND PLAN AND SECTION JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-28 (SHEET 1 OF 2)
 
REV 21  5/08 INTAKE STRUCTURE AT STORAGE POND PLAN AND SECTION JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-28 (SHEET 2 OF 2)
 
REV 21  5/08 POND SPILLWAY STRUCTURE PLAN AND SECTIONS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-29 (SHEET 1 OF 4)
 
REV 21  5/08 POND SPILLWAY STRUCTURE PLAN AND SECTIONS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-29 (SHEET 2 OF 4)
 
REV 21  5/08 POND SPILLWAY STRUCTURE PLAN AND SECTIONS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-29 (SHEET 3 OF 4)
 
REV 21  5/08 POND SPILLWAY STRUCTURE PLAN AND SECTIONS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-29 (SHEET 4 OF 4)
 
REV 21  5/08 EQUIPMENT HATCH BOUNDARY LINES FOR THE SAP ANALYSIS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-30
 
REV 21  5/08 SAP FINITE ELEMENT MESH FOR THE EQUIPMENT HATCH JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-31
 
REV 21  5/08 SAP ANALYSIS OF EQUIPMENT HATCH JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-32
 
REV 21  5/08 SAP ANALYSIS OF EQUIPMENT HATCH JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-33
 
REV 21  5/08 SAP ANALYSIS OF EQUIPMENT HATCH JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-34
 
REV 21  5/08 SAP ANALYSIS OF EQUIPMENT HATCH JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-35
 
REV 21  5/08 LOCATION PLAN - FOUNDATIONS FOR CATEGORY 1 STRUCTURES JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-36
 
REV 21  5/08 CONTAINMENT BASE SLAB DETAILS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-37
 
REV 21  5/08 AUXILIARY BUILDING BASE SLAB DETAILS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-38
 
REV 21  5/08 FOUNDATION DETAILS FOR DIESEL GENERATOR BUILDING, RIVER INTAKE STRUCTURE, AND INTAKE STRUCTURE AT STORAGE POND JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-39
 
REV 21  5/08 GEOMETRY OF PERSONNEL LOCK AND AUXILIARY ACCESS LOCK JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-40
 
REV 21  5/08 AUXILIARY BUILDING CASK WASH AND CASK STORAGE AREAS PLAN AND SECTION JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-41
 
REV 21  5/08 AUXILIARY BUILDLING CASK WASH AND CASK STORAGE AREAS SECTION JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.8-42
 
FNP-FSAR-3
 
3.9-1 REV 21  5/08 3.9 MECHANICAL SYSTEMS AND COMPONENTS 3.9.1 DYNAMIC SYSTEM ANALYSIS AND TESTING 
 
3.9.1.1  Vibration Operational Test Program Piping vibration and thermal expansion tests were performed during the startup program to conform to Regulatory Guide 1.68 and as outlined in section 14.1. Criteria for the test satisfy the requirements of the applicable portions of ASME Section III Code for Class 1 and 2 components. 
 
Farley Nuclear Plant systems included in this program are: reactor coolant system (RCS), power conversion system (PCS), emergency core coolant systems (ECCS), and chemical and volume control systems (CVCS). These system tests are integrally performed during the hot functional and power ascension test programs. These programs allow system operation at normal operating temperature and including flow modes, temperature plateaus, valve operations, pump starts and stops. 
 
Criteria for these tests, including the basic monitoring locations and the type of monitoring, were coordinated with design groups and the test results were evaluated by the design groups for acceptability. If the acceptance criteria established by the design groups were not satisfied during these tests, then corrective measures were taken to achieve an acceptable system response. Further retests were performed as required to verify the acceptability of design following modifications. 
 
3.9.1.2  Dynamic Testing Procedures A description of the analyses or tests used in the design of safety related mechanical equipment such as pumps and heat exchangers to withstand seismic loadings is given in subsections 3.7.2 and 3.7.3.
Most of this mechanical equipment is isolated from the effects of the faulted plant condition and, therefore, will see negligible accident loadings. For equipment which is not isolated from the effects of the faulted plant condition, the dynamic accident loads are evaluated. 
 
The tubes in the steam generator are subject to a possible flow induced vibration that does not exist in the primary coolant loop. This vibration could result from flow across the tubes due to vortex shedding. To ensure that no sympathetic vibration is generated by the vortex shedding, there is a wide frequency separation between the vortex frequency of the fluid and the beam frequency of the tube. Parallel flow vibration is analyzed using the correlations of Burgreen, and the amplitude of vibration is shown to be low enough that neither stress, banging, nor fatigue is a problem. 
 
FNP-FSAR-3
 
3.9-2 REV 21  5/08 3.9.1.3  Dynamic System Analysis Methods for Reactor Internals The reactor internals are modeled dynamically for loads produced by a pipe rupture of the largest branch lines attached to the main reactor coolant loop; the design basis accident (DBA),
for both cold and hotleg breaks; response due to a safe shutdown earthquake (SSE); and for the most unfavorable combination of LOCA and SSE. Seismic analysis of the reactor vessel and its internals is described in subsections 3.7.2 and 3.7.3. 
 
The upper internals support structure is made of two plates much like a sandwich. The upper support assembly is a plate reinforced by a weldment of a skirt and a grid of beams. The upper core plate is connected to the upper support assembly by hollow columns bolted to the plates. The guide tubes are pinned to the upper core plate and bolted to the upper support assembly. This structure compresses the fuel assembly springs during assembly and is subjected to vertical upward forces from these springs. During operation, normal and abnormal transverse flow drag forces are applied to the columns and guide tubes, and differential pressure exists across the horizontal plates. The forces on the columns and guide tubes vary with the distance from the outlet nozzles. Because of the complexity of the upper package geometry and loading conditions, the modeling of the reactor internals was performed by the structure and matrix displacement for each finite element. This finite element structural analysis computer program permits static elastic, non linear dynamic and plastic analysis. Descriptions of the techniques used to model the various parts of the internals are given in the following paragraphs. 
 
The top structure, deep beam, and upper core plate have been modeled with flat shell elements, the support columns with "three dimensional" beam elements and the fuel assemblies and hold down springs with "three dimensional" spring elements. Because of symmetry, a one-eighth slice of the upper package has been modeled. The core plate is perforated and is modeled as a geometrically equivalent solid plate which has modified elastic constants according to the theory of perforated plates. 
 
Columns of two different lengths are modeled, the long columns between the upper support and upper core plates and the short columns between the beam grid and the upper core plate. 
 
Under the loads used for design, according to the operating condition under study, the previously described computer program provides stresses and deflections at all nodal points. 
 
There is no change in the configuration of the reactor internals core support structures from the 15 x 15 fuel assembly configuration due to the incorporation of the 17 x 17 fuel assembly. The mechanical properties of the 17 x 17 fuel assembly, such as fuel assembly weight and beam stiffness, are similar to the 15 x 15 fuel assembly. Their input to the reactor internals core support structures is similar and the response of the total reactor internals core support structural model is also essentially similar. 
 
3.9.1.3.1 Preoperational Tests The program used to establish the integrity of reactor internals has involved extensive design analysis, model testing, and post hot functional inspection. Additionally, a full size reactor has been instrumented (2) to measure the dynamic behavior of a Farley size plant and has compared measurements with predicted values.
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3.9-3 REV 21  5/08 This program was instituted as part of a basic philosophy of instrumenting the internals of the "first-of-a-kind" of the current nuclear steam supply system designs for power plants. The magnitude of this test program was much greater than the intent of the philosophy, and was established as part of an extensive plan to develop theories and basic concepts related to internals vibration under various operating conditions.
 
Thus, not only is added assurance obtained that all of the hardware will operate in the manner for which it was designed, but these data also assist in the development of increased capability for the prediction of the dynamic behavior of pressurized water reactor (PWR) internals. The previous "first-of-a-kind" plants that were instrumented are R. E. Ginna (two loops), H. B.
Robinson No. 2 (three loops) and Indian Point Unit II (four loops).
The H. B. Robinson No. 2 reactor has been established as the prototype for the Westinghouse three-loop plant internals verification program. Subsequent three-loop plants are similar in design. Past experience with other reactors indicates that plants of similar designs behave in a similar manner. For these reasons an instrumentation program was conducted on the H. B. Robinson No. 2 to confirm the behavior of the reactor internal components. The main objectives of this test were to increase confidence in the adequacy of the internals by determining stress or deflection levels at key locations. 
 
In the final analysis, the proof that the internals are adequate, free from harmful vibrations, and have performed as intended is through component observations and examinations during service. With this thought, Robinson, the 3-loop prototype, was subjected to a thorough visual and dye penetrant examination by a qualified Westinghouse quality assurance engineer before and after the hot functional test. This inspection was in addition to the normal inspection of the internals in the shop, and before and after shipment. A visual inspection of the internals was also conducted during the Robinson unit's first refueling in March of 1973. This inspection was performed with the aid of television cameras and borescopes 
 
Also, for the particular case of the three loop plants, the following operating experiences offer additional assurance of the adequacy of this design: 
 
A. Southern California Edison's San Onofre plant is a three-loop plant with a slightly different design. This plant has been in operation since 1967 with no internals vibration problems. The internals have been inspected on
 
various occasions. 
 
B. H. B. Robinson No. 2, after completion of the hot functional inspection, has been at power operation since 1970 with no internals vibration problems. 
 
C. Florida Power and Light's Turkey Point No. 3 and No. 4 have successfully completed the post hot functional inspection, with the results indicating no internals vibration problems.
D. Virginia Electric and Power Company's Surry No. 1 and No. 2 have also successfully completed the post hot functional inspection with similar
 
results. 
 
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3.9-4 REV 21  5/08 The only significant differences between the Farley plant internals and the Robinson Plant internals are the replacement of the annular thermal shield with neutron shield panels, and the substitution of 17 x 17 fuel assemblies for 15 x 15 assemblies. In addition, the Farley Unit 1 upper internals were modified after hot functional testing to add additional instrumentation to measure temperatures in the reactor vessel head plenum as described in subsection 4.4.5.4.
The design of the special instrumentation stanchion has been reviewed analytically by Westinghouse, using very conservative assumptions for both flow loading and seismic loading under normal, upset, and faulted conditions.  (See table 3.9-4.)  With the conservative assumptions made, all stresses were found to be within ASME code allowable values as shown in table 3.9-5. No excitation is expected due to vortex shedding since the ratio of natural frequency to the shielding frequency is greater than 3.0.
In addition, the inclusion of the special stanchions and associated hardware increases the weight of the upper internals by approximately 0.5 percent. The design is such that it does not change the structural stiffness of the upper internals nor does it change the normal and upset forcing functions imposed on the internals. Consequently, a negligible effect on the internals vibratory response will be realized, and therefore, no additional preoperational testing is required. 
 
The replacement of the thermal shield with segmented neutron shield panels results in a reduction of the flow induced vibrations of the reactor core structures. This conclusion was confirmed in tests with a 1/24 th scale model.
(111)  The flow test was first conducted on a model with a thermal shield and indicated that the vibration levels of the internals were low and levels on the neutron shield panel were negligible. Appendix B of reference 1 presents the test results. Reference 12 justifies in more detail the comparison of the relative effects of replacing the annular thermal shield with neutron shielding pads.
 
The change to 17 x 17 fuel assemblies results in the use of newly designed guide tubes which are stronger and more rigid than the 15 x 15 guide tubes and hence will be less susceptible to flow induced vibration problems. The remainder of the core structure design has not been changed, and consequently remains identical to the prototype, which has been tested and proven to be well within design expectations and limits. 
 
The Portland General Electric Company's Trojan plant internals were instrumented for strain measurements on the core barrel, and on the 17 x 17 guide tube subject to highest cross flow.
The Trojan plant is the lead plant featuring neutron panels and 17 x 17 style internals. The data obtained in this program provides verification of Westinghouse analysis and scale model predictions of 17 x 17 and neutron panel behavior in a full size plant and is applicable to Farley Nuclear Plant. 
 
The Three Loop Internals Assurance Program conducted on H. B. Robinson No. 2, supplemented by the Trojan data on neutron panels and 17 x 17, jointly satisfies the intent of
 
Regulatory Guide 1.20. 
 
The core support structures will receive, in addition to the testing discussed above, the normal pre- and post-hot functional testing examination for integrity per paragraph D, "Regulations for Reactor Internals Similar to the Prototype Design," of Regulatory Guide 1.20. This examination will include the points in figure 3.9-1, summarized as follows: 
 
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3.9-5 REV 21  5/08  A. All major load bearing elements of the reactor internals relied upon to retain the core structure in place 
 
B. The lateral, vertical, and torsional restraints provided within the vessel 
 
C. Those locking and bolting devices whose failure could adversely affect the structural integrity of the internals 
 
D. Those other locations on the reactor internals components that are similar to those which were examined on the prototype H. B. Robinson No. 2
 
design.
The inside of the vessel was inspected before and after the hot functional test, with all the internals removed, to verify that no loose parts or foreign material were in evidence. 
: 1. Lower Internals A particularly close inspection was made on the following items or areas, using a 5X or 10X magnifying glass where applicable. The locations of these areas are shown in figure 3.9-1. 
: a. Upper barrel to flange girth weld. 
: b. Upper barrel to lower barrel girth weld. 
: c. Upper core plate aligning pin. Examined bearing surfaces for any shadow marks, burnishing, buffing, or scoring. Inspected welds for integrity. 
: d. Irradiation specimen guide screw locking devices and dowel pins. Checked for lockweld integrity. 
: e. Baffle assembly locking devices. Checked for lockweld integrity. 
: f. Lower barrel to core support girth weld. 
: g. Neutron shield panel screw locking devices and dowel pin cover plate welds. Examined the interface surfaces for evidence of tightness and for lockweld integrity. 
: h. Radial support key welds. 
: i. Insert screw locking devices. Examined soundness of lockwelds. 
: j. Core support columns and instrumentation guide tubes. Checked all the joints for tightness and soundness of the
 
locking devices.
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3.9-6 REV 21  5/08    k. Secondary core support assembling welds. 
: l. Lower radial support keys and inserts. 
    (Examined for any shadow marks, burnishing, buffing, or scoring. Checked the integrity of the lockwelds.)  These members supply the radial and torsion constraint of the internals at the bottom relative to the reactor vessel while permitting axial growth between the two. One would expect to see, on the bearing surfaces of the key and keyway, burnishing, buffing, or shadowing marks that would indicate pressure loading and relative motion between the two parts. Some scoring of engaging surfaces is also possible and  acceptable. 
: m. Gaps at baffle joints.  (Checked for gaps between baffle and top former, and at baffle to baffle joints.) 
: 2. Upper Internals
 
A particularly close inspection was made on the following items or areas, using a magnifying glass of 5X or 10X magnification, where necessary.
The locations of these areas are shown in figure 3.9-1. 
: a. Thermocouple conduits, clamps, and couplings. 
: b. Guide tube, support column, and thermocouple column assembly locking devices. 
: c. Support column and conduit assembly clamp welds. 
: d. Upper core plate alignment inserts. Examined for any shadow marks, burnishing, buffing, or scoring. Checked the locking devices for integrity of lockwelds. 
: e. Connections of the support columns mixing devices and orifice plates to the upper core plate. Checked for tightness and lock device integrity. 
: f. Thermocouple conduit gusset and clamp welds. 
: g. Thermocouple end plugs.  (Checked for tightness.) 
: h. Guide tube closure welds, tube transition plate welds and card welds. 
 
Acceptance standards are the same as required in the shop by the original design drawings and specifications.
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3.9-7 REV 21  5/08 During the hot functional test, the internals were subjected to a total operating time at greater than normal full flow conditions (three pumps operating) of at least 240 hours. This provides a cyclic loading of approximately 10 7 cycles on the main structural elements of the internals. In addition there was some operating time with only one and two pumps operating. 
 
When no signs of abnormal wear, no harmful vibrations are detected, or no apparent structural changes take place, the three-loop core support structures are considered to be structurally adequate and sound for operations. 
 
3.9.1.4  Correlation of Test and Analytical Results The dynamic behavior of reactor components has been studied, using experimental data obtained from operating reactors, along with results of model test and static and dynamic tests in the fabricators shops and at plant site. Extensive instrumentation programs to measure vibration on reactor internals (including protot ype units of various reactors) have been carried out during preoperational flow tests, and reactor operation. 
 
From scale model tests, information on stresses, displacements, flow distribution and fluctuating differential pressures is obtained. Studies have been performed (2) to verify the validity and to determine the prediction accuracy of models for determining reactor internals vibration due to
 
flow excitation. Similarity laws need to be satisfied to ensure that the model response can be correlated to the real prototype behavior. 
 
Vibration of structural parts during preoperational tests is measured using displacement gauges, accelerometers, and strain transducers. The signals are recorded with magnetic tape recorders. Onsite offsite signal analysis is done using hybrid real time and digital techniques to determine the approximate frequency and phase content. In some structural components the spectral content of the signals include nearly discrete frequency or very narrow band, usually due to excitation by the main coolant pumps and other components that reflect the response of the structure at a natural frequency to broad bands, mechanically or flow induced excitation.
Damping factors are also obtained from wave analyses. 
 
In general, the study follows two parallel procedures. Frequencies and spring constants are obtained analytically, and these values are confirmed from the results of the tests.
Damping coefficients are established experimentally, and forcing functions are estimated from pressure fluctuations measured during operation and in models. Once these factors are established, the response can be computed analytically. In parallel, the responses of important reactor structures are measured during preoperational reactor tests and the frequencies and mode shapes of the structures are obtained. Once all the dynamic parameters are obtained, as explained above, the forcing functions can be estimated. These two procedures are not independent; both are performed simultaneously and, when combined, they provide indications of the internals behavior during reactor operation. Internals behavior during reactor operation also is measured using mechanical devices and nuclear noise methods. The last method involves the frequency spectral analysis of signals from out-of-core ion chambers. Information is obtained on the frequency, amplitude, and damping of their vertical and lateral vibrations of the core, because relative motions of the core cause reactivity perturbations and fluctuations in the neutron flux signal level. 
 
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3.9-8 REV 21  5/08 Some components, such as control rod guide tubes, fuel rods, and incore instrumentation tubes, are subjected to cross flow and parallel flow with respect to the axis of the structure. In these cases there are numerous theoretical and experimental studies directed toward establishing the response of the structure.
(2) These studies also provide information on the added apparent mass of the water, which has the effect of decreasing the natural frequency of the component. For both cases, cross and parallel, the response is obtained after the forcing function and the damping of the system is determined. 
 
Cross flow may excite the structure with periodic vortex shedding, which gives rise to a lateral oscillatory lift force perpendicular to the flow direction and a drag force in the flow direction. The dimensionless vortex shedding frequency, or Strouhal number S = fD/V, is a function of the Reynolds number and known for different cross sections. The structure is usually designed in such a manner that its natural frequency in water is considerably higher than the vortex shedding frequency so as to avoid coincidence. The lateral force per unit length is given by 
 
F(x,t) = C L [1/2 P f V(x)2]D cos t  where C L is the oscillatory lift coefficient including correlation length effects (C L depends on the Reynolds number); P f is fluid density; V is cross flow velocity; D is the characteristic diameter, and  is the vortex shedding circular frequency. Data obtained from preoperational and shop tests are used to confirm the coefficients used. 
 
The preoperational vibration monitoring test on H. B. Robinson No. 2, the three-loop prototype plant, has been completed. The pre- and postoperational flow test examination of the internals bas been completed indicating that all the components performed as predicted. No evidence of damage or incipient failure has been found. 
 
The testing programs consisted of measurements of the stresses, deflections, and responses of select key points in the internals structures during hot functional and low power physics tests.
The main purpose of this testing program was to ensure that no unexpected large amplitudes of vibration existed in the internals structure during operation. The tests were intended to provide data and results on indications of overall core support structure performance and to verify particular stress and deflection quantities. 
 
3.9.1.5  Analysis Methods Under LOCA Loadings The scope of the different dynamic analysis techniques and methods used to evaluate mechanical systems and components of the Westinghouse pressurized water reactor for loads produced by a double ended pipe rupture of the largest branch lines attached to the main coolant loop (LOCA) is very extensive.
A. Reactor Internals Analysis
 
Analysis of the reactor internals for blowdown loads resulting from a loss-of-coolant accident is based on the time history response of the internals to simultaneously applied blowdown forcing functions. The forcing functions are defined at points in the system where changes in cross section or direction of flow occur so that differential loads are generated during blowdown FNP-FSAR-3
 
3.9-9 REV 21  5/08 transient. The dynamic analysis can employ the displacement method, lumped parameters, and stiffness matrix formulations and assumes that all components behave in a non-linear manner, due to the presence of gaps at certain interfaces, such as gaps at the reactor vessel to core barrel flange, reactor vessel to upper support flange, lower radial keys, upper core plate alignment pins and core barrel outlet nozzles. 
 
In addition, because of the complexity of the system and the components, it is necessary to use finite element stress analysis codes to provide more detailed information at various points. 
 
A comprehensive explanation of all the techniques and analytical methods used cannot be included in the scope of the FSAR. The more important and relevant methods are presented as an overview in paragraph 3.9.1.3 and summarized in the following. 
 
B. Blowdown Forces Due to Cold and Hot Leg Break
 
A digital computer program called MULTIFLEX (9), which is developed for the purpose of calculating local fluid pressure, flow, and density transients that occur in pressurized water reactor coolant systems during a loss-of-coolant accident, is applied to the subcooled, transition, and saturated two-phase blowdown regimes. This is in contrast to programs such as WHAM (3) which are applicable only to the subcooled region and which, due to their method of solution, could not be extended into the region in which large changes in the sonic velocities and fluid densities take place. This MULTIFLEX (9) code is based on the method of characteristics wherein the resulting set of ordinary differential equations, obtained from the laws of conservation of mass, momentum, and energy, are solved numerically using a fixed mesh in both space and time. 
 
Although spatially one dimensional conservation laws are employed, the code can be applied to describe three dimensional system geometries by use of the equivalent piping networks. Such piping networks may contain any number of pipes or channels of various diameters, dead ends, branches (with up to six pipes connected to each branch), contractions, expansions, orifices, pumps, and free surfaces (such as in the pressurizer). System losses such as friction, contraction, expansion, etc., are considered. 
 
The MULTIFLEX (9) code evaluates the pressure and velocity transients for a maximum of 2000 locations throughout the system. These pressure and velocity transients are stored as a permanent tape file and are made available to the programs LATFORC and FORCE2, which utilize detailed geometric descriptions in evaluating the horizontal and vertical loadings on the reactor internals.
Each reactor component for which FORCE2 calculations are required is designated as an element and assigned an element number. Vertical forces acting upon each of the elements are calculated summing the effects of: 
: 1. The pressure differential across the element. 
: 2. Flow stagnation on, and unrecovered orifice losses across the element. 
: 3. Friction losses along the element. 
 
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3.9-10 REV 21  5/08 Input to the code, in addition to the MULTIFLEX (9) pressure and velocity transients, includes the effective area of each element on which the force acts due to the pressure differential across the element, a coefficient to account for flow stagnation and unrecovered orifice losses, and the total area of the element along which the shear forces act. 
 
In addition to the vertical forces calculated by FORCE2, the horizontal forces on the vessel, core barrel, and thermal shield are calculated by LATFORC. The horizontal forces are calculated by summing the lateral force components around the vessel, core barrel, and thermal shield, based on the pressure differential across each section, multiplied by the area of each section. This is done at ten different elevations. The total lateral force is calculated by summing the forces over
 
the ten elevations.
The mechanical analysis has been performed using conservative assumptions in order to obtain results with extra margin. Some of the most significant are: 
: 1. When applying the hydraulic forces, no credit is taken for the stiffening effect of the fluid environment which will reduce the deflections and stresses in the structure. 
: 2. The multi-mass model for the Reactor Pressure Vessel (RPV) system described below is considered to have a sufficient number of degrees of freedom to represent the most important modes of vibration of the system.
The RPV system finite element model for the nonlinear time history dynamic analysis consists of three concentric structural sub-models connected by nonlinear impact elements and linear stiffness matrices. The first sub-model represents the reactor vessel shell and its associated components. The reactor vessel is restrained by six reactor vessel supports (situated beneath each nozzle) and by the attached primary coolant piping.
 
The second sub-model represents the reactor core barrel assembly, lower support plate, tie plates, and the secondary support components. These sub-models are physically located inside the first, and are connected to them by stiffness matrices at the vessel/internals interfaces. Core barrel to reactor vessel shell impact is represented by nonlinear elements at the core barrel flange, upper support plate flange, core barrel outlet nozzles, and the lower radial restraints.
The third and innermost sub-model represents the upper support plate assembly consisting of guide tubes, upper support columns, upper and lower core plates, and the fuel. The fuel assembly simplified structural model incorporated into the RPV system model preserves the dynamic characteristics of the entire core. For each type of fuel design, the corresponding simplified fuel assembly model is incorporated into the system model. The third sub-model is connected to the first and second by stiffness matrices and nonlinear elements.
 
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3.9-11 REV 21  5/08 The appropriate dynamic differential equations for finite element system model describing the aforementioned phenomena are formulated and the results are obtained using a general purpose finite element computer code which computes the response at each mode point of the RPV system model. The system model is excited by a set of time dependent horizontal and vertical forces generated by the LATFORC and FORCE2 programs.
The results from the computer program provide time history nodal displacements and nonlinear impact forces at various locations of the reactor vessel and reactor internals interfaces. The methodology used in the RPV system LOCA/seismic analyses is the NRC approved methodology (Reference 23).
C. Reactor Coolant Loop (RCL) Analysis
 
A flow diagram representing the procedure for the complex time history dynamic solution is shown in figure 3.9-2. The procedure for dynamic solution is iterative in nature since the definition of support stiffness matrices for dynamic behavior (to be incorporated in the reactor coolant loop (RCL) model) depends upon the response of the support points which is not known
 
a priori.
The initial displacement configuration of the mass points is defined by applying the initial steady state hydraulic forces to the unbroken RCL model. For this calculation, the support stiffness matrices for the static behavior are incorporated into the RCL model. For dynamic solution, the unbroken RCL model is modified to simulate the physical severance of the pipe due to the postulated LOCA under consideration. The static support cases (i.e., steam generator columns and reactor coolant pump columns) are included in the dynamic model as stiffness matrices.
Other supports such as tie rods, bumper blocks, and hydraulic snubbers, which go directly to ground, are represented in FIXFM by nonlinear elements which correctly define the restraint of the physical element. For supports which cannot be represented by nonlinear elements, the stiffness matrix for dynamic behavior is selected on the basis of anticipated displacement response at the support points. 
 
The natural frequencies and normal modes for the modified RCL dynamic model are determined. The time history hydraulic forces at appropriate node points are combined to determine the forces and moments at structural lumped mass points of interest. After proper coordinate transformation to the RCL global coordinate system, the hydraulic forcing functions are stored on magnetic tape for later use as input to the FIXFM program. 
 
The initial displacement conditions, natural frequencies, normal modes, and the time history hydraulic forcing functions form the input to the FIXFM program which calculates the dynamic time history displacement response for the dynamic degrees of freedom in the RCL model. The displacement response at support points is reviewed to validate the use of support stiffness matrices for dynamic behavior. If the calculated support point response does not match with the anticipated response, the dynamic solution is revised using a new set of support stiffness matrices for dynamic behavior. This procedure is repeated until a valid dynamic solution is obtained. 
 
The time history displacement response from the valid solution is stored on magnetic tape for later use to compute the support loads and to analyze the RCL piping stresses.
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3.9-12 REV 21  5/08 The support loads, {F}, are computed by multiplying the support stiffness matrix, [K], and the displacement vector, {}, at the support point. The support loads are stored on magnetic tape for use in the support member evaluation. The time history displacement response from the FIXFM program is used as input to the WESDYN-2 program. The program treats this input as an imposed deflection condition on the RCL model and computes the time history of internal forces, deflections, and stresses at each end of the members of the RCL piping system.
The results of this solution are stored on magnetic tape for later use in piping stress evaluation. 
 
3.9.1.6  Analytical Methods for ASME Code Class 1 Components No plastic instability allowable limits given in ASME Section III are used when dynamic analysis is performed. The limit analysis methods have the limits established by ASME Section III for normal, upset, and emergency conditions. For these cases, the limits are sufficiently low to assure that the elastic system analysis is not invalidated. For ASME Code Class I components, the stress limits for faulted loading conditions are specified in section 5.2. For ASME components other than Class 1 and components not covered by the ASME code, the stress limits for faulted loading conditions are specified in subsections 3.9.2 and 3.9.3, respectively.
These faulted condition limits are established in such a manner that there is equivalence with the adopted elastic limits and consequently will not invalidate the elastic system analysis.
Particular cases of concern are checked by readjusting the elastic system analysis.
3.9.2 ASME CODE CLASS 2 AND 3 COMPONENTS The design loading combinations and design stress limits for ASME Code Class 2 and 3 components are given below. 
 
3.9.2.1  Plant Conditions and Design Loading Combinations ASME Code Class 2 and 3 components were not designed to specific plant conditions. However, the design loading combinations used for the design are given in subsection 3.9.2.2, below. 
 
3.9.2.2  Design Loading Combination Table 3.9-1 presents the various loading combinations for ASME Class 2 and 3 components. 
 
3.9.2.3  Design Stress Limits The design stress limits for the various design loading combinations are given in table 3.9-1. 
 
Where no design stress limits were defined at the time of purchase, the design limits were specified by the vendors for the various ASME Class 2 and 3 components. These limits were based on having no gross deformation of the components that would render the components incapable of performing their intended safety functions.
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3.9-13 REV 21  5/08 For the ASME Class 2 and 3 piping, where no design stress limits were defined at the time of design, the stress allowables were based on having no gross deformation of the piping. The analysis also demonstrated that the piping would not transmit loads to the components connected to the piping which would exceed vendor allowable loads. 
 
3.9.2.4  Analytical and Empirical Methods for Design of Pumps and Valves The design methods for pumps and valves are described in table 3.9-1. 
 
The methods used to assure operability are provided in subsection 3.9.4.
3.9.2.5  Design and Installation Criteria, Pressure-Relieving Devices All overpressure relief valves and their connected piping (i.e. headers, header connections, and discharge piping) are designed to withstand the following conditions without exceeding the applicable code's primary stress allowable. The maximum loads due to valve discharge thrust internal pressure, deadweight, and earthquake are applied simultaneously. When more than one relief valve is attached to a piping system, the loads due to all relief valves discharging
 
simultaneously are applied to the system along with the above mentioned primary loads. In addition, the loads from the most critical combination of valves discharging are applied. The local stresses in the main steam line outside the containment at the connection of the relief valves were computed as specified in "Welding Research Council Bulletin", No. 107, and held below the allowable stress level S h  defined in Section NC-3611.1-(b.4) of Section III, 1971 Edition, and modified according to Section NC-3612.3. 
 
A static analysis was initially used in the analysis of safety and safety relief valves and a dynamic analysis performed to verify the adequacy of the design. 
 
3.9.2.6  Stress Levels for Category I Components
 
Methods used to analyze Category I systems are discussed in subsections 3.9.1 and 3.9.2. Stress analysis results are documented in the applicable piping system stress calculations.
 
3.9.2.7  Field Run Piping System Piping classified under ASME Section III, Classes 2 and 3, and analysis, was routed on the piping design drawings, but dimensioned in the field. Detail isometrics were prepared for those pipes dimensioned in the field and forwarded to the project engineer for review and analysis of seismic stress, thermal stress, shielding, and thermal insulation requirements as needed. The approved isometrics were then released for permanent installation. 
 
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3.9-14 REV 21  5/08 3.9.2.8  Class 2 and 3 Component Supports The stress limits used for ASME Class 2 and 3 component supports are identical to those used for the supported component. These allowed stresses are such that the design requirements for the components and the system structural integrity are maintained. 
 
3.9.3 COMPONENTS NOT COVERED BY ASME CODE Core and Internals Integrity Analysis (Mechanical Analysis)
The response of the reactor core and vessel internals under excitation produced by a simultaneous complete severance of a reactor coolant pipe and seismic excitation for a typical Westinghouse pressurized water reactor plant internals has been determined. The following mechanical functional performance criteria apply: 
 
A. Following the design basis accident the basic operational or functional criterion to be met for the reactor internals is that the plant shall be shutdown and cooled in an orderly fashion so that fuel cladding temperature is kept within specified limits. This criterion implies that the deformation of certain critical reactor internals must be kept sufficiently small to allow core cooling.
B. For large breaks, the reduction in water density greatly reduces the reactivity of the core, thereby shutting down the core whether the rods are tripped or not. The subsequent refilling of the core by the emergency core cooling system uses borated water to maintain the core in a subcritical state. Therefore, the main requirement is to ensure effectiveness of the emergency core cooling system. Insertion of the control rods, although not needed, gives further assurance of ability to shut the plant down and keep it in a safe shutdown condition. 
 
C. The functional requirements for the core structures during the design basis accident are shown in table 3.9-2. The inward upper barrel deflections are controlled to ensure no contacting of the nearest rod cluster control guide tube. The outward upper barrel deflections are controlled in order to maintain an adequate annulus for the coolant between the vessel inner diameter and core barrel outer diameter.
D. The rod cluster control guide tube deflections are limited to ensure operability of the control rods. 
 
E. To ensure no column loading of rod cluster control guide tubes, the upper core plate deflection is limited to the value shown in table 3.9-2.
F. The reactor has mechanical provisions that are sufficient to maintain the design core and internals and to ensure that the core is intact with acceptable heat transfer geometry following transients arising from the design basis accident operating conditions.(4)(8)
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3.9-15 REV 21  5/08  G. The core internals are designed to withstand mechanical loads arising from 1/2 SSE, SSE, and pipe ruptures.(4)(5)(6)(8)
 
3.9.3.1  Faulted Conditions
 
The following events are considered in the faulted conditions category: 
 
A. Loads produced by a double ended pipe rupture of the largest branch lines attached to the main coolant loop design basis accident, for both cases: cold and hot leg break. Branch line breaks, rather than main loop piping breaks are analyzed in accordance with the leak-before-break exemptions to GDC-4 discussed in FSAR chapter 3.6 references(3)( 4)(5). The methods of analysis adopted are related to the type of accident assumed (cold leg break or hot leg break).
B. Response due to an SSE. 
 
C. Most unfavorable combination of a safe shutdown earthquake and a design basis accident. Maximum stresses obtained in each case are conservatively added using the square root of the sum of the squares
 
method.
Maximum stress intensities are compared to allowable stresses for each of the above conditions. Elastic analysis on each component is performed on an elastic basis. For faulted conditions, stresses may be above yield in a few locations. For these cases only, when deformation requirements exist, a plastic analysis is independently performed to ensure that functional requirements are maintained (guide tubes deflections and core barrel expansion). The elastic limit allowable stresses are used to compare with the result of the analysis.   
 
The above described analyses show that the stresses and deflections that would result following a faulted condition are less than those that would adversely affect the integrity of the structures.
Also, the natural and applied frequencies are such that resonance problems should not occur.
3.9.3.2  Structural Response of Reactor Vessel Internals During LOCA and Seismic Conditions
 
3.9.3.2.1 Structural Model and Methods of Analysis The response of reactor vessel internals due to an excitation produced by a complete severance of auxiliary loop piping is analyzed. With the acceptance of Leak-Before-Break (LBB) by USNRC, References(3)(4)(5) of Chapter 3.6, the dynamic effects of main coolant loop piping no longer have to be considered in the design basis analysis. Only the dynamic effects of the next most limiting breaks of auxiliary lines need to be considered, and consequently, the components will experience considerably less loads than those from the main loop line breaks.
 
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3.9-16 REV 21  5/08 Assuming that such a pipe break in cold leg occurs in a very short period of time (1 millisecond), the rapid drop of pressure at the break produces a disturbance that propagates through the reactor vessel nozzle into the down-comer (vessel and barrel annulus) and excites the reactor vessel and the reactor internals. The characteristics of the hydraulic excitation combined with those of the affected structures present a unique dynamic problem. Because of the inherent gaps that exist at various interfaces of the reactor vessel/reactor internals/fuel, the problem becomes that of nonlinear dynamic analysis of the RPV system. Therefore, nonlinear dynamic analyses (LOCA and seismic) of the RPV system include the development of LOCA and seismic forcing functions which are also discussed here.
 
3.9.3.2.2 Structural Model The RPV system finite element model for the nonlinear time history dynamic analysis consists of three concentric structural sub-models connected by nonlinear impact elements and linear stiffness matrices. The first sub-model represents the Farley reactor vessel shell and its associated components. The reactor vessel is restrained by six reactor vessel supports (situated beneath each nozzle) and by the attached primary coolant piping.
 
The second sub-model represents the Farley reactor core barrel assembly, lower support plate, tie plates, and the secondary support components.
These sub-models are physically located inside the first, and are connected to them by stiffness matrices at the vessel/internals interfaces. Core barrel to reactor vessel shell impact is represented by nonlinear elements at the core barrel flange, upper support plate flange, core barrel outlet nozzles, and the lower
 
radial restraints.
 
The third and innermost sub-model represents the Farley upper support plate assembly consisting of guide tubes, upper support columns, upper and lower core plates, and the fuel. The fuel assembly simplified structural model incorporated into the RPV system model preserves the dynamic characteristics of the entire core. For each type of fuel design, the corresponding simplified fuel assembly model is incorporated into the system model. The third sub-model is connected to the first and second by stiffness matrices and nonlinear elements.
 
3.9.3.2.3 Analysis Technique The Westinghouse Electric Computer ANalysis (WECAN) Computer Code, Reference (18), which is used to determine the response of the reactor vessel and its internals, is a general purpose finite element code. In the finite element approach, the structure is divided into a finite number of discrete members or elements. The inertia and stiffness matrices, as well as the force array, are first calculated for each element in the local coordinates. Employing appropriate transformations, the element global matrices and arrays are assembled into global structural matrices and arrays and used for dynamic solution of the differential equation of motion for the
 
structure.
The WECAN Code solves equation of motions using the nonlinear modal superposition theory.
Initial computer runs such as dead weight analyses and the vibration (modal) analyses are made to set the initial vertical interface gaps and to calculate eigenvalues and eigenvectors.
The modal analysis information is stored on magnetic tapes and is used in subsequent FNP-FSAR-3
 
3.9-17 REV 21  5/08 computer runs which solve equations of motions. The first time step performs the static solution of equations to determine the steady state solution under normal operating hydraulic forces.
After the initial time step, WECAN calculates the dynamic solution of equations of motions and nodal displacements, and the impact forces are stored on tape for post-processing.
 
The fluid-solid interactions in the LOCA analysis are accounted through the hydraulic forcing functions generated by MULTIFLEX Code, Reference (9). Following a postulated LOCA pipe rupture, forces are imposed on the reactor vessel and its internals. These forces result from the release of the pressurized primary system coolant. The release of pressurized coolant results in traveling depressurization waves in the primary system. These depressurization waves are characterized by a wave front with low pressure on one side and high pressure on the other.
Depressurization waves propagate from the postulated break location into the reactor vessel through either a hot leg or a cold leg nozzle. After a postulated cold leg break, the depressurization path for waves entering the reactor vessel is through the nozzle that contains the broken pipe and into the region between the core barrel and the reactor vessel (i.e., down-comer region). The initial wave propagates up, around, and down the down-comer annulus, then up through the region circumferentially enclosed by the core barrel, that is, the fuel region.
In the case of a cold leg break, the region of the down-comer annulus close to the break depressurizes rapidly, but because of the restricted flow areas and finite wave speed (approximately 3000 feet per second), the opposite side of the core barrel remains at a high pressure. This results in a net horizontal force on the core barrel and the reactor vessel. As the depressurization wave propagates around the down-comer annulus and up through the core, the core barrel differential pressure reduces, and similarly, the resulting hydraulic forces drop.
 
In the case of a postulated break in the hot leg, the wave follows a similar depressurization path, passing through the outlet nozzle and directly into the upper internals region, depressurizing the core, and entering the down-comer annulus from the bottom exit of the core barrel. Thus, after an RPV outlet nozzle break, the down-comer annulus would be depressurized with very little difference in pressure forces across the outside diameter of the core barrel. A hot leg break produces less horizontal force because the depressurization wave travels directly to the inside of the core barrel (so that the down-comer annulus is not directly involved), and internal differential pressures are not as large as for a cold leg break of the same size. Since the differential pressure is less for a hot leg break, the horizontal force applied to the core barrel is less for hot leg break than for a cold leg break. For breaks in both the hot leg and cold leg, the depressurization waves continue to propagate by reflection and translation through the reactor
 
vessel and loops.
 
The MULTIFLEX (9) computer code calculates the hydraulic transients within the entire primary coolant system. It considers subcooled, transition, and early two-phase (saturated) blowdown regimes. The MULTIFLEX code employs the method of characteristics to solve the conservation laws, and it assumes one-dimensionality of flow and homogeneity of the liquid-vapor mixture. As mentioned earlier, the MULTIFLEX code considers a coupled fluid-structure interaction by accounting for the deflection of constraining boundaries, which are represented by a separate spring-mass oscillator system. A beam model of the core support barrel has been developed from the structural properties of the core barrel. In this model, the cylindrical barrel is vertically divided into equally spaced segments, and the pressure as well as the wall motions are projected onto the plane parallel to the broken nozzle. Horizontally, the barrel is divided into 10 segments; each segment consists of three separate walls. The spatial pressure variation at FNP-FSAR-3
 
3.9-18 REV 21  5/08 each time step is transformed into 10 horizontal forces which act on the 10 mass points of the beam model. Each flexible wall is bounded on either side by a hydraulic flow path. The motion of the flexible wall is determined by solving the global equations of motions for the masses representing the forced vibration of an undamped beam.
 
In order to obtain the response of reactor pressure vessel system (vessel/internals/fuel), the LOCA horizontal and vertical forces obtained from the LATFORC and FORCE2 Codes are applied to the finite element system model. The transient response of the reactor internals consists of time history nodal displacements and time history impact forces.
 
3.9.3.2.4 Seismic Analysis The basic mathematical model for seismic analysis is essentially the same as the LOCA model except for some minor differences. In the LOCA model, the fluid-structure interactions are accounted through the MULTIFLEX Code; whereas, in the seismic model, the fluid-structure interactions are included through the hydrodynamic mass matrices in the down-comer region. Another modeling difference is the difference between loop stiffness matrices. The seismic model uses the unbroken loop stiffness matrix, whereas, the LOCA model uses the broken loop stiffness matrix. Except for these two differences , the RPV system seismic model is identical to that of LOCA model.
 
The horizontal fluid-structure or hydroelastic interaction is significant in the cylindrical fluid flow region between the core barrel and the reactor vessel annulus. Mass matrices with off-diagonal terms (horizontal degrees-of-freedom only) attach between nodes on the core barrel, thermal shield and the reactor vessel. The mass matrices for the hydroelastic interactions of two concentric cylinders are developed using the work of reference (19). The diagonal terms of the mass matrix are similar to the lumping of water mass to the vessel shell, thermal shield, and core barrel. The off-diagonal terms reflect the fact that all the water mass does not participate when there is no relative motion of the vessel and core barrel. It should be pointed out that the hydrodynamic mass matrix has no artificial virtual mass effect and is derived in a straight forward, quantitative manner.
 
The matrices are a function of the properties of two cylinders with the fluid in the cylindrical annulus, specifically, inside and outside radius of the annulus, density of the fluid and length of the cylinders. Vertical segmentation of the reactor vessel and the core barrel allows inclusion of radii variations along their heights and approximates the effects of beam mode deformation.
These mass matrices were inserted between the selected nodes on the core barrel, thermal shield, and the reactor vessel. The seismic evaluations are performed by including the effects of simultaneous application of time history accelerations in three orthogonal directions. The WECAN computer code is also used to obtain the response for the RPV system under seismic
 
excitations.
 
3.9.3.3  Results and Acceptance Criteria The reactor internals behave as a highly nonlinear system during horizontal and vertical oscillations of the LOCA forces. The nonlinearities are due to the coulomb friction at the sliding surfaces and due to gaps between components causing discontinuities in force transmission.
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3.9-19 REV 21  5/08 The frequency response is consequently a function not only of the exciting frequencies in the system but also of the amplitide. Different break conditions excite different frequencies in the system. This situation can be seen clearly w hen the response under LOCA forces is compared with the seismic response. Under seismic excitations, the system response is not as nonlinear as LOCA response because various gaps do not close during the seismic excitations.
The results of the nonlinear LOCA and seismic dynamic analysis include the transient displacements and impact loads for various elements of the mathematical model. These displacements and impact loads and the linear component loads (forces and moments) are then used for detailed component evaluations to assess the structural adequacy of the reactor vessel, reactor internals, and the fuel.
 
A. Structural Adequacy of Reactor Internals Components 
 
The Farley reactor internal components are not ASME Code components. This is due to the fact that subsection NG of the ASME Boiler and Pressure Code edition applicable to Farley reactor internals did not include design criteria for the reactor internals since its design preceded subsection NG of the ASME Code. However, these components were originally designed to meet the intent of the 1971 Edition of Section III of the ASME Boiler and Pressure Vessel Code with addenda through the Winter, 1971. As mentioned earlier, with the acceptance of Leak-
 
Before-Break by USNRC, Reference (3)(4)(5) of Chapter 3.6, the dynamic effects of the main reactor coolant loop piping no longer have to be considered in the design basis analysis. Only the dynamic effects of the next most limiting breaks of the auxiliary lines (accumulator line and pressurizer surge or RHR lines) are considered. Consequently, the components experience considerably less loads and deformations than those from the main loop breaks which were considered in the original design of the reactor internals.
 
B. Allowable Deflection and Stability Criteria The criteria for acceptability with regard to mechanical integrity analyses are that adequate core cooling and core shutdown must be ensured. This implies that the deformation of reactor internals must be sufficiently small so that the geometry remains substantially intact.
Consequently, the limitations established on the reactor internals are concerned principally with the maximum allowable deflections and stability of the components. For faulted conditions, deflections of critical reactor internal components are limited to the values given in table 3.9-2. In a hypothesized vertical displacement of internals, energy absorbing devices limit the displacement to 1.25 inches by contacting the vessel bottom head.
 
Core Barrel Response Under Transverse Excitations - In general, there are two possible modes of dynamic response of the core barrel during LOCA conditions: a) during a cold leg break, the inside pressure of the core barrel is much higher than the outside pressures, this subjecting the core barrel to outward deflections, and b) during hot leg break, the pressure outside the core barrel is greater than the inside pressure thereby subjecting the core barrel to compressive loading. Therefore, this condition requires the dynamic stability check of the core barrel during hot leg break.
  (1) To ensure shutdown and cooldown of the core during cold leg blowdown, the basic requirement is a limitation on the outward deflection of the barrel at the locations of the inlet nozzles connected to unbroken lines. A large outward deflection of the upper barrel in front of the inlet nozzles, FNP-FSAR-3
 
3.9-20 REV 21  5/08 accompanied with permanent strains, could close the inlet area and restrict the cooling water coming from the accumulators. Consequently, a permanent barrel deflection in front of the unbroken inlet nozzles larger than a certain limit, called "no loss of function" limit, could impair the efficiency of the ECCS.
(2) During the hot leg break, the rarefaction wave enters through the outlet nozzle into the upper internals region and thus depressurizes the core and then enters the down-comer annulus from the bottom exit of the core barrel.
This depressurization of the annulus region subjects the core barrel to external pressures, and this condition requires a stability check of the core barrel during hot leg break. Therefore, to ensure rod insertion and to avoid disturbing the control rod cluster guide structure, the barrel should not interfere with the guide tubes.
Table 3.9-2 summarizes the allowable and no loss of function displacement limits of the core barrel for both the cold leg and hot leg breaks postulated in the main line loop piping. With the acceptance of LBB, the reactor internal components such as core barrel will experience much less loads and deformations than those obtained from main loop piping.
 
Control Rod Cluster Guide Tubes - The deflection limits for the guide tubes (to be consistent with conditions under which the ability to trip has been tested), and for fuel assembly thimbles cross-section distortion (to avoid interference between the control rods and the guides) are given in table 3.9-2.
 
Upper Package - The local vertical deformation of the upper core plate, where a guide tube is located, shall be below 0.100 inch. This deformation will cause the plate to contact the guide tube since the clearance between the plate and the guide tube is 0.100 inch. This limit will prevent the guide tubes from undergoing compression. For a plate local deformation of 0.150 inch, the guide tube will be compressed and deformed transversely to the upper limit previously established. Consequently, the value of 0.150 inch is adopted as the "no loss of function" local
 
deformation limit with an allowable limit of 0.100 inch. The deformation limits are given in table 3.9-2. 3.9.3.4  Method of Analysis The internals structures are analyzed for loads corresponding to normal, upset, emergency, and faulted conditions. The analysis performed depends on the mode of operation under consideration. 
 
The scope of the stress analysis problem is large, requiring many different techniques and methods, both static and dynamic. The more important and relevant methods are presented in subsections 3.9.1 and 3.9.3.2. 
 
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3.9-21 REV 21  5/08 3.9.3.5  Evaluation of Reactor Internals for Accumulator Line Cold Leg and Pressurizer Surge Line Hot Leg Breaks This section contains an evaluation of the effects of a 90.75 in 2 accumulator line cold leg safe end break and a 103.87 in 2 pressurizer surge line hot leg safe end break on the reactor internals. Both breaks are assumed to have a break opening time of 1 millisecond.
The main operational requirement to be met is that the plant be shutdown and cooled down in an orderly fashion so that the fuel cladding temperature is kept within the specified limits. This implies that the deformation of the reactor internals must be kept sufficiently small to allow core cooling and assure effectiveness of the emergency core cooling system. A detailed description of LOCA methodology and the acceptance criteria for the components is given in subsections 3.9.3.2 and 3.9.3.3. Use of LBB methodology was approved for the pressurizer surge line (Reference 25).
 
3.9.3.6  Baffle-Former Bolt Replacement Analysis In order to satisfy the concern of possible degradation in the reactor vessel baffle-former bolts due to long term irradiation, a selected number of bolts have been replaced in Units 1 and 2. To justify replacing a selected number of these bolts, an analysis was performed by Westinghouse to
 
determine the acceptability of a bolt replacement pattern that would require a limited number of bolts to be replaced, while maintaining the functionality of the reactor vessel during normal and accident conditions.
 
This new reactor vessel analysis used for the baffle-former bolt replacement utilized sophisticated tools such as the computer application code MULTIFLEX Version 3.0. This code utilizes a detailed network to represent the vessel downcomer, and allow for vessel motion and for non-linear boundary conditions at the vessel and downcomer junctions at the radial keys and upper core barrel flange.
 
While MULTIFLEX Version 3.0 was used for the single-phase blowdown portion of the transient (first five hundred milliseconds), W COBRA/TRAC was used for the two-phase portion. Loads from the two-phase portion were derived and compared to the loads obtained from the single-phase portion of the transient. Loads from the single-phase portion were determined to be the limiting faulted event conditions for this analysis.
 
The ANSYS computer program was used to develop a finite element model of the baffle-former region, which was used in LOCA, seismic, thermal growth, and flow induced vibration analyses. The modeled baffle plates and former plates were si mulated by elastic plate elements, the bolts were simulated by pipe elements, and baffle-former, barrel-former, and baffle-fuel nozzle interfaces were simulated by gap elements.
 
A. Criteria These analysis programs were used to calculate loads on the fuel assemblies during LOCA and seismic conditions (singularly and combined), and then evaluated to the following acceptance
 
criteria:
 
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3.9-22 REV 21  5/08 1. Fuel rods must remain intact such that fuel pellets are not allowed to escape where they could achieve a configuration in which a core coolable geometry cannot be demonstrated.
: 2. Control rod guide tubes must not be deformed to the point at which control rod insertion cannot be demonstrated where it is credited in accident analysis consequences.
: 3. Fuel grid loads must be below allowable grid crush strength limits.
 
In addition to the above criteria, core bypass flow, fuel rod stability (momentum flux), high cycle fatigue, low cycle fatigue, and structure stress limits were factors that were also examined in the analysis.
 
B. Break Opening Time For the project of replacing baffle-former bolts, the NRC-approved Westinghouse methodology to invoke the leak-before-break (LBB) concept, which allowed for a break opening time greater than 1 millisecond, was utilized. This methodology was based on the results of break opening experiments, calculations of break openings by Westinghouse and others, engineering practices of domestic and foreign nuclear suppliers, conservatism inherent within the computer analysis software, and regulatory considerations.
C. Acceptability
 
The final configuration of the baffle-former bolts replaced within the reactor was analyzed as being acceptable for the following reasons:
* The normalized fuel grid impact loads were found acceptable for both peripheral and interior fuel assemblies.
* There is adequate stress margin in maintaining the guide thimble structural integrity.
* Control rod insertability was maintained.
* Fuel rod integrity was maintained, and fuel rod fragmentation will not occur.
* Low-cycle fatigue for the design lifetime of the replaced baffle-former bolts was less severe than the fatigue of the original bolts.
* Momentum flux margin of safety was found to be acceptable with the installed baffle-former bolt pattern.
* The design core bypass flow can be maintained.
 
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3.9-23 REV 21  5/08
* High-cycle fatigue stresses were found to be below the bolt material endurance limits.
* Alternating stresses due to flow induced vibration were determined to be below bolt material allowables.
 
3.9.3.7  Heating, Ventilation, and Air-Conditioning (HVAC) Equipment
 
Table 3.9-3 presents a list of safety related heating, ventilation, and air-conditioning (HVAC) equipment, and the applicable standards and codes to which they are designed. This table also presents report numbers containing test results for the equipment. 
 
3.9.4 OPERABILITY ASSURANCE Equipment for the Farley Nuclear Plant was designed to comply with the intent of Regulatory Guide 1.48; i.e., it was designed/analyzed to ensure structural integrity and operability.
However, the load combinations and stress limits that were used reflect NRC requirements that were in effect when the construction permit for this plant was issued and when the components were purchased and subsequently designed. Furthermore, the codes and procedures which were available when the components were purchased are based on conservative design requirements rather than detailed stress analysis and actual testing. These codes and procedures have been used by the nuclear industry for the design of components that are installed in plants that are presently operating. 
 
3.9.4.1  ASME Code Class Valves A tabulation of all active valves in the reactor pressure boundary whose operation is relied upon either to assure safe plant shutdown or to mitigate the consequences of a transient or accident is provided in table 5.2-8. 
 
The requirements of the (draft) ASME Code for Pumps and Valves were adhered to in the design of Active Code Class 1 valves. For faulted conditions, stress intensities in the valves and extended structures were limited to 1.0 S m for general membrane and 1.5 S m for general membrane plus bending. These limits ensure that the valve stresses will remain within elastic limits and that no plastic deformation will occur. 
 
The requirements of Section III of the ASME Boiler and Pressure Vessel Code were adhered to in the design of Code Class 1 manually operated gl obe valves and check valves, 2 in. in size or less. Class 2 and 3 active valves were designed to the requirements of ANSI B16.5 Code. In addition, an analysis of the extended structure was performed with loads of 3.0 g in the horizontal and vertical directions, simultaneously for valves specified by Bechtel and Southern Company Services Specifications. For this analysis, stresses were limited to values that restrict FNP-FSAR-3
 
3.9-24 REV 21  5/08 the maximum stress in the extended structure. Deflections of the extended structure will thus be small and operability of the valves will not be impaired. 
 
Prior to installation, the valves are subjected to shell hydrostatic tests, seat leakage tests, and functional tests. After installation, the valves undergo cold hydrostatic tests, hot functional tests to verify operation, and those under the Far ley ISI program undergo periodic inservice inspection and operation to ensure the continued ability of the valves to operate. 
 
3.9.4.2  ASME Code Class Pumps
 
Active pumps were designed in accordance with the ASME Code for Pumps and Valves or the ASME Boiler and Pressure Vessel Code for Nuclear Power Plants, depending on which code was in effect at the time the purchase order was issued. The stress levels in the pumps did not exceed those allowed by the applicable code. Forces resulting from seismic accelerations in the horizontal and vertical directions are included in the analysis of the pumps and their supports. The supports were designed to have natural frequencies in excess of 20 Hz. 
 
The pumps are subjected to a series of tests prior to installation and after installation in the plant. In-shop tests include hydrostatic tests to 150 percent of the design pressure, seal leakage tests, net positive suction head (NPSH) tests to qualify the pumps for the minimum available NPSH, and functional performance tests. For the NPSH and functional performance tests, the pumps are placed in a test loop and subjected to operating conditions. After installation, the pumps undergo cold hydrostatic tests, hot functional tests to verify operation, and periodic inservice inspection and operation. 
 
The above design procedures and qualification tests are, therefore, adequate to ensure the structural integrity and operability of the pumps and valves for this plant. 
 
3.9.4.3  Qualification of Vital Appurtenances The following typical appurtenances that were identified to be vital to the operation of active pumps and valves were qualified for operation during a seismic event by dynamic testing procedures as described below. 
 
Seismic Qualification Test of National ACME Company Snap-Lock Electric Switch No. D 2400X-2 A seismic qualification test program of National ACME Snap-Lock electric switch No. D 2400X-2 was conducted by Fisher Controls Co. and reported in document No. 1529 dated 11/2/72.
Testing was conducted with the switch assembly fastened to a metal plate which in turn was attached to a shaker table. All tests were conducted with the switch in an operating condition.
The following is a summary of the test procedure and results: 
 
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3.9-25 REV 21  5/08 Test Procedure
 
A. Conduct a continuous frequency sweep for each of the three axes, from 5 to 60 Hz at an acceleration level of 1.0g in no less than 31 seconds.
B. If the resonant frequency is less than 33 Hz, conduct a 4g 1-min dwell at the resonant frequency and at 10 and 33 Hz. 
 
C. If the resonant frequency is greater than 33 Hz, conduct a 4g 1-min dwell at 10, 17, 25 and 33 Hz and at the resonant frequency if it is less than 60 Hz. 
 
Test Results
 
The Snap-Lock electric switch performed satisfactorily with no malfunctions noted and meets or exceeds the specifications outlined in the test procedure. 
 
Seismic Qualification of Valve Motor Operators
 
A seismic qualification test program of valve motor operators, manufactured by Limitorque
 
Corporation, was conducted by Lockheed Electronics Company, Inc., Environmental Laboratory and reported in Report No. 2785-3-4785 dated 2/6/73; Report No. 2786-4786, Issue 2, dated 9/5/72; Report No. 2773C-4773, dated 5/3/72; and Report No. 2785-4-4785, dated 2/1/73. The test specimens were electrically monitored and operated during testing. The test procedure consisted of subjecting the specimen to the vibration test referenced in Limitorque Co. Purchase Order No. 600374, dated 6/2/72, and is summarized as follows for each of the three orthogonal axes: 
: a. Two exploratory scans were performed over the frequency range of 5 to 60 Hz at the amplitudes specified in Table 1.
TABLE 1  Freq. Range Vib. Amplitude (Hz)      (In., Double Amplitude) 5 - 33  .020 +/- .004    34 - 50  .006 + .000 - .002 51 - 60  .004 + .000 - .002
: b. Two 1-min dwells were performed at the resonant frequency at a nominal vibration input of 3 to 5.8 g's. The first minute of vibration was followed by one minute of rest. 
 
Test results for the SMB-0/H3BC, SMB-0-25/H3BC, SMB-000-2/HOBC and SMB-3/H5BC valve operators indicated that the vibration test was completed with no visible evidence of any FNP-FSAR-3
 
3.9-26 REV 21  5/08 external damage or performance degradation. There were no resonances detected during the vibration test except for a resonance at 44 Hz in the Y axis, 46 Hz in the Z axis, and 39 Hz in the X axis for valve operator No. SMB-3/H5BC. 
 
A review of the test data indicates that the valve motor operator performed satisfactorily when subjected to the dynamic environment.
Seismic Qualification of Solenoid Valves
 
A seismic qualification test program for the solenoid valves used on Westinghouse supplied air operated valves has been completed. The components tested were ASCO valve models 8300C58RV, 8300B64RU, and 831654. The test dynamic input forces, frequency limits etc., are discussed in reference 13. 
 
Instrumentation and Control Panel (Series 7300, Westinghouse) - Balance of Plant The test involved subjecting a 7300 Series nuclear power plant control system to seismic conditions for qualification and evaluation of performance. The seismic test was run in three parts for horizontal (front to back and side to side) and vertical conditions. These parts consisted of: Part 1 - Low Present Seismic, Part 2 - High Present Seismic, and Part 3 - High
 
Future Seismic. 
 
The control system tested included at least one of each type of printed circuit card used in all the various protection and safeguard actuation channels. 
 
The equipment performed satisfactorily with no malfunctions noted and meets or exceeds the specifications outlined in the detailed procedures.
Class 2 and 3 Air-Operated Control Valves
 
The valve with diaphragm actuator was analyzed in accordance with the customer's specifications, following acceptable analytical methods and allowable stress limits as set forth in the appropriate design standards and codes.
The stress developed by gravity loads, operating loads, applicable temperature and pressure, combined with simultaneously applied horizontal and vertical seismic loads, shall not cause loss of function of this valve. 
 
The analysis demonstrated that the design adequately satisfies the requirements of all the specifications. 
 
Valve Motor Operators
 
Valve motor operators for applicant specified valves are seismically tested at g-loadings higher than those specified in the valve design specification. These tests demonstrate that the operators experience no physical damage as a result of the postulated seismic event, and that the activating mechanisms undergo no change in position during the test and remain operable after the test. 
 
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3.9-27 REV 21  5/08 REFERENCES 
: 1. Kraus, S., "Neutron Shielding Pads," WCAP-7870, May 1972. 
: 2. Kuenzel, A. J., "Westinghouse PWR Internals Vibration Summary 3-Loop Internals Assurance," WCAP-7765, September 1971. 
: 3. Fabic, S., "Computer Program WHAM for Calculation of Pressure, Velocity, and Force Transients in Liquid Filled Piping Networks," Kaiser Engineers Report No. 67-49-R , November 1967. 
: 4. Bohn, G. J., "Indian Point Unit No. 2 Internals Mechanical Analysis for Blowdown Excitation," WCAP-7822 , December 1971. 
: 5. Olsen, B. E., et al, "Indian Point No. 2 Primary Loop Vibration Test Program," WCAP-7920 , September 1972. 
: 6. Moore, J. S., "Westinghouse PWR Core Behavior Following a Loss-of-Coolant Accident," WCAP-7422, August 1971. 
: 7. Deleted
: 8. Gesinski, L. T., Fuel Assembly Safety Analysis for Combined Seismic and Loss-of-Coolant Accident , WCAP-7950, July 1972. 
: 9. Takeuchi, K., "MULTIFLEX, A FORTRAN-IV Computer Program for Analyzing Thermal-Hydraulic-Structure System Dynamics," WCAP-8708-PA , WCAP-8709-A , (Non-Proprietary), September, 1977.
: 10. Deleted 
: 11. Supplemental Information Supplied on WCAP-7870 by letter from Westinghouse, NES, R. Salvatori, to AEC, D. Vassallo, NS-RS-145 (February 25, 1974). 
: 12. Lee, H., Prediction of the Flow-Induced Vibration Reactor Internals by Scale Model Tests, WCAP-8303 and WCAP-8317, May 1974. 
: 13. Plant, E. K., ASCO Seismic Test Report, Report No. 103, Job No. 30633, August 27, 1975.   
: 14. WCAP-15030, Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distribution Under Faulted Load Conditions, Westinghouse Non-Proprietary Class 3, 1998 (WCAP-15029 is the Westinghouse Proprietary Class 2 version of this document).
: 15. Letter from Thomas H. Essig (Nuclear Regulatory Commission), to Lou Liberatori (Westinghouse Owners Group), regarding Safety Evaluation Related to Topical Report WCAP-14748/14749, "Justification for Increasing Break Opening Times in Westinghouse PWRs," (TAC No. M98031), dated October 1, 1998.
FNP-FSAR-3
 
3.9-28 REV 21  5/08 16. Westinghouse Interoffice Letter NSD-E-MSI-98-340, regarding "Farley Unit #1 Final Baffle Bolt Replacement Pattern Reconciliation," dated December 22, 1998.
: 17. Letter from Thomas H. Essig (Nuclear Regulatory Commission), to Lou Liberatori (Westinghouse Owners Group), regarding Safety Evaluation of Topical Report WCAP-15029, "Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions Under Faulted Load Conditions," (TAC No. MA1152), dated November 10, 1998.
: 18. "Benchmark Problem Solutions Employed for Verification of the WECAN Computer Program," WCAP-8929, June 1977.
: 19. Fritz, R. J., "The Effects of Liquids on the Dynamic Motions of Immersed Solids," Trans. ASME, Journal of Engineering for Industry, 1972, pp. 167-173.
: 20. WCAP-5890, Rev. 1, "Ultimate Strength Criteria to Ensure No Loss of Function of Piping and Vessel Under Seismic Loading," October 1967.
: 21. WCAP-15102 Volume 2, "Electicite de France 1300 Mwe Plants Reactor Internals Functional Criteria," December 1997.
: 22. WCAP-7332-L-AR, "Topical Report - Indian Point Unit 2 Reactor Internals Mechanical Analysis for Blowdown Excitations," November 1973.
: 23. WCAP-9401-P-A "Verification Testing and Analysis of 17x17 Optimized Fuel Assemblies - Approved Version," August 1981.
: 24. "Safety Evaluation of Elimination of Dynamic Effects of Postulated Primary Loop Pipe Ruptures from Design Basis for Joseph M. Farley Nuclear Plant, Units 1 and 2 (TAC
 
Nos. 79660 and 79661)," August 12, 1991.
: 25. "Safety Evaluation of Elimination of Dynamic Effects of Postulated Pipe Ruptures in the Pressurizer Surge Line from Structural Design Basis for Joseph M. Farley Nuclear Plant, Units 1 and 2 (TAC Nos. 80367 and 80368), " January 15, 1992.
 
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.9-1 DESIGN CRITERIA FOR ASME CLASS 2 AND 3 COMPONENTS Vessel/
Loads Tanks  (Note 1) Pumps (Note 1) Valves Piping Pressure + Deadweight ASME III/ASME VIII ASME III/Performance ASME III/ANSI B 16.5 ASME III + Thermal (nozzle  Testing in accordance with loads only)  standards of the Hydraulic Institute Procedures 
 
Pressure + Deadweight ASME III/ASME VIII ASME III/Performance ASME III/ANSI B 16.5 ASME III + Thermal (nozzle loads  Testing in accordance with only) +  standards of the Hydraulic Transients (Note 2)  Institute Procedures 
 
Pressure + Deadweight ASME III/ASME VIII Structural Functional Structural Functional (Note 4) + SSE + Dynamic (Note 3) Assured by Rigid (f n>20) Assured by Rigid (f n>20)  Effects (where  integrity of within working integrity of within working  applicable) (Note 6)  connecting  conditions by connecting  conditions by    piping dynamic piping dynamic analysis  analysis 
 
  (Note 3) (Note 5) 
: 1. Allowable nozzle loads are contained in the equipment specifications or specified by the vendor for pumps and tanks. The pi ping is designed so that the loads generated on the components nozzles are no greater than the allowable loads specified for that component. The allowables and s tress calculations for the components are reviewed by the designer.
: 2. The transients considered in the piping analyses are: (1) a relief valve-closed system (transient). (2) a fast valve closure.
(3) a relief valve-open system (sustained) +1/2 SSE.
 
  (The loads in the piping generated by t he individual transients are considered in the design of the components, where applica ble). 
: 3. The design limits for tanks are specified by the vendors. These design limits are based on having no gross deformation of t he components.
: 4. The design limits for piping are based on having no gross deformation of the piping.
: 5. ASME Class 2 and 3 valves are designed such that the section m odules of the valves is greater than that of the pipe connecte d to the valve. For valves that do not meet the selection modules criteria, a stress analysis  will be performed to verify adequacy.
: 6. The loads in the piping generated by the dynamic e ffects are considered in t he design of the components.
 
FNP-FAR-3 REV 21  5/08 TABLE 3.9-2 MAXIMUM DEFLECTIONS SPECIFIED FOR REACTOR INTERNAL SUPPORT STRUCTURES No Loss-of Allowable (1) Function Component Limit(in.) Limit(in.)
 
Upper Barrel, Expansion/Compression (to ensure sufficient inlet flow area/and to prevent the barrel from touching any guide tube to avoid disturbing the rod cluster control 
 
guide structure) 2,3      Radial Inward 4.1 8.2
 
Radial Outward 1.0 1.0 Upper Package, Axial Deflection (to 0.1 0.15 maintain the control rod guide   
 
structure geometry) 2,3 Rod Cluster Control Guide Tube  1.0 1.75 Deflection As a Beam (to be consistent with conditions under which ability 
 
to trip has been tested) 3 Fuel Assembly Thimbles Cross-Section 0.036 0.072 Distortion (to avoid interference between the control rods and the   
 
guides)3   
 
Notes: 
: 1. The allowable limit deflection values giv en above correspond to stress levels for internals structure well below the limiting criteria giv en by the collapse curves in WCAP-5890 (Reference 20). Consequently, for the internals, the geomet ric limitations established to ensure safe shutdown capability are more restrictive than t hose given by the failure stress criteria.
: 2. See Reference 21.
: 3. See Reference 22.
 
FNP-FSAR-3 REV 21  5/08 TABLE 3.9-3 (SHEET 1 OF 2)
DESIGN CRITERIA FOR COMPONENTS NOT COVERED BY ASME CODE Systems - Components Controlling Standards and/or Codes Test Report Number (type)
 
Auxiliary Building Ventilation System ES air cooling units: 
 
Heat exchangers ARI Standard 410-64 CVI Seismic Analysis Report (Dynamic/Seismic Analysis)
 
Fan AMCA Test Code 300-67, 211A-67 CVI Seismic Analysis Report (Dynamic/Seismic Analysis)
Penetration Room Filtration System Fans AMCA Test Code 300-67, 211A-67 AAF Report on Seismic Analysis PEP-497 (Dynamic/Seismic Analysis)
 
Filters (HEPA and AACC CS-IT (HEPA), AAF Report on Seismic Analysis charcoal) ORNL-NSIC-65 (charcoal) PEP-497 (Dynamic/Seismic Analysis)
Ductwork SMACNA High Velocity Duct Stress Analysis Construction, 2nd Edition, 1969.
Containment Cooling System Fans AMCA Test Code 300-67, 211A-67 AAF Report on Seismic Analysis, PEP-495 (Seismic-Dynamic Analysis)
 
Heat Exchangers ARI Standard 410-64 AAF Report on Seismic Analysis, PEP-495 (Seismic-Dynamic Analysis)
 
Fusible linked plate UL Standard of Safety AAF Report on Seismic Analysis, UL555-1970, UL33-1968 PEP-495 (Seismic - Dynamic Analysis)
 
Control Room Ventilation System Fans AMCA Test Code 300-67, 211A-67 Joy Certification or Dynamic Analysis Filters (HEPA and AACA, CS-IT (HEPA), AAF Report on Seismic Analysis charcoal) ORNL, NSIC - 65 (charcoal) PEP-497 (Dynamic - Seismic Analysis)
FNP-FSAR-3 REV 21  5/08 TABLE 3.9-3 (SHEET 2 OF 2)
Systems - Components Controlling Standards and/or Test Report Number (type)
Codes Air conditioning ANSI B9-1971, Safety Code for AAF Report on Seismic Analysis    unit Mechanical Refrigeration PEP 648 (Seismic Dynamic Analysis)
 
Ductwork SMACNA Low Velocity Duct Stress Analysis  Construction, 2nd Edition,    1969 
 
FNP-FSAR-3 REV 21  5/08 TABLE 3.9-4 COMPARISON OF BEST ESTIMATE AND DESIGN VALUES OF PEAK SEISMIC ACCELERATIONS (g)
Maximum Acceleration (g)
 
Best Estimate Design Earthquake Type    Value      Value 
 
Operational basis earthquake (OBE) 1.68 3.20
 
Safe shutdown earthquake (SSE) 2.20 4.20
 
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.9-5
 
==SUMMARY==
OF STRESS AND MARGIN OF SAFETY TO CODE ALLOWABLES Maximum Allowable Margin Item Stress (lb/in.
: 2) Stress (lb/in.
: 2) of Safety
 
Stanchion 5,090.0 24,795.0 3.87
 
Stanchion bolts 44,412.0 45,000.0 0.013
 
Stanchion flange 12,397.5 24,795.0 1.0
 
Conduit tube cantilevered 1,590.0 24,795.0 14.70
 
Conduit tube attached to stanchion 1,375.0 24,795.0 17.0
 
Conduit tube support clip welds 104.0 9,918.0 Large
 
NOTE:  Margin of safety = allowable stress - 1.0    maximum stress
 
REV 21  5/08 VIBRATION CHECKOUT FUNCTIONAL TEST INSPECTION POINTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.9-1
 
REV 21  5/08 TIME-HISTORY DYMAMIC SOLUTION FOR LOCA LOADING JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.9-2
 
FNP-FSAR-3
 
3.10-1 REV 21  5/08 3.10 SEISMIC DESIGN OF CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT  Equipment and components of the reactor protec tion system and the engineered safety feature actuation system meet the seismic Category I requirements and are identified in section 3.2. 
 
3.10.1  SEISMIC DESIGN CRITERIA A. Category I Electrical Equipment
 
Category I electrical equipment has been designed to withstand, without exceeding normal
 
allowable working stresses and without loss of function, the forces resulting from the 1/2 SSE
 
caused by a horizontal ground acceleration of 0.05g and a vertical ground acceleration of
 
0.033g. The equipment is also designed to withstand, without exceeding 90 percent of the yield
 
stresses, or without loss of function, the forces resulting from the safe shutdown earthquake (SSE) caused by a horizontal ground acceleration of 0.10g and a vertical ground acceleration of
 
0.067g. 
 
The seismic response spectra, based on the synthesized time history spectra, have been
 
developed for the specific equipment location and appropriate damping factors. A detailed
 
discussion of the seismic design criteria is given in section 3.7. The electrical equipment under
 
Category I was qualified in one of the following ways: 
: 1. The natural frequencies of the equipment (as it would be installed in service) were determined in the horizontal and vertical directions based
 
on a multi-degree of freedom lumped mass system. From the
 
appropriate response spectra curve, the acceleration levels were
 
selected corresponding to the natural frequency. Forces due to this
 
acceleration level are used in the seismic analysis. 
: 2. If it was not practical to calculate the natural frequency, the maximum acceleration of the spectra curves was used for seismic analysis. 
: 3. Prototype equipment was subjected to a test demonstrating its ability to perform its intended function during and after SSE. 
 
When simulated seismic testing was not entirely practical, proof of performance was obtained by a combination of mathematical analysis
 
and simulated testing. 
 
The following items, which were part of the test reports submitted by the manufacturer, conform
 
to the requirements of IEEE 344-1971. 
: 1. Equipment identification. 
: 2. Equipment specification. 
 
FNP-FSAR-3
 
3.10-2 REV 21  5/08  3. Test facility:
location. 
 
test equipment. 
: 4. Test method. 
: 5. Test data. 
: 6. Results and conclusions (pertaining, in particular, to natural frequencies and maximum accelerations). 
: 7. Signature of manufacturer's authorized representative and date. 
 
For the analytical approach, the manufacturer was required to submit complete seismic design
 
calculations in step-by-step form. Preference was given to actual testing. The manufacturer
 
was required to furnish documentation justifying his selection of the analytical method over
 
simulated testing. These requirements are in accordance with the stipulations of IEEE 344-
 
1971. 
 
Cable tray supports are designed using the appropriate instructure response spectra. The
 
calculated stresses from dead load, live load, and earthquake loads are less than 50 percent of
 
yield stress for the 1/2 SSE and 90 percent of yield stress for the SSE. 
 
The location and performance requirements of class IE switchgear, motor control centers, and
 
distribution panels are such that post accident conditions do not impose any additional stresses
 
over and above those experienced due to a safe shutdown earthquake (SSE). The capability of
 
the equipment to withstand seismic disturbances, established under nonaccident conditions, is
 
considered adequate to meet the requirements during post accident operation. 
 
B. Category I Instrumentation and Control Equipment
 
Equipment specifications for Category I instrumentation control equipment required that
 
equipment be designed to withstand without loss of function the forces resulting from the 1/2
 
safe shutdown earthquake and the safe shutdown earthquake. The following procedures were
 
used. 
: 1. Equivalent static acceleration factors for the horizontal and vertical directions were provided to the equipment manufacturers. 
: 2. The seismic response spectra and the appropriate damping factor, based on the synthesized time history spectra, were provided for the
 
specific equipment location. 
 
The equipment vendor was given two ways by wh ich the equipment could be qualified:  by dynamic analysis and/or by testing. The manufacturer was permitted to use: 
: 1. Test reports of the particular component(s). 
 
FNP-FSAR-3
 
3.10-3 REV 21  5/08  2. Performance data of equipment, with applicable supporting data, which under specified conditions has been subjected to equal or
 
greater dynamic loads. 
: 3. Analysis. 
 
The choice of the method was based on the practicability of either method (test or analysis) for
 
the size, type, shape, and complexity of the instruments or equipment, and reliability of results. 
 
Table 3.7-4 indicates the type of procedure, qualification method (test or analysis), as well as
 
the applicable stress or information criteria. 
 
Following submission of the results of the test or analysis, the methods, procedures, and results
 
were examined for compliance with the specification requirements. Test and analytical
 
procedures, as well as submitted reports, conform to the requirements of IEEE-344-1971. 
 
Component Testing
:
Testing of components, such as relays, was performed as part of the primary equipment being
 
tested. The relays were tested in the energized state and the output contacts were monitored
 
for continuity. A change in continuity would be indicative of a malfunction. Where a
 
representative component was qualified by previous testing, the test results were reviewed and, if found acceptable, a certificate of conformance to the FNP seismic specification from the
 
vendor was considered adequate in lieu of a repeat test. 
 
C. Category I Equipment and Components 
 
Seismic qualification for Category I equipment and components supplied by the NSSS vendor
 
was originally described in WCAP-7817 , "Seismic Testing of Electrical and Control Equipment (Low Seismic Plants)" and its supplements 1, 2, 3, and 4. The AEC, in a letter dated January
 
12, 1973, indicated its acceptance of this Westinghouse report. However, in a letter dated
 
December 23, 1974, additional information was requested. 
 
In order to address these additional concerns, Westinghouse conducted a supplemental
 
qualification program and submitted the results to the NRC. The NRC conducted a seismic
 
audit of Westinghouse to evaluate the results of the supplemental program as well as on-site
 
item by item inspection of equipment at the Salem, Farley, and Sequoyah plants. 
 
For equipment to be tested after May 1974, and for equipment to be installed in plants having a
 
construction permit docketed after October 1972, Westinghouse has committed to conduct
 
seismic qualification testing in conformance with IEEE 344-1975. However, for equipment
 
tested prior to May 1974, the following conclusion was drawn by the NRC in Section IV "Conclusion and Regulatory Position of Report on Seismic Audit of Westinghouse Electrical
 
Equipment (TAR's 3678-1, 3683-1, 0706, 0921-1, 0788-2, 1111-2, and 3000-2)" dated August
 
26, 1976. 
 
"The Mechanical Engineering Branch, Division of Systems Safety has
 
completed the seismic audit of Westinghouse electrical equipment tested
 
prior to May 1974. Based on our evaluation of topical reports, inspection
 
of equipment on the plant site, numerous meeting discussions, laboratory FNP-FSAR-3
 
3.10-4 REV 21  5/08 visits, and our evaluation of confirmatory retesting for equipment in question, we conclude that adequate assurance is achieved for this
 
equipment to sustain seismic excitations to their designated SSE levels." 
 
These test and analytical procedures, as well as the submitted reports, conform to the
 
requirements of IEEE 344-1971. The seismic design criteria applicable to NSSS scope
 
equipment are addressed in the Westinghouse generic qualification program. 
 
3.10.2  SEISMIC ANALYSES, TESTING PROCEDURES AND RESTRAINT MEASURES  Table 3.7-4 indicates which of the methods given in subsection 3.10.1 have been used for
 
seismic qualification of the equipment. The procedures for seismic qualification, either by
 
analysis or by testing, are in conformance with the requirements of IEEE 344-1971. 
 
The seismic design criteria discussed in subsection 3.10.1 form a part of all specifications for
 
Category I equipment. Certification was obtained from each manufacturer to ensure that his
 
equipment will perform without loss of function in accordance with the stipulations of the
 
specifications. 
 
For seismic qualification by analysis, the certification requires that the calculations have been
 
checked by an engineer knowledgeable in the design of such equipment. 
 
The specification includes applicable floor response spectra where the equipment is located, and the values indicated in the curves are used for seismic qualification of that equipment. The
 
floor response spectra take into consideration the loading amplification of floors. 
 
Equipment hold down details, specifically size and spacing of weld or bolt, are obtained from the
 
manufacturer for designing the foundation to be compatible with the seismic withstanding
 
capability of the equipment. 
 
FNP-FSAR-3
 
3.11-1 REV 21  5/08 3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRICAL EQUIPMENT This section provides information on the environmental conditions and design bases for which
 
the mechanical, instrumentation, and electrical portions of the engineered safety features, the
 
reactor protection systems, and other safety-re lated systems are designed to ensure acceptable performance during normal and design basis accident (DBA) environmental conditions. 
 
3.11.1  EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL CONDITIONS Safety-related equipment which is required to function during and subsequent to a DBA, is
 
identified in section 3.2 of the FSAR. Active pumps and valves are discussed in section 3.9 of
 
the FSAR. 
 
The original specifications for safety-related elec trical equipment which is subject to a post DBA harsh environment and required to function during and subsequent to a DBA required
 
qualification to IEEE 323-1971. Subsequently, the Farley Nuclear Plant Environmental
 
Qualification (EQ) Program was implemented to comply with the requirements of NRC Inspection and Enforcement Bulletin (IEB) 79-01B, NUREG-0588, Revision 1, Interim Staff
 
Position on Environmental Qualification of Safety-Related Electrical Equipment, and 10 CFR
 
50.49. Based on the dates of the Farley plant operating licenses, Unit 1 was required to comply
 
with the requirements of IEB 79-01B, which provides the NRC Division of Operating Reactors (DOR) Guidelines, and Unit 2 was required to comply with the requirements of NUREG-0588, Category II. The requirements set forth under these programs supplement the requirements of
 
IEEE 323-1971. After implementation of these programs, 10 CFR 50.49 was issued and
 
mandated environmental qualification requirements for safety related electrical equipment.
Regulatory Guide 1.89, Revision 1, followed and established IEEE 323-1974 as an acceptable
 
standard to comply with the requirements of 10 CFR 50.49. The provisions of 10 CFR 50.49
 
waive the need to requalify components previously qualified under the DOR Guidelines or
 
NUREG-0588 unless the components are replaced. The replacement components must comply
 
with the provisions of 10 CFR 50.49 unless there are sound reasons to the contrary. These
 
reasons, when required, will be documented. A ccordingly, the EQ program implements the requirements of 10 CFR 50.49 as documented in the EQ master lists and the associated EQ
 
packages. The EQ packages document which version of the IEEE-323 standard was used for
 
the qualification.
 
Normal operating environmental conditions are defined as conditions existing during routine
 
plant operations. These environmental conditions, as listed in table 3.11-1, represent the
 
normal, maximum, and minimum conditions expected during routine plant operations. 
 
Accident environmental conditions are defined as those deviating significantly from the normal operating environmental conditions as a result of a DBA. These conditions are specified in
 
table 3.11-1 for the postulated accident duration of 30 days. Compatibility of equipment with the
 
specified environmental conditions is provided to fulfill the following design criteria: 
 
A. For normal operation, systems and components required to mitigate the consequences of a DBA or to provide for safe shutdown are designed to remain
 
functional after exposure to the environmental conditions listed in table 3.11-1. 
 
FNP-FSAR-3
 
3.11-2 REV 21  5/08  Where possible, all safety-related systems and components are designed to withstand the maximum expected 40-year (a) integrated radiation dose at their respective locations within the plant. If it cannot be assured that equipment is
 
designed for the 40-year (a) dose, a replacement maintenance program for that equipment is established. The replacement maintenance program ensures
 
operational integrity of the equipment throughout the life of the plant. 
 
B. In addition to the normal operation environmental requirements given in A.
above, systems and components required to mitigate the consequences of a
 
DBA or to provide for safe shutdown of the reactor are designed to remain
 
functional after exposure to the following environmental conditions. Qualification
 
time is based on the operating duration following a DBA. 
: 1. Such components inside the containment are designed for the temperature, pressure, humidity, and chemical environment inside
 
the containment after a design basis LOCA or main steam line
 
break accident (MSLB). 
: 2. Such components inside the containment which are required after a LOCA are designed for the post-LOCA radiation dose. 
: 3. Such components outside the containment which are required to mitigate the consequences of a design basis LOCA are designed for the expected
 
integrated accident radiation dose at the equipment location. 
: 4. Such components outside the containment are designed for the temperature, pressure, and humidity environmental conditions
 
resulting from a postulated high energy line break (HELB) in areas
 
where such components are located. 
 
________________
: a. The renewed operating licenses authorize an additional 20-year period of extended
 
operation for both FNP units, resulting in a plant operating life of 60 years. The EQ program is
 
credited to continue to manage aging effects associated with EQ equipment for the period of
 
extended operation (see chapter 18, subsections 18.3.1 and 18.4.4). Applicable EQ
 
evaluations based on a 40-year design life were evaluated as time-limited aging analyses (TLAAs) for license renewal and will be revised as necessary to reflect the 60-year plant
 
operating life before the units enter the period of extended operation.
FNP-FSAR-3
 
3.11-3 REV 21  5/08 3.11.2 QUALIFICATION TESTS AND ANALYSES Qualification is based on simulated environmental testing where feasible. If qualification test
 
data was inadequate, and if sufficiently reliable data and proven analytical methods were
 
available, environmental adequacy was based on analysis. 
 
Testing consists of simulation of actual physical conditions on an actual component or
 
prototype, analyses, or a combination of tests and analyses, as applicable. Qualification testing
 
is performed under conditions of temperature, pressure, humidity, chemistry, and radiation in
 
excess of the design basis conditions. The testing period is sufficient to ensure the capability to
 
function during and for the required interval after an accident (30 days). 
 
3.11.2.1 Equipment Inside Containment Equipment listed in table 3.2-1 is designed for 40 years (a) of operation in the most severe temperature, pressure, humidity, and radiati on environment which exists at the equipment location during normal operation. In some cases, a 40-year (a) life under such conditions is not within the state-of-the-art; therefore, a replacement program is established to ensure
 
continuous, reliable operation. Furthermore, the safety-related equipment listed in table 3.2-1 is
 
designed to remain functional in the most severe temperature, pressure, humidity, radiation, and
 
chemical environment which exists at the equipment location at the time it is required to perform after a design  basis loss-of-coolant or main steam line break accident. Such
 
equipment required after a design basis LOCA is also designed for the integrated radiation
 
exposure after the LOCA. The temperature, pressure, radiation, and humidity environment inside the containment after such accidents is presented in table 3.11-1. The containment
 
spray characteristics are given in subsection 6.2.2. 
 
3.11.2.2 Equipment Outside Containment Active safety-related equipment located outside the containment normally operates in ambient
 
temperatures up to 104°F. Normal operating radiation environments are provided in table 3.11-1. The design environmental conditions, including cumulative radiation exposure, are also
 
given in table 3.11-1. 
 
________________
: a. The renewed operating licenses authorize an additional 20-year period of extended
 
operation for both FNP units, resulting in a plant operating life of 60 years. The EQ program is
 
credited to continue to manage aging effects associated with EQ equipment for the period of
 
extended operation (see chapter 18, subsections 18.3.1 and 18.4.4). Applicable EQ
 
evaluations based on a 40-year design life were evaluated as time-limited aging analyses (TLAAs) for license renewal and will be revised as necessary to reflect the 60-year plant
 
operating life before the units enter the period of extended operation.
FNP-FSAR-3
 
3.11-4 REV 21  5/08 3.11.2.3 Equipment Supplied by Bechtel and Southern Company Services Descriptions of the qualification tests and analyses that have been performed on the
 
components of safety-related systems are cont ained in the sections indicated below: 
 
A. Containment isolation system in paragraph 6.2.4.4. 
 
B. Containment cooling system in paragraph 6.2.2.4.2. 
 
C. Penetration room filtration system in paragraph 6.2.3.4.2.
 
D. Control room ventilation system in paragraph 9.4.1.4. 
 
E. Auxiliary feedwater system in subsection 6.5.4. 
 
F. Component cooling system in subsection 9.2.2. 
 
G. Service water system in subsection 9.2.1. 
 
H. Diesel building ventilation system in subsection 9.4.7. 
 
In the auxiliary building ventilation systems, 11 coolers and fan units are designated as engineering safeguards. They are: 
 
A. High head injection pump rooms (3 cooling units). 
 
B. Low head injection pump rooms (2 cooling units). 
 
C. Auxiliary feed pump rooms (2 cooling units). 
 
D. Containment spray pump rooms (2 cooling units). 
 
E. Component cooling pump rooms (2 cooling units). 
 
The test and analysis requirements for these cooling units are the same as required for the
 
containment heat removal system, as given in paragraph 6.2.2.4.2. 
 
The main steam isolation valves are safety-rela ted components. They are hydrostatically tested in the manufacturer's facilities in accordance with the applicable code. Test and inspection
 
requirements are contained in subsection 10.3.4. 
 
3.11.2.4 Equipment Supplied By Westinghouse Temperature in the control room and computer room is maintained for personnel comfort
 
between 60 and 80°F, with the exact range in each room being controlled by a procedure. 
 
Design specifications for this equipment require that no loss of protective function should result 
 
FNP-FSAR-3
 
3.11-5 REV 21  5/08 when operating in temperatures up to 120°F and humidity up to 95 percent, which may occur upon the loss of air conditioning and/or the ventilation system. Thus there is a wide margin
 
between the design limit and the normal operating environment for the protective equipment. 
 
The normal operating temperature for the protective equipment in the containment will be
 
maintained below 120°F, (except that for out of core neutron detectors the normal operating
 
temperature will be maintained below 135°F). The protective equipment is designed for
 
continuous operation within design tolerance in this environment. 
 
The neutron detectors will be designed for continuous operation at 135°F (the normal operating
 
environment is always designed to be below this value) and will be capable of operation at
 
175°F for 8 hours. The power range detector has been tested in temperatures in excess of
 
175°F for a period of 16 hours with negligible decrease in insulation resistance. The insulation
 
resistance is the governing factor for severe environments. 
 
Type testing has been performed on safety-related equipment required to operate in the post
 
design basis accident environment. This testing has demonstrated that Westinghouse supplied
 
safety-related equipment has been designed to complete its protective functions in the
 
environments in which it must operate. 
 
A reconfirmation program of this testing as described in letter NS-CE-692, Eicheldinger to
 
Vassallo, July 10, 1975 has been completed for Farley. 
 
Applicable subprograms for Westinghouse supplied safety-related electrical equipment on the
 
Farley plant are: 
 
Subprogram Equipment B Process instrumentation and control  equipment, Parts 1, 3.
 
C Post accident hydrogen control system.
D Valve motor operators, Parts 1, 2, 3.
 
Solenoid valves, Part 1.
 
The electric hydrogen recombiners used for post-accident protection have been type tested to
 
demonstrate their compatibility of design for post accident operation. This test series is
 
documented in WCAP 7820 , and WCAP 7709-L Supplements 1-4 (1) which were accepted by the NRC in a letter from Vassallo to Eicheldinger dated May 1, 1975. 
 
The safety-related motor-operated valves which are required to operate in the design basis
 
accident environment are protected by Class H insulation. The insulation is used regardless of
 
the brevity of time for which the valves must operate after the design basis accident. 
 
The environmental qualification of the safety re lated motor operated valves is demonstrated in reference 2 and Westinghouse submittals NS-CE-692, Eicheldinger to Vassallo, dated July 10, FNP-FSAR-3
 
3.11-6 REV 21  5/08 1975 and NS-CE-756, Eigheldinger to Vassallo, dated August 15, 1975. The operability under severe environmental conditions of Westinghouse supplied solenoid valves is documented in
 
NS-CE-755, Eicheldinger to Vassallo, dated August 15, 1975. 
 
Westinghouse furnished process instrumentation and control equipment which is located inside
 
the containment and which must function in a post DBA environment, have been identified in
 
responses to IEB 79-01B and NUREG 0588. With the exception of the pressurizer pressure
 
channel installed in Unit 1, which was tested as described below, instruments for each of these
 
applications have been tested by exposing them to a steam and chemical spray environment, as described in the test references. As a result of this testing, the Unit 1 pressurizer level, steam generator level (W/R and N/R) and RCS pressure (W/R) instruments were replaced with
 
modified instruments during the first Unit 1 refueling. Modified transmitters for these
 
applications have been installed for Unit 2. 
 
The pressurizer pressure instruments have also been type tested for the design basis accident
 
environment. This test consisted of exposing a pressurizer pressure instrument similar to the ones used at Farley to saturated steam at 60 psig and 300°F for a 2-hour period and then to 20
 
psig and 244°F for 22 hours. The proper performance of the tested instrument was verified by
 
monitoring its output signal and then comparing it to a reference transmitter which was outside
 
the test chamber at room conditions. A simila r instrument was exposed to an integrated dose of 7.6 x 10 7 rads.
The pressurizer pressure instruments were not ex posed to sodium hydroxide. The instruments' protective functions of guaranteeing engineered safety features (ESF) actuation will be
 
completed prior to containment spray initiation which is the source of NaOH. ESF actuation will
 
occur when the containment pressure reaches the high setpoints of the containment pressure
 
instruments, while the sprays are not initiated until the high-high setpoints of the containment
 
pressure instruments are reached. The effect of chemical sprays on the transmitters has been
 
shown to be negligible in the short term as reported in NS-CE 719, Eicheldinger to Vassallo, dated July 25, 1975. Thus NaOH will not be part of the design basis accident environment prior
 
to ESF actuation by the pressurizer instruments or by attainment of the high setpoints of the
 
containment pressure instruments. 
 
While the instruments which are typical of the ones used on FNP were being tested, their output
 
voltages were monitored. The pressurizer pressure instruments had a change of 6.5 percent of
 
span while being subjected to the DBA environment. For those instruments assumed to
 
function in the safety analyses, the reactor protection system setpoints will be compatible with
 
the recorded accuracies of environmental testing, normal operational accuracies, and the
 
accident analyses. 
 
Analyses have been performed which include taking into account those short-term
 
environmental inaccuracies reported in letter NS-CE-792, Eichelding to Vassallo, dated October
 
1, 1975. The corresponding setpoint modifications have been provided in the Farley Technical
 
Specifications. These analyses demonstrate that the design bases are still met for all chapter
 
15 analyses. 
 
FNP-FSAR-3
 
3.11-7 REV 21  5/08 3.11.3  QUALIFICATION TEST RESULTS A final rule on environmental qualification of electr ical equipment important to safety for nuclear power plants became effective on February 22, 1983. This rule, 10 CFR 50.49, established the
 
NRC acceptance criteria and specified the requirements to be met for demonstrating the
 
environmental qualification of electrical equipment important to safety located in a harsh environment. In accordance with this rule, equipment may be qualified to the criteria specified
 
in either the DOR Guidelines, "Guidelines for Evaluating Environmental Qualification of Class
 
1E Electrical Equipment in Operating Reactors," or NUREG-0588, "Interim Staff Position on
 
Environmental Qualification of Safety-Related El ectrical Equipment," except for replacement equipment. Replacement equipment installed subsequent to February 22, 1983, must be
 
qualified in accordance with the provisions of 10 CFR 50.49, using the guidance of Regulatory
 
Guide 1.89, unless there are sound reasons to the contrary.  (Reference NRC letter dated
 
December 13, 1984)
 
In order to address the question for environmental qualification of electrical equipment for the
 
Farley Nuclear Plant, Alabama Power Company organized a task force to review the
 
qualification of electrical equipment. The equipment covered in this review included Class 1E
 
equipment inside containment and Class 1E equipment outside containment which is required
 
to mitigate a postulated accident and is subjected to a harsh environment. Harsh environment
 
is defined as LOCA/MSLB inside the containment and HELB areas outside the containment. 
 
Additionally, this review addressed the effects of radiation on equipment outside the
 
containment building during post-LOCA recirculation of containment sump fluids. This scope of
 
review assured that equipment necessary to protect the public health and safety is capable of
 
performing its function when subjected to a harsh environment. 
 
This review of environmental qualification was based on IE Bulletin 79-01B - Environmental
 
Qualification of Class 1E Equipment dated January 14, 1980 and the guidelines outlined for
 
Category II plants as defined by NUREG 0588, which was issued to operating license applicants
 
by NRC letter on February 5, 1980. The review was conducted by a task force composed of
 
personnel experienced in reactor systems safety analysis and design, plant operations, emergency operating procedures, nuclear safety, and environmental qualification. A critical
 
review of all documentation was conducted, using criteria established from IEB 79-01B and
 
NUREG 0588, resulting in an auditable record with appropriate procedures documented to
 
identify the specific equipment, the criteria used in reviewing the report, the reviewer, and the
 
specific report reference. 
 
As a part of this review effort, the task force reviewed the Plant Emergency Procedures to
 
ensure that equipment required by the procedures that could be subjected to a harsh
 
environment is qualified to operate for the time necessary to mitigate the particular accident. 
 
The results of the Farley Nuclear Plant environmental qualification review for each item of
 
safety-related electrical equipment subject to a harsh environment are documented in a submittals to the NRC dated July 30, 1980, for IEB 79-01B and September 15, 1980, as revised
 
and amended. These submittals consisted of tabular listings of all such equipment and
 
appropriate qualification-related data for each item in accordance with the NRC guidelines. 
 
Documentation was also provided, for a com parison, of the environmental qualification data against the requirements set forth in IEB 79-01B and NUREG 0588, on report evaluation sheets
 
for each type of equipment to identify the degree to which the qualification complies with the FNP-FSAR-3
 
3.11-8 REV 21  5/08 NRC staff position. Outstanding items were defined as being those for which discrepancies in meeting the guidelines of IEB 79-01B and NUREG 0588 have been identified. A summary of
 
these discrepancies was provided as part of the submittals and included corrective actions and
 
schedules together with justification for interim operation. 
 
3.11.4  LOSS OF VENTILATION The control room is provided with redundant air conditioning and filtration systems, as described
 
in subsection 9.4.1. This ensures that there will be no loss of ventilation to the control and
 
electrical equipment located within the control room. 
 
All safeguard pumps and motors in the auxiliary building are located in rooms equipped with
 
pump room coolers to provide adequate ventilation for the motors. These coolers are provided as a redundant system, so that if any one pum p room cooler fails, the corresponding pump would be shut down, except for the component cooling pumps. An engineering analysis has
 
been performed for all engineered safety feature pump rooms with room coolers. This analysis demonstrates that the equipment in the CCW pump rooms are capable of performing their
 
specified function during the temporary unavailability of one or both room coolers with the plant
 
encountering a design basis accident (DBA). The pump room cooler fan in a safeguard pump
 
room is powered from the same emergency bus as the pump motor. Thus, no single active or
 
passive failure can result in the loss of the safety function of both pumps of a redundant system.
 
FNP-FSAR-3
 
3.11-9 REV 21  5/08 REFERENCES 
: 1. Wilson, J. F., "Electric Hydrogen Recombiner for PWR Containments Equipment Qualification Report," WCAP 7709-L Supplements 1-4 (Proprietary), October 1973, WCAP 7820 (Non-proprietary), October 1973. 
: 2. Locante, J. and Igne, E. G., "Environmental Testing of Engineered Safety Features-Related Equipment (NSSS Non-standard Scope)," WCAP 7744 , Volumes 1 and 2, September 1970. 
 
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.11-1 (SHEET 1 OF 5)
EQ PROGRAM ENVIRONMENTAL CONDITIONS Environmental Conditions Location Inside Outside Containment Containment
 
Aux Bldg. Main Aux Bldg.
Steam Room Other Areas Normal Category A Category B Category C Pressure, (psig) 0 0 0 Temperature, (°F) 120 (a) 104 104 (j) Humidity, (%) 60 50 50
 
Post Accident Conditions For EQ DBA   
 
Pressure (psig) 52 (b) 1.3 (c) 0 (k) Temperature (°F) 367 (d) 325 (e) 104 (l) Humidity (%) 100 100 50 (k) Cumulative radiation  1.89 X 10 7(g) 3.36 X 10 6 See sheets dose (rads)(f,i)  2 and 3    Chemical additive Trisodium phosphate NA NA with pH 10.5 2,500 ppm boron 
 
Submergence  el 115'0" and el 130'5" NA below and below (h)     
 
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.11-1 (SHEET 2 OF 5)
Unit 1 Auxiliary Building Harsh Environment Room Radiation Doses (i)
 
Room  Total Number Room Description  Dose (rads) 111 Containment Spray Pump Room 1.34E6 113 Valve Encapsulation Room 4.04E5 120 Corridor 1.35E5 124 Valve Encapsulation Room 4.04E5 125 Containment Spray Pump Room 2.68E6 128 RHR Heat Exchanger Room 4.10E6 129 RHR Low Head Pump Room 3.35E6 131 RHR Low Head Pump Room 3.35E6 161 Corridor 2.99E5 162 Hallway 3.01E3 172 Hallway 2.78E6 173 Charging/SI Pump Room 2.12E6 174 Charging/SI Pump Room 2.14E6 175 Hallway 1.85E5 181 Charging/SI Pump Room 2.15E6 182 Containment Storage Room 1.24E6 184 Piping Penetration Room 4.68E6 218 BTRS Chiller Unit Room 5.53E3 223 Piping Penetration Room 5.85E6 317 Penetration and Filtration    System and Equipment Room 6.90E6 322 Hallway 2.74E5 333 Electrical Penetration Room 5.97E5 334 Electrical Penetration Room 1.34E6 347 Electrical Penetration Room 1.03E6
 
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.11-1 (SHEET 3 OF 5)
Unit 2 Auxiliary Building Harsh Environment Room Radiation Doses (i)
 
Room    Total Number Room Description  Dose (rads) 2111 Containment Spray Pump Room 1.44E6 2113 Valve Encapsulation Room 4.04E5 2120 Corridor 2.56E5 2124 Valve Encapsulation Room 4.04E5 2125 Containment Spray Pump Room 2.68E6 2128 RHR Heat Exchanger Room 4.80E6 2129 RHR Low Head Pump Room 3.29E6 2131 RHR Low Head Pump Room 3.29E6 2161 Corridor 1.96E5 2162 Hallway 3.54E3 2172 Hallway 2.78E6 2173 Charging/SI Pump Room 2.10E6 2174 Charging/SI Pump Room 2.10E6 2175 Hallway 5.25E5 2181 Charging/SI Pump Room 2.10E6 2182 Containment Storage Room 1.85E6 2184 Piping Penetration Room 4.68E6 2218 BTRS Chiller Unit Room 7.34E3 2223 Piping Penetration Room 5.65E6 2317 Penetration and Filtration System and Equipment Room 6.90E6 2322 Hallway 2.74E5 2333 Electrical Penetration Room 5.97E5 2334 Electrical Penetration Room 1.34E6 2347 Electrical Penetration Room 1.03E6
 
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.11-1 (SHEET 4 OF 5)
NOTES: 
: a. Normal temperature for the out-of-core neutron detectors is 135°F. The neutron detectors are capable of operation at 175°F for 8 h.
: b. Peak containment pressure is 52 psig for an MSLB inside containment and 44 psig for a LOCA. The LOCA/MSLB containment composit e pressure profile is shown in figure 3.11-1 and covers the DBA duration of 30 days.
: c. Peak pressure for an MSLB in the main steam room is 1.3 psig. The main steam room accident pressure profile is shown on FSAR figure 3J-15.
: d. Peak containment atmosphere temperature is 367°F for an MSLB inside containment and 264°F for LOCA. The containment accident temperature profiles cover the DBA duration of 30 days. The LOCA/MSLB contai nment composite temperature profile is shown in figure 3.11-2 and the LOCA-specific profile is shown in figure 6.2-40. 
 
Qualification testing envelopes either the figure 3.11-2 profile by itself or the figure 6.2-
 
40 profile, with consideration of worst-case MSLB surface temperatures from the
 
spectrum of MSLBs.
: e. Peak main steam room atmosphere temper ature for an MSLB in the main steam room is 325°F. The main steam room accident temperature profile is shown in figure 3J-10.
: f. Radiation dose is cumulative for 40 years (a) and 30 days post accident. Individual components may be required to operate less than 30 days as specified in operating time
 
of APC responses to IE Bulletin 79-01B and NUREG 0588. 
 
The operating licenses for both FNP units have been renewed ad the original licensed
 
operating terms have been extended by 20 years.
Radiation dose values used in EQ evaluations will be updated before the units enter the period of extended operation.
: g. A radiation dose of 1.89 x 10 7 rads is outside the secondary shield wall. The radiation dose for the out-of-core neutron detectors, inside the secondary shield wall, is 1 x 10 10 rads. The area of the head of the narrow and wide range RCS RTDs from which the
 
RTD pigtails emerge is exposed to a radiation dose of 4.11 x 10 7 rads.   
 
________________
: a. The renewed operating licenses authorize an additional 20-year period of extended operation for both FNP units, resulting in a plant operating life of 60 years. The EQ program
 
is credited to continue to manage aging effects associated with EQ equipment for the period
 
of extended operation (see chapter 18, subsections 18.3.1 and 18.4.4). Applicable EQ
 
evaluations based on a 40-year design life were evaluated as time-limited aging analyses (TLAAs) for license renewal and will be revised as necessary to reflect the 60-year plant
 
operating life before the units enter the period of extended operation.
FNP-FSAR-3
 
REV 21  5/08 TABLE 3.11-1 (SHEET 5 OF 5)
: h. Submergence is considered below el 130 ft 5 in. for an MSLB in the main steam room. 
: i. The auxiliary building rooms listed on sheets 2 and 3 of table 3.11-1 are those rooms for which 1) the total integrated dose (TID) is greater than 1E3 rads 2) in which there is a
 
significant radiation level increase due to an accident and 3) in which EQ equipment is
 
located. In the context of this note, cable is not considered to be equipment. Therefore, there are no rooms listed on sheets 2 and 3 which have only EQ cable in them. The
 
rooms having EQ cable but no other EQ equipment were not included in the list
 
because a detailed calculation to arrive at a precise value was not performed. Instead, it was assumed that the total radiation dose is 1E8 rads. From the calculated values
 
listed in sheets 2 and 3, this is clearly a conservative assumption. 
 
Per the recommendations of EPRI NP-2129, dated November 1981, the threshold values are taken to be 1E3 rads for electronic components, 1E4 rads for equipment
 
utilizing Teflon, and 1E5 rads for all other electrical equipment. Thus, any room listed
 
with a TID less than 1E5 rads is considered a harsh environment only if it has electronic
 
equipment or equipment utilizing Teflon; any room listed with a TID less than 1E4 rads
 
is considered harsh only if it has electronic equipment. 
: j. Temperatures above 104°F are evaluated to ensure that equipment operability is maintained and is appropriately documented.
: k. Post-accident pressure and humidity for all rooms in the auxiliary building would at no time be significantly more severe than the environment that would occur during normal
 
plant operation, including anticipated operational transients.
: l. For rooms not cooled by service water t he post-accident temperature would at no time be significantly more severe than the environment that would occur during normal plant
 
operation, including anticipated operational transients. For rooms cooled by service
 
water the post-accident temperatures are documented in Table 9.4-6A. The electrical
 
equipment in these rooms has been evaluated and found acceptable for operation at
 
the anticipated temperatures. Therefore, the increase in the ambient temperature of
 
these rooms during the 30 day post-accident period is not considered significant as
 
defined in 10CFR50.49 Section c.(iii). The equi pment in rooms cooled by service water is thus considered to be in the mild env ironment except for the rooms that have been previously identified harsh due to the post-accident radiation.
 
REV 21  5/08 FNP COMPOSITE LOCA/MSLB CONTAINMENT PRESSURE PROFILES JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.11-1
 
REV 21  5/08 FNP COMPOSITE LOCA/MSLB CONTAINMENT TEMPERATURE PROFILES JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3.11-2
 
FNP-FSAR-3A
 
3A-i REV 21  5/08 3A CONFORMANCE WITH NRC REGULATORY GUIDES TABLE OF CONTENTS
 
Page Regulatory Guide 1.1 .......................................................................................................3A-1.1-1 Regulatory Guide 1.2 .......................................................................................................3A-1.2-1 Regulatory Guide 1.3 .......................................................................................................3A-1.3-1 Regulatory Guide 1.4 .......................................................................................................3A-1.4-1 Regulatory Guide 1.5 .......................................................................................................3A-1.5-1 Regulatory Guide 1.6 .......................................................................................................3A-1.6-1 Regulatory Guide 1.7 .......................................................................................................3A-1.7-1 Regulatory Guide 1.8 .......................................................................................................3A-1.8-1 Regulatory Guide 1.9 .......................................................................................................3A-1.9-1 Regulatory Guide 1.10 ...................................................................................................3A-1.1 0-1 Regulatory Guide 1.11 ...................................................................................................3A-1.1 1-1 Regulatory Guide 1.12 ...................................................................................................3A-1.1 2-1 Regulatory Guide 1.13 ...................................................................................................3A-1.1 3-1 Regulatory Guide 1.14 ...................................................................................................3A-1.1 4-1 Regulatory Guide 1.15 ...................................................................................................3A-1.1 5-1 Regulatory Guide 1.16 ...................................................................................................3A-1.1 6-1 Regulatory Guide 1.17 ...................................................................................................3A-1.1 7-1 Regulatory Guide 1.18 ...................................................................................................3A-1.1 8-1 Regulatory Guide 1.19 ...................................................................................................3A-1.1 9-1 Regulatory Guide 1.20 ...................................................................................................3A-1.2 0-1 Regulatory Guide 1.21 ...................................................................................................3A-1.2 1-1 Regulatory Guide 1.22 ...................................................................................................3A-1.2 2-1 Regulatory Guide 1.23 ...................................................................................................3A-1.2 3-1 Regulatory Guide 1.24 ...................................................................................................3A-1.2 4-1 Regulatory Guide 1.25 ...................................................................................................3A-1.2 5-1 Regulatory Guide 1.26 ...................................................................................................3A-1.2 6-1 Regulatory Guide 1.27 ...................................................................................................3A-1.2 7-1 [HISTORICAL]
[Regulatory Guide 1.28] ........................................................................3A-1.28-1 Regulatory Guide 1.29 ...................................................................................................3A-1.2 9-1 Regulatory Guide 1.30 ...................................................................................................3A-1.3 0-1 Regulatory Guide 1.31 ...................................................................................................3A-1.3 1-1 Regulatory Guide 1.32 ...................................................................................................3A-1.3 2-1 Regulatory Guide 1.33 ...................................................................................................3A-1.3 3-1 Regulatory Guide 1.34 ...................................................................................................3A-1.3 4-1 [HISTORICAL] [Regulatory Guide 1.35]........................................................................
3A-1.35-1 Regulatory Guide 1.36 ...................................................................................................3A-1.3 6-1 Regulatory Guide 1.37 ...................................................................................................3A-1.3 7-1 Regulatory Guide 1.38 ...................................................................................................3A-1.3 8-1 Regulatory Guide 1.39 ...................................................................................................3A-1.3 9-1 Regulatory Guide 1.40 ...................................................................................................3A-1.4 0-1 Regulatory Guide 1.41 ...................................................................................................3A-1.4 1-1 FNP-FSAR-3A
 
3A-ii REV 21  5/08 TABLE OF CONTENTS Page Regulatory Guide 1.42 ...................................................................................................3A-1.4 2-1 Regulatory Guide 1.43 ...................................................................................................3A-1.4 3-1 Regulatory Guide 1.44 ...................................................................................................3A-1.4 4-1 Regulatory Guide 1.45....................................................................................................3A-1.4 5-1 Regulatory Guide 1.46....................................................................................................3A-1.4 6-1 Regulatory Guide 1.47....................................................................................................3A-1.4 7-1 Regulatory Guide 1.48....................................................................................................3A-1.4 8-1 Regulatory Guide 1.49....................................................................................................3A-1.4 9-1 Regulatory Guide 1.50....................................................................................................3A-1.5 0-1 Regulatory Guide 1.51....................................................................................................3A-1.5 1-1 Regulatory Guide 1.52....................................................................................................3A-1.5 2-1 Regulatory Guide 1.53....................................................................................................3A-1.5 3-1 Regulatory Guide 1.54....................................................................................................3A-1.5 4-1 Regulatory Guide 1.55....................................................................................................3A-1.5 5-1 Regulatory Guide 1.56....................................................................................................3A-1.5 6-1 Regulatory Guide 1.57....................................................................................................3A-1.5 7-1 Regulatory Guide 1.58....................................................................................................3A-1.5 8-1 Regulatory Guide 1.59....................................................................................................3A-1.5 9-1 Regulatory Guide 1.60....................................................................................................3A-1.6 0-1 Regulatory Guide 1.61....................................................................................................3A-1.6 1-1 Regulatory Guide 1.62....................................................................................................3A-1.6 2-1 Regulatory Guide 1.63....................................................................................................3A-1.6 3-1 Regulatory Guide 1.64....................................................................................................3A-1.6 4-1 Regulatory Guide 1.65....................................................................................................3A-1.6 5-1 Regulatory Guide 1.66....................................................................................................3A-1.6 6-1 Regulatory Guide 1.67....................................................................................................3A-1.6 7-1 Regulatory Guide 1.68....................................................................................................3A-1.6 8-1 Regulatory Guide 1.69....................................................................................................3A-1.6 9-1 Regulatory Guide 1.70....................................................................................................3A-1.7 0-1 Regulatory Guide 1.70.1..............................................................................................3A-1.70.1-1 Regulatory Guide 1.70.2..............................................................................................3A-1.70.2-1 Regulatory Guide 1.70.3..............................................................................................3A-1.70.3-1 Regulatory Guide 1.70.4..............................................................................................3A-1.70.4-1 Regulatory Guide 1.70.5..............................................................................................3A-1.70.5-1 Regulatory Guide 1.70.6..............................................................................................3A-1.70.6-1 Regulatory Guide 1.70.7..............................................................................................3A-1.70.7-1 Regulatory Guide 1.71....................................................................................................3A-1.7 1-1 Regulatory Guide 1.72....................................................................................................3A-1.7 2-1 Regulatory Guide 1.73....................................................................................................3A-1.7 3-1 Regulatory Guide 1.74....................................................................................................3A-1.7 4-1 Regulatory Guide 1.75....................................................................................................3A-1.7 5-1 Regulatory Guide 1.76....................................................................................................3A-1.7 6-1 Regulatory Guide 1.77....................................................................................................3A-1.7 7-1 Regulatory Guide 1.78....................................................................................................3A-1.7 8-1 FNP-FSAR-3A
 
3A-iii REV 21  5/08 TABLE OF CONTENTS
 
Page Regulatory Guide 1.79....................................................................................................3A-1.7 9-1 Regulatory Guide 1.80....................................................................................................3A-1.8 0-1 Regulatory Guide 1.81....................................................................................................3A-1.8 1-1 Regulatory Guide 1.82....................................................................................................3A-1.8 2-1 Regulatory Guide 1.83....................................................................................................3A-1.8 3-1 Regulatory Guide 1.84....................................................................................................3A-1.8 4-1 Regulatory Guide 1.85.................................................................................................. 3A-1.8 5-1 Regulatory Guide 1.86.................................................................................................. 3A-1.8 6-1 Regulatory Guide 1.87.................................................................................................. 3A-1.8 7-1 Regulatory Guide 1.88.................................................................................................. 3A-1.8 8-1 Regulatory Guide 1.95.................................................................................................. 3A-1.9 5-1 Regulatory Guide 1.99.................................................................................................. 3A-1.9 9-1 Regulatory Guide 1.108.............................................................................................. 3A-1.108-1 Regulatory Guide 1.109.............................................................................................. 3A-1.109-1 Regulatory Guide 1.111................................................................................................3A-1.111-1 Regulatory Guide 1.112................................................................................................3A-1.112-1 Regulatory Guide 1.113................................................................................................3A-1.113-1 Regulatory Guide 1.127................................................................................................3A-1.127-1 Regulatory Guide 1.155.............................................................................................. 3A-1.155-1 Regulatory Guide 1.163.............................................................................................. 3A-1.163-1 Regulatory Guide 1.182.............................................................................................. 3A-1.182-1 Regulatory Guide 1.190................................................................................................3A-1.190-1 Regulatory Guide 1.194................................................................................................3A-1.194-1 Regulatory Guide 1.195................................................................................................3A-1.195-1 Regulatory Guide 1.196.............................................................................................. 3A-1.196-1 Regulatory Guide 1.197.............................................................................................. 3A-1.197-1 
 
FNP-FSAR-3A
 
3A-1 REV 21  5/08 APPENDIX 3A CONFORMANCE WITH REGULATORY GUIDES
 
This appendix discusses the extent of conformance of the Farley Nuclear Plant with Division 1
 
NRC Regulatory Guides, which were issued through the end of August 1974 -- that is, through
 
Regulatory Guide 1.88. Regulatory Guide 1.70 is discussed through 1.70.7. A description of
 
Farley conformance with Regulatory Guides issued subsequent to Regulatory Guide 1.88 is included in this appendix. Specific revision numbers and dates of issue are identified in the title
 
of each guide.
 
Page numbering for appendix 3A has been changed in Amendment 40 to reflect the appendix
 
letter, division number, particular Guide and page number.
 
Example: 3A-1.3-1 (one page)
 
3A-1.18-2 (two pages)
 
FNP-FSAR-3A
 
3A-1.1-1 REV 21  5/08 Regulatory Guide 1.1 - NET POSITIVE SUCTION HEAD FOR ECCS AND CONTAINMENT HEAT REMOVAL PUMPS (SAFETY GUIDE 1, 11/2/70) 
 
CONFORMANCE
 
The NRC Regulatory Guide 1.1 states that the emergency core cooling and containment heat
 
removal systems should be designed so that adequate net positive suction head (NPSH) is
 
provided to system pumps assuming maximum expected temperatures of pumped fluids and no increase in containment pressure from that present prior to postulated loss-of-coolant accidents.
 
As discussed in subsection 6.3.2.14, the emergency core cooling and containment heat removal
 
systems are designed to provide an availabl e NPSH which is greater than pump vendor specified minimum NPSH requirements assuming either of the following conditions: 
: 1. If expected maximum pumped fluid temperatures are less than 212°F, NPSH is determined assuming a liquid temperature of 212°F and no increase in
 
containment pressure from that present prior to postulated loss-of-coolant
 
accidents.
: 2. If expected maximum pumped fluid temperatures exceed 212°F, NPSH is determined assuming the temperature of the pumped liquid is at saturation for
 
the containment pressure.
 
Adequate net positive suction head is provided to both the RHR and containment spray pumps.
 
In calculating the available NPSH during injection, no credit is taken for any water within the
 
RWST and full penalty is taken for head losses based on actual piping layouts. During the
 
recirculation mode, credit is taken for a minimum available water level above the top of the inlet
 
to the containment spray pump suction piping and the RHR suction piping. No credit is taken
 
for any increase in containment pressure from a postulated accident.
 
Additional conservatism is introduced by assuming that all recirculated sump water is at
 
saturation conditions.
 
The methods utilized in calculating NPSH for the Farley Nuclear Plant, as described above and
 
in subsection 6.3.2.14, are adequately conservative and meet Regulatory Guide 1.1 by ensuring
 
adequate NPSH with adequate margin for the centrifugal charging, safety injection, residual
 
heat removal, and containment spray pumps.
 
FNP-FSAR-3A
 
3A-1.2-1 REV 21  5/08 Regulatory Guide 1.2 - THERMAL SHOCK TO REACTOR PRESSURE VESSELS (SAFETY GUIDE 2, 11/2/70) 
 
CONFORMANCE
 
The reactor pressure vessel design supported by current research programs in the area of
 
fracture toughness of reactor vessel materials conforms to the intent of Regulatory Guide 1.2. 
 
Fracture toughness is discussed in subsection 5.2.4; the capability for annealing the reactor
 
vessel is discussed in subsection 5.4.3.7.
 
FNP-FSAR-3A
 
3A-1.3-1 REV 21  5/08 Regulatory Guide 1.3 - ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT
 
ACCIDENT FOR BOILING WATER REACTORS (Rev. 2, 6/74) 
 
CONFORMANCE
 
The Guide is not applicable to the Farley Nuclear Plant.
 
FNP-FSAR-3A
 
3A-1.4-1 REV 21  5/08 Regulatory Guide 1.4  - ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT
 
ACCIDENT FOR PRESSURIZED WATER REACTORS (Rev. 2, 6/74) 
 
CONFORMANCE
 
The assumptions of the regulatory position of Regulatory Guide 1.4 are used without
 
exception in the analysis of the design basis loss-of-coolant accident in subsection 15.4.1.
 
FNP-FSAR-3A
 
3A-1.5-1 REV 21  5/08 Regulatory Guide 1.5  - ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A STEAM LINE BREAK
 
ACCIDENT FOR BOILING WATER REACTORS (SAFETY GUIDE
 
5, 3/10/71) 
 
CONFORMANCE
 
The Guide is not applicable to the Farley Nuclear Plant.
 
FNP-FSAR-3A
 
3A-1.6-1 REV 21  5/08 Regulatory Guide 1.6  -  INDEPENDENCE BETWEEN REDUNDANT STANDBY (ONSITE)
POWER SOURCES AND BETWEEN THEIR DISTRIBUTION
 
SYSTEMS (SAFETY GUIDE 6, 3/10/71) 
 
CONFORMANCE
 
As described in subsection 8.3.1.2(c), the applicant's design conforms to the regulatory position
 
without exception.
 
FNP-FSAR-3A
 
3A-1.7-1 REV 21  5/08 Regulatory Guide 1.7  -  CONTROL OF COMBUSTIBLE GAS CONCENTRATIONS IN CONTAINMENT FOLLOWING A LOSS-OF-COOLANT
 
ACCIDENT (SAFETY GUIDE 7, 3/10/71)
CONFORMANCE
 
The design guidance and assumptions for analysis of the regulatory position of Regulatory
 
Guide 1.7 are used without exception for control of combustible gas concentrations in
 
containment following a loss-of-coolant accident, as described in subsections 6.2.5, 7.6.4, and
 
15.4.1.
 
FNP-FSAR-3A
 
3A-1.8-1 REV 21  5/08 Regulatory Guide 1.8  - PERSONNEL SELECTION AND TRAINING (SAFETY GUIDE 8, 3/10/71) 
 
CONFORMANCE
 
Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI
 
N18.1-1971 for comparable positions and the supplemental requirements specified in
 
10 CFR 55, except for the Health Physics manager who shall meet or exceed the
 
qualifications of Regulatory Guide 1.8, September 1975. Personnel who complete an
 
accredited program which has been endorsed by the NRC shall meet the requirements of the
 
accredited program in lieu of the above. Since the occurrence of the TMI-2 accident, additional shift manpower has been added and an upgrading process has been instituted for
 
the training and qualification of operating personnel. These changes are in conformance
 
with NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident," as
 
modified by NUREG-0737, "Clarification of TMI Action Plan Requirements."  Details of the
 
training program are presented in Section 13.2.
 
FNP-FSAR-3A
 
3A-1.9-1 REV 21  5/08 Regulatory Guide 1.9  - SELECTION OF DIESEL GENERATOR SET CAPACITY FOR STANDBY POWER SUPPLIES (SAFETY GUIDE 9, 3/10/71) 
 
CONFORMANCE
 
The standby power system is discussed fully in subsection 8.3.1.2(d), AC Power Systems, and meets the recommendations of Regulatory Guide 1.9.
 
FNP-FSAR-3A
 
3A-1.10-1 REV 21  5/08 Regulatory Guide 1.10  - MECHANICAL (CADWELD) SPLICES IN REINFORCING BARS OF CONCRETE CONTAINMENTS (Rev. 1, 1/2/73) 
 
CONFORMANCE
 
The recommendations of Regulatory Guide 1.10 were the basis for testing and inspecting all
 
mechanical splices utilized at the facility. However, a disadvantage of using production splices
 
for testing was that each production splice removed had to be replaced by two cadwelds, thus
 
introducing additional splices into the structure. Further, production splices in hoop bars were
 
not tested, since the test itself imposes bending in addition to tension and would not represent
 
the actual stressing of the bars. Accordingly, splices in curved bars were tested by the "sister
 
splice" method while straight bar splices were tested by either method.
 
Additional information is contained in appendix 3C and subsection 4.4.2.
 
FNP-FSAR-3A
 
3A-1.11-1 REV 21  5/08 Regulatory Guide 1.11  - INSTRUMENT LINES PENETRATING PRIMARY REACTOR CONTAINMENT (SAFETY GUIDE 11, 3/10/71) 
 
CONFORMANCE
 
Sensing lines penetrating containment are provided with isolation valves in accordance with
 
NRC General Design Criteria 55 or 56 or meet regulatory position C.2 of Regulatory Guide 1.11.
 
The sensing lines, configurations are discussed in subsection 6.2.4.
 
The containment pressure sensing lines for the Post-Accident Monitoring (PAM) and
 
Engineered Safety Features Actuation (ESF)/Reactor Protection System (RPS) must be open to
 
containment at all times. The transmitters are located on unvalved lines outside containment
 
with remote seal sensors located inside containment. This arrangement is considered to be in
 
compliance with General Design Criterion 56 as the sensors are located as close to
 
containment boundary as practical and their pressure boundaries are either ASME Code Class
 
2 or are a double pressure boundary rated higher than the containment design pressure.
 
FNP-FSAR-3A
 
3A-1.12-1 REV 21  5/08 Regulatory Guide 1.12  - INSTRUMENTATION FOR EARTHQUAKES (Rev. 1, April 1974) 
 
The Regulatory Guide 1.12 guidelines for instrumentation to monitor earthquakes has been
 
replaced by EPRI Reports. The NRC has accepted the EPRI Reports listed in section 3.7.4 as
 
an acceptable approach to meet the seismic monitoring requirements and determine plant
 
action following an earthquake.
 
CONFORMANCE
 
With the exceptions listed below, and as described in subsection 3.7.4, Seismic Instrumentation
 
Program, the instrumentation for measuring the seismic response of the plant due to
 
earthquakes complies with the recommendations of Regulatory Guide 1.12.
 
A. No seismic instruments are mounted on piping or pipe supports, where mechanical vibration can obscure a seismic event.
 
B. A peak response accelerograph is employed instead of a response spectrum recorder for a floor location in a seismic Category 1 structure (Regulatory Guide
 
1.12, paragraph 1.C (2).(a).).
 
C. Two triaxial time history recorders are used instead of response spectrum recorders on two seismic Category 1 structures that differ from the containment
 
in seismic response (Regulatory Guide 1.12, paragraph 1.c.(3).).
 
D. The present seismic trigger on the containment foundation performs the function of the seismic switch recommended in paragraph 4.1.3 of ANSI N18.5-1974.
 
FNP-FSAR-3A
 
3A-1.13-1 REV 21  5/08 Regulatory Guide 1.13  - FUEL STORAGE FACILITY DESIGN BASES (SAFETY GUIDE 13, 3/10/71) 
 
CONFORMANCE
 
Design of the wet spent-fuel facility complies with the recommendations of Regulatory Guide
 
1.13 in the manner described in subsections 9.1.2, Wet Spent-Fuel Storage; 9.1.3, Spent-Fuel
 
Pool Cooling and Cleanup; 9.1.4, Fuel Handling and Spent-Fuel Cask Crane; section 3.5, Missile Protection; subsection 3.8.4, Design of Other Category I Structures; and section 9.4, Air
 
Conditioning, Heating, Cooling and Ventilation.
 
FNP-FSAR-3A
 
3A-1.14-1 REV 21  5/08 Regulatory Guide 1.14  - REACTOR COOLANT PUMP FLYWHEEL INTEGRITY (SAFETY GUIDE 14, 10/27/71) 
 
CONFORMANCE
 
This topic is addressed in subsection 5.2.6.
 
FNP-FSAR-3A
 
3A-1.15-1 REV 21  5/08 Regulatory Guide 1.15  -  TESTING OF REINFORCING BARS FOR CONCRETE STRUCTURES (Rev. 1, 12/28/72) 
 
CONFORMANCE
 
The methodology for testing reinforcing bars used in Seismic Category I concrete structures
 
conforms with the recommendations of Regulatory Guide 1.15 as described in subsection
 
3.8.1.6.2.
 
FNP-FSAR-3A
 
3A-1.16-1 REV 21  5/08 Regulatory Guide 1.16  - REPORTING OF OPERATING INFORMATION (Rev. 1, 10/73) 
 
CONFORMANCE
 
In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the
 
program for reporting of Farley, Units 1 & 2, operating information is in accordance with Generic
 
Letter 97-02, "Revised Contents of the Monthly Operating Report," dated May 15, 1997.
 
Reporting Requirements are contained within the Technical Specifications.
 
FNP-FSAR-3A
 
3A-1.17-1 REV 21  5/08 Regulatory Guide 1.17 - PROTECTION AGAINST INDUSTRIAL SABOTAGE (Rev. 0, 6/73) 
 
CONFORMANCE
 
The requirements of the Regulatory Guide are met as described in section 13.7 and the Security
 
Plan.
 
FNP-FSAR-3A
 
3A-1.18-1 REV 21  5/08 Regulatory Guide 1.18  - STRUCTURAL ACCEPTANCE TEST FOR CONCRETE PRIMARY REACTOR CONTAINMENTS (Rev. 1, 12/28/72) 
 
CONFORMANCE
 
Regulatory Guide 1.18 establishes a systematic approach to testing wherein quantitative
 
information is obtained concerning structural response to pressurization. The following
 
discussion is provided to clarify the extent of compliance to this Regulatory Guide.
: 1.
 
==Reference:==
Paragraph C.1 of the Regulatory Guide 
 
A continuous increase in containment pressure, rather than incremental pressure increases, is considered acceptable, since data collection is made rapidly at
 
each pressure datum.  "Rapidly" is defined as requiring a time interval for data
 
collection sufficiently short so that the change in pressure while the data are
 
being collected would cause a  change in the structural response of less than 5 
 
percent of the total anticipated change. For example, assume a pressure datum
 
at each 15-lb/in 2 interval, a  test pressure of 60 lb/in 2 , and a total expected strain of 200 microstrains (microinches/in.). The time interval for data collection, therefore, is required to be equal to, or less than, the time during which
 
pressurization would create a 10-microstrain change. 
 
When the test pressure reached its maximum in the containment, a hold period of at least 1 hour, or of such duration as necessary for recording crack patterns, was provided. 
: 2.
 
==Reference:==
Paragraph C.5 of the Regulatory Guide 
 
The design of the Farley Containment closely follows those of the Arkansas Nuclear One Unit 1 (Docket No. 50-313) and Millstone Unit 2 (Docket No. 50-
 
336), and therefore is not a prototype containment. Consequently, this
 
paragraph is not considered to be applicable. 
: 3.
 
==Reference:==
Paragraph C.9 of the Regulatory Guide 
 
The structural integrity test was scheduled for periods in which extremely inclement weather is not forecast. However, due to the state-of-the-art of
 
weather forecasting, and the time involved in the preparation and performance of
 
the test, should snow, heavy rain, or strong wind occur during the test, it may be
 
continued and the results considered valid unless evidence indicates otherwise. 
 
A retest will be made if the results are found to be invalid. 
: 4.
 
==Reference:==
Paragraph C.10 of the Regulatory Guide 
 
Due to the amount of time involved in preparing for and performing the test, should the test pressure drop due to unexpected condition to or below the next
 
lower pressure level, it is intended to continue the test, without a restart at
 
atmospheric pressure, unless the structural response deviates significantly from
 
that expected.
FNP-FSAR-3A
 
3A-1.18-2 REV 21  5/08
: 5.
 
==Reference:==
Paragraph C.12 of the Regulatory Guide 
 
It is believed that this paragraph applies to PSAR only. However, the type of information which will be included in the final test report will conform to
 
Paragraph C.13 of the Regulatory Guide. 
: 6.
 
==Reference:==
Appendix A.2.a of the Regulatory Guide 
 
The design of the Farley Containment closely follows those of the Arkansas Nuclear One Unit 1 and the Millstone Unit 2, and therefore is not a prototype
 
containment. Consequently, this paragraph is not considered to be applicable. 
: 7.
 
==Reference:==
Appendix A.2.g of the Regulatory Guide 
 
The design of the Farley Containment closely follows those of the Arkansas Nuclear One Unit 1 and the Millstone Unit 2, and therefore is not a prototype
 
containment. Consequently, this paragraph is not considered to be applicable. 
 
A description of the containment structural acceptance test is presented in appendix 3H. 
 
FNP-FSAR-3A
 
3A-1.19-1 REV 21  5/08 Regulatory Guide 1.19  - NON-DESTRUCTIVE EXAMINATION OF PRIMARY CONTAINMENT LINER WELDS (SAFETY GUIDE 19, Rev. 1, 8/11/72) 
 
CONFORMANCE
 
The recommendations of Regulatory Guide 1.19 have been met. See appendix 3G for details. 
 
FNP-FSAR-3A
 
3A-1.20-1 REV 21  5/08 Regulatory Guide 1.20  - VIBRATION MEASUREMENTS ON REACTOR INTERNALS (SAFETY GUIDE 20, 12/29/71) 
 
CONFORMANCE
 
This topic is addressed in subsection 3.9.1.3.1. 
 
FNP-FSAR-3A
 
3A-1.21-1 REV 21  5/08 Regulatory Guide 1.21 - MEASURING AND REPORTING OF EFFLUENTS FROM NUCLEAR POWER PLANTS (SAFETY GUIDE 21, 12/29/71) 
 
CONFORMANCE
 
The Farley units are in compliance with Regulatory Guide 1.21 with the following exceptions:
: 1. There are no continuous monitors on the turbine building drains. Grab samples are taken for composite prior to or during its discharge for each batch released.
: 2. Gamma spectroscopy measurements are used as the basis for estimating the quantity of low-level particulate activity released.
: 3. I-135 is not monitored in gaseous effluents due to its short half-life.
: 4. The steam jet air ejector is monitored by monthly grab samples in accordance with the Offsite Dose Calculation Manual.
 
FNP-FSAR-3A
 
3A-1.22-1 REV 21  5/08 Regulatory Guide 1.22 - PERIODIC TESTING 0F PROTECTION SYSTEM ACTUATION FUNCTIONS (SAFETY GUIDE 22, 2/17/72) 
 
CONFORMANCE
 
The design recommendations of Regulatory Guide 1.22 have been met. Design of the
 
protection system is discussed fully in chapter 7, Instrumentation and Controls. Some tests
 
must be performed as sequential steps on isolated portions of the system so that an actual
 
reactor scram does not occur as a result of the testing. Specific discussion is found in
 
subsections 7.1.2, 7.2.3, and 7.3.2. 
 
FNP-FSAR-3A
 
3A-1.23-1 REV 21  5/08 Regulatory Guide 1.23 - ONSITE METEROLOGICAL PROGRAMS (SAFETY GUIDE 23, 2/17/72) 
 
CONFORMANCE
 
Meteorological programs are discussed in Section 2.3 in detail. 
 
Regulatory Guide 1.23, which was issued in 1972, states in paragraph 4 of its Regulatory
 
Position that temperature difference measur ements should have an accuracy of +/-0.1°C. 
 
The thermistors used in the Farley tower temperature difference circuits were ordered in late
 
1970 and have given excellent servic e (well over 90-percent recovery) for the past 12 years.
Inspection of the analog charts has shown very few questionable temperature difference
 
readings or calibration or drift problems. 
 
At the factory, each thermistor is checked against a calibration curve to ensure the accuracy is
 
0.15°C over its entire temperature range. Most of the measurements at the Farley site are
 
taken over a small portion of this total range; therefore, the accuracy is better than 0.15°C. In
 
view of the reliability and quality of the data recorded using this system, modifications are not
 
necessary. 
 
Regulatory Guide 1.23 makes no recommendation for maintaining the area surrounding the met
 
tower free from obstructions. The Farley Nuclear Plant met tower will be maintained free from
 
obstructions to wind flow in accordance with proposed Revision 1 to Regulatory Guide 1.23 as
 
recommended by NUREG-0654, Revision 1, Appendix 2.
 
FNP-FSAR-3A
 
3A-1.24-1 REV 21  5/08 Regulatory Guide 1.24 - ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL-RADIOLOGICAL CONSEQUENCES OF A PRESSURIZED
 
WATER REACTOR RADIOACTIVE GAS STORAGE TANK
 
FAILURE (SAFETY GUIDE 24, 3/23/72) 
 
CONF0RMANCE
 
The assumptions of the regulatory position of Regulatory Guide 1.24 are used without exception
 
in the analysis of the potential radiological consequences of the failure of a radioactive gas
 
storage tank in section 15.3. 
 
FNP-FSAR-3A
 
3A-1.25-1 REV 21  5/08 Regulatory Guide 1.25 - ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING
 
ACCIDENT IN THE FUEL HANDLING AND STORAGE FACILITY
 
FOR BOILING AND PRESSURIZED WATER REACTORS (SAFETY GUIDE 25, 3/23/72) 
 
CONFORMANCE
 
The assumptions of the regulatory position of Regulatory Guide 1.25 as used in the analysis of
 
the potential radiological consequences of a fuel handling accident, are discussed in subsection
 
15.4.5. 
 
FNP-FSAR-3A
 
3A-1.26-1 REV 21  5/08 Regulatory Guide 1.26 - QUALITY GROUP CLASSIFICATION AND STANDARDS (SAFETY GUIDE 26, 3/23/72) 
 
CONFORMANCE
 
Equipment classification and code requirements are given in subsection 3.2.2. The
 
classification system of ANSI "Nuclear Safety Criteria for the Design of Stationary Pressurized
 
Water Reactor Plants," August 1970 draft is an alternate acceptable method of meeting the
 
intent of Regulatory Guide 1.26. Since there was no established commercial standard for
 
pumps at the time of the license application, ASME Boiler and Pressure Vessel Code
 
Subsection VIII, Division 1, and ANSI B31.1.0, Power Piping, represented related available
 
standards that, while intended for other applications, were used for guidance and
 
recommendations in determining quality group D pump construction requirements, such as
 
allowable stresses, steel casting quality factors, wall thicknesses, materials compatibility and
 
specifications, temperature pressure environment restrictions, fittings, flanges, gaskets, bolting, and installation procedures. 
 
FNP-FSAR-3A
 
3A-1.27-1 REV 21  5/08 Regulatory Guide 1.27 - ULTIMATE HEAT SINK (Rev. 2, 1/76) 
 
CONFORMANCE
 
The ultimate heat sink meets the recommendations of Regulatory Guide 1.27. Compliance is
 
discussed fully in subsection 9.2.5. 
 
FNP-FSAR-3A
 
3A-1.28-1 REV 21  5/08
[HISTORICAL] [Regulatory Guide 1.28 - QUALITY ASSURANCE PROGRAM REQUIREMENTS (DESIGN AND CONSTRUCTION) (SAFETY GUIDE 28, 6/7/72)
CONFORMANCE This topic is addressed in section 17.1.] 
 
FNP-FSAR-3A
 
3A-1.29-1 REV 21  5/08 Regulatory Guide 1.29 - SEISMIC DESIGN CLASSIFICATION (Rev. 1, 8/73) 
 
CONFORMANCE
 
Design of structures, systems, and components complies with the recommendations of
 
Regulatory Guide 1.29. Seismic design classification is discussed in subsection 3.2.1, Seismic
 
Classification. 
 
FNP-FSAR-3A
 
3A-1.30-1 REV 21  5/08 Regulatory Guide 1.30 - QUALITY ASSURANCE REQUIREMENTS FOR THE INSTALLATION, INSPECTION, AND TESTING OF
 
INSTRUMENTATION AND ELECTRIC EQUIPMENT (SAFETY GUIDE 30, 8/11/72) 
 
CONFORMANCE
  [HISTORICAL] [This topic is addressed in chapter 17.]
Regulatory Guide 1.30 provided NRC endorsement of ANSI N45.2.4 (IEEE 336-1971). The SNC Quality Assurance Topical Report (QATR) is based on NQA-1-1994 which incorporates IEEE 336-1985. Accordingly, the quality assurance requirements for the installation, inspection, and testing of instrumentation and electric equipment are described in the QATR.
 
FNP-FSAR-3A
 
3A-1.31-1 REV 21  5/08 Regulatory Guide 1.31 - CONTROL OF STAINLESS STEEL WELDING (Rev. 1, 6/73) 
 
CONFORMANCE
 
For Westinghouse scope of supply Farley Nuclear Plant conforms to the intent of Regulatory
 
Guide 1.31 as discussed in Subsection 5.2.5.5, paragraphs 3 through 7, and Subsection
 
5.2.5.7. The tests described in these sections are in accordance with the requirements of
 
ASME III, NB2430. In general, production welds were not checked for delta ferrite
 
measurements, but the combination of specifying ferrite content in material procurement
 
together with ASME Code required examinations provide assurance of the integrity of stainless
 
steel welds. 
 
Outside of Westinghouse scope of supply, conformance is as follows: 
: 1. As an equivalent alternate method to Regulatory Guide 1.31 for the prevention of potential fissures in austenitic stainless steel welds for service temperatures up
 
to 650°F, all welding material pertaining to AWS classification ER308L or E308L
 
has a ferrite content within the range of 8 to 25 percent, and all welding material
 
pertaining to AWS classification ER309 or E309 has a ferrite content within the
 
range of 5 to 15 percent. For bare electrode, rod, or wire filler metal used with
 
gas metal arc welding or gas tungsten arc welding processes, the chemical
 
analysis was performed on the electrode, rod, wire, or consumable insert, or an
 
undiluted weld deposit made with the bare filler materials. For all other
 
processes and filler metal, the chemical analysis was performed on an undiluted
 
weld deposit. The delta ferrite content was determined by analysis of each heat
 
or lot as applicable as required by the ASME Boiler and Pressure Vessel Code
 
Subsection III, Subarticle NB-2400.
: 2. Control of the chemical analysis of the weld filler metal in this range ensured adequate ferrite in the weld deposit in order to prevent fissuring in austenitic
 
stainless steel welds. This made it unnecessary to make magnetic
 
measurements of welding procedure samples or production welds. Magnetic
 
measurements were subject to mass effects and were unreliable on groove
 
welds such as root passes on piping where the potential for fissuring may be
 
highest. At best, the magnetic measurements on actual welds were only
 
semiquantitative and are subject to a variety of interpretations. The details of a
 
welding procedure have minor effects on the quantity of ferrite in a weld, but the
 
degree of variation was well within the margin of error of measurement. 
: 3. Meaningful control of ferrite was accomplished by control of weld filler metal based on chemical analysis with an adequate margin above the minimum
 
required to prevent fissuring under all conditions. There was no technical basis
 
to control the upper limit of ferrite to 15 percent because the more ferrite that
 
exists up to the point that it becomes the continuous phase (about 40 percent),
the more resistant the welds are to fissuring. For example, type 312 electrodes, nominally 29 percent chromium and 9 percent nickel, are noted for excellent
 
resistance to fissuring, and the ferrite content is about 30 percent. Many
 
austenitic stainless steel castings, particularly grades CF3A, CF8A, CF3MA and
 
CF8MA, purposely contain up to 30 percent ferrite to prevent fissures, hot tears FNP-FSAR-3A
 
3A-1.31-2 REV 21  5/08 and shrinkage. Weld metal of similar composition and ferrite content is also desirable.
: 4. The restriction of using only one heat of weld rod in a particular joint presented significant problems. Where a number of weldments require 600-700 lb of rod of
 
one heat of weld per joint, setting aside sufficient rods from a separate heat for
 
each joint was not considered necessary; chemical analysis of weld material
 
provided required assurance for integrity of the welds.
 
FNP-FSAR-3A
 
3A-1.32-1 REV 21  5/08 Regulatory Guide 1.32 - USE OF IEEE STD 308-1971, "CRITERIA FOR CLASS IE ELECTRICAL SYSTEMS FOR NUCLEAR POWER
 
GENERATING STATIONS" (SAFETY GUIDE 32, 8/11/72) 
 
CONFORMANCE
: 1. Design of the electric power system complies with IEEE Standard 308-1971.
The degree of conformance is discussed in subsection 8.3.1.2(e). 
: 2. Availability of power from the transmission network is of preferred design as recommended by Regulatory Guide 1.32. The design is discussed in subsection
 
8.2.1.3, Compliance with NRC Design Criteria. 
: 3. Battery charger capacity complies with that recommended by Regulatory Guide 1.32 and is discussed in subsection 8.3.2.1.2.
 
FNP-FSAR-3A
 
3A-1.33-1 REV 21  5/08 Regulatory Guide 1.33 - QUALITY ASSURANCE PROGRAM REQUIREMENTS (OPERATION) (SAFETY GUIDE 33, 11/3/72) 
 
CONFORMANCE
  [HISTORICAL] [During original plant licensing, a 24-month review process for all safety-related and security procedures was developed to meet the intent of Safety Guide 33 and ANSI N18.7-1972. Since the procedural process has now matured and adequate progr ams to assure procedural revisions consistent with plant design, operational, and regulatory requiremen ts are in place, this original commitment has been modified to require biennial quality assu rance audits of the procedural development and maintenance program utilizing a representative samp ling process. Therefore, the 24-month review process is no longer required.
Conduct of operations of the Safety Review Board, including audits performed under the cognizance of the Safety Review Board, meet the requirements of ANSI N18.7-1976, Section 4, Review and Audit.
Regulatory Guide 1.33, Section 4 provides that the following program elements should be audited at the indicated frequencies:  the results of actions taken to correct deficiencies that affect nuclear safety and occur in facility equipment, structures, systems, or method of operation - at least once per 6 months; the conformance of facility operation to provisions contained within the Technical Specifications and applicable licensing conditions - at least once p er 12 months; and the performance, training, and qualifications of the facility staff - at least once p er 12 months. Audit frequencies for each of these program elements are now established as at least once per 24 months.]
The SNC Quality Assurance Topical Report (QATR) is based on ASME NQA-1-1994 and incorporates the applicable requirements of ANSI N18.7-1976. Accordingly, SNC complies with the applicable requirements of ANSI N18.7-1976 via compliance with the QATR without an explicit (or implied) commitment to ANSI N18.7-1976.
 
FNP-FSAR-3A
 
3A-1.34-1 REV 21  5/08 Regulatory Guide 1.34 - CONTROL OF ELECTROSLAG WELD PROPERTIES (Rev. 0, 12/28/72) 
 
CONFORMANCE
 
Regulatory Guide 1.34 is not applicable to the Farley Nuclear Plant, inasmuch as electroslag
 
welds were not used in the fabrication of core support structures nor in Class 1 and 2 vessels
 
and components. 
 
FNP-FSAR-3A
 
3A-1.35-1 REV 21  5/08
[HISTORICAL] [Regulatory Guide 1.35 - INSERVICE INSPECTION OF UNGROUTED TENDONS IN PRESTRESSED CONCRETE CONTAINMENT STRUCTURES (Rev. 2, 1/76)  CONFORMANCE The containment tendon surveillance program for th e Farley containment prestressing system complies with Regulatory Guide 1.35, "Inservice Inspecti on of Ungrouted Tendons in Prestressed Concrete Containment Structures," Rev. 2. The cont ainment tendon surveillance program is discussed in subsection 3.8.1.
]   
 
FNP-FSAR-3A
 
3A-1.36-1 REV 21  5/08 Regulatory Guide 1.36 - NONMETALLIC THERMAL INSULATION FOR AUSTENITIC STAINLESS STEEL (Rev. 0, 2/23/73) 
 
CONFORMANCE
 
Regulatory Guide 1.36 is not applicable for components within the reactor coolant pressure
 
boundary, since only stainless steel metal reflective insulation was used on reactor coolant
 
pressure boundary austenitic stainless steel piping and equipment. 
 
For austenitic stainless steel piping and components outside the reactor coolant pressure
 
boundary, Regulatory Guide 1.36 was followed. 
 
FNP-FSAR-3A
 
3A-1.37-1 REV 21  5/08 Regulatory Guide 1.37 - QUALITY ASSURANCE REQUIREMENTS FOR CLEANING OF FLUID SYSTEMS AND ASSOCIATED COMPONENTS OF
 
WATER COOLED NUCLEAR POWER PLANTS (Rev. 0, 3/16/73) 
 
CONFORMANCE
  [HISTORICAL] [Preoperational cleaning and layup and associated activities involving the cleanliness of safety-related fluid systems were performed in a ccordance with this Regulat ory Guide as discussed in chapter 17.]
Regulatory Guide 1.37 provides NRC endorsement of ANSI N45.2.1. The SNC Quality Assurance Topical Report (QATR) is based on ASME NQA-1-1994 which incorporates the requirements of ANSI N45.2.1. Accordingly, quality assurance requirements for cleaning of fluid systems and associated components are described in the SNC QATR.
 
FNP-FSAR-3A
 
3A-1.38-1 REV 21  5/08 Regulatory Guide 1.38 - QUALITY ASSURANCE REQUIREMENTS FOR PACKAGING, SHIPPING, RECEIVING, STORAGE, AND HANDLING OF ITEMS
 
FOR WATER COOLED NUCLEAR POWER PLANTS (ANSI
 
N45.2.2-1972) 
 
CONFORMANCE
  [HISTORICAL] [Conformance with applicable sections of ANSI N45.2.2-1972 and hence with this Regulatory Guide is discussed in sections 17.1 and 17.2.
Certification of involved personnel in accordan ce with a recommended practice such as SNT-TC-1A-1968 was considered to be excessive. Personnel involved in conducting activities governed by this ANSI standard were properly qualified by reason of experience and training.
Exception is taken to paragraph 3 of the Regulatory Guid
: e. Tape used to secure caps to stainless steel pipe was "essentially chloride free and approved by the purchaser prior to use." Po lyethylene was used in combination with a wooden plug to protect all nonflanged stainless steel pipe openings larger than 2 in.]
The SNC Quality Assurance Topical Report (QATR) is based on ASME NQA-1-1994 which incorporates the requirements of ANSI N45.2.2. Accordingly, quality assurance requirements for packaging, shipping, receiving, storage, and handling are described in the SNC QATR.
 
FNP-FSAR-3A
 
3A-1.39-1 REV 21  5/08 Regulatory Guide 1.39 - HOUSEKEEPING REQUIREMENTS FOR WATER COOLED NUCLEAR POWER PLANTS (Rev. 0, 3/16/73) 
 
CONFORMANCE
  [HISTORICAL] [The housekeeping requirements during the construction phase were established prior to issuance of Regulatory Guide 1.39. These requirem ents were structured to meet the standards of Nuclear Electric Insurance Limited (NEIL) and the Occupational Safety and Health Act (OSHA).
Housekeeping activities during operation meet the re quirements of ANSI Standard N45.2.3-1973, except with regard to the general fire protection guidelin es of subdivision 3.2.3. The FNP fire protection program is described in Appendix 9B.]
The SNC Quality Assurance Topical Report (QATR) is based on ASME NQA-1-1994 which incorporates the requirements of ANSI N45.2.3. Accordingly, housekeeping requirements are described in the SNC QATR.
 
FNP-FSAR-3A
 
3A-1.40-1 REV 21  5/08 Regulatory Guide 1.40 - QUALIFICATION TESTS OF CONTINUOUS DUTY MOTORS INSTALLED INSIDE THE CONTAINMENT OF WATER COOLED
 
NUCLEAR POWER PLANTS (Rev. 0, 3/16/73) 
 
CONFORMANCE
 
Continuous-duty, Class 1 (a) motors installed within the containment were purchased on the basis that prototypes were type-tested to the requirements of IEEE Standard 334-1971, "IEEE
 
Trial-Use Guide for Type Tests of Continuous-Duty Class I Motors Installed Inside the
 
Containment of Nuclear Power Generating Stati ons."  Qualification reports from the vendors covering the tests were reviewed for full compliance. The tests covered, as far as practicable, any auxiliary equipment that was a part of the complete motor assembly. There was no special
 
condition to meet for the auxiliary equipment, as was exampled by Regulatory Position 2. 
: a. In later IEEE standards, called Class 1E.
FNP-FSAR-3A
 
3A-1.41-1 REV 21  5/08 Regulatory Guide 1.41 - PREOPERATIONAL TESTING OF REDUNDANT ONSITE ELECTRIC POWER SYSTEMS TO VERIFY PROPER LOAD
 
GROUP ASSIGNMENTS (Rev. 0, 3/16/73) 
 
CONFORMANCE
 
This topic is discussed in subsection 14.1.3 and also in the conformance to Regulatory Guide
 
1.79. 
 
FNP-FSAR-3A
 
3A-1.42-1 REV 21  5/08 Regulatory Guide 1.42 - INTERIM LICENSING POLICY ON AS LOW AS PRACTICABLE FOR GASEOUS RADIOIODINE RELEASES FROM LIGHT
 
WATER COOLED NUCLEAR POWER REACTORS (Rev. 1, 3/74)
 
CONFORMANCE
 
This topic is addressed in subsection 11.3.6. 
 
FNP-FSAR-3A
 
3A-1.43-1 REV 21  5/08 Regulatory Guide 1.43 - CONTROL OF STAINLESS STEEL WELD CLADDING OF LOW-ALLOY STEEL COMPONENTS (Rev. 0, 5/73) 
 
The Farley Units are in compliance with Regulatory Guide 1.43 as demonstrated by the
 
following: 
 
The reactor vessel bottom head, intermediate, and lower shell courses were fabricated from
 
A-533 Grade B Class 1 plate material and clad by use of the 3-wire submerged arc process. 
 
The closure head is fabricated from an SA-508 Grade 3 Class 1 forging and is strip clad with the
 
electroslag process and manual welding as dictated by the geometry. These materials and
 
cladding processes are not restricted by Regulatory Guide 1.43 and, consequently, are in
 
compliance with the Guide. 
 
The closure head flange, vessel flange, nozzles, lower transition ring, and upper shell course
 
were fabricated from A-508 Class 2 forging material and clad by use of either the single-wire or
 
3-wire submerged arc process or the manual metal arc process or combinations of these
 
processes. Since these welding methods are not considered as high heat input processes, the
 
above components are in compliance with the Guide. 
 
Stainless steel weld cladding was applied to the steam generator channel head surface in
 
contact with primary coolant. The head, including the nozzles and manway opening, are
 
integrally forged SA-508 Class 3 material. The head hemispherical surface was clad by the
 
shielded metal arc process or strip cladding submerged arc process with controlled dilution of
 
the deposit, and the channel head nozzles and manway openings were clad by the shielded
 
metal arc or tungsten arc welding process. Both processes are low heat input techniques. The
 
material and the weld processes are not restricted by the Guide. Therefore, the steam
 
generator stainless steel cladding meets the recommendations of the Guide. 
 
Stainless steel weld cladding was applied to the pressurizer shell courses, heads, spray nozzle, and manway opening surfaces in contact with primary coolant. The pressurizer shell and upper
 
and lower heads are constructed of SA-533 Grade A Class 2 material. The shell courses are
 
clad by the plasma process. The heads are clad by the 2-wire series submerged arc process
 
with controlled dilution of the deposit. 
 
The pressurizer lower head nozzle and the upper head manway openings are constructed of
 
SA-508 Class 2 material, use of which is restricted by the Guide. 
 
However, the manual metal arc process, a low heat input process, was used to apply the weld
 
cladding on these surfaces. Because of the low heat input processes and/or materials used for
 
the stainless steel weld cladding, the intent of Regulatory Guide 1.43 is met for the pressurizer. 
 
FNP-FSAR-3A
 
3A-1.44-1 REV 21  5/08 Regulatory Guide 1.44 - CONTROL OF THE USE OF SENSITIZED STAINLESS STEEL (Rev. 0, 5/73) 
 
CONFORMANCE
 
The Farley Nuclear Plant meets the intent of Regulatory Guide 1.44. It is Westinghouse
 
practice to use processing, packaging and shipping controls, and preoperational cleaning to
 
preclude adverse effects of exposure to contaminants on all stainless steel materials as
 
recommended by Position 1 of the Regulatory Guide.
 
All austenitic stainless steel starting materials were procured from the raw material suppliers in
 
the final heat-treated condition required by the respective ASME Code subsection II material
 
specification for the particular type or grade of alloy, as recommended by the Guide, Regulatory
 
Position 2. 
 
Westinghouse met the intent of the remaining positions of Regulatory Guide 1.44 by using the
 
most practicable and conservative methods and techniques to avoid partial or local severe
 
sensitization. Methods and material techniques that were employed to avoid partial or local
 
severe sensitization are discussed in subsection 5.2.5. 
 
Moreover, Westinghouse technical background and service experience, as detailed in WCAP-
 
7735 (reference 7 of section 5.2), support the conclusion that serious intergranular attack of
 
sensitized stainless steel is unlikely in Westinghouse PWR nuclear steam supply systems, since
 
water chemistry and contamination are kept under control. 
 
FNP-FSAR-3A
 
3A-1.45-1 REV 21  5/08 Regulatory Guide 1.45 - REACTOR COOLANT PRESSURE BOUNDARY LEAK DETECTION SYSTEMS (Rev. 0, 5/73) 
 
CONFORMANCE
 
The leakage detection system as described in subsection 5.2.7 meets the intent of Regulatory
 
Guide 1.45. The instrumentation available prov ides approximately 1 gal/min RCS leak detection and a response time of approximately 1 hour under conditions described in subsection 5.2.7. 
 
Calibration of the leakage detection system is performed during plant shutdown. Experience on similar equipment has shown calibration performed during plant shutdown is reliable and
 
sufficient. 
 
Although the leak detection system does not incl ude a sump monitoring system as required by the Regulatory Guide, an acceptable alternative is the plant's condensate measuring system, as documented in NUREG-75/034, section 5.6.
 
Farley's reactor coolant pressure boundary is designed to withstand the design basis
 
earthquake (DBE) loads; therefore, no failure in t he system following such an event is expected.
The present leakage detection system in FNP forewarns the operator of minor leakages that
 
may develop during normal operation; however, the system neither performs a safeguard function nor is required to operate during or after a seismic event. Therefore, the system is not
 
seismically qualified. 
 
FNP-FSAR-3A
 
3A-1.46-1 REV 21  5/08 Regulatory Guide 1.46 - PROTECTION AGAINST PIPE WHIP INSIDE CONTAINMENT (Rev. 0, 5/73) 
 
CONFORMANCE
 
Conformance is discussed in section 3.6. 
 
FNP-FSAR-3A
 
3A-1.47-1 REV 21  5/08 Regulatory Guide 1.47 - BYPASSED AND INOPERABLE STATUS INDICATION FOR NUCLEAR POWER PLANT SAFETY SYSTEMS (Rev. 0, 5/73) 
 
CONFORMANCE
 
Extensive control room indication is employed in the protection and engineered safety features systems to provide status and availability in formation to plant personnel, as discussed in chapter 7. Specifically, the existing design of the main control board provides adequate
 
information to the operator on the component level. Component level instruments associated
 
with the same system are grouped together on the main control board. This information is
 
applied by the trained operator to establish the effects of component status, as it relates to
 
overall system status and availability. 
 
A system to provide indication on a system le vel was not a requirement when the plant was designed, and the specific recommendations of Regulatory Guide 1.47, particularly the
 
requirement to provide automatic bypass indicati on at the system level, was not anticipated at the time of the design. 
 
A manually operated light display board was subsequently installed in the main control room on
 
a single panel to show those engineered safety features that are bypassed by deliberate
 
operator action. These systems are train oriented and arranged on the board to indicate clearly
 
which system in a single train is bypassed. The systems displayed on this board are as follows: 
: 1. Containment spray. 
: 2. RHR. 
: 3. High-head safety injection. 
: 4. Component cooling water. 
: 5. Auxiliary feedwater. 
: 6. Post-LOCA combustible gas control. 
: 7. Main steam line isolation. 
: 8. Containment isolation. 
: 9. Safety-related heating, ventilation, and air conditioning (HVAC). 
: a. Penetration room HVAC. 
: b. Control room emergency HVAC. 
: c. Containment cooling. 
: d. Spent fuel pool. 
: 10. Service water.
FNP-FSAR-3A
 
3A-1.47-2 REV 21  5/08
: 11. Emergency power. 
 
These features coupled with the administrative procedures outlined in the technical specifications provide adequate insurance that systems important to safety are not inadvertently bypassed. 
 
FNP-FSAR-3A
 
3A-1.48-1 REV 21  5/08 Regulatory Guide 1.48 - DESIGN LIMITS AND LOADING COMBINATIONS FOR SEISMIC CATEGORY I FLUID SYSTEM COMPONENTS (Rev. 0, 5/73) 
 
CONFORMANCE
 
Sections 3.7, 3.9, and 5.2 include the design criteria for Seismic Category I fluid system
 
components. 
 
FNP-FSAR-3A
 
3A-1.49-1 REV 21  5/08 Regulatory Guide 1.49 - POWER LEVELS OF NUCLEAR POWER PLANTS (Rev. 1, 12/73) 
 
CONFORMANCE
 
The rated thermal power (i.e., core thermal power) level of the Farley Nuclear Plant is 2775
 
MWt, which is below the maximum limit of 3800 MWt recommended in Regulatory Guide 1.49. 
 
FNP-FSAR-3A
 
3A-1.50-1 REV 21  5/08 Regulatory Guide 1.50 - CONTROL OF PREHEAT TEMPERATURE FOR WELDING OF LOW-ALLOY STEEL (Rev. 0, 5/73) 
 
CONFORMANCE
 
Regulatory Guide 1.50 describes an acceptable method of implementing the requirements of 10
 
CFR 50 in regard to the control of welding for low-alloy steel components during initial
 
fabrication. 
 
The Farley Nuclear Plant conforms to Regulatory Guide 1.50 in the following manner: 
: 1. Preheat for welding of low-alloy steel was controlled in accordance with the regulatory position of Regulatory Guide 1.50 except as described in sections 2
 
through 4 below. 
: 2. The position of Regulatory Guide Part C, Paragraph l.a, was met when impact testing, in accordance with ASME Boiler and Pressure Vessel Code, Subsection
 
III, Subarticle 2300, was required. When impact testing was not required, specifying a maximum interpass temperature in the welding procedure was not
 
necessary in order to ensure that the other required mechanical properties of the
 
weld were met. 
: 3. Compliance with Regulatory Guide Part C, Paragraph 2, was required for pressure vessels with nominal thicknesses greater than 1 in. Maintaining
 
preheat after welding until postweld heat treatment (PWHT) was not required for
 
thinner sections, since experience indicated that delayed cracking in the weld or
 
heat- affected zone (HAZ) was not a problem. 
: 4. Usage of low-alloy steel in piping, pumps, and valves was minimal and was normally limited to Class 3 construction. When low-alloy steel piping, pumps, and valves were used, preheat was maintained until welding was complete, but
 
not until PWHT was performed. 
 
FNP-FSAR-3A
 
3A-1.51-1 REV 21  5/08 Regulatory Guide 1.51 - INSERVICE INSPECTION OF ASME CODE CLASS 2 AND 3 NUCLEAR POWER PLANT COMPONENTS (Rev. 0, 5/73) 
 
CONFORMANCE
 
This topic is discussed in subsection 5.2.8. 
 
FNP-FSAR-3A
 
3A-1.52-1 REV 21  5/08 Regulatory Guide 1.52 - DESIGN, TESTING, AND MAINTENANCE CRITERIA FOR ATMOSPHERE CLEANUP SYSTEM AIR FILTRATION AND
 
ADSORPTION UNITS OF LIGHT-WATER-COOLED NUCLEAR
 
POWER PLANTS (Rev. 0, 6/73) 
 
DESIGN, INSPECTION, AND TESTING CRITERIA FOR AIR FILTRATION AND ADSORPTION UNITS OF POST ACCIDENT
 
ENGINEERED-SAFETY-FEATURE ATMOSPHERE CLEANUP
 
SYSTEMS IN LIGHT-WATER-COOLED NUCLEAR POWER
 
PLANTS (Rev. 3, 6/01) 
 
CONFORMANCE
 
Regulatory Guide 1.52 provides detailed information on design, testing, and maintenance for air
 
filtration and adsorption units of atmosphere cleanup systems in light-water-cooled nuclear
 
power plants. Conformance with this guide is not complete because at the time of the design of
 
the atmosphere cleanup systems, such as penetra tion room filtration system, the guide ORNL-NSIC-65, "Design, Construction and Testing of High Efficiency Air Filtration Systems for Nuclear
 
Application," was the only available design reference. However, each engineered safety feature
 
air filtration system is designed to safely and reliably mitigate the consequences of the
 
postulated accidents. 
 
The Farley Nuclear Plant conforms to the Regulatory Guide with the following exceptions: 
: 1. Penetration Room Filtration Unit
 
Reference Paragraph C.2.a of the Regulatory Guide. No demister was provided because the unit is located outside the containment and no entrained water
 
droplets are anticipated. No high efficiency particulate air (HEPA) filters are
 
provided downstream of the charcoal sinc e radioactive fines carryover is very unlikely. This is true because the charcoal trays are pressure tested at high
 
velocity in the manufacturer's shop prior to delivery, thereby removing fines. 
 
Also, during system operation, air is passing through the charcoal at a very low
 
velocity. 
: 2. Control Room Filtration, Recirculation, and Pressurization Units
 
Reference Paragraph C.2.a of the Regulatory Guide. No demister was provided because the unit is located outside of any high humidity area and no entrained
 
water droplets are anticipated. No electric heater was provided for the filtration
 
and recirculation units since humidity control is unnecessary. No HEPA filters
 
are provided downstream of the charcoal si nce radioactive fines carryover is very unlikely. This is true because the charcoal trays are pressure tested at high
 
velocity in the manufacturer's shop prior to delivery, thereby removing fines. 
 
Also, during system operation, air is passing through the charcoal at a very low
 
velocity. 
 
FNP-FSAR-3A
 
3A-1.52-2 REV 21  5/08  3. All Engineered Safety Feature Filtration Units
 
Reference Paragraph C.2.b of the Regulatory Guide. No physical separation was provided for each filtration unit since there are no units located in areas
 
where missiles are postulated.
 
As discussed above, each engineered safety feature (ESF) air filtration system is designed to
 
safely and reliably mitigate the consequences of postulated accidents and includes the majority
 
of the recommendations of Regulatory Guide 1.52, Revision 0. The testing of these systems is
 
performed in accordance with the Plant Technical Specifications. Through license amendments
 
adopted since the Technical Specifications were originally issued, surveillance testing
 
associated with the ESF filtration systems is now conducted in accordance with the
 
recommendations of Regulatory Guide 1.52, Revision 3 as discussed below. 
 
Reference Paragraphs C.6.3 and C.6.4 of the Regulatory Guide. These paragraphs recommend
 
an in-place leak test removal efficiency of 99.95 percent for HEPA filters and charcoal adsorbers in
 
filter systems. As noted above, design and construction were completed prior to issuance of this
 
guide, and conformance with this guide is not complete. Therefore, the in-place leak tests on
 
HEPA filters and charcoal adsorbers are performed to demonstrate a 99.5 percent removal
 
efficiency. These efficiencies were reviewed and approved by the NRC in License Amendment
 
No. 46 for Unit 1 and License Amendment No. 37 for Unit 2.
 
Regulatory Guide 1.52, Revision 3 references ASME N510-1989 for testing air cleaning
 
systems for Nuclear Power Plants. FNP has adopted ASME N510-1989 with errata dated
 
January 1991 as the appropriate standard for guidance in testing ESF filtration units. The details
 
of testing conformance to ASME N510-1989 are documented in FSAR tables 9.4-15 through
 
9.4-18. As stated in the NRC SER dated May 1, 1997, Farley Nuclear Plant is not required to
 
perform all the acceptance tests as identified in ASME N510-1989. These type tests will be
 
performed after major modification or major repair to the systems as identified in tables 9.4-15
 
through 9.4-18. Inspections following system modification or repair will be those inspections
 
required on only those components affected by the modification or repair and not the complete system.
 
FNP-FSAR-3A
 
3A-1.52-3 REV 21  5/08 Post-Accident Containment Venting Filter Unit
 
The post-accident containment venting filter unit is not an ESF filtration system. This system was designed prior to issuance of Regulatory Guide 1.52, Rev. 0, therefore, conformance with
 
this guide is not complete. Because of the time of the design and the system not being an ESF
 
filtration system, the post-accident containment venting filter unit has unique design features which do not allow strict application of the design, maintenance, and testing requirements of this
 
Regulatory Guide. Maintenance and periodic testing will be provided with guidance from ASME
 
N510-1989 (see Section 6.2.5.4.2 for description of testing program). Design conformance is
 
summarized as follows:
 
Reference Paragraph C.2.a of Regulatory Guide 1.52, Rev. 0. The post-accident containment
 
venting filter unit is provided with an iodine adsorber and a HEPA filter downstream of the
 
adsorber. The system components and arrangement are described in FSAR Section 6.2.5, and
 
shown on drawings D-175019 and D-205019. No credit was taken for the system as an ESF
 
filtration system in DBA analyses; therefore, redundant filter units are not installed. No demister
 
was provided because the unit is located outside the containment and no entrained water
 
droplets are anticipated. No prefilters or HEPA filters before the adsorbers are provided since
 
this is a standby system and, if utilized, it is expected to have low operating time. No fan was
 
provided since motive force is provided by containment pressure. No electric heater was
 
provided since humidity is not to be controlled.
 
FNP-FSAR-3A
 
3A-1.53-1 REV 21  5/08 Regulatory Guide 1.53 - APPLICATION OF THE SINGLE FAILURE CRITERION TO NUCLEAR POWER PLANT PROTECTION SYSTEMS (Rev. 0, 6/73) 
 
CONFORMANCE
 
The principles described in IEEE Trial Use Guide 379-72 were used in the design of the
 
Westinghouse protection system. Although this guide had not been issued at the time of the
 
design of the Farley Plant, the system does comply with the intent of this guide and the
 
additional requirements of Regulatory Guide 1.53. The formal analyses required by the trial use
 
guide have not been documented exactly as outlined although parts of such analyses are
 
published in various documents, such as references 1, 2, and 3. Failure analysis results are
 
given in tables 7.3-5 through 7.3-15. Failure analysis of the plant cooling water system is given
 
in subsection 9.2.1.
: 1. W. C. Gangloff, "An Evaluation of Anticipated Operational Transients in Westinghouse Pressurized Water Reactors," WCAP-7486 , May 1971.
: 2. W. C. Gangloff and W. D. Loftus, "An Evaluation of Solid State Logic Reactor Protection in Anticipated Transients," WCAP-7706 , July 1971. 
: 3. "Anticipated Transients Without Reactor Trip in Westinghouse Pressurized Water Reactors," WCAP-8096 April 1973. 
 
FNP-FSAR-3A
 
3A-1.54-1 REV 21  5/08 Regulatory Guide 1.54 - QUALITY ASSURANCE REQUIREMENTS FOR PROTECTIVE COATINGS APPLIED TO WATER COOLED NUCLEAR POWER
 
PLANTS (Rev. 0, 6/73) 
 
CONFORMANCE
 
Regulatory Guide 1.54 (June 1973) and related ANSI Standard N101.4 (November 1972)
 
postdate the construction permit for the Farley Nuclear Plant, which was issued in August 1972.
 
Consequently, these requirements were not available for application to the nuclear steam
 
supply system (NSSS) equipment for the FNP. 
 
For Westinghouse scope of supply equipment, however, a process specification was applied to
 
the NSSS equipment. This required that protective coatings for use on system components in
 
the reactor containment be demonstrated to withstand the design basis accident conditions and
 
meet all the criteria given in ANSI Proposed Standard N-101.2-1971, "Protective Coatings (Paints) For Light Water Nuclear Reactor Containment Facilities." 
 
Regulatory Guide 1.54 and ANSI N101.4-1972 postdate the construction permit. Therefore, specifications and procedures relative to coatings on Category I structures did not reference the
 
ANSI Standard. However, specifications and quality procedures used to control the application
 
processes for these structures are such that they ensured proper application of these coatings.
(See subsection 3.8.1.6.6.) 
 
The protective coating systems specified for Seismic Category I structures are discussed in
 
subsection 3.8.1.6.6, Interior Coating Systems. 
 
FNP-FSAR-3A
 
3A-1.55-1 REV 21  5/08 Regulatory Guide 1.55 - CONCRETE PLACEMENT IN CATEGORY I STRUCTURES (Rev. 0, 6/73) 
 
CONFORMANCE
  [HISTORICAL] [Concrete placement in Seismic Catego ry I structures was in accordance with Regulatory Guide 1.55, "Concrete Placement in Ca tegory I Structures," except as discussed below: 
: 1. Regulatory Guide 1.55, Appendix A, Reference 11, "ACI/ASME Proposed Standard-Code for Concrete Reactor Vessels and Containments," and Reference 12, "ANSI N45.2.5-1972 (proposed) Supplementary QA Requirements for Installation, Inspection, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants," were not used since they had not received final approval by their sponsoring organizations. 
: 2. Creep tests for concrete were performed fo r the containment structure only. Loss of prestress through creep was not applicabl e to non-prestressed structures.
Concrete placement and testing are discussed in subsection 3.8.1.6.1, Reinforced Concrete.]
The requirements for concrete placement in Category I structures applicable to operation-phase activities are contained in ASME NQA-1-1994, as described in the SNC Quality Assurance Topical Report (QATR).
 
FNP-FSAR-3A
 
3A-1.56-1 REV 21  5/08 Regulatory Guide 1.56 - MAINTENANCE OF WATER PURITY IN BOILING WATER REACTORS (Rev. 0, 6/73) 
 
CONFORMANCE
 
Regulatory Guide 1.56 is not applicable to the Farley Nuclear Plant. 
 
FNP-FSAR-3A
 
3A-1.57-1 REV 21  5/08 Regulatory Guide 1.57 - DESIGN LIMITS AND LOADING COMBINATIONS FOR METAL PRIMARY REACTOR CONTAINMENT SYSTEM COMPONENTS (Rev. 0, 6/73) 
 
CONFORMANCE
 
Regulatory Guide 1.57 is not applicable to the Farley Nuclear Plant. 
 
FNP-FSAR-3A
 
3A-1.58-1 REV 21  5/08 Regulatory Guide 1.58 - QUALIFICATION OF NUCLEAR POWER PLANT INSPECTION, EXAMINATION, AND TESTING PERSONNEL (Rev. 1, 9/80) 
 
CONFORMANCE
  [HISTORICAL] [Personnel involved with examinati on of items on the site are qualified and certified in accordance with the requirements of ANSI N45.2.
6-1978. Personnel performing inspection and testing were qualified for those specific tasks on the basis of experience and specific training (education or on-the-job-training or a combination of both). Co mpliance was verified by periodic audits by quality assurance personnel. However, two exceptions have b een taken to Regulatory Guide 1.58, Revision 1, and ANSI N45.2.6-1978. A description and justification of these exceptions is presented below.
Regulatory Position C.2 of Regulatory Guide 1.58, Revision 1, endorses the 1975 edition of SNT-TC-1A as acceptable guidance for qualifications of nondestructi ve examination (NDE) personnel. In lieu of this, the version of SNT-TC-1A or other similar docum ent used for qualifying personnel to perform nondestructive inspection, examination, or testing shall be in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a, except where specific written relief has been granted by the NRC.
The document used for the qualification of such personnel shall be specifically identified in the Inservi ce Inspection Program for FNP. In addition, FNP shall supplement these requirements by replacing the "shoulds" contained in SN T-TC-1A with "shalls" where they occurred in the 1975 version. A change to th is paragraph shall be trea ted as a change to the FNP QAP in accordance with NRC SER dated March 17, 1998.
Subsection 2.3 of ANSI N45.2.6-1978 requires that the job performance of inspection, examination, and testing personnel be reevaluated at least every 3 years and that any person who has not performed inspection, examination, and testing activities in hi s qualified area for a period of 1 year shall undergo requalification in accordance with subsection 2.2.
Inspection, examination, and testing activities are inherently integrated into the FNP staff's routin e job responsibilities such that when a person holds a given position, he routinely performs those inspecti on, examination, and testing activities for which that position is responsible. Therefore, an annual demons tration of proficiency has no meaning under the FNP Quality Control Program. Moreover, the requa lification per ANSI N45.2.6-1978, paragraph 2.2, is based on the individual's education and experience.
Since each person's initia l certification was also based on that individual's accumulated education and experience, and since an individual's accumulated education and experience cannot be revoked, there is no purpose in performing the requalification exercise simply because an individual did not p erform or document an annual demonstration of proficiency in the inspection, examination, and tes ting activities for which he is certified (assuming, of course, that the individual's job performance of inspection, examination, and testing activities remained satisfactory). For these reasons an annual demonstrati on of proficiency in inspection, examination, and testing activities is fruitless and exception is taken to this aspect of ANSI N45.2.6-1978. The present practice, which is expected to be continued, of conducting job performance evaluations as a basis for recertification for inspection, examination, and tes ting personnel on a 2-year cycle satisfies the ANSI N45.2.6-1978 requirement that these evaluations be conducted at periodic intervals not to exceed 3 years.
On September 9, 1996, the NRC amended its regul ations to incorporate by reference the 1992 edition with the 1992 addenda of Subsections IWE and IWL of S ection XI, Division 1, of the ASME Boiler and Pressure Vessel Code with specified limitations in 10 CFR 50.55a. The new rules require certain containment liner and concrete inspections/examinations to be performed prior to September 9, 2001 and to be repeated on a regular basis thereafter. C ontainment repair and replacement requirements of the new rules including preservice examinations after repai r or replacement were effective on September 9, FNP-FSAR-3A
 
3A-1.58-2 REV 21  5/08 1996. Relief from this effective date until March 15, 1997 was requested by FNP. The 1992 edition with 1992 addenda of Section XI requires personnel perf orming NDE examinations to be qualified and certified using a written practice prepared in acco rdance with ANSI/ASNT CP-189. However, current certification based on SNT-TC-1A remains valid until recertification is required.]
The SNC Quality Assurance Topical Report (QATR) is based on ASME NQA-1-1994 which incorporates the requirements of ANSI N45.2.6. Accordingly, the requirements for qualification of inspection, examination, and testing personnel are described in the QATR.
 
FNP-FSAR-3A
 
3A-1.59-1 REV 21  5/08 Regulatory Guide 1.59 - DESIGN BASIS FLOODS FOR NUCLEAR POWER PLANTS (Rev. 0, 8/73) 
 
CONFORMANCE
 
The design of Category I structures for the protection of safety-related equipment from external
 
flooding as discussed in subsection 3.4, Water Level (Flood) Design Criteria, complies with
 
Regulatory Guide 1.59. 
 
The material in the FSAR complies with Regulatory Position No. 1 of Regulatory Guide 1.59 as
 
detailed in Appendix A and explained below. 
 
A.1 Introduction
 
No comment. 
 
A.2 Probable Maximum Flood (PMF)
 
Flood design is addressed in subsections 2.4.2.2 and 2.4.3.4. 
 
A.3 Hydrologic Characteristics
 
A topographic map of drainage basin showing sub-basins and isohyetal pattern of PMP is shown on figures 2.4-3 and 2.4-12. A list of upstream river control
 
structures is shown in figure 2.4-14. The historical flood profiles for 1929 and
 
1961 floods are shown on figure 2.4-10. Major storms and resulting floods were
 
considered by the Corps of Engineers in the study for the spillway design flood
 
for Walter F. George Project. 
 
A.4 Flood Hydrograph Analysis
 
Analysis of observed hydrographs was done by the Corps of Engineers for Walter F. George Project. The results of this study are summarized in
 
subsections 2.4.3.2 and 2.4.3.3. 
 
A.5 Precipitation Losses and Base Flow
 
Precipitation losses are given in subsection 2.4.3.2. A base flow of approximately 2.5 ft 3/s/mi 2 drainage area was added to obtain the total discharge hydrograph from each of the inflow areas. 
 
A.6 Runoff Model
 
The analysis of rainfall runoff records was done by the Corps of Engineers for the Walter F. George Project. The runoff model used for FNP is given in subsection
 
2.4.3.3. 
 
FNP-FSAR-3A
 
3A-1.59-2 REV 21  5/08  A.7 Probable Maximum Precipitation Estimate
 
The results of the PMP estimates are given in subsection 2.4.3.1. The adjustment factors are shown on figure 2.4-2 and the isohyetal map used in
 
study is shown on figure 2.4-3. 
 
A.9 PMF Hydrograph Estimates
 
The antecedent reservoir level for Walter F. George Dam was taken at elevation 185, and the induced surcharge envelope shown on figure 2.4-65 was used in
 
routing by the dam. As the highest power pool during summer months is at
 
elevation 190, a routing of PMF was made using this initial pool elevation. The
 
peak flood level at the FNP was elevation 144.3, 0.1 ft higher than the elevation
 
used in chapter 2. 
 
In this report, no antecedent storm before PMF was considered. A review of the flow record of four of the highest floods indicates that an average flow of 50,000
 
ft 3/s for the fifth day after the peak would be a reasonable flow to apply to peak PMF flow as the effect of an antecedent storm. This would increase the peak
 
flow of PMF from 642,000 ft 3/s to 692,000 ft 3/s and raise the peak water surface elevation from 144.2 to 145.9, or 1.7 ft. 
 
Regulatory Guide 1.59, paragraph A.12, states that a 40-mph wind would be an acceptable postulate. However, as stated in subsection 2.4.3.6, a 50-mph wind
 
was used for wind wave activity. By using a 40-mph wind rather than 50-mph
 
wind, a reduction of the runup of 1.9 ft could be made. The net change resulting
 
from a peak flow of 692,000 ft 3/s and 40- mph wind would be a reduction of peak water surface elevation of about 0.2 ft. 
 
A.10 Seismically Induced Floods
 
This study is given in subsections 2.4.4.1, 2.4.4.2, 2.4.4.3, and 2.4.4.4. 
 
A.11 Water Level Determinations
 
The stage discharge relation at the site, as shown in figure 2.4-11, was determined by using the Corps of Engineers program HEC-2 as stated in
 
subsection 2.4.3.5. 
 
A.12 Coincident Wind Wave Activity
 
The studies made for wind waves in the river are stated in subsection 2.4.3.6. As stated under A.9 above, a 50-mph wind was used in the report when a  40-mph
 
wind would be acceptable. 
 
FNP-FSAR-3A
 
3A-1.60-1 REV 21  5/08 Regulatory Guide 1.60 - DESIGN RESPONSE SPECTRA FOR SEISMIC DESIGN OF NUCLEAR POWER PLANTS (Rev. 1, 12/73) 
 
CONFORMANCE
 
Regulatory Guide 1.60 is intended to apply to nuclear power plants docketed after April 1, 1973;
 
consequently, it was not applicable to the Farley plant. 
 
The design response spectra for seismic design for Seismic Category I structures are discussed
 
in subsection 3.7.1.1, Design Response Spectra, and subsection 3.7.1.2, Design Response
 
Spectra Derivation. 
 
FNP-FSAR-3A
 
3A-1.61-1 REV 21  5/08 Regulatory Guide 1.61 - DAMPING VALUES FOR SEISMIC DESIGN OF NUCLEAR POWER PLANTS (Rev. 0, 10/73) 
 
CONFORMANCE
 
Regulatory Guide 1.61 is intended to apply to nuclear power plants docketed after April 1, 1973; consequently, it was not considered applicable to the Farley Plant. 
 
The damping values for seismic design for Seismic Category I structures are discussed in
 
paragraph 3.7.1.3, Critical Damping Values. However, as documented in table 3.7-1, Regulatory Guide 1.61 damping values are applied in the analysis of the reactor vessel head
 
assembly structure. These values are endorsed by the NRC in the Standard Review Plan (NUREG-0800).
 
FNP-FSAR-3A
 
3A-1.62-1 REV 21  5/08 Regulatory Guide 1.62 - MANUAL INITIATION OF PROTECTION ACTIONS (Rev. 0, 10/73) 
 
CONFORMANCE
 
The protection system for the Farley Nuclear Plant meets the intent of IEEE-279-71 as
 
discussed in subsection 7.1.2.1. Regulatory Guide 1.62 presents an acceptable method for
 
complying with the requirements of subsection 4.17 of IEEE-279- 71. The protection system
 
does not, however, fully comply with Item 1 of Regulatory Guide 1.62. There are six manual, steam line isolation switches in the control room. 
 
The switchover from injection to recirculation is performed at the component level following an
 
accident when the refueling water storage tank (RWST) low level alarm is sounded. 
 
Since there are two trains of safeguards, this system also has the ability to accept a single
 
failure. 
 
The Protection System complies with all other portions of Regulatory Guide 1.62. 
 
FNP-FSAR-3A
 
3A-1.63-1 REV 21  5/08 Regulatory Guide 1.63 - ELECTRIC PENETRATION ASSEMBLIES IN CONTAINMENT STRUCTURES FOR WATER COOLED NUCLEAR POWER
 
PLANT (Rev. 0, 10/73) 
 
CONFORMANCE
 
Each penetration assembly was designed to withstand, without loss of assembly integrity, the
 
maximum short circuit current for a duration compatible with Insulated Power Cable Engineering
 
Association (IPCEA) standards for the size of the applicable conductor. In addition, the circuits
 
associated with the penetration assemblies are provided with overcurrent protection. Only Unit
 
2 power and control electrical containment penetrations are provided with primary and backup
 
overcurrent protection to meet the single failure criteria. Those Unit 2 power and control
 
electrical penetrations that are de-energized during normal plant operations and can be
 
energized only under administrative control ar e excluded from the requirements of dual overcurrent protection to meet the single failure criteria.
 
Each penetration assembly is designed to withstand the maximum containment internal
 
pressure of Item 2 under the Regulatory Position.
 
The penetration assemblies were installed, inspected, and tested in accordance with subsection
 
8.3.1.3, and particularly in accordance with the ASME Boiler and Pressure Vessel Code, Section III, Subsection NE, Class MC vessels. In addition, leak tests and electrical tests to
 
verify conductor continuity and insulation resistance were performed at the site. 
 
FNP-FSAR-3A
 
3A-1.64-1 REV 21  5/08 Regulatory Guide 1.64 - QUALITY ASSURANCE REQUIREMENTS FOR THE DESIGN OF NUCLEAR POWER PLANTS (Rev. 0, 10/73)(a)
CONFORMANCE
  [HISTORICAL] [The plant was designed with appropriate qua lity assurance provisions to meet the requirements of Appendix B to 10 CFR Part 50. This subject is discussed in Chapter 17.
Regulatory Guide 1.64, dated October 1973, provides NRC endorsement of ANSI N45.2.11.
The SNC Quality Assurance Topical Report (QATR) is based on ASME NQA-1-1994 which incorporates the requirements of ANSI N45.2.11. Accordingly, quality assurance requirements applicable to design activities are described in the SNC QATR.
 
[HISTORICAL] [a. Regulatory Guide 1.64, Rev. 0 applies to the design and construction phase of FNP.
Additional Quality Assurance compliance with Regulatory Guide 1.64, Rev. 1 is given in Chapter 17.2.]
 
FNP-FSAR-3A
 
3A-1.65-1 REV 21  5/08 Regulatory Guide 1.65 - MATERIAL AND INSPECTIONS FOR REACTOR VESSEL CLOSURE STUDS (Rev. 0, 10/73) 
 
CONFORMANCE
 
Regulatory Guide 1.65 was published after procurement of the Farley Units 1 and 2 reactor
 
vessel bolting material. However, this material meets the intent of the guide as follows. 
 
All the reactor vessel closure stud bolts, nuts, and washers for the Farley Unit 1 Plant were
 
fabricated from a single heat of SA540 Grade B24 Material. 
 
The reactor vessel closure bolts for Unit 2 were machined from bars of SA540 Grade B24
 
material. The Unit 2 closure nuts and washers were machined from tubes of SA540 Grade B23
 
material. 
 
The bolting material qualification tests were performed per the ASME Section III Code and
 
Addenda (1970 Summer Addenda for both units) which required meeting an average of 35 ft-lb
 
energy with no lateral expansion tests required and no maximum tensile strength limitation. All
 
bolting material for both units met these ASME Code requirements. 
 
Charpy tests were performed at 10°F on bolting material bar and tube specimens (three impact tests per bar or tube end tested) as required by the ASME Code referred to above. With the
 
exception of one end of a single bar, all impact tests of the Unit 1 reactor vessel closure stud
 
bar and tube material showed data equal to or greater than 45 ft-lb conforms with the guide
 
position on Charpy impact energy. The single exc eption had three impact energy values of 39, 40, and 40 ft-lb on one end of the bar and 41, 48, and 50 ft-lb on the opposite end. 
 
For the Unit 2 reactor vessel closure stud material, the required three impact tests on each end
 
of each bar and tube that was tested showed Char py impact energy values that ranged from a low of 39, 40, and 40 ft-lb to a high of 48, 50 and 51 ft-lb. Three of the bars and one of the
 
tubes that were tested showed Charpy impact values that were less than the 45 ft-lb proposed
 
by the guide. 
 
For the bars and tubes showing 10°F impact data averaging below 45 ft-lb, Alabama Power
 
Company believes that the intent of the guide is met, inasmuch as sufficient fracture toughness
 
is obtained at the preload temperature or at the lo west service temperature (as specified by the Guide's reference to Appendix G to 10 CFR 50, Paragraph IV.A.4), both of which are
 
significantly above the 10°F Charpy test temperat ure. Also, the impact energies at the preload or lowest service temperatures are higher than was obtained at the lower, actual Charpy test
 
temperature. 
 
For the bar material from which reactor vessel closure studs were made for Unit l, one end of
 
one bar that was tested showed an ultimate tensile strength of 178,000 psi. Ultimate tensile
 
strengths ranging from 157,000 psi to 170,000 psi were obtained for all the other bars, and the
 
tubes that were tested, conforming with the guide position. 
 
For Unit 2, the corresponding data showed a tensile strength of 178,000 psi reported for the end
 
of one bar. Ultimate tensile strengths ranging from 155,000 psi to 167,500 psi were reported for
 
all the other bars and tubes tested.
FNP-FSAR-3A
 
3A-1.65-2 REV 21  5/08 Alabama Power Company believes that the bar material tensile datum in excess of 170,000 psi
 
meets the intent of the Guide, since Reference 2 of the Guide shows that the yield strength
 
should be below 170,000 psi; the guide position concerning a maximum ultimate tensile
 
strength should be based on the yield strength. Units 1 and 2 bar and tube bolting material
 
yield strength data were below 170,000 psi. 
 
The closure stud bolting material was procured to a minimum yield strength of 130,000 psi and
 
a minimum tensile strength of 145,000 psi. This strength level is compatible with Appendix G to
 
10 CFR 50 (July 1973, Paragraph I.C), although higher strength bolting materials are permitted
 
in the ASME Code. Stress corrosion has not been observed in reactor vessel closure stud
 
bolting manufactured from material of this strength level. Accelerated stress corrosion test data
 
do exist for materials of 170,000 psi minimum yield strength exposed to marine water
 
environments stressed to 75 percent of the yield strength (given in Reference 2 of the Guide). 
 
That these data show much more severe stress corrosion than is applicable because of the
 
specified yield strength differences and the less severe environment to which Farley-reactor-
 
vessel-closure-stud bolting is exposed is demonstr ated by many reactor-years of satisfactory experience at nuclear power plants. 
 
The reactor-vessel-stud-bolt materials for both plants were inspected to the requirements of the
 
ASME Code Section III (through summer 1970) and meet the intent of Regulatory Guide C. 2 "Inspection." 
 
Ultrasonic inspection of the Farley stud material was examined radially based upon the first
 
back reflection from an indication-free area of the basic bar stock for each stud and of each bar
 
from which the nuts and washers are machined per ASME Specification SA-388. The
 
indication-free section is selected for calibration by a preliminary scan. Calibration for the axial
 
examination conducted from the end faces of the studs and the bars for nuts and washers was
 
established on a calibration block or standard per ASME Specification SA-388. The axial testing
 
calibration block was manufactured from a representative, indication-free section of the stud, nut, and washer bar stock, based on a preliminary scan conducted for that purpose.
 
Magnetic particle examination of all exterior closure stud surfaces was conducted after
 
threading, using continuous circular and longitudinal magnetization. The Farley Units reactor
 
vessel design and Westinghouse practice and recommendations meet the recommendations of
 
Regulatory Guide Positions 3, "Protection Against Corrosion," and 4, "Inservice Inspection." 
 
FNP-FSAR-3A
 
3A-1.66-1 REV 21  5/08 Regulatory Guide 1.66 - NONDESTRUCTIVE EXAMINATION OF TUBULAR PRODUCTS (Rev. 0, 10/73) 
 
CONFORMANCE
 
Regulatory Guide 1.66 was published after procurement of tubular products for Farley Nuclear
 
Plant Units 1 and 2. In addition, the licensee takes exception to Regulatory Guide 1.66. The
 
Guide position concerning defect detection capability is impractical and the axial testing
 
recommendations to meet the Guide position are technically unnecessary. 
 
The Guide states that, "Nondestructive examination applied to tubular products used for
 
components of the reactor coolant pressure boundary and other safety related systems. . .
 
should be capable of detecting unacceptable defects regardless of defect shape, orientation, or
 
location in the product." This is impractical to attain. 
 
In addition, the guide position regarding ultrasonic angle beam scanning in the axial direction is
 
technically unnecessary, since any flaws that developed from the processes employed in tubular product manufacture are invariably oriented in the axial direction. Any circumferential or
 
transverse flaws that developed were mechanically induced surface defects detected by normal
 
QC procedures. However, the nondestructive examinations performed on tubular products
 
covered by the Guide (reactor vessel nozzles, control rod drive mechanism (CRDM) housings, core support columns) met the purposes of the Guide. 
 
The tubular products used in the Farley units' equipment and systems, as listed above, were
 
fabricated and inspected to high quality standards, suitable for nuclear equipment as were
 
required by the applicable contemporary codes and standards. (See table 3.2-1.) 
 
The examinations were performed to higher sensitivity levels than required by the codes, and
 
included 100 percent volumetric nondestructive examination. 
 
For the reactor vessel nozzle forgings, the ultrasonic examination included end to end axial
 
testing, from the end faces, and angle beam testing in two circumferential directions. The angle
 
beam testing was repeated to the fullest extent possible after machining. After heat treatment
 
and prior to cladding, magnetic particle examinations were conducted over all nozzle surfaces. 
 
The CRDM housings were ultrasonically angle beam tested in two circumferential directions, axially tested from the end faces, and radially te sted from the circumferential surfaces of the bars or the tubular product machined from the bars. 
 
The core support structure tubular products were ultrasonically tested axially from the end faces, and radially tested from the circumferential surfaces. 
 
FNP-FSAR-3A
 
3A-1.67-1 REV 21  5/08 Regulatory Guide 1.67 - INSTALLATION OF OVERPRESSURE PROTECTION DEVICES (Rev. 0, 10/73) 
 
CONFORMANCE (a)
All structures, systems, and components of over pressure protection systems important to safety, including the main steam safety and relief valves and associated piping and valve
 
headers, were designed, analyzed, and qualified in accordance with the recommendations of
 
this Guide. 
 
Details for the pressurizer overpressure protection system are addressed in subsection 5.2.2. 
: a. The Winter 1978 Addenda to the 1977 Edition of the ASME Boiler and Pressure Vessel
 
Code, Appendix O, Section III, Division 1, "Rules for Design of Safety Valve Installations,"
included requirements equivalent to the recommendations of Regulatory Guide 1.67. These
 
changes to the Code were incorporated by reference to 10 CFR 50.55a on April 3, 1981. 
 
Subsequently, the NRC withdrew Regulatory Guide 1.67 on April 15, 1983.
 
However, the withdrawal of this Regulatory Guide does not alter any prior or existing licensing
 
commitments based on its use. As noted above, Farley Nuclear Plant meets the
 
recommendations of Regulatory Guide 1.67.
FNP-FSAR-3A
 
3A-1.68-1 REV 21  5/08 Regulatory Guide 1.68  - PREOPERATIONAL AND INITIAL STARTUP TEST PROGRAM FOR WATER COOLED POWER REACTORS (Rev. 0, 11/73)
 
CONFORMANCE
 
The conformance with Regulatory Guide 1.68 is discussed in chapter 14.
 
The standard Westinghouse nuclear steam supply system contains online analog protective
 
circuits designed to provide continuous online protection against excessive power density and
 
DNB. The plant process computer is not a part of this standard system but was purchased as
 
an option. The plant process computer does not perform any safety-related function nor is it
 
required for the operation of the plant; therefore, Alabama Power Company takes exception to
 
Item D.1.r of Appendix A in Regulatory Guide 1.68.
 
FNP-FSAR-3A
 
3A-1.69-1 REV 21  5/08 Regulatory Guide 1.69 - CONCRETE RADIATION SHIELDS FOR NUCLEAR POWER PLANTS (Rev. 0, 12/73)
 
CONFORMANCE
 
This topic is addressed in subsection 12.1.2.1.
 
FNP-FSAR-3A
 
3A-1.70-1 REV 21  5/08 Regulatory Guide 1.70 - STANDARD FORMAT AND CONTENT OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS (Rev. 1, 10/72) 
 
CONFORMANCE
 
The FSAR for the Farley Nuclear Plant was prepared in accordance with Revision 1, October
 
1972. Recommendations for additional information that have been issued as new regulatory guides with numbers in the form 1.70.X dated up to and including an issue of August 1974 are
 
addressed in this appendix. Some plant project drawings are included in the FSAR by
 
reference to the drawing identification number (e.g., D-177024) in lieu of inclusion in the FSAR
 
as a figure. Some pages and figures in the FSAR are referenced by the date of change or
 
revision number or both in the lower right-hand corner per 10 CFR 50.71(e)(5).
 
FNP-FSAR-3A
 
3A-1.70.1-1 REV 21  5/08 Regulatory Guide 1.70.1 - ADDITIONAL INFORMATION - HYDROLOGICAL CONSIDERATIONS FOR NUCLEAR POWER PLANTS (Rev. 0, 12/73) 
 
CONFORMANCE
 
Water quality is addressed in subsection 2.4.13.5. 
 
The design bases for groundwater-induced hydrostatic loadings on safety-related structures are
 
addressed in subsections 2B.4.4 and 2B.7.1 and section 2B.6. The design for hydrostatic
 
loadings is based either on the normal groundwater level or on the design flood level, whichever
 
is more critical for the particular structure. Dewatering during construction is not critical to the
 
integrity of safety-related structures. 
 
Groundwater conditions are addressed in subsection 2.5.4.6 
 
A history of groundwater fluctuations beneath the site is provided in subsection 2.4.13.2.2. The
 
water levels in the piezometers are shown on figures 2.4-26 through 2.4-60. Discussions of
 
groundwater conditions during and after construction of the plant are included in subsections
 
2.4.13.1.3, 2.4.13.2.5, and 2B.4.4. 
 
FNP-FSAR-3A
 
3A-1.70.2-1 REV 21  5/08 Regulatory Guide 1.70.2 - ADDITIONAL INFORMATION - AIR FILTRATION SYSTEMS AND CONTAINMENT SUMPS FOR NUCLEAR POWER PLANTS (Rev. 0, 12/73) 
 
CONFORMANCE
 
B.(1) The analyses of the engineered safety features air filtration systems with respect to Regulatory Guide 1.52 are provided in the following sections: 
: a. Fuel handling building - subsection 9.4.2.2.2. 
: b. Control room - subsection 9.4.1.2. 
: c. Penetration room - subsection 6.2.3.2.2. 
 
B.(2) The information requested dealing with the containment sumps and sump intake screens is provided in subsections 6.2.2.2.1 and 6.2.2.3.1. 
 
FNP-FSAR-3A
 
3A-1.70.3-1 REV 21  5/08 Regulatory Guide 1.70.3 - ADDITIONAL INFORMATION - RADIOACTIVE MATERIALS SAFETY FOR NUCLEAR POWER PLANTS (Rev. 0, 2/74) 
 
CONFORMANCE
 
The additional information described in the Regulatory Guide is provided in section 12.4. 
 
FNP-FSAR-3A
 
3A-1.70.4-1 REV 21  5/08 Regulatory Guide 1.70.4 - ADDITIONAL INFORMATION - FIRE PROTECTION CONSIDERATIONS FOR NUCLEAR POWER PLANTS (February 1974) 
 
CONFORMANCE
 
Several specific portions of this regulatory guide spell out information to be provided in FSARs.
 
All required information is provided in t he Fire Protection Program Reevaluation.
 
FNP-FSAR-3A
 
3A-1.70.5-1 REV 21  5/08 Regulatory Guide 1.70.5 - ADDITIONAL INFORMATION - WATER LEVEL (FLOOD)
DESIGN FOR NUCLEAR POWER PLANTS (Rev. 0, 5/74) 
 
CONFORMANCE
 
The additional information requested in the Regulatory Guide is provided in subsection 3.4. 
 
FNP-FSAR-3A
 
3A-1.70.6-1 REV 21  5/08 Regulatory Guide 1.70.6 - ADDITIONAL INFORMATION QUALITY ASSURANCE DURING DESIGN AND CONSTRUCTION (Rev. 0, 6/74) 
 
CONFORMANCE
 
Conformance with sections of Appendix B to 10 CFR Part 50, and hence with this Regulatory
 
Guide, is discussed in subsection 17.1.1. 
 
FNP-FSAR-3A
 
3A-1.70.7-1 REV 21  5/08 Regulatory Guide 1.70.7 - ADDITIONAL INFORMATION, GEOGRAPHY AND DEMOGRAPHY CONSIDERATIONS FOR NUCLEAR POWER
 
PLANTS (Rev. 0, 8/74) 
 
CONFORMANCE
 
Sufficient information is presented in section 2.1 to respond to the considerations of the
 
Regulatory Guide, although the format is based on Revision 1 of the Standard Format and
 
Content of Safety Analysis Report for Nuclear Power Plants. 
 
FNP-FSAR-3A
 
3A-1.71-1 REV 21  5/08 Regulatory Guide 1.71  - WELDER QUALIFICATION FOR AREAS OF LIMITED ACCESSIBILITY (Rev. 0, 12/73) 
 
CONFORMANCE
 
The recommendation of the Regulatory Guide for limited accessibility qualification or requalification, in addition to ASME Section III and IX requirements, is an unduly restrictive
 
requirement for shop fabrication, where the welder's physical position relative to the welds is
 
controlled and does not present any significant problems. In addition, shop welds of limited
 
accessibility were repetitive due to multiple production of similar components, and such welding
 
was closely supervised.
 
Field welding procedures and personnel were qualified in accordance with the requirements of ASME Section IX. As far as is practicable, welds were located to provide physical and visual
 
accessibility for welding and inspection. In the event that a weld must be located in an
 
unfavorable position, considerations were given to the preparation of a mockup simulating the production weld, using welders qualified to Section IX.
 
The above practices with associated quality control will meet the intent of the Regulatory Guide. 
 
FNP-FSAR-3A
 
3A-1.72-1 REV 21  5/08 Regulatory Guide 1.72  - SPRAY POND PLASTIC PIPING (12/73) 
 
CONFORMANCE
 
Regulatory Guide 1.72 is not applicable to the Farley Nuclear Plant. 
 
FNP-FSAR-3A
 
3A-1.73-1 REV 21  5/08 Regulatory Guide 1.73  - QUALIFICATION TESTS OF ELECTRIC VALVE OPERATORS INSTALLED INSIDE THE CONTAINMENT OF NUCLEAR
 
POWER PLANTS (Rev. 0, 2/74)
CONFORMANCE
 
Westinghouse safety-related-active-valve-motor operators used inside containment are
 
environmentally qualified by having passed a comprehensive testing program. The testing included heat, live steam, heat aging, shock and vibration, cycle life tests, radiation, and
 
postaccident steam and chemical spray testing. Although the test sequences employed and
 
documentation of the tests vary somewhat from requirements issued in 1974, these factors do not detract from the intent of the qualification, i.e., to provide assurance of operability under
 
accident conditions. 
 
Test results for Westinghouse supplied equipment are addressed in section 3.11. 
 
FNP-FSAR-3A
 
3A-1.74-1 REV 21  5/08 Regulatory Guide 1.74  - QUALITY ASSURANCE TERMS AND DEFINITIONS (Rev. 0, 2/74)
CONFORMANCE
  [HISTORICAL] [The NRC Regulatory Guide 1.74 identifies forms and acceptable definitions that are important to the understanding of quality assurance requirements for the design, construction, and operation of nuclear power plant structures, systems, and components.
Terms and definitions contained in Quality Assur ance Procedures, References, or Reports are in accordance with ANSI N45.2.10-1973, an acceptable standard for use in describing and implementing quality assurance programs, as described in subsection 17.2.1.]
The SNC Quality Assurance Topical Report (QATR) is based on ASME NQA-1-1994 which incorporates the requirements of ANSI N45.2.10. Accordingly, terms and definitions used in the quality assurance program are provided in the SNC QATR.
 
FNP-FSAR-3A
 
3A-1.75-1 REV 21  5/08 Regulatory Guide 1.75 - PHYSICAL INDEPENDENCE OF ELECTRIC SYSTEMS (Rev. 0, 2/74)
 
CONFORMANCE
 
The extent of compliance of Farley design with the guide is given below: 
 
Separation is provided to maintain independence between redundant safety-related circuits so
 
that a failure in one circuit does not jeopardize the protective function of other safety-related
 
circuits during and following any design basis event. Equipment and circuits requiring
 
separation have the safety train designation assigned as part of the equipment and/or scheme
 
cable number. 
 
The Farley design is in compliance with the Regulatory Guide recommendations except for the
 
separation recommendations between associated circuits. The "associated circuits" as defined
 
in the guide, though uniquely identified, were not designed to meet the separation requirements
 
of Class 1E circuits. On Farley, a non-Class lE circuit associated with Class 1E cables and equipment is assigned either an "X" or a "Y" scheme cable number in accordance with
 
subsection 8.3.1.5. The routing procedures ensure that a non-Class 1E circuit routed with Class
 
1E circuits of one safety train will not be routed with a Class 1E circuit of the opposite train. The "X" or "Y" circuits, however, may run together in a non-Class 1E circuit raceway. Also, cables with "X" and "Y" scheme cable numbers are permitted to enter non-Class 1E equipment
 
provided that they do not form an electrically continuous circuit within the equipment. Any
 
exceptions are reviewed for acceptability on a case-by-case basis. 
 
"Associated cables" of one train entering Class 1E equipment of the opposite train are run
 
separately from other opposite train cables. In this way, a failure of a circuit does not defeat the
 
protective function requirements of redundant Class 1E circuits. The cables used at Farley are
 
of flame-retardant construction and have been manufactured to meet the quality assurance
 
requirements of subsection 17B.1.2. In addition, power cables run in trays have interlocked
 
armor that further prevents the propagation of fire. 
 
The underlying philosophy of physical independence requirements for Farley is that fires are primarily caused by insulation failures due to overheating of cables. A fire in an "X" cable in an "A" train tray could affect the adjacent safety-related cables but only in the "A" train. Because of
 
the flame- retardant qualities of cable insulation it is considered inconceivable that this fire would propagate to the non-Class 1E raceway carrying X and Y cables, then to an adjacent "Y" cable and further propagate via the "Y" cable to the "B" train safety-related cables and
 
jeopardize both trains. 
 
Specific areas of Farley design positions are given below: 
 
A. Isolation Devices (Paragraph 3.8) 
 
Interrupting devices actuated by fault current are isolation devices when justified by test or analysis. 
 
FNP-FSAR-3A
 
3A-1.75-2 REV 21  5/08  B. Non-Class 1E Circuits (Paragraph 4.6) 
 
Non-class 1E circuits are not separated by minimum separation distances, nor are non-class 1E circuits separated from associated circuits by minimum
 
separation distances, nor are all non-class 1E circuits treated as associated
 
circuits. 
 
C. Routing of Class 1E Circuits (Paragraph 5.1.1.1) 
 
Opposite sides of rooms or areas, if confined or otherwise incapable of dissipating heat from fires, are Class 1 separation if provided with fire protection
 
or otherwise incapable of supporting combustion. 
 
D. Cable Spreading Area and Main Control Room (Paragraph 5.1.3) 
 
With regard to the requirement for a minimum of 1-in. separation between redundant Class 1E circuits and between Class 1E and non-Class 1E circuits, FNP requirements for cable routing meet the intent of this separation requirement
 
as follows:
: 1. The outside edges of instrumentation and control conduits are separated by a minimum of 1/2 in. This separation is adequate since:
: a. These conduits contain low energy circuits which are provided with fuse/breaker interrupting devices that would preclude a fault
 
in these cables from generating sufficient heat to ignite the cable
 
insulation and jacketing material before being interrupted.
: b. All cables are IEEE 383 qualified which, even if faulted, will not sustain combustion.
: 2. For 600-V power circuits, spacing is no less than 1/4 of the largest conduit diameter. This separation is adequate since:
: a. The short-circuit coordination schemes preclude a fault in these cables from generating sufficient heat to ignite the cable insulation
 
and jacketing material before being interrupted.
: b. IEEE 383 cables will not sustain fire propagation.
 
Based on these separation requirements and the fault protection provided, an internally faulted cable would not produce sufficient
 
heat energy to degrade cables in another conduit within the
 
distances specified. Thus, these separation requirements meet
 
the intent of this requirement. 
 
FNP-FSAR-3A
 
3A-1.75-3 REV 21  5/08  E. General Plant Areas (Paragraph 5.1.4) 
 
Solid enclosed raceways may be more detrimental than nonsolid raceways because of flue effects. 
 
With regard to the requirement for a 1-in. separation between redundant Class 1E circuits and between Class 1E circuits and non-Class 1E circuits, the
 
discussion provided above in response to paragraph 5.1.3 also applies to
 
paragraph 5.1.4. Additionally, 4-kV power circuits are separated by a minimum
 
of 1 conduit diameter spacing and all 4-kV power conduits are greater than 1-in.
 
in diameter. This meets the 1-in. separation requirement for these cables.
 
F. Instrument Cabinets (Paragraph 5.7) 
 
Separation requirements should not be the same for instrumentation racks and control boards because functional requirements are different. The IEEE draft
 
criteria are adequate. 
 
FNP-FSAR-3A
 
3A-1.76-1 REV 21  5/08 Regulatory Guide 1.76 - DESIGN BASIS TORNADO FOR NUCLEAR POWER PLANTS (Rev. 0, 7/74) 
 
CONFORMANCE
 
The design basis tornado for the Farley Nuclear Plant is given in subsection 3.3.2, Tornado
 
Loadings. The maximum rotational speed, maximum translational speed, and the rate of
 
pressure drop differ somewhat from those specified in Regulatory Guide 1.76 as indicated
 
below. However, the maximum windspeed, radi us of maximum rotation, and the maximum pressure drop are in conformance with the Guide. 
 
Rate of Rotational Translational Pressure drop Speed (mph) Speed (mph)      psi/s   
 
Regulatory Guide 1.76 290 70 2.0
 
Farley 300 60 1.0
 
FNP-FSAR-3A
 
3A-1.77-1 REV 21  5/08 Regulatory Guide 1.77 - ASSUMPTIONS USED FOR EVALUATING A CONTROL ROD ACCIDENT FOR PRESSURIZED WATER REACTORS (Rev. 0, 5/74) 
 
CONFORMANCE
 
Methods for evaluating rod ejection accidents are in compliance with Regu1atory Guide 1.77. 
 
The methods employed to evaluate postulated rod ejection accidents are described in
 
subsection 15.4.6. 
 
FNP-FSAR-3A
 
3A-1.78-1 REV 21  5/08 Regulatory Guide 1.78 - EVALUATING THE HABITABILITY OF A NUCLEAR POWER PLANT CONTROL ROOM DURING A POSTULATED
 
HAZARDOUS CHEMICAL RELEASE (Rev. 1, 12/01)
 
CONFORMANCE
 
The design guidance and assumptions of Regulatory Guide 1.78 are used in the evaluation of
 
control room habitability except as noted below:
: a. Hazardous chemicals in the vicinity of the site are discussed in section 2.2. 
: b. Instead of release and transport models of paragraphs 3.2 and 3.3, the evaluation model used conforms to the guidance of NUREG-0570.
: c. The design of the isolation system conforms to IEEE-279(1971) in lieu of IEEE-603 described in paragraph 4.2.
: d. Analysis of a chlorine release from its storage locations onsite is discussed in section 2.2 and subsection 9.4.1. This discussion is provided for historical
 
purposes only since all significant quantities of chlorine, i.e., single containers
 
greater than 150 lb, have been removed from the plant site. Control of chlorine is
 
in accordance with Regulatory Guide 1.95.
 
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3A-1.79-1 REV 21  5/08 Regulatory Guide 1.79 - PREOPERATIONAL TESTING OF EMERGENCY CORE COOLING SYSTEMS FOR PRESSURIZED WATER REACTORS (Rev. 0, 6/74) 
 
CONFORMANCE
 
These tests were intended to evaluate the performance of the components of the emergency
 
core cooling system (ECCS) to ensure that the ECCS accomplished its required function. 
 
The preoperational test program covered the following tests: 
: 1. Tests under ambient temperature conditions. 
 
The reactor vessel was open and flooded; thus the reactor coolant system (RCS) was essentially at atmospheric pressure. 
: 2. Tests under hot operating conditions. 
 
The reactor vessel was closed and the reactor coolant system was at pressure and temperature conditions obtainable under preoperational testing. 
: 3. ECCS component testing. 
: 1. Tests Under Ambient Temperature Conditions
 
1.1 Integrated System Test 
 
The objective of this test was to ensure that the diesel generators had the capability to start and accelerate the engineered safety
 
features (ESF) loads to rated speed without exceeding the
 
specified safety limits. 
 
A safety injection signal concurrent with a loss of offsite power was simulated in one safety train, and it gave a signal to
 
simultaneously start the diesel generators and shed the safety-
 
related loads from the buses. After the diesel generators reached
 
rated speed and voltage, the safety-related loads indicated below
 
were automatically sequenced by the diesel generator sequencer.
 
The response time from initiation of the safety injection signal to
 
starting these loads was measured manually (with a stopwatch)
 
and checked for acceptability. 
 
Safety-Related Loads
: a. High-head safety injection pump. 
: b. Low-head safety injection (residual heat removal) pump. 
: c. Component cooling water pump. 
: d. Service water pumps. 
: e. Auxiliary feedwater pump.
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3A-1.79-2 REV 21  5/08    f. Containment spray pump.      g. Control room AC. 
: h. Containment coolers. 
: i. Battery charger. 
: j. River water pumps.(a)
: k. Motor control centers associated with emergency power systems.      l. Safety-related valves. 
 
For this test, all valves which, if operated, would have a detrimental effect on the subsequent commissioning of the plant, were blocked from operation. An example of this case was the
 
accumulator isolation valve. Similarly, where full flow conditions
 
could not be achieved, the pumps were operated on miniflow or
 
on bypass. An example of the latter case was the containment
 
spray pump. 
 
The following tests were associated specifically with the ECCS: 
 
1.2 High-Pressure Safety Injection Test - Flow Test 
 
Fluid from the refueling water storage tank was injected into the open reactor vessel through various combinations of injection legs
 
and pumps by operating the high head safety injection pumps. 
 
Flow was demonstrated to be within the design specifications. 
 
Response time data were obtained (manual measurement) for
 
components under test to demonstrate that they met or exceeded
 
the acceptance criteria. 
 
1.3 Low-Pressure Safety Injection Test - Flow Test
 
The test under 1.2 above was repeated but with the operation of the residual heat removal pumps instead of the safety injection
 
pumps.
1.4 Recirculation Test 
 
This test, as required by paragraph C.3.b.(2), was not performed because, in view of the advanced stage of construction, substantial plant modifications would have been required to do the
 
testing. This mode of operation was checked by analyses and
 
operation of individual components, checked separately. 
: a. Although river water pumps were originally tested as safety-related loads, they are not relied
 
upon in any analysis and are therefore not autom atically sequenced by the diesel generator sequencer.
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3A-1.79-3 REV 21  5/08    1.5 Core Flooding - Flow Test 
 
Each accumulator was filled and pressurized with the motor-operated isolation valve closed. The accumulator discharge was
 
initiated by opening the accumulator isolation valves with the RCS
 
at reduced pressure. Discharge flowrate was calculated from the
 
change of accumulator pressure with time. 
: 2. Test Under Hot Operating Conditions
 
2.1 High-Pressure Safety Injection - Flow Test 
 
The capability of the high-head safety injection (HHSI) pumps to deliver emergency core cooling water from the refueling water
 
storage tank to the RCS was checked by analysis. The operation
 
of the safety injection check valves was shown to be satisfactory
 
by brief operation of the HHSI pumps. 
 
During this test, flow of auxiliary feedwater to the main feedwater system was blocked to avoid temperature and pressure transients.
 
The pumps were started and run on recirculation. Flow from the
 
auxiliary feedwater pumps was verified as part of feedwater system tests. 
 
2.2 The HHSI pumps were used to produce flow through the check valves in the HHSI system. 
 
The operation (partially open) of the check valves was verified by recording the flow in each branch line using installed orifices. 
: 3. ECCS Component Testing
 
3.1 Accumulator Isolation Valve Test 
 
The operation of the accumulator isolation valves was tested along with Test 1.5 above. Since the isolation valve operators are
 
supplied from motor control centers that receive power from either
 
normal or emergency power source, the test was performed using
 
normal power only. The valve operation was initiated by the
 
simulation of a safety injection signal. 
 
3.2 Testing of Valves and Pumps of ECCS 
 
Routine periodic testing of ECCS components is performed at power as discussed in subsection 6.3.4. Valves are operated
 
through a complete cycle and their operation observed in the
 
control room. The response times are measured manually. 
 
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3A-1.79-4 REV 21  5/08    Operation of pumps and motors are routinely checked at power by operation on miniflow or bypass. 
 
3.3 Initiating Instrumentation 
 
Safeguard system logic tests were performed in accordance with subsection 7.3.2. 
 
3.4 Testing of Onsite Power System During Refueling Shutdown 
 
The objective is to check the integrity of the onsite power system to start the safety related loads. Each safety train is tested in two
 
steps so that the normal process of plant shutdown is not affected. 
 
Step 1: A safety injection signal concurrent with a loss of offsite power is simulated in one safety train at the sequencer, which eventually loads the pump motors on the diesel
 
generators in the manner of test 1.1. The safety
 
injection signal is not transmitted to the safety-related
 
valves, but the valves are positioned to ensure the
 
normal shutdown procedure. The following pumps are
 
operated on bypass:  HHSI pumps and the
 
containment spray pumps. This test provides a means
 
of ensuring that the starting and operating of pumps
 
and their response times (measured manually) are
 
within acceptable limits. 
 
Step 2: With the pump motor breakers in the test position, a safety injection signal is initiated that operates the
 
safety-related valves. Indication in the control room
 
provides a check of valve operation without simulating
 
flow conditions. The response times, determined
 
manually, for these valves are also checked for
 
acceptability. Since the sequencing of motor control
 
center loads was checked in step 1, the operation of
 
the valves is tested with the diesel generator operating
 
from step 1. 
 
Periodic testing requirements are addressed in the Technical Specifications and the Technical Requirements Manual. 
 
3.5 System Piping and Supports 
 
The acceptability of system piping movements is discussed in subsection 3.9.1. 
 
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3A-1.80-1 REV 21  5/08 Regulatory Guide 1.80 - PREOPERATIONAL TESTING OF INSTRUMENT AIR SYSTEMS (Rev. 0, 6/74) 
 
CONFORMANCE
 
The Regulatory Guide addresses safety-related instrument air systems and thus is not
 
applicable to the Farley instrument air system. 
 
The acceptance test for the instrument air sy stem generally included the recommendations of subsections Cl through C7 of the Guide. 
 
Operational tests of those air operated safety related valves required to assume the safe
 
operating position upon 1oss of instrument air were included on an individual basis in the
 
preoperational tests of the systems in which the valves are located. 
 
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3A-1.81-1 REV 21  5/08 Regulatory Guide 1.81 - SHARED EMERGENCY AND SHUTDOWN ELECTRIC SYSTEMS FOR MULTIUNIT NUCLEAR POWER PLANTS (6/74) 
 
CONFORMANCE
 
Paragraph C.1 
 
Each of the two units is provided with separate and redundant dc electrical systems. The
 
sharing of the dc electrical systems is limit ed to control power requirements of components in the service water intake structure and diesel generators 1-2A, 1C, and 2C, which are shared
 
between Unit 1 and Unit 2. The dc control power supplies to these diesels are mechanically
 
interlocked so that only one source furnishes control power requirement at any time.(a)  See drawings D-177082, D-177083, D-207082, and D-207083 for the dc distribution system for
 
diesels. Details of the dc electrical system are discussed in subsection 8.3.2. 
 
Paragraph C.2 
 
The Onsite ac Power System is described in subsection 8.3.1, and, more specifically, the onsite
 
emergency power system is described in 8.3.1.1.7. The details of conformance with this
 
paragraph follow: 
 
Item a. The sharing of onsite ac electrical systems is limited to two units. 
 
Items b. The sizing of diesels together with the control circuitry design is adequate and c. considering a single failure capability to automatically supply the ESF loads for the accident unit and the safe shutdown requirements for the
 
other unit. Details of diesel operation under various conditions are
 
provided in paragraph 8.3.1.1.7 and take into consideration the most
 
severe condition of a DBE on one unit and a failure of one diesel
 
generator. A single failure includes a false or spurious accident signal in
 
the non-accident unit.
 
Item d. The control circuits for shedding and loading the ESF loads are essentially separate in that each ESF 4160V bus is provided with its load
 
sequencer. The interaction between control circuits for Unit 1 and 2 is
 
limited to automatic starting signal from the 4160-V buses and
 
instrumentation for the shared diesels. However, maintenance and
 
testing of these starting signals in one unit will not prevent the diesel
 
generator from supplying the minimum ESF loads on the other unit. 
 
Item e. All diesel generators are controlled from the emergency power board common to both Units 1 and 2 and located in the control room. 
 
Coordination between unit operators is not necessary for meeting
 
recommendations of Regulatory Positions 2b, 2c, and 2d. 
: a. No common mode failures exist which could fail dc systems in both units.
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3A-1.81-2 REV 21  5/08  Item f. Complete information in regard to the diesel generator, the load sequences, and the associated 4160-V breakers is displayed on the
 
emergency power board in the control room 
 
Item g. The design conforms to Regulatory Guides 1.6 and 1.9 as discussed in the appropriate areas of this appendix. Information regarding bypassed
 
and inoperable systems is provided, although detailed conformance to Regulatory Guide 1.47 is not  achieved. This is detailed in the discussion
 
of Regulatory Guide 1.47 elsewhere in this appendix. 
 
Paragraph C.3 
 
This paragraph is not applicable to the Farley Nuclear Plant. 
 
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3A-1.82-1 REV 21  5/08 Regulatory Guide 1.82 - SUMPS FOR EMERGENCY CORE COOLING AND CONTAINMENT SPRAY SYSTEMS (Rev. 3) 
 
CONFORMANCE
 
Plant Farley complies with the regulatory positions on design criteria, performance standards, and analysis methods that relate to PWRs except, or as clarified in this section:
C.1.1.1.2: The sumps are located outside the missile barrier and are physically separated from each other by structural barriers to the extent practical. No additional protection from high-energy piping is required.
C.1.1.1.3: The sumps are located on the lowest possible floor elevation in the containment exclusive of the reactor vessel cavity. Each pump intake is protected by one strainer assembly. The sump strainer assemblies are not depressed below the floor. No trash rack is installed. Exception is based on:
* There are no high-energy line breaks postulated to occur near the strainers that would result in the required post-accident recirculation and there are no missiles generated in the vicinity of the strainer assemblies; therefore, there are no jet loads, no pipe whip restraint loads, and no missiles applicable to the strainer assemblies.
* The design of the stacked disk strainer prevents pieces of debris larger than 1.75 in. in diameter from reaching the perforated area due to the small slots between the strainer disks.
* The stress analysis results show the strainer assemblies can meet the design requirements of ASME Section III, subsections NC, ND, and NF. The loads used in the analyses include the pressure from debris, seismic loads and dead weight, and all other pertinent parameters.
Plant-specific transport analysis and testing has been performed.
C.1.1.1.4: The floor level in the vicinity of each coolant sump is sloped toward a drain through the missile barrier. In some cases the slope is away from the sump; in some cases it is towards the sump. Plant-specific transport analysis and testing has been performed. Therefore, floor slope and curb are not credited to limit debris transport.
C.1.1.1.6: The sumps are located outside the missile barrier and are physically separated from each other by structural barriers to the extent practical. No additional protection from high-energy piping is required. The strainers are designed to withstand loading for the largest postulated debris pieces and types resulting from a LOCA that requires post accident recirculation.
C.1.1.1.7: The strainers are designed to withstand loading for the largest postulated debris pieces and types. The stacked disc strainer design with perforated plate is self venting. A single solid cover is not practical.
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3A-1.82-2 REV 21  5/08 C.1.1.1.12: Unit 2 has to replace the high-head safety injection throttle valves in order to comply with this position.
C.1.1.3: N/A C.1.1.4: N/A C.1.2:  N/A C.1.3.1.2: N/A C.1.3.1.3: N/A C.1.3.1.9: Minimum projected LBLOCA sump water level used in all NPSH calculations.
C.1.3.2.1: Main steam and main feedwater line breaks were not evaluated since it is assumed that recirculation is not credited for these situations.
C.1.3.2.2: A zone of influence (ZOI) of 4 for acceptable coating has been used in the plant specific debris generation and transport analysis and testing. This is based on the guidance of WCAP-16568-P.
C.1.3.2.6: Chemical effects evaluations have not been completed.
C.1.3.3.8: N/A C.1.3.4.4: N/A
[HISTORICAL] [CONFORMANCE (Prior to December 2007)
C. 1: The requirements of this paragraph are met. See appendix 6C for a further discussion.
C. 2: The sumps are located outside the mi ssile barrier and are physically separated from each other by structural barriers to the extent practical. No additional protection from high-energy piping is required.
C. 3: The sumps are located on the lowest floor elevation in the containment exclusive of the reactor vessel cavity. Each sump inta ke is protected by two screens:  an inner grating and a fine outer screen. The sump screens are not depressed below the floor elevation. Figure 6C-6 shows a containment sump intake and screen arrangement.
C. 4: The floor level in the vicinity of each coolant sump is sloped toward a drain through the missile barrier. In some cases the slope is away from the sump; in some cases it is towards the sump.
C. 5: The requirements of this paragraph are met.
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3A-1.82-3 REV 21  5/08 C. 6: An outer trash rack is provided.
C. 7: A vertically mounted fine screen is prov ided. The present effec tive sump screen height which was selected to ensure sump submerg ence during recirculation is such that a total effective sump peripheral length approximately 150 ft (corresponding to overall sump dimensions of approximately 40 ft x 40 ft, or equivalent) would be required to yield screen velocities of 0.2 ft/s for the sump serving both the ECCS and the containment spray systems, based on the a ssumptions of Regulatory Guide 1.82.
Existing Farley containment layout p recludes the location of sumps of these dimensions in protected locations. Th e present sumps were designed to yield low velocities of approach in the vicinity of the sump to promote the settling out of debris, and to yield negligible pressure drops th rough the sump screen. Materials inside containment that could cause sump screen blockage post-LOCA are eliminated or minimized by design. Present liquid vel ocities through the fine screen, based on the assumptions required by this guide, are between 1.16 and 2.17 ft/s C. 8: A 3-ft-wide solid plate covers most of th e top of the 5-ft-wide sump.  (See figure 6C-6.)
The top deck is designed to be fully submerged after a LOCA and completion of the safety injection.
C. 9: The recommendations of this paragraph are met.
C.10: The recommendations of this paragraph are met. Some nuclear fuel used at Farley Nuclear Plant may contain design features that provide for flow paths smaller than the inner containment sump screen. However, these flowpaths are evaluated as part of the fuel design and will provide adequate ECCS flow to ensure long term core cooling. C.11: The recommendations of this paragraph are met, as there is a vortex breaker provided at each pump inlet. (See figure 6C-6.)
C.12: The recommendations of this paragraph a re met by having the trash racks made of galvanized steel and the screens of stainless steel.
C.13: The recommendations of this paragraph are met by means of a removable 1/4-in.
solid plate (see figure 6C-6) on the top of each sump.
C.14: The recommendations of this paragraph are met.]
 
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3A-1.83-1 REV 21  5/08 Regulatory Guide 1.83 - INSERVICE INSPECTION OF PRESSURIZED WATER REACTOR STEAM GENERATOR TUBES (Rev. 1, 7/75) 
 
CONFORMANCE
 
The Farley Nuclear Plant meets the intent of the Regulatory Guide. Surveillance requirements
 
for the steam generator tubes are conducted in accordance with the Steam Generator Tube
 
Surveillance Program, as required by the plant Technical Specifications.
 
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3A-1.84-1 REV 21  5/08 Regulatory Guide 1.84 - CODE CASE ACCEPTABILITY ASME III DESIGN AND FABRICATION (Rev. 0, 6/74) 
 
CONFORMANCE
 
Of the design and fabrication code cases used by the licensee, all were either annulled, adopted in later versions of the Code, or endorsed in the Regulatory Guide, except Code Case
 
1360. Tubes are not explosively welded to tube sheets in Code Class 1 components. However, in the absence of other references to explosive welding in Section III or Section XI of the ASME
 
Code, this Code Case was cited as the basis for explosive tube plugging. This process was
 
used only after installation of steam generators when they were found to have defective tubes. 
 
ASME Code Case N-411 is used in the analysis of piping as accepted and endorsed in Rev. 28
 
of Regulatory Guide 1.84. See Section 3.7.1.3 for use of Code Case N-411.
 
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3A-1.85-1 REV 21  5/08 Regulatory Guide 1.85 - CODE CASE ACCEPTABILITY ASME SECTION III MATERIAL (Rev. 0, 6/74) 
 
CONFORMANCE
 
All the materials code cases that may have been used by the licensee were either annulled, adopted in later versions of the Code, or endorsed in the Regulatory Guide. Therefore, the
 
Farley Nuclear Plant conforms to the Guide. 
 
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3A-1.86-1 REV 21  5/08 Regulatory Guide 1.86 - TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS (6/74) 
 
CONFORMANCE
 
The termination of the operating license and the subsequent decommissioning of the Farley
 
Nuclear Plant will be carried out in conformance with the regulations applicable at that time. 
 
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3A-1.87-1 REV 21  5/08 Regulatory Guide 1.87 - CONSTRUCTION CRITERIA FOR CLASS 1 COMPONENTS IN ELEVATED TEMPERATURE REACTORS (SUPPLEMENT TO
 
ASME SECTION III CODE CASES 1592, 1593, 1594, 1595 AND
 
1596) (Rev. 0, June 1974) 
 
CONFORMANCE
 
This guide is not applicable to the Farley Nuclear Plant. 
 
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3A-1.88-1 REV 21  5/08 Regulatory Guide 1.88 - COLLECTION, STORAGE, AND MAINTENANCE OF NUCLEAR POWER PLANT QUALITY ASSURANCE RECORDS (Rev. 0, 8/74) 
 
CONFORMANCE
  [HISTORICAL] [Compliance with ANSI N45.2.9-1974 pr ovisions, which constitutes generally acceptable requirements for collection, storage , and maintenance of nuclear power plant quality assurance records, as stated in Regulatory Guide 1.88, is discussed in subsection 17.2.17.] 
 
The SNC Quality Assurance Topical Report (QATR) is based on ASME NQA-1-1994 which incorporates the requirements of ANSI N45.2.9. Accordingly, the requirements  for collection, storage, and maintenance of quality assurance records are described in the QATR.
 
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3A-1.95-1 REV 21  5/08 Regulatory Guide 1.95 - PROTECTION OF NUCLEAR POWER PLANT CONTROL ROOM OPERATORS AGAINST AN ACCIDENTAL CHLORINE
 
RELEASE (Rev. 0, 2/75)
 
CONFORMANCE
 
Chlorine storage locations onsite are discussed in section 2.2 and subsection 3.4.1. This
 
discussion is provided for historical purposes only since all significant quantities of chlorine, i.e.,
single containers greater that 150 pounds, have been removed from the plant site. Control of
 
chlorine is in accordance with Regulatory Guide 1.95.
 
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3A-1.99-1 REV 21  5/08 Regulatory Guide 1.99, Rev.2 - RADIATION EMBRITTLEMENT OF REACTOR VESSEL MATERIALS (MAY 1988) 
 
CONFORMANCE
 
The methodology for determining the effect of neutron irradiation on the reactor vessel beltline
 
materials conforms with the recommendations of Regulatory Guide 1.99, Revision 2, as
 
described in subsection 5.2.4.3.
 
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3A-1.108-1 REV 21  5/08 Regulatory Guide 1.108 - PERIODIC TESTING OF DIESEL GENERATOR UNITS USED AS ONSITE ELECTRIC POWER SYSTEMS AT NUCLEAR POWER
 
PLANTS (Rev. 1, 8/77)
 
CONFORMANCE
 
The conformance of diesel generator test frequencies to Regulatory Guide 1.108 is discussed in
 
subsection 8.3.1.1.8.
 
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3A-1.109-1 REV 21  5/08 Regulatory Guide 1.109 - CALCULATION OF ANNUAL DOSES TO MAN FROM ROUTINE RELEASES OF REACTOR EFFLUENTS FOR THE PURPOSE
 
OF EVALUATING COMPLIANCE WITH 10 CFR PART 50, APPENDIX I (Rev. 1, 10/77)
 
CONFORMANCE
 
Compliance with Regulatory Guide 1.109 is discussed in detail in paragraphs 11.2.8, 11.2.9, and 11.3.9.
 
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3A-1.111-1 REV 21  5/08 Regulatory Guide 1.111 - METHODS FOR ESTIMATING ATMOSPHERIC TRANSPORT AND DISPERSION OF GASEOUS EFFLUENTS IN ROUTINE
 
RELEASES FROM LIGHT-WATER-COOLED REACTORS (Rev. 1, 7/77)
 
CONFORMANCE
 
Compliance with Regulatory Guide 1.111 is discussed in detail in paragraph 2.3.5.2, and Table
 
2.3-17.
 
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3A-1.112-1 REV 21  5/08 Regulatory Guide 1.112 - CALCULATION OF RELEASES OF RADIOACTIVE MATERIALS IN GASEOUS AND LIQUID EFFLUENTS FROM LIGHT-WATER-
 
COOLED POWER REACTORS (Rev. O-R, 4/76)
 
CONFORMANCE
 
Compliance with Regulatory Guide 1.112 is discussed in detail in paragraph 11.1.1.2, and Table
 
11.1-7.
 
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3A-1.113-1 REV 21  5/08 Regulatory Guide 1.113 - ESTIMATING AQUATIC DISPERSION OF EFFLUENTS FROM ACCIDENTAL AND ROUTINE REACTOR RELEASES FOR THE PURPOSE OF IMPLEMENTING APPENDIX I (Rev. 1, 4/77)
 
CONFORMANCE
 
Compliance with Regulatory Guide 1.113 is discussed in detail in paragraph 11.2.8.
 
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3A-1.127-1 REV 21  5/08 Regulatory Guide 1.127 -  INSPECTION OF WATER-CONTROL STRUCTURES ASSOCIATED WITH NUCLEAR POWER PLANTS (Rev. 1, 03/78)
 
CONFORMANCE
 
The service water pond dam and spillway inspecti ons during the period of extended operation meet the intent of the guidance provided in NRC Regulatory Guide 1.127, Revision 1. See
 
chapter 18, subsection 18.2.3.
 
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3A-1.155-1 REV 21  5/08 Regulatory Guide 1.155 - STATION BLACKOUT (August 1988)
 
CONFORMANCE
 
Compliance with Regulatory Guide 1.155 is discussed in detail in paragraph 8.3.1.2.F.
 
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3A-1.163-1 REV 21  5/08 Regulatory Guide 1.163 - PERFORMANCE-BASED CONTAINMENT LEAK-TEST PROGRAM (September 1995)
 
CONFORMANCE
 
Farley Nuclear Plant has established a Containment Leakage Rate Testing Program to
 
implement the requirements of 10 CFR 50 Appendix J, Option B, consistent with Regulatory
 
Guide 1.163.
 
Regulatory Guide 1.163 endorses Nuclear Energy Institute (NEI) 94-01 Revision 0 dated July
 
26, 1995, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50
 
Appendix J", with some exceptions. NEI 94-01 endorses ANSI/ANS - 56.8 - 1994, "Containment System Leakage Testing Requirements" for detailed descriptions of the technical methods and techniques for performing containment leakage tests with some exceptions. In
 
addition, SNC maintains the option to use the Bechtel Topical Report BN-TOP-1, "Testing
 
Criteria for Integrated Leak - Rate Testing of Primary Containment Structures for Nuclear Power
 
Plants," Revision 1, November 1972, method for performing Type A tests. 
 
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3A-1.182-1 REV 21  5/08 Regulatory Guide 1.182 - Assessing and Managing Risk before Maintenance Activities at Nuclear Power Plants (May 2000)
CONFORMANCE FNP conforms to the guidance provided in this Regulatory Guide for complying with the provisions of 10 CFR 50.65(a)(4). This Regulatory Guide states that section 11, "Assessment of Risk Resulting from Performance of Maintenance Activities," dated February 11, 2000, of NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," provides methods that are acceptable to the NRC staff for complying with the provisions of 10 CFR 50.65(a)(4).
 
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3A-1.190-1 REV 21  5/08 Regulatory Guide 1.190 - CALCULATIONAL AND DOSIMETRY METHODS FOR DETERMINING PRESSURE VESSEL NEUTRON FLUENCE (March 2001)
 
CONFORMANCE
 
Neutron fluence calculation procedures and dosimetry methods conform to the
 
recommendations of Regulatory Guide 1.190.
 
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3A-1.194-1 REV 21  5/08 Regulatory Guide 1.194 - ATMOSPHERIC RELATIVE CONCENTRATIONS FOR CONTROL ROOM RADIOLOGICAL HABITABILITY
 
ASSESSMENTS AT NUCLEAR POWER PLANTS (June 2003)
 
CONFORMANCE
 
The design guidance and assumptions of Regulatory Guide 1.194 are used with plant
 
meteorological data for years 2000 through 2003 to calculate control room and technical support center X/Q for releases from the containment, containment equipment hatch, and plant
 
vent. The plant meteorological data have been supplemented by a data set constructed to
 
generate a 4.5-year data set with an overall frequency from the SSE, S, and SSW directions
 
similar to that in the 1971 through 1975 data set presented in section 2.3.
 
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3A-1.195-1 REV 21  5/08 Regulatory Guide 1.195 - METHODS AND ASSUMPTIONS FOR EVALUATING RADIOLOGICAL CONSEQUENCES OF DESIGN BASIS
 
ACCIDENTS AT LIGHT-WATER NUCLEAR POWER REACTORS (May 2003)
 
CONFORMANCE
 
The design guidance and assumptions of Regulatory Guide 1.195 are used as applicable in the
 
evaluation of a fuel handling accident in the containment except:
 
Dose conversion factors described in subsections 4.1.2 and 4.1.4 are not used. Instead, dose conversion factors from ICRP 30 are used as described in table 15B-1.
 
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3A-1.196-1 REV 21  5/08 Regulatory Guide 1.196 - CONTROL ROOM HABITABILITY AT LIGHT-WATER NUCLEAR POWER REACTORS (May 2003)
 
CONFORMANCE
 
Regulatory Guide 1.196 provided detail guidance regarding control room habitability and the
 
documentation of the control room habitability licensing bases as well as a program for testing
 
the systems. FNP complies with this RG with the following exceptions:
: 1. FNP implemented a Technical Specification (TS) change that is an acceptable alternative to TS change identified in RG 1.196 which includes control room
 
envelope (CRE) integrity testing and periodic assessments. These TS changes
 
implement a Control Room Integrity Program.
: 2. FNP continues to use RG 1.52, Revision 0, for the design. FNP will update to Revision 3, dated June 2001, for testing only.
: 3. FNP uses RG 1.140 as information only for nonsafety-related air filtration systems. 4. FNP uses ASHRAE Guideline 1-1996 as reference only for maintenance programs for systems that handle hazar dous chemicals and smoke challenges.
: 5. FNP uses RG 1.197, dated May 2003, for testing of the CRE. FNP has completed initial testing and will perform future testing of the Control Room
 
Habitation System (CRHS) in accordance with RG 1.197, with the exceptions as
 
defined in NEI 99-03, Rev. 1, Appendix EE, "ASTM E741 Exceptions."
: 6. Surveys of hazardous chemicals will continue to be performed on a frequency of every 6 years. Offsite will be conducted in the year the tracer gas testing is
 
scheduled, and the onsite will be offset by 3 years.
 
FNP-FSAR-3A
 
3A-1.197-1 REV 21  5/08 Regulatory Guide 1.197 - DEMONSTRATING CONTROL ROOM ENVELOPE INTEGRITY AT NUCLEAR POWER REACTORS (May 2003)
 
CONFORMANCE
 
Regulatory Guide 1.197 provides an approach acceptable to the NRC staff for measuring
 
inleakage into the control room and associated rooms and areas at nuclear power reactors. 
 
The amount of inleakage is an input to the design of the control room, and periodic verification
 
of the inleakage provides assurance that the control room will be habitable during normal and
 
accident conditions. This guide provides guidance on methods acceptable to the staff for
 
determining control room envelope (CRE) integrity for the purpose of confirming that the reactor
 
meets GDC-19.
 
Farley Nuclear Plant conforms to this Regulatory Guide with the following exceptions:
 
The following exceptions as defined in NEI 99-03, Rev. 1, Appendix EE, "ASTM E-741
 
Exceptions."
* These paragraphs may be totally excluded from implementation:  6.6.1, 6.6.2, 6.7,.6.7.1, 6.7.2, 8.5.4, 9.5.3, 9.5.4, 11.1.1, 12.3.2, 12.3.2.2, 13.2.1.2, 13.2.2, 13.4.2.
 
Other editions are acceptable and may require similar exceptions.
* Use sections 1 through 5 only to define the test method and the equipment to be used.
* In section 8.5.3.1 a decay test using the regression method may be used to obtain confidence intervals as a part of the regression calculation.
* In section 9.2.1 the standard is not typically used when there is a nonsteady flow since such a test would only permit establishing bounds on the inleakage.
* Sections 9.2.3.1, 9.2.3.2, 9.2.3.3, and 9.2.3.4 are not typically used since makeup flowrate is typically used to estimate the anticipated concentration for an assumed
 
tracer gas injection flowrate.
* Section 9.4.2 is not followed since a statistically significant number of samples are usually taken over 1 or 2 hours following the establishment of equilibrium.
* Sections 9.5.3.1 and 9.5.3.2 calculations are not used since the vendor demonstrates that concentration in CRE is not changing before making
 
measurements designed to calculate total inleakage.
* Section 10 is not used in total.
 
FNP-FSAR-3A
 
3A-1.197-2 REV 21  5/08
* Section 11.1 is not used to measure indoor and outdoor temperatures or wind speed and direction, unless there is a direct need for the information.
* Section 15 is not used in total.
* Section 16 is not used in total. The vendor's report is to present the theory, data analysis, sampling locations, operating conditions, procedures, quality assurance
 
records for the particular plant work order, data, calculations, and references.
* Section 17 is not used in total. The information is useful, but most vendors performing the test are highly experienced in many industrial settings and are
 
familiar with these cautions and conditions. Uncertainty analysis or precision analysis may use the ANSI PT 19.1 Standard to calculate the 95-percent confidence
 
intervals. The ANSI PT 19.1 Standard is not listed in Table D-1 since it is unrelated
 
to the actual leakage determination.
 
FNP-FSAR-3
 
3B-1 REV 21  5/08
[HISTORICAL] [APPENDIX 3B CONTAINMENT PROOF TESTS
 
The basic purpose of a structural proof test is to subs tantiate that the containment can, in fact, carry the pressure load for which it is designed. By subjecting the containment to some degree of overpressure, the test can show that the containment has that degree of margin over design pressure and there would not be
 
an incipient failure as might be the case if it were only tested at design pressure. Most previous steel and reinforced concrete containments and a number of the European prestressed concrete reactor vessels have been tested to 115 percent of design pressure.
The FNP containment is proof tested at 115 percent of design pressure. 
 
The prestressed containment relies mainly upon the te nsile strength of the tendons for its ultimate strength. The secondary stresses of the containment are isolated from the tendons. At ultimate capacity of the containment, the secondary stresses and the therm al stresses are relieved by local cracking of the concrete and the tendons are generally subjected to internal pressure and dead load only. 
 
In an evaluation of the containment overall margin of safety it is recognized that an exact determination is not a feasible requirement; however, it is possible to predict what is a reasonable value for the margin of safety. 
 
As pointed out in Appendix 3D, the load factors a ssociated with the pressure resulting from a LOCA will be the largest considered in the analysis and design of the containment. This is due to the fact that the degree of certainty for the magnitude of the pressure is l ess than that of the other loads. In view of this, the calculations to evaluate the margin of safety for the containment is based on the determination of what magnitude of pressure could be resisted at ultimate as a function of the design pressure. 
 
Consequently, the margin of safety is defined as a safety factor which, when multiplied by the design pressure, will result in the projected pressure that can be resisted by the ultimate capacity of the containment. 
 
The calculations to determine this safety factor, which defines the margin of safety, are based on the strength of the prestressing steel, reinforcing steel, and the concrete. The additive strength that might be
 
gained by the resistance afforded by the steel liner pl ate is not included because the leaktight membrane is not considered as a structural component member. 
 
Limit strength calculations are developed to ascertai n the minimum factor of safety. They are then compared to actual laboratory test results on prestressed concrete model structures for primary containments which are similar in form except for th e end closure which in the model studies have been flat slabs.
]
FNP-FSAR-3 3C-i REV 21  5/08 
[HISTORICAL] [3C. MECHANICAL SPLICING REINFORCING BAR USING THE CADWELD PROCESS Page  3C.1  SCOPE.......................................................................................................................................3C-1
 
3C.2  PROCESS...................................................................................................................................3C-1
 
3C.3  QUALIFICATIONS OF OPERATORS......................................................................................3C-1
 
3C.4  PROCEDURE............................................................................................................................3C-1
 
3C.5  ONSITE USER TESTS...............................................................................................................3C-3
 
3C.6  JOINT ACCEPTANCE STANDARDS........................................................................................3C-4
 
3C.7  REPAIRS...................................................................................................................................3C-4
] 
 
FNP-FSAR-3
 
3C-1 REV 21  5/08 
[HISTORICAL] [APPENDIX 3C MECHANICAL SPLICING REINFORCING BAR USING THE CADWELD PROCESS 3C.1 SCOPE Mechanical splicing of deformed reinforcing bar for full tensile loading is accomplished with Cadweld
 
connectors. The average tensile strength of the C adweld joints is greater than the minimum tensile strength for the particular grade of reinforcing steel as specified in the appropriate ASTM standard. The minimum tensile strength of the splices exceeds 125 percen t of the minimum yield strength for each grade
 
of reinforcing steel as specified in the appropriate ASTM standard. 
 
3C.2 PROCESS All splices are made by the Cadweld process (Erico Products, Inc.), using clamping devices, sleeves, charges, and so on, as specified by the Cadwel d Instruction Sheets for "T" series connections. 
"C" series and C-16 series materials are not permitted. 
 
3C.3 QUALIFICATIONS OF OPERATORS Prior to the production splicing of reinforcing bars, each operator or crew, including the foreman or supervisor for that crew, prepares and tests a joint fo r each of the positions used in production work.
These splices are made and tested in strict accordance with the specification using the same ASTM grade and size of bar spliced in the production work. To qualify, the completed splices must meet the "Joint Acceptance Standards" for workmanship, visual quality, and minimum tensile strength. A list containing
 
the names of qualified operators and their qualification t est results is maintained at the jobsite. 
 
3C.4 PROCEDURE All joints are made in strict accordance with the manufacturer's instructions as presented in Erico
 
Products Bulletin RB10M-670, 1970 Cadweld Rebar Splicing," plus the following additional requirements: 
 
A. A manufacturer's representative, experienced in Cadweld splicing of reinforcing bar, is present at the jobsite at the outset of the work to demonstrate the equipment and
 
techniques used for making quality splices. He is also present for at least the first 50 production splices to observe and verify that the equipment is being used correctly
 
and that quality splices are being obtained. 
 
B. The splice sleeves, cartridges, asbestos wicking, ceramic inserts, and graphite parts are stored in a clean, dry area with adequate protection to prevent absorption of moisture. 
 
C. Each splice sleeve is visually examined imme diately prior to use to ensure the absence of rust and other foreign material on the inner surface. 
 
FNP-FSAR-3
 
3C-2 REV 21  5/08  D. The graphite molds are preheated with an oxyacetylene torch to 300°F minimum to drive off moisture immediately prior to use.
E. Bar ends to be spliced are in good c ondition with full-size, undamaged deformations. The bar ends are power brushed to remove all loose mill scale, rust, concrete and other
 
foreign material. Prior to power brushing, all water, grease, and paint is removed by
 
heating the bar ends with an oxyacetylene or propane torch. 
 
F. A permanent line, marked 12 inches back from the end of each bar, serves as a reference point to confirm that the bar ends are properly centered in the splice sleeve. 
 
G. Immediately before the splice sleeve is pla ced in final position, the previously cleaned bar ends are preheated with an oxyacetylen e or a propane torch to insure complete absence of moisture. 
 
H. Special attention is given to maintaini ng the alignment of sleeve and guide tube to insure a proper fill. 
 
I. When the temperature is below freezing or the relative humidity is above 65 percent, the splice sleeve is externally preheated with an oxyacetylene or propane torch after all materials and equipment are in position. 
 
J. The reinforcing bar deformations which b ecome engaged in the Cadweld splice are not ground, flame cut, or altered in any way except for the longitudinal ribs, which are ground to a diameter not less than the other bar deformations. 
 
K. An adequate escape route is provided for gases generated during the casting of horizontal splices. For splices in bars smaller than No. 11, this is done by inserting a
 
hairpin piece of soft, twisted wire at the top of the splice between the rebar and the
 
sleeve. 
 
L. The packing material at the ends of the horizontal splices and at the top of the vertical splices is not hard packed. The material is firmly in place but loose enough to allow the escape of gases. 
 
3C.5 ONSITE USER TESTS The onsite user test program for reinforcing steel splices is described below: 
 
A. Every operator is required to pass a qualification test. 
 
B. All splices are visually inspected. As i ndicated in section 3C.7, unsatisfactory splices are replaced. 
 
C. For each crew, after qualification, test s are made for each position as follows: 
 
FNP-FSAR-3
 
3C-3 REV 21  5/08  Sister Splice Program The following tensile program is used: 
 
One out of the first lot of 10 production splices for each position, bar size and grade of bar.
One production splice and three "sister spli ces" from the next 90 splices, for each position, bar size, and grade of bar. 
 
Three splices out of the next and subsequent lots of 100 splices for each position, bar size, and grade of bar. One-fourth of these splices will be from production splices and three fourths from "sister splices."
 
A "sister splice" is defined as a 3-foot-l ong test bar spliced in sequence with, and in an otherwise identical manner as, the production splices. 
 
3C.6 JOINT ACCEPTANCE STANDARDS The following criteria are used for judging the acceptability of Cadweld joints: 
 
A. Sound, nonporous filler metal must be visible at both ends of the splice sleeve and at the tap hole in the center of the splice sleeve.
Filler metal is usually recessed 1/4 inch from the end of the sleeve due to the packing material. This recess is not considered as poor
 
fill.
B. Splices which contain slag or porous metal in the riser, tap hole, or at the ends of the sleeves (general porosity) are rejected. A si ngle shrinkage bubble present below the riser is not detrimental and is distinguished fr om general porosity as described above. 
 
C. The Cadweld splices, both horizontal and vertical, may contain voids at either or both ends of the Cadweld splice sleeve. At the end of the Cadweld splice sleeves, the
 
acceptable size void for a No. 18 splice does not exceed 3 square inches per end of splice
 
sleeve. The area of the void is assumed to be the circumferential length as measured at the inside face of the sleeve multiplied by the maximum depth of wire probe minus 3/16 inch. 
 
D. The average tensile strength of the Cadw eld joints shall be greater than the minimum tensile strength for the particular grade of reinforcing steel as specified in the appropriate ASTM standard. The minimum strength of the Cadweld joints must be
 
greater than 125 percent of the specified minimum yield strength for the particular bar. 
 
3C.7 REPAIRS Joints which do not meet the quality acceptance standards of section 3C.6 are rejected and completely removed. The bars are then rejoined with a new splice.
]
FNP-FSAR-3
 
3D-1 REV 21  5/08 APPENDIX 3D JUSTIFICATION FOR LOAD FACTORS AND LOAD COMBINATIONS USED IN DESIGN EQUATIONS FOR THE CONTAINMENT The load factors and load combinations in the factored load design equations represent the consensus of the individual judgments of a group of Bechtel engineers and consultants who are
 
experienced in both structural and nuclear power plant design. Their judgment has been
 
influenced by current and past practice, by the degree of conservativeness inherent in the basic
 
loads, and particularly by the probabilities of coincident occurrences in the case of accident, wind, and seismic loads. 
 
The following discussions explain the justification for the individual factors, particularly as they
 
apply to containments.
 
A. Dead Load
--Dead load in a large structure such as this is easily identified and its effect can be accurately determined at each point in the containment. For dead
 
load in combination with accident and seismic or wind loads, a load factor of
 
1.0 is used for all load combinations. 
 
B. Live Load
--The live load that is present along with accident and seismic or wind loads produces a very small portion of the stress at any point. Also, it is
 
extremely unlikely that the full live load is present over a large area at the time of
 
an unusual occurrence. For these reasons, live loads are not included in the
 
factored load design equations. 
 
C. Seismic
--The one-half safe shutdown earthquake (SSE) that has been selected is considered to be the strongest possible earthquake which could occur during
 
the life of the plant. In addition to the one-half SSE, a safe shutdown earthquake
 
which defines the maximum credible earthquake that could occur at the site, is
 
considered in design. Category I structures are designed so that no loss of
 
function results from the safe shutdown earthquake. Consequently, the
 
probability of an SSE causing the loss-of- coolant accident (LOCA) is very small.
 
For this reason, the two events, SSE and LOCA, are considered together, but at
 
much lower load factors than those applied to the events separately. The
 
earthquake load factors of 1.25 and 1.0 are conservative for one-half SSE and
 
safe shutdown earthquake combination with the factored LOCA. 
 
D. Wind--Loads are determined from the design tornado wind speed. Since the containment is designed for this extreme wind it is inconceivable that the wind
 
would cause a LOCA. Therefore, wind loads are being considered with accident
 
loads. A load factor of 1.0 is applied to the tornado load. 
 
F. LOCA--The design pressure and temperature will be based on the operation of partial safeguards equipment using emergency diesel power. 
 
European practice has been to use a load factor of 1.5 on the design pressure.(a) This factor is reasonable and is adopted for this design. The probabilities of a
 
LOCA occurring simultaneously with a maximum wind or seismic disturbance are FNP-FSAR-3
 
3D-2 REV 21  5/08 very small; therefore, a reduced load factor of 1.25 is used for the combination of events. 
 
In all cases the design temperature is defined as that corresponding to the unfactored pressure. At 1.5 P the temperature is somewhat higher than the
 
temperature at 1.0 P. It would be unrealistic to apply a corresponding
 
temperature factor of 1.5 since this could occur only with a pressure much
 
greater than a pressure of 1.5 P. 
: a. Refer T. C. Waters and N. T. Barrett, "Prestressed Concrete Pressure Vessels for
 
Nuclear Reactors," J. Brit. Nucl. Soc.
2, 1963.
FNP-FSAR-3
 
3E-1 REV 21  5/08 
[HISTORICAL] [APPENDIX 3E JUSTIFICATION FOR CAPACITY REDUCTION FACTORS ( - FACTORS) USED IN DETERMINING CAPACITY OF CONTAINMENTS The -factors are provided to allow for variations in materials and workmanship. In the ACI Code,  varies with the type of stress or member considered
; that is, with flexure, bond or shear stress, or compression. 
 
The -factor is multiplied into the basic strength equati on or, for shear, into the basic permissible unit shear to obtain the dependable strength. The basic strength equation gives the "ideal" strength assuming materials are as strong as specified, as shown on the drawings, the workmanship is excellent, and the strength equation itself is theoretically correct.
The practical, dependable strength may be something less since all these factors vary. 
 
The ACI Code provides for these variables by using these -factors: 
  = 0.90 for concrete in flexure 
  = 0.85 for diagonal tension, bond, and anchorage 
  = 0.75 for spirally reinfo rced, concrete compression members 
  = 0.70 for tied compression members is larger for flexure because th e variability of steel is less than that of concrete. The  values for columns are lower (favoring the toughness of spiral columns over tied columns) because columns fail in compression where concrete strength is critical. Also, it is possible that the analysis might not combine the worst combination of axial load and moment, and sin ce the member is critical in the gross collapse of the containment, a lower value is used. 
 
The additional  values used represent Bechtel's best judgm ent of how much understrength should be assigned to each material and condition not covered directly by the ACI Code. The additional -factors have been selected based on material quality in relation to the existing -factors. 
 
Conventional concrete design of beams requires that th e design be controlled by yielding of the tensile reinforcing steel. This steel is generally spliced by lapping in an area of reduced tension. For members in flexure, ACI uses  = 0.90. The same reasoning is applied in assigning a value of  = 0.90 to reinforcing steel in tension, which now includes axial tension. However, the code recognizes the possibility of reduced bond of bars at the laps by specifying a  of 0.85. For lap splices in the auxiliary building and structures other than the containment, a -factor of 0.90 is used because of the excessive lap used to splice. Mechanical and welded splices devel op at least 125 percent of the yield strength of the reinforcing steel. Therefore,  = 0.90 is recommended for this type of splice.
The only significantly new value introduced is  = 0.95 for prestressed tendons in direct tension. A higher  value than for conventional reinforcing steel is allowed because (1) during installation the tendons are each jacked to about 94 percent of their yi eld strength, so in effect each tendon has been proof tested, and (2) the method of manufacturing prestressing steel (cold drawing and stress relieving) ensures a higher quality product than conventional reinforcing steel.
]
FNP-FSAR-3F
 
3F-i REV 21  5/08 3.F COMPUTER PROGRAMS USED IN STRUCTURAL ANALYSES TABLE OF CONTENTS Page 3F.1 INTRODUCTION....................................................................................................3F-1
 
3F.2 COMPUTER PROGRAMS USED FOR THE STRUCTURAL ANALYSES BY BECHTEL POWER CORPORATION.....................................................................3F-1
 
3F.2.1 CE 316 FINITE ELEMENT STRESS ANALYSIS (FINEL)...............3F-2 3F.2.2 CE 639-2 FORCES AND PRESSURES ACTING ON THE DOME DUE TO PRESTRESSING OF TENDONS.............................. 3F-2 3F.2.3 CE 779 STRUCTURAL ANALYSIS PROGRAM (SAP)....................... 3F-3 3F.2.4 ME 620 HEAT CONDUCTION.............................................................3F-4 3F.2.5 AXISYMMETRIC SHELL AND SOLID COMPUTER PROGRAM (ASHSD)...............................................................................................3F-5
 
3F.3 COMPUTER PROGRAMS USED IN THE SEISMIC ANALYSES BY BECHTEL POWER CORPORATION.....................................................................3F-6
 
3F.3.1 CE 309 STRUCTURAL ENGINEERING SYSTEMS SOLVER (STRESS).............................................................................................3F-6 3F.3.2 CE 591 SPECTRA ANALYSIS.............................................................3F-7 3F.3.3 CE 611 TIME-HISTORY RESPONSE ANALYSIS................................3F-8 3F.3.4 CE 617 MODES AND FREQUENCIES EXTRACTION........................3F-9 3F.3.5 CE 641 EARTHQUAKE RESPONSE SPECTRUM ANALYSIS OF STRUCTURES.....................................................................................3F-9 3F.3.6 CE 655 STRESS-DYNAMIC ANALYSIS............................................3F-10 3F.3.7 CE 785 SPECTRUM SUPPRESSING................................................3F-10 3F.3.8 CE 786 SPECTRUM RAISING...........................................................3F-11 3F.3.9 CE 791 SPECTRA ANALYSIS...........................................................3F-12 3F.3.10 CE 792 GENERATION OF RESPONSE SPECTRA FROM STRONG-MOTION EARTHQUAKE RECORDS................................3F-12
 
3F.4 COMPUTER PROGRAMS USED IN THE SEISMIC ANALYSES BY SOUTHERN COMPANY SERVICES, INC...........................................................3F-13
 
3F.4.1 EN 3426 ICES STRUDL II, THE STRUCTURAL DESIGN LANGUAGE.........................................................................3F-13 3F.4.2 EN 3423 MODAL ANALYSIS.............................................................3F-14 3F.4.3 EN 8043 RESPONSE FOR SPECTRA PLOTS..................................3F-16 3F.4.4 EN 8045 RESPONSE SPECTRA.......................................................3F-16 3F.4.5 EN 8046 SPECTRA PLOTTER..........................................................3F-17 3F.4.6 PS + CAEPIPE...................................................................................3F-17
 
FNP-FSAR-3F
 
3F-ii REV 21  5/08 TABLE OF CONTENTS Page
 
3F.5 COMPUTER PROGRAMS USED IN THE STRUCTURAL ANALYSES BY VENDORS AND SUBCONTRACTORS.....................................................................3F-18
 
3F.5.1 COMPUTER PROGRAM USED BY CHICAGO BRIDGE AND IRON COMPANY.....................................................................................3F-18
 
3F.5.1.1 CB&I Program 7-81.................................................................3F-18
 
3F.5.2 COMPUTER PROGRAMS USED BY INLAND-RYERSON CONSTRUCTION PRODUCTS COMPANY............................................3F-19
 
3F.5.2.1 Program WDINT.....................................................................3F-19 3F.5.2.2 Program POCKET..................................................................3F-20 3F.4.2.3 Program NUFRCOHO............................................................3F-21
 
3F.5.3 COMPUTER PROGRAM USED BY WHITING CORPORATION.............3F-22
 
3F.5.3.1 STARDYNE DYNRE 4............................................................3F-23
 
FNP-FSAR-3F
 
3F-iii REV 21  5/08 LIST OF TABLES 3F-1 Computer Programs Used for Category I Structural Analyses by Bechtel Power Corporation 
 
3F-2 Computer Programs Used in Seismic Analyses by Bechtel Power Corporation 
 
3F-3 Computer Programs Used in Seismic Analys es by Southern Company Services, Inc. 
 
3F-4 Computer Programs Used in Category I Structural Analysis by Vendors and Subcontractors 
 
FNP-FSAR-3F
 
3F-1 REV 21  5/08 APPENDIX 3F COMPUTER PROGRAMS USED IN STRUCTURAL ANALYSES
 
3F.1 INTRODUCTION A number of computer programs are used in the structural analyses of the Category I structures.
 
They are described and documented in this appendix. These computer programs are divided
 
into four groups, as follows: 
 
A. Computer programs used in the structural analyses by Bechtel Power Corporation. 
 
B. Computer programs used in the seismic analyses by Bechtel Power Corporation. 
 
C. Computer programs used in the seismic analyses by Southern Company Services, Inc. (SCS). 
 
D. Computer programs used in the structural analyses by vendors and subcontractors. 
: 1. Chicago Bridge and Iron Company. 
: 2. Inland-Ryerson Construction Products Company. 
: 3. Whiting Corporation. 
 
3F.2 COMPUTER PROGRAMS USED FOR THE STRUCTURAL ANALYSES BY BECHTEL POWER CORPORATION Several computer programs are used by Bechtel Power Corporation in the structural analyses of
 
the Category I structures. They are listed below and described and documented in the following
 
sections: 
 
A. Bechtel CE 316-4 Finite Element Stress Analysis (FINEL). 
 
B. Bechtel CE 639-2 Forces and Pressures Acting on the Dome due to Prestressing of Tendons. 
 
C. Bechtel CE 779 Structural Analysis Program (SAP). 
 
D. Bechtel ME 620 Heat Conduction. 
 
E. Axisymmetric Shell and Solid Computer Program (ASHSD).
 
A summary of these computer programs is presented in table 3F.1. 
 
FNP-FSAR-3F
 
3F-2 REV 21  5/08 3F.2.1  CE 316-4 FINITE ELEMENT STRESS ANALYSIS (FINEL)
A. Description 
 
The program performs the static analyses of plane or axisymmetric structures using the finite element method, in which a structure is idealized as an
 
assemblage of finite elements. The finite elements are of either triangular or
 
quadrilateral shape, connected at their corners (nodal points). The applied loads
 
may be concentrated, uniformly distributed, or inertial, or may be temperature
 
distributions. At boundaries, displacements may be forced. 
 
The program develops the force displacement relationship (element stiffness matrix) for each individual element from its geometry and material properties. 
 
The element relationships are then assembled into an overall structure force
 
displacement relationship (structure stiffness matrix). Equilibrium equations are
 
developed for each degree of freedom at each nodal point in terms of the
 
structure force displacement relationship, the unknown nodal point displacement, and the externally applied nodal point forces. Finally, these equations are solved
 
simultaneously for the unknown nodal point displacements by a modified
 
Gaussian elimination scheme. Once the nodal point displacements are known, element stresses are calculated. 
 
B. Assumptions 
 
The stress and the strain are assumed to be constant within each element. 
 
C. Validation 
 
The program has been verified by manual calculations. Document traceability is available at Pacific International Computing Corporation. 
 
D. Extent of Application 
 
The program is used to compute stresses in the containment structure due to gravity, pressure, and thermal loads. 
 
3F.2.2  CE 639-2  FORCES AND PRESSURES ACTING ON THE DOME DUE TO PRESTRESSING OF TENDONS A. Description 
 
The program performs an analysis of forces and stresses that act on a dome due to prestressed tendons. The shape of the dome may be sphere-torus, sphere-cone, hemisphere, cone, or ellipsoid; and the tendons may be in one, two, or three directions. The program is capable of analyzing the prestress loss
 
of the dome tendons caused by frictional resistance, and the variation of the
 
prestressing forces due to seating of the anchorages. In addition, the program
 
may be used in preparing prestressing sequences.
FNP-FSAR-3F
 
3F-3 REV 21  5/08  B. Validation 
 
The necessary validation of CE 639-2 has been completed. The program was validated by an independent calculation performed in November 1974. 
 
C. Extent of Application 
 
The program is used to analyze the forces and pressures acting on the dome due to prestressing of tendons. It is also used in the preparation of the
 
prestressing sequences. 
 
3F.2.3  CE 779  STRUCTURAL ANALYSIS PROGRAM (SAP)
A. Description 
 
The program performs the static and dynamic analyses of linear, elastic, three-dimensional structures using the finite element method. The finite element
 
library contains truss and beam elements, plane and solid elements, plate and
 
shell elements, axisymmetric (torus) elements, and special boundary (spring)
 
elements. 
 
Element stresses and displacements are solved for either applied loads or temperature distributions. Concentrated loads, pressures or gravity loads may
 
be applied. Temperature distributions are assigned as an appropriate uniform
 
temperature change in each element. Prestressing may be simulated by using
 
artificial temperature changes on rod elements. 
 
Dynamic response routines are availabl e for solving arbitrary dynamic loads or seismic excitations using either modal superposition or direct integration. The
 
program can also perform response spectrum and time- history analyses. 
 
B. Validation 
 
The solutions to test problems have been demonstrated to be essentially identical to the results obtained using the ASKA program, which was developed
 
by Prof. A. J. Argyris (Institut fur Statik und Dynamik, Stuttgart) and to the Chan
 
and Fermin program. The test problem solutions have also been compared to, and found to be in agreement with, the solutions of the programs from the ASME
 
Library of Benchmark Computer Problems and Solutions. Document traceability
 
is available at Pacific International Computing Corporation. 
 
C. Extent of Application 
 
The program is used in the structural analysis of the containment shell at the region of the equipment hatch opening. 
 
FNP-FSAR-3F
 
3F-4 REV 21  5/08  D. Reference 
: 1. "A Refined Quadrilateral Element for Analysis of Plate Bending," Proc. (second) Conference on Matrix Methods in Structural Mechanics , Wright Patterson AFB, Ohio, 1968. 
 
3F.2.4  ME 620 HEAT CONDUCTION A. Description 
 
The program performs the transient or steady-state temperature analyses of plane or axisymmetric solids. Regional temperature distributions may be
 
determined due to prescribed boundary temperatures or boundary and internal
 
heat fluxes. 
 
The thermal analyses are carried out by the finite element technique coupled with a step-by-step time integration procedure. 
 
B. Validation 
 
The program has been verified by manual calculations. Document traceability is available at Pacific International Computing Corporation. 
 
C. Extent of Application 
 
The program is used to compute the temperature distribution for the containment at locations where the geometry of the structure is too complex for manual
 
calculation, such as wall and base slab intersections, and wall and ring girder
 
intersection, etc. 
 
3F.2.5  AXISYMMETRIC SHELL AND SOLID COMPUTER PROGRAM (ASHSD)
A. Description 
 
The program performs the static and dynamic analyses of linear, elastic, axisymmetric structures with axisymme tric or nonaxisymmetric loadings, utilizing the finite element technique. The program computes the element stresses and
 
nodal displacements due to uniform, concentrated, or pressure loads, or
 
temperature distributions, either over the surface area or through the wall
 
thickness. Prestress forces may be simulated by applying the forces as
 
equivalent concentrated temperature gradients. 
 
B. Validation 
 
The solutions of the program for various loadings have been demonstrated to be essentially identical to the results obtained by manual calculations and to those FNP-FSAR-3F
 
3F-5 REV 21  5/08 obtained from accepted experimental tests or analytical results published in technical literature. (See references 1 and 2.) 
 
C. Extent of Application 
 
The program is used in the analysis of the containment for nonaxisymmetric loadings. 
 
D.
 
==References:==
: 1. Ghosh, S. and Wilson, E. L., "Dynamic Stress Analysis of Axisymmetric Structures under Arbitrary Loading", Report No. EERC 69-10 , Univ. of California, Berkeley, Sept. 1969, pp 69-81.
: 2. "Topical Report on Dynamic Analysis of Reactor Vessel Internals under Loss-of-Coolant Accident Conditions with Application of Analysis to
 
CE 800 MWe Class Reactors", Combustion Engineering Report CENPD-42 , Combustion Engineering, Inc., Nuclear Power Department, Combustion Division,  Windsor, Conn., Appendix A. 
 
3F.3 COMPUTER PROGRAMS USED IN THE SEISMIC ANALYSIS BY BECHTEL POWER CORPORATION A number of computer programs are used by Bechtel Power Corporation in the seismic
 
analyses of the Category I structures. They are listed below and described and documented in
 
the following sections. 
: 1. Bechtel CE 309 Structural Engineering Systems Solver (STRESS). 
: 2. Bechtel CE 591 Spectra Analysis. 
: 3. Bechtel CE 611 Time-History Response Analysis. 
: 4. Bechtel CE 617 Modes and Frequencies Extraction. 
: 5. Bechtel CE 641 Earthquake Response Spectrum Analysis of Structures. 
: 6. Bechtel CE 655 Stress Dynamic Analysis. 
: 7. Bechtel CE 785 Spectrum Suppressing. 
: 8. Bechtel CE 786 Spectrum Raising. 
: 9. Bechtel CE 791 Spectra Analysis. 
: 10. Bechtel CE 792 Generation of Response Spectra from strong-motion Earthquake Records. 
 
FNP-FSAR-3F
 
3F-6 REV 21  5/08 A summary of these computer programs is presented in table 3F.2. 
 
3F.3.1  CE 309  STRUCTURAL ENGINEERING SYSTEMS SOLVER (STRESS)
A. Description 
 
STRESS is a programming system for the solution of structural engineering problems. The system is capable of executing the linear, elastic, static analyses
 
of two- and three-dimensional framed structures of the following types: 
: 1. Plane truss. 
: 2. Plane frame. 
: 3. Plane grid. 
: 4. Space truss. 
: 5. Space frame. 
 
The programming system was originally developed at Massachusetts Institute of Technology in 1964 and is now in the public domain. 
 
B. Validation 
 
The program has been verified by the ICES STRUDL II program. A sample problem of space frame analysis was run using the CE 309 program and the
 
commercially available versions (Version 1 and Version 2) of the ICES STRUDL
 
II program. The results from these runs were found to be identical. Document
 
traceability is available at Pacific International Computing Corporation. 
 
C. Extent of Application 
 
The program is used to obtain the flexibility matrices of the Category I structures.
The flexibility matrices are used in the dynamic analyses of the structures. 
 
D. Reference 
 
Fenjes, S.J., Logcher, R.D., and Mauch, S.P., Stress Reference Manual , The M.I.T. Press, Cambridge, Mass., 1964. 
 
3F.3.2  CE 591  SPECTRA ANALYSIS A. Description 
 
The program computes and plots the response spectra for any ground excitation described in acceleration time coordinates, such as earthquakes, blasts, etc.
FNP-FSAR-3F
 
3F-7 REV 21  5/08  B. Validation 
 
The solutions to the program have been verified to be essentially identical to the results obtained by manual calculations. Document traceability is available at
 
Bechtel Power Corporation. 
 
C. Extent of Application 
 
The program is used to compute and plot the response spectra for the seismic analyses of Category I structures. 
 
3F.3.3  CE 611  TIME-HISTORY RESPONSE ANALYSIS A. Description 
 
The program performs the response time-history analysis of a structure subjected to an earthquake motion using the modal superposition technique. 
 
The response is calculated in terms of displacement, velocity, and acceleration
 
time-histories at selected points. Inertia forces, moments, and shears may be
 
computed for cantilevered structures. Maximum response values and time of
 
occurrence may be found. 
 
B. Validation 
 
The solutions of the program have been verified to be substantially identical to the results of manual calculations. Document traceability is available at Pacific
 
International Computing Corporation. 
 
C. Extent of Application 
 
The program is used to generate the time histories at all Category I equipment locations in the structures. 
 
D. References 
: 1. Biggs, J. M., Introduction to Structural Dynamics , McGraw-Hill, 1964. 
: 2. Hildebrand, F. B., Introduction to Numerical Analysis , McGraw-Hill, 1956. 
: 3. Hurty, W. C., Rubinstein, M. F., Dynamics of Structures , Prentice Hall, Inc., 1964. 
: 4. Kuo. S. S., Numerical Methods and Computers , Addison Wesley, 1965. 
 
FNP-FSAR-3F
 
3F-8 REV 21  5/08 3F.3.4  CE 617  MODES AND FREQUENCIES EXTRACTION A. Description 
 
This program provides a means for obtaining the natural frequencies and mode shapes of structural models. Input to the program consists of the model's
 
lumped masses and either the stiffness matrix or the flexibility matrix. If the
 
flexibility matrix is entered, the program provides for automatic inversion to a
 
stiffness matrix. 
 
The program extracts eigenvalues and ei genvectors from the input, using the Jacobi diagonalization method by successive rotations. 
 
B. Validation 
 
The program has been verified by manual calculations. Document traceability is available at Bechtel Power Corporation. 
 
C. Extent of Application 
 
The program is used to obtain the mode shapes and natural frequencies of all Category I structures. 
 
D. Reference 
 
Crandall, S., Engineering Analyses, A Survey of Numerical Procedures , McGraw-Hill, 1966. 
 
3F.3.5  CE 641  EARTHQUAKE RESPONSE SPECTRUM ANALYSIS OF STRUCTURES A. Description 
 
The program computes the response of a structure subjected to an earthquake motion, utilizing the response spectrum technique. The structure is defined in
 
terms of natural frequencies, mode shapes, lumped weights, and elevations. 
 
The earthquake is described in terms of a response spectrum curve. The
 
response values computed for each mode are combined, using the sum of the
 
absolute values and the square root of the sum of the squares (SRSS). 
 
B. Validation 
 
The solutions to the problem have been verified to be substantially identical to the results obtained by manual calculations. Document traceability is available at
 
Bechtel Power Corporation. 
 
FNP-FSAR-3F
 
3F-9 REV 21  5/08  C. Extent of Application 
 
The program is used to obtain the modal responses of all Category I structures. 
 
3F.3.6  CE 655  STRESS-DYNAMIC ANALYSIS A. Description 
 
The program is used in conjunction with CE 309, STRESS, to obtain the flexibility matrix or stiffness matrix of a structure. The matrix is then used with programs
 
such as CE 617, Modes and Frequencies Ex traction, to evaluate the dynamic characteristics of the structure. 
 
B. Validation 
 
The program by itself does not have the capability to analyze a structure. It merely extracts the output from the CE 309 STRESS program to build up a
 
flexibility/stiffness matrix for the same structure. Consequently, validation is not
 
necessary. 
 
C. Extent of Application 
 
The program is used to extract a flexib ility/stiffness matrix from CE 309 STRESS program. 
 
3F.3.7  CE 785  SPECTRUM SUPPRESSING A. Description 
 
The program generates a synthetic time-history to fit closely a given response spectrum curve. This is accomplished by modifying a given accelerogram so that the acceleration response spectrum may be locally suppressed at any given
 
frequency without significantly changing the remaining portions of the spectrum. 
 
The program computes the modified accelerogram and adjusts its maximum
 
acceleration to match the specified value, plots the modified accelerogram, and
 
calculates the modified accelerogram response spectrum. 
 
B. Validation 
 
The results of the spectrum comput ation of the program have been compared with those obtained by Bechtel programs CE 786 and CE 791. It was found that
 
the spectra of the same motion as computed by the three different programs are
 
essentially identical. Document traceability is available at Bechtel Power
 
Corporation. 
 
FNP-FSAR-3F
 
3F-10 REV 21  5/08  C. Extent of Application 
 
The program is used to generate the synthetic time-history used in the seismic analyses of Category I structures. 
 
3F.3.8  CE 786  SPECTRUM RAISING A. The program generates a synthetic time history to fit closely a given response spectrum curve. This is accomplished by modifying a given accelerogram so that the acceleration response spectrum may be locally raised at any given frequency
 
without significantly changing the remaining portions of the spectrum. The
 
program computes the modified accelerogram and adjusts its maximum
 
acceleration to match the specified value, plots the modified accelerogram, and
 
calculates the modified accelerogram response spectrum. 
 
B. Validation 
 
The results of the spectrum comput ation of the program have been compared with those obtained by Bechtel programs CE 785 and CE 791. It was found that
 
the spectra of the same motion as computed by the three different programs are
 
essentially identical. Document traceability is available at Bechtel Power
 
Corporation. 
 
C. Extent of Application 
 
The program is used to generate the synthetic time-history used in the seismic analyses of Category I structures. 
 
3F.3.9  CE 791  SPECTRA ANALYSIS A. Description 
 
The program computes and plots the response spectra for any ground excitation described in acceleration time coordinates, such as earthquakes, blasts, etc. 
 
B. Validation 
 
The solutions to the program have been verified to be essentially identical to the results obtained by manual calculations. Document traceability is available at
 
Bechtel Power Corporation. 
 
C. Extent of Application 
 
This program is used to compute and plot the response spectra for the seismic analyses of Category I structures. 
 
FNP-FSAR-3F
 
3F-11 REV 21  5/08 3F.3.10 CE 792  GENERATION OF RESPONSE SPECTRA FROM STRONG-MOTION EARTHQUAKE RECORDS A. Description 
 
The program computes the response spectra at specified values of frequencies from the input time-history which are generated by CE 611, Time-History
 
Response Analysis, and for given values of damping ratios. 
 
B. Assumptions 
 
The numerical method used for integration is based on the exact solution to the governing differential equation, assuming that the input acceleration time-history
 
varies linearly between consecutive data points. 
 
C. Validation 
 
The solutions to the program have been demonstrated to be substantially identical to the results obtained by manual calculations. Document traceability is
 
available at Bechtel Power Corporation. 
 
D. Extent of Application 
 
The program is used to generate response spectra at all equipment locations in Category I structures. 
 
E. Reference 
 
Nigam, N. C. and Jennings, P. C., "Digital Calculation of Response Spectra From Strong-motion Earthquake Records," C.I.T., 1968. 
 
3F.4 COMPUTER PROGRAMS USED IN SEISMIC ANALYSES BY SOUTHERN COMPANY SERVICES, INC.
Several computer programs are used by Sout hern Company Services, Inc., (SCS) in the seismic analysis of the Category I structures. They are listed below and are described and
 
documented in the following sections: 
 
A. Program EN 3426, ICES STRUDL II. 
 
B. Program EN 3423, Modal Analysis. 
 
C. Program EN 8043, Response Spectra Plots. 
 
D. Program EN 8045, Response Spectra. 
 
E. Program EN 8046, Spectra Plotter. 
 
FNP-FSAR-3F
 
3F-12 REV 21  5/08 A summary of these computer programs is presented in table 3F-3. 
 
3F.4.1  EN 3426  ICES STRUDL II, THE STRUCTURAL DESIGN LANGUAGE A. Description 
 
This program computes the flexibility ma trix of a mathematical beam model of a structure. This is accomplished by applying a unit load at one mass point and
 
calculating displacements at all mass points. The procedure is repeated for all
 
mass points. This matrix is used as input for program EN 3423 to generate
 
mode shapes and frequencies. 
 
B. Assumptions 
 
The beam theory is used in the analytical procedure of the stiffness analysis. It assumes a linear, elastic, static, small-displacement analysis. Member
 
properties are required, and the program treats the joint displacements as
 
unknowns. 
 
C. Validation 
 
The solutions of the program have been proven to be substantially identical to the results obtained by another program, SAP IV. Document traceability is
 
available at Southern Company Services, Inc. The STRUDL program is in the
 
public domain and was originally issued March 1968. It was obtained from the
 
Massachusetts Institute of Technology. The version used is ICES-VI-M2, STRUDL2-VI-MO, issued 1970. 
 
SAP IV is a Structural Analysis Program that originated at the University of California, Berkeley. The original version was issued in September 1970 and is in
 
the public domain. Southern Company Services, Inc., obtained SAP IV in
 
October 1973 from the University of California. 
 
The computer used is the IBM 370 155, together with its operation systems. 
 
D. Extent of Application 
 
The program is used to generate the flexibility matrix of all Category I structures. 
 
E.
 
==References:==
: 1. Fenves, S. J. and Branin, F., "Network - Topological Formulation of Structural Analysis", Journal of the Structural Division , ASCE, August 1963.   
: 2. Fenves, S. J., Mauch, S. P., and Kinra, R. J., "Treatment of Releases and Constraints in the Network Formulation of Structural Analysis,"
FNP-FSAR-3F
 
3F-13 REV 21  5/08 Technical Report, Structural Research Series, No. 299 , University of Illinois, Urbana, Ill., October 1965. 
 
3F.4.2  EN 3423  MODAL ANALYSIS A. Description 
 
The flexibility matrix of the mathemat ical model is obtained from program EN 3426. A diagonal mass matrix is used. Basically, the program solves for the
 
natural frequencies and mode shapes for the mathematical model. It calculates
 
the composite damping as a percent of critical damping by the modal weighing method. From the ground spectra the spectral acceleration is obtained for each
 
mode. The program then calculates the square root of the sum of the squares of
 
deflections, shears, moments, inertial forces, and spectral accelerations for each
 
mass point. These calculations are made for safe shutdown earthquake and
 
one-half safe shutdown earthquake. 
 
B. Assumptions 
 
A three-dimensional structure is repr esented by a mathematical model of lumped masses with weightless elastic columns acting as spring restraints to obtain the
 
response of the structure. The program solves for the natural frequencies and
 
mode shapes of a freely vibrating, undamped linear elastic system. 
 
C. Validation 
 
The solutions of the program have been proven to be substantially identical to the results obtained by another program, SAP IV. Document traceability is
 
available at Southern Company Service, Inc. SAP IV is a Structural Analysis
 
Program that originated at the University of California, Berkeley. The original
 
version was issued in September 1970 and is in the public domain. Southern
 
Company Services, Inc., obtained SAP IV in October 1973 from the University of California. 
 
The computer used is the IBM 370 155, together with its operation systems. 
 
D. Extent of Application 
 
The program is used to generate the response of the structure and obtain inertial forces that are applied to the original structure of all Category I structures. 
 
E.
 
==References:==
: 1. Invert matrix - Standard Gauss - Jordan Method. 
: 2. Computation of eigenvalues and eigenvectors - Diagonalization method originated by Jacobi and adapted by Von Neumann. 
 
FNP-FSAR-3F
 
3F-14 REV 21  5/08  3. Ralston, A. and Wilf, H. W., eds., Mathematical Methods for Digital Computers , John Wiley and Sons, N.Y., 1962, Chapter 7. 
 
3F.4.3  EN 8043  RESPONSE FOR SPECTRA PLOTS A. Description 
 
The program uses the modal superposition method as a solution technique.
Mode shapes, natural frequencies, participation factors, and viscous damping
 
obtained from program EN 3423 are used as input. The program determines the
 
response of each mode separately and then calculates the total response by
 
superposition. The time-histories of displacement and acceleration of each mass
 
point are calculated. The acceleration time-histories for desired mass points are
 
stored on tape and used as input for program EN 8045. 
 
B. Validation 
 
The solutions of the program have been proven to be substantially identical to the results obtained by another program, DYNAL. DYNAL (Dynamic Analysis Computer Program) is in the public domain and has been in use since 1970. 
 
Document traceability is available at Southern Company Services, Inc. 
 
C. Extent of Application 
 
The program is used to calculate the deflections and accelerations of the structure. It is also used to generate the time-history acceleration at required
 
mass points. 
 
D. Reference 
 
Scarborough, J. B., Numerical Mathematical Analysis , Johns Hopkins Press, Baltimore, 1956 
 
3F.4.4  EN 8045  RESPONSE SPECTRA A. Description 
 
The time-history of floor acceleration generated in program EN 8043 is used as input. EN 8045 is then used in the generation of floor response spectra
 
computed from the time-history motions at the floor desired. The floor response
 
spectra give the maximum response of single degree of freedom bodies of
 
different natural frequencies as a function of damping when these bodies are
 
subjected to a floor time- history. 
 
FNP-FSAR-3F
 
3F-15 REV 21  5/08  B. Validation 
 
The solutions of the program have been proven to be substantially identical to the results obtained by another program, DYNAL. DYNAL (Dynamic Analysis Computer Program) is in the public domain and has been in use since 1970. 
 
Document traceability is available at Southern Company Services, Inc. 
 
C. Extent of Application 
 
The program is used to generate floor response spectra at all equipment locations. 
 
D. Reference 
 
Scarborough, J. B., Numerical Mathematical Analysis , Johns Hopkins Press, Baltimore, 1956. 
 
3F.4.5  EN 8046  SPECTRA PLOTTER 
 
A. Description 
 
The program plots the floor response spectra. The maximum acceleration for a given frequency generated by program EN 8045 is used as input. The
 
acceleration versus frequency is plotted on semilogarithmic, three- cycle paper. 
 
B. Validation 
 
The graph plotted by the program has been reproduced by a manual method.
Both graphs were compared and found to be identical. Document traceability is
 
available at Southern Company Services, Inc. 
 
C. Extent of Application 
 
The program is used to plot all floor response spectra. 
 
D. Reference 
 
Programming Calcomp Pen Plotters , California Computer Products, Inc., Anaheim, Calif., June 1968. 
 
3F.4.6  PS+CAEPIPE A. Description 
 
The PS+CAEPIPE program is a finite el ement computer program which performs linear elastic analysis of piping systems using the stiffness method of finite
 
element analysis; the displacements of the joints of a given structure are FNP-FSAR-3F
 
3F-16 REV 21  5/08 considered basic unknowns. The dynamic analysis by the modal synthesis method utilizes known maximum accelerations produced in a single degree of
 
freedom model of a certain frequency. The principal program assumptions are
 
as follows:
 
It is a linearly elastic structure. Simultaneous displacement of all supports is described by a single time-dependent function. Lumped mass model
 
satisfactorily replaces the continuous structure. Modal synthesis is applicable.
 
Rotational inertia of the masses has negligible effect.
 
B. Validation
 
The results obtained from pipe stress program PS+CAEPIPE have been compared with the following: 
 
ASME Benchmark problem results, Pressure Vessel and Piping 1972 computer programs verification, American Society of Mechanical Engineers. Longhand
 
calculations--PS+CAEPIPE is compatible with NRC Regulatory Guide 1.92. A
 
synthesis of closely spaced modes is provided based on equation 4 of
 
Regulatory Guide 1.92. 
 
Verification problems were prepared by SST, Inc. and reviewed by SCS. 
 
C. Extent of Application
 
The PS+CAEPIPE program is used to determine stresses and loads in the piping systems due to restrained thermal expansion, deadweight, seismic inertia and
 
anchor movements, externally applied loads such as jet-loads, and transient
 
forcing functions such as created by fast relief valve opening and closing, fast
 
check valve closure after pipe breaks in main feedwater line, fast valve closure in
 
main steam line, etc. PS+CAEPIPE analyzes piping systems in accordance with
 
ANSI and ASME codes. 
 
D. Reference
 
PS+CAEPIPE Program is a software licensed by SST System, Inc. 
 
3F.5 COMPUTER PROGRAMS USED IN THE STRUCTURAL ANALYSES BY VENDORS AND SUBCONTRACTORS A number of computer programs are used in the structural analyses of Category I structures by
 
the following vendors and subcontractors: 
 
A. Chicago Bridge and Iron Company. 
 
B. Inland-Ryerson Construction Products Company. 
 
C. Whiting Corporation.
FNP-FSAR-3F
 
3F-17 REV 21  5/08 The computer programs that they have used are listed, described, and documented in the following sections. 
 
A summary of these programs is presented in table 3F-4. 
 
3F.5.1  COMPUTER PROGRAM USED BY CHICAGO BRIDGE AND IRON COMPANY The computer program described and documented in the following section is used by Chicago
 
Bridge and Iron Company in the analyses of the equipment hatch and personnel lock for the
 
containment structure. 
 
3F.5.1.1 CB&I Program 7-81 A. Description 
 
This Shells of Revolution program, which is based on the ASME paper "Analysis of Shells of Revolution Subjected to Symmetrical and Non-Symmetrical Loads" by A. Kalnins, is a standard computer program in the industry. The program
 
computes the stresses and displacements in thin-walled, elastic shells of
 
revolution when they are subjected to static edge loads, surface loads, or
 
arbitrary temperature distribution over the surface of the shell. The geometry of
 
the shell must be symmetrical; however, the shape of the median may be
 
arbitrary. The shell wall may consist of four layers of different orthotropic
 
materials, and the thickness and elastic property of each layer may vary along
 
the median. 
 
The program numerically integrates the eight ordinary first-order differential equations of the thin-shell theory derived by H. Reissner. 
 
The CB&I version of the Shells of Revolution program incorporated modifications on the method of input and the format of output. 
 
B. Validation 
 
The results of the program were compared with those obtained by other shell programs, such as Seal and Cerl II, and were found to be in excellent agreement.
 
Document traceability is available at Chicago Bridge and Iron Company. 
 
C. Extent of Application 
 
The program is used in the analyses of the equipment hatch and personnel lock for the containment structure. 
 
D. Reference 
 
Kalnins, A., "Analysis of Shells of Revolution subjected to Symmetrical and Non-Symmetrical Loads," presented at the Summer Conference of the Applied FNP-FSAR-3F
 
3F-18 REV 21  5/08 Mechanics Division, Boulder, Colorado, June 9-11, 1964, of the American Society of Mechanical Engineers. 
 
3F.5.2  COMPUTER PROGRAMS USED BY INLAND-RYERSON CONSTRUCTION PRODUCTS COMPANY The computer programs described and documented in the following sections are used by
 
Inland-Ryerson Construction Products Company in computing the geometry and prestress losses of the containment structure post-tensioning system. 
 
A. Program WDINT. 
 
B. Program POCKET. 
 
C. Program NUFRCOHO. 
 
3F.5.2.1 Program WDINT A. Description 
 
Program WDINT is a proprietary computer program of Inland-Ryerson Construction Products Company. The program deals with the spatial relationship
 
between tendons of the containment structure post tensioning system, by the
 
usual and familiar methods and formulas of three-dimensional analytic geometry. 
 
Input parameters include the defined locations of the dome tendons, the desired locations of the vertical tendons, and the dimensions of both the dome and the
 
vertical tendons. 
 
The program examines each vertical tendon in turn for interference with any dome tendons. If an interference is detected, a new location for the vertical
 
tendon is examined until the closest location to either side of the desired location
 
which does not interfere with the dome tendons is found. 
 
The output of the program provides the necessary information for detailing the tendon placement drawings. 
 
B. Validation 
 
The results of the program have been verified by manual calculations. Document traceability is available at Inland-Ryerson Construction Products Company. 
 
C. Extent of Application 
 
The program is used to detect interference between the vertical and the dome tendons of the containment structure post tensioning system. 
 
FNP-FSAR-3F
 
3F-19 REV 21  5/08 3F.5.2.2 Program POCKET A. Description 
 
Program POCKET is a proprietary computer program of Inland-Ryerson Construction Products Company. The program deals with the spatial relationship
 
between the containment structure dome tendon anchorages and the
 
surrounding concrete surfaces, by the use of three-dimensional analytic
 
geometry. 
 
Input parameters include the locations and trajectories of the dome tendons and the geometry of the concrete surfaces at the anchorages. 
 
The program computes the necessary dimensions and angles to define the locations, orientations, and dimensions of the pockets for the anchorage assemblies. 
 
Output of the program is used to detail the pockets on the erection shop drawings. 
 
B. Validation 
 
The results of the program have been verified by manual calculations. Document traceability is available at Inland-Ryerson Construction Products Company. 
 
C. Extent of Application 
 
The program is used to compute the required angles and dimensions of the pockets at the containment structure ring girder for the dome tendon anchorage assemblies. 
 
3F.5.2.3 Program NUFRCOHO A. Description 
 
Program NUFRCOHO is a proprietary co mputer program of the Inland-Ryerson Construction Products Company. The program deals with the prestress losses of
 
the containment structure post tensioning system. 
 
Input parameters include the geometry of the tendon trajectories, physical properties of materials, friction coefficients, and jacking characteristics. 
 
The program computes the prestress losses for each tendon of the containment structure. For purposes of analysis, the geometry of each tendon trajectory is
 
considered in segments, i.e. straight, curved, and transitional. The force at the
 
end of the tendon is the jacking force, and the force at the remote end of the first
 
segment is the jacking force reduced by friction. The force at the beginning of
 
the second segment is then the force at the end of the first segment, etc., for the FNP-FSAR-3F
 
3F-20 REV 21  5/08 length of the tendon. The minimum force in the tendon is either at the fixed end (for tendons stressed from one end) or near the middle of the tendon (for tendons
 
stressed from both ends). The location and magnitude of the minimum force for
 
tendons stressed from both ends are found by computing, from each end, the
 
point of intersection of the lines graphing the influences of each jack. 
 
Output consists of data on elongations, forces, and stresses at various points along the tendon, and certain dimensional information. 
 
B. Assumptions 
: a. The prestress elements behave in accordance with Hook's Law within the stress level to which they are subjected. 
: b. The force used in computing the elongation of any segment of a tendon is the average of the forces at the ends of the segment. 
: c. Friction acts in accordance with Coloumb's friction formula. In addition, the friction coefficients for use in Coloum b's formula are derived experimentally for each containment structure and are assumed to be the same for all
 
similar tendons in the structure. 
 
C. Validation 
 
The results of the program have been verified by manual calculations. Document traceability is available at Inland-Ryerson Construction Products Company. 
 
D. Extent of Application 
 
The program is used to prepare field stressing cards, which provide the necessary information required during stressing of tendons, such as hydraulic
 
pressures, pressure ranges, and elongations at various specified force levels. 
 
E. Reference 
 
American Concrete Institute, "Building Code Requirements for Reinforced Concrete," ACI 318-71. 
 
3F.5.3  COMPUTER PROGRAMS USED BY WHITING CORPORATION The computer program described and documented in the following section is used by Whiting
 
Corporation in the seismic analysis of the polar crane inside the containment structure. 
 
FNP-FSAR-3F
 
3F-21 REV 21  5/08 3F.5.3.1 STARDYNE DYNRE 4 A. Description 
 
The STARDYNE Static and Dynamic Stru ctural Analysis System, developed by Mechanics Research, Inc., Los Angeles, California, can perform static and
 
dynamic analyses of complex three-dimensional structures using the finite
 
element method. DYNRE 4 is one of the six major programs of the STARDYNE
 
system and is used for computing the response of a general STARDYNE
 
modeled structure to an arbitrarily oriented shock spectra input. The program
 
includes two superposition techniques for displacements, velocities, and
 
accelerations, as well as element loads and stresses. Shock input consists of
 
user-furnished spectral values or averaged spectra, normalized to the 1940 El
 
Centro earthquake (N-S component). The averaged spectra include data from
 
the 1934 El Centro, 1940 El Centro, 1949 Olympia, and 1952 Taft earthquakes. 
 
B. Validation 
 
The program is a recognized program in the public domain and has had sufficient history of use to justify its applicability and validity. 
 
C. Extent of Application 
 
The program is used in the seismic analyses of the polar crane inside the containment. 
 
D. Reference 
 
  "MRI/STARDYNE-Static and Dynamic Stru ctural Analysis System-Theoretical Manual," Publication No. 866 16300 , Control Data Corporation, June 15, 1973. 
 
FNP-FSAR-3f
 
TABLE 3F-1 COMPUTER PROGRAMS USED FOR CATEGORY I STRUCTURAL ANALYSES BY BECHTEL POWER CORPORATION
 
REV 21  5/08 Program Title Document Traceability Program Capabilities
 
CE 316-4 FINEL PICC (a) Analyzes complex axisymmetric structures for both elastic and inelastic behavior by the finite element method CE 639 BPC (b) Performs analysis to determine the forces and stresses on a dome caused by prestressing of tendons, including the effects of friction losses CE 779 SAP PICC Performs linear elastic analyses of three-dimensional structural systems ME 620 Heat Induction PICC Determines the temperature distribution within a plane or axisymmetric body    ASHSD - PICC Analyzes axisymmetric structures by the finite element method for axisymmetric and asymmetric static and dynamic
 
loads 
 
_________________ a. Pacific International Computing Corporation, San Francisco, California. 
: b. Bechtel Power Corporation, Gaithersburg, Maryland.
FNP-FSAR-3F
 
REV 21  5/08 TABLE 3F-2 (SHEET 1 OF 2)
COMPUTER PROGRAMS USED IN SEISMIC ANALYSES BY BECHTEL POWER CORPORATION Document Program Program Title Traceability Capabilities
 
CE 309 STRESS PICC (a) Generates flexibility    matrix or stiffness matrix for structural models 
 
CE 591 Spectra PICC Calculates and plots Analysis  response spectra for earthquake accelerogram 
 
CE 611 Time-History PICC Computes time-Response Analysis  history response for structures subjected to earthquake 
 
CE 617 Modes and BPC (b) Extracts modes    Frequencies  and frequencies Extraction  from stiffness or flexibility matrix and diagonal mass matrix 
 
CE 641 Response Spectrum BPC Spectral response Analysis  analysis of simple cantilever structures 
 
CE 655 Stress Dynamic PICC Extracts flexibility Analysis  or stiffness matrix from CE 309, "STRESS" 
 
CE 785 Spectrum BPC Suppresses locally Suppression  the response spectrum of a given accelerogram 
 
CE 786 Spectrum BPC Raises locally the Raising  response spectrum of a given accelerogram 
 
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REV 21  5/08 TABLE 3F-2 (SHEET 2 OF 2)
Document Program Program Title Traceability Capabilities
 
CE 791 Spectra BPC Calculated and plots Analysis  response spectra for earthquake accelerogram 
 
CE 792 Response BPC Computes response Spectra  spectra at specified values of frequencies and damping ratios using CE 611, "Time-History Response Analysis" 
: a. Pacific International Computing Corporation, San Francisco, California. 
: b. Bechtel Power Corporation, Gaithersburg, Maryland.
 
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REV 21  5/08 TABLE 3F-3 COMPUTER PROGRAMS USED IN SEISMIC ANALYSES BY SOUTHERN COMPANY SERVICES, INC.
Document Program Program Title Traceability Capabilities
 
EN 3426 ICES SCS Analyzes two- or three-dimensional STRUDL II  framed structures and continuous mechanics problems. Flexibility or stiffness matrix may be generated. 
 
EN 3423 Modal SCS Computes the weights of mass Analysis  points, damping values, mode shapes, and frequencies from the flexibility matrix generated by EN 3426 and EN 3429 
 
EN 8043 Response SCS Computes the maximum response of Spectra Plots  a single degree of freedom oscillator for various frequencies and damping values
 
En 8045 Response SCS Computes the time acceleration of Spectra  the desired mass points using the mode shapes and frequencies obtained from EN3423, "Modal Analysis," and the damping values of the structure
 
EN 8046 Response SCS Plots the maximum response for Spectra  computed by EN 8043, "Response Plots  Spectra Plots"
 
PS+CAEPIPE  SCS Analysis of Piping System per ANSI and ASME Codes
 
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REV 21  5/08 TABLE 3F-4 (SHEET 1 OF 2)
COMPUTER PROGRAMS USED IN CATEGORY I STRUCTURAL ANALYSIS BY VENDORS AND SUBCONTRACTORS Document Program Company Program Title Traceability Capabilities
 
CB&I (a) 7-81 Shells of CB&I Performs the    Revolution  analysis of shells of revolution subjected to symmetrical and nonsymmetrical loads 
 
Inland- WDINT -- Inland- Computes the 
 
Ryerson (b)  Ryerson spatial    relationship between tendons of the containment structure post-tensioning system 
 
Inland- POCKET -- Inland- Computes the Ryerson  Ryerson spatial relationship between dome tendon anchorages and the surrounding concrete surfaces of the containment structure post-tensioning system 
 
Inland- NUFRCOHO -- Inland- Computes the Ryerson  Ryerson prestress losses of the containment structure post-tensioning system 
 
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REV 21  5/08 TABLE 3F-4 (SHEET 2 OF 2)
Document Program Company Program Title Traceability Capabilities
 
Whiting (c) STARDYNE- - Recognized Computes    DYNRE 4  Program in responses Public caused by Domain arbitrarily-oriented shock spectra 
: a. Chicago Bridge & Iron Company, Birmingham, Alabama. 
: b. Inland-Ryerson Construction Products Company, Melrose Park, Illinois. 
: c. Whiting Corporation, Harvey, Illinois.
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3G-i  3G QUALITY CONTROL PROCEDURES FOR FIELD WELDING AND NONDESTRUCTIVE EXAMINATIONS OF CONTAINMENT LINER PLATE TABLE OF CONTENTS
 
Page
 
3G.1 SCOPE................................................................................................................................3G-1
 
3G.2 WELDING PERFORMED BY SUBCONTRACTOR............................................................3G-1
 
3G.2.1 WELDING PROCEDURES................................................................................3G-1  3G.2.2 WELDER QUALIFICATION...............................................................................3G-1
 
3G.3 NONDESTRUCTIVE EXAMINATIONS OF WELDMENTS.................................................3G-2
 
3G.3.1 VISUAL INSPECTION.......................................................................................3G-2 3G.3.2 RADIOGRAPHIC EXAMINATION.....................................................................3G-2  3G.3.3 DYE PENETRANT EXAMINATION...................................................................3G-3  3G.3.4 MAGNETIC PARTICLE EXAMINATION...........................................................3G-4  3G.3.5 VACUUM BOX EXAMINATION.........................................................................3G-4  3G.3.6 HALOGEN LEAK INSPECTION (LEAK CHASE SYSTEM)..............................3G-5  3G.3.7 PRESSURE DECAY TEST (LEAK CHASE SYSTEM)......................................3G-6
 
3G.4 REPAIRS............................................................................................................................3G-7
 
3G.5 RECORDS..........................................................................................................................3G-7
 
3G.6 UNIT 2.................................................................................................................................3G-7
 
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3G-1  [HISTORICAL] [QUALITY CONTROL PROCEDURES FOR FIELD WELDING AND NONDESTRUCTIVE EXAMINATIONS OF CONTAINMENT LINER PLATE 3G.1 SCOPE These procedures outline the general qua lity control requirements for field welding of the containment steel liner plate, to ensure that all welding is performed in full compliance with the applicable job specifications. 
 
3G.2 WELDING PERFORMED BY SUBCONTRACTOR
 
3G.2.1  WELDING PROCEDURES 
 
All welding is performed in strict accordance with th e applicable job specifications. All welding procedures, including the procedure qualification test records, used for the liner plate and penetrations are submitted to Southern Nuclear Operating Company (SNC) for approv al. Production welding is not permitted without prior approval of those procedure and qualification records. 
 
In all cases, the general contractor's welding inspector is responsible for ensuring that all subcontractors' welding is performed in accordance with th e appropriate qualified welding procedures. 
 
3G.2.2  WELDER QUALIFICATION 
 
All welders and welding operators who perform weldi ng under a code, standard, or specification that requires qualification of welders are tested and qualified accordingly prior to production welding. The subcontractor is responsible for tes ting and qualifying his own welders. 
 
The general contractor's welding inspector is, in all cases, responsible for ensuring that the subcontractors' welders have successfully passed the n ecessary qualification tests, and that the subcontractor has the proper qualification test records for each qualified welder on file.
 
3G.3 NONDESTRUCTIVE EXAMINATIONS OF WELDMENTS
 
3G.3.1  VISUAL INSPECTION 
 
The general contractor's welding inspector is responsible for carrying out the necessary surveillance to ensure that all welding meets the following requirements for visual inspection and general workmanship. 
 
A. Each weld is uniform in width and size thr oughout its full length. Each layer of weldment is smooth and free of cracks, pinholes, and undercut, and is completely fused to the adjacent
 
weld beads and metals. 
 
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3G-2    In addition, the cover pass is free of coa rse ripples, irregular surface, nonuniform bead pattern, high crown, and deep ridges or valleys between beads. 
 
B. Butt welds are of multipass construction, s lightly convex, of uniform height, and have full penetration. 
 
C. Butt welds made entirely from one side of the joint are welded against a backing strip. 
 
D. Butt welds without backing strips are we lded from both sides of the joints and are back-gouged to sound metal by arc-air, chippi ng, or grinding, before welding from the opposite side. 
 
E. Fillet welds are of the specified sizes with full throat and legs of uniform length. 
 
F. Where different base metal thi cknesses are joined by welding, th e finished joints are tapered no greater than one to four (1:4) between the thick and the thin sections. 
 
3G.3.2  RADIOGRAPHIC EXAMINATION 
 
The general contractor's welding inspector is respons ible for ensuring that all radiographic inspection is properly performed. He ensures that the correct radi ographic techniques are followed and that the completed films are submitted to the licensee (SNC) fo r review, interpretation, and record. 
 
The techniques for identification of radiographic examina tions of welds are in accordance with Paragraph UW-51 of Section VIII of the 1968 ASME Code, using fine grain X-ray films. 
 
The criteria for radiographic acceptance of welds a re in accordance with Paragraph UW-52 of the 1968 ASME Code. 
 
For quality control purposes for each we lder's work, one spot radiograph 12 in. long is taken of the first 10 ft of welding completed on the liner in the flat, vertical , horizontal, and overhead positions. No further welding is permitted until this initial radiographic inspection has been completed and the welding is found to be acceptable by the general contractor's welding inspector. 
 
Thereafter, a minimum of 2 percent of the welding is spot radiographed progressively as welding is performed, using 12- in.-long film, on a random basis as specified by th e general contractor's welding inspector. This is done in such a manner that an appr oximately equal number of spot radiographs is taken from the work of each welder. Under conditions in which two or more welders make weld layers in a joint or on the two sides of a double-welded butt joint, one s pot radiograph may represent the work of both welders. 
 
If a radiograph discloses welding that does not comply with the minimum quality requirements, two additional spot radiographs 12 in. long are taken in the same weld seam at locations away from the original spot. The locations of these additional spots are determined by the general contractor's welding inspector. If the two additional spot radiographs show we lding that meets the minimum quality requirements, the entire weld represented by the three radiographs is acceptable; however, th e defective welding disclosed by the first of the three radiographs is removed and repaired by welding. 
 
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3G-3  If either of the two additional spot radiographs shows we lding that does not comply with the minimum quality requirements, the entire portion of the weld seam represented is rejected or, at the subcontractor's option, the entire weld seam represented is completely radiographed, and defective welding is corrected to meet the minimum quality requirements. 
 
The rewelded joints or weld repai red areas are completely radiographed and must meet the minimum quality requirements. 
 
3G.3.3  DYE PENETRANT EXAMINATION 
 
The general contractor's welding inspector is responsib le for ensuring that all dye penetrant inspection is properly performed. He sees to it that the correct t echniques are followed, and that the results are properly interpreted. 
 
Dye penetrant examination is used to examine 18 p ercent of all liner welds and 100 percent of all welds not vacuum box tested. 
 
All dye penetrant inspection of welds is in accor dance with Appendix VIII, "Methods for Liquid Penetrant Examination," of Section VIII of the 1968 ASME Code. 
 
3G.3.4  MAGNETIC PARTICLE EXAMINATION 
 
The general contractor's welding inspector is responsible for ensuring that all magnetic particle inspection is properly performed, that the correct techniques are follo wed, and that the results are properly interpreted. 
 
Magnetic particle examination is used as an alternative to dye penetrant examination, at the subcontractor's option. 
 
All magnetic particle inspection of welds is in accordance with Appendix VI, "Methods for Magnetic Particle Examinations," (dry particle) of Section VIII of the 1968 ASME Code. 
 
3G.3.5  VACUUM BOX EXAMINATION 
 
The general contractor's welding inspector is responsib le for ensuring that all vacuum box inspection is properly performed, that the correct techniques are follo wed, and that the results are properly interpreted. 
 
Vacuum box testing is used to examine 100 percent of a ll welds that must maintain leaktightness. Those welds that cannot be vacuum boxed because of configurati on or space limitations are dye penetrant or magnetic particle examined. 
 
The procedures for vacuum box inspection are as follows: 
 
A. The vacuum box is a portable unit and has a viewing window large enough to view the complete test area and to allow sufficient light to enter for proper examination. The box is capable of producing and holding a pressure differential of at least 8 psi. 
 
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3G-4  B. The leak detector solution is "Seam-Test Concentrate," manufactured by Winton Products Company, or "Test- Point" manufactured by United States Gulf Corporation. 
 
C. Prior to testing, the test area is cleane d free of slag, scale, grease, paint, or any other materials that interfere with the testing pro cedures or the interpretation of the test. 
 
D. Within 1 minute prior to testing, the leak detector solution is applied evenly over the entire test area. 
 
E. The vacuum box then is put in place and eva cuated to at least 5-psi pressure differential with respect to the atmospheric pressure. 
 
The 5-psi minimum pressure differential is ver ified by a gage attached to the vacuum box unit and maintained for a minimum time of 20 seconds. The leak detector solution is
 
continuously observed for bubbles that indicate lea ks, from the time evacuation of the box is begun until 20 seconds after the required vacuum has been obtained. 
 
F. All leaks regardless of size are repaired by completely removing the defect and rewelding the repaired area. A minimum of 2 in. on each side thereof is reinspected by vacuum box testing. 
 
3G.3.6  HALOGEN LEAK INSPECTION (LEAK CHASE SYSTEM) 
 
The general contractor's welding inspector is responsib le for ensuring that all halogen leak inspection is properly performed, that the correct techniques are follo wed, and that the results are properly interpreted. 
 
The halogen sniffer test is used at all welds that are fitted with a leak chase channel system. This system is installed over all welds that will not be accessible fo r inservice inspection during the plant's life. The halogen sniffer test is not used to examine th e liner seam welds; it is used to verify the leaktightness of the liner to leak chase welds. 
 
The procedure for halogen leak inspection of the leak chase system is as follows: 
 
A. The standard high sensitivity industrial halogen le ak detector is the General Electric Control Unit with a type H-2 gun probe. The refrigerant is R-12. 
 
B. The test area is clean and dry, free of ha lides, cigarette smoke, welding smoke, and explosive vapors and is protected from sudden gusts of wind, foreign materials, or turbulent conditions. 
 
C. The leak chase system test zone is pressurized with refrigerant to 15 psig. The pressure then is increased with air to 62 psig.
 
D. The halogen leak detector is calibrated against the halogen standard at the start of the testing and about every 2 hours during testing to determine that the instrument is capable of
 
detecting the required minimum leakage of 1.0 x 10
-5 standard cm 3/s.
E. The channel to liner plate weld joints are t ested with the halogen leak detector at a sniffer rate of about 2-1/2 ft/min. with the halide gun held within 1/2 in. of the test surface.
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3G-5  F. All leaks detected at the established test sensitivity of 1.0 x 10
-5 standard cm3/s are repaired and reinspected. 
 
3G.7 PRESSURE DECAY TEST (LEAK CHASE SYSTEM) 
 
The general contractor's welding inspector is respons ible for ensuring that a ll pressure decay tests are properly performed, that the correct techniques are follo wed, and that the results are properly interpreted. 
 
The pressure decay test is performed after the halogen sniffer test is conducted on the leak chase system. This is the test that verifies the leaktightness of all seam welds that are fitted with leak chases. 
 
The procedure for pressure deca y test is as follows: 
 
A. After the halogen leak inspection of the leak chase system, the channel test zone is purged of the refrigerant a minimum of three times. 
 
B. The channel test zone is pressurized to 62 psig and allowed to stand for 25 to 30 minutes prior to monitoring the pressure hold test. 
 
C. Surface thermometers are installed on the channel test zone. 
 
D. The pressure decay test is performed by hol ding the pressure for a minimum of 30 minutes. 
 
E. The acceptance criterion is that the temp erature corrected pressure lo ss is not greater than 2 psig per 25 to 30 min. 
 
F. All channel test zones with unacceptable temperature- corrected p ressure loss are repaired and retested. 
 
3G.4 REPAIRS
 
It is the responsibility of the general contractor's weld ing inspector to determine that all weld defects are removed, repaired, and reinspected in accordan ce with the applicable job specifications. 
 
3G.5 RECORDS
 
It is the responsibility of the general contractor's weld ing inspector to ensure that proper records of welding and nondestructive testing of the containment liner plate are kept on file. 
 
3G.6 UNIT 2
 
Because the contract for Unit 2 was issued on Febr uary 1, 1972, some changes were made to the Unit 1 design documents to incorporate new or updated c odes. The 1971 (with summer 1971 addenda) edition of the ASME Code was invoked, and sections of NRC Regulato ry Guide No. 1-19 were included. Consequently, the following revisions to section 3.2 of appendix 3G are applicable to Unit 2 only:
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3G-6  A. The criteria for radiographic acceptance of welds are in accordance with paragraph UW-51 of the 1971 ASME Code. 
 
B. For quality control purposes the first 10 feet of welding is 100-percent radiographically examined for the flat, vertical, horizontal, and o verhead positions of weld performed by each welder. No further welding is permitted un til this initial radiographic inspection has been completed and the welding found to be acceptable.
]   
 
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3H-i REV 21  5/08 
[HISTORICAL] [3H. CONTAINMENT STRUCTURAL ACCEPTANCE TEST TABLE OF CONTENTS Page 3H.1 INTRODUCTION......................................................................................................................3H-1
 
3H.2 TEST DESCRIPTION................................................................................................................3H-1
 
3H.2.1 GENERAL..............................................................................................................3H
-1  3H.2.2 TEST PRESSURE...................................................................................................3H-1 3H.2.3 DEFLECTION MEASUREMENTS........................................................................3H-1  3H.2.4 CRACK PATTERNS...............................................................................................3H-2 3H.2.5 STRAIN MEASUREMENTS...................................................................................3H-2 3H.2.6 TESTING ENVIRONMENT....................................................................................3H-2 3H.2.7 CONDITIONS FOR REPEATING TEST................................................................3H-3
 
3H.3 STRUCTURE RESPONSE.........................................................................................................3H-3 3H.4 TEST REPORT .........................................................................................................................3H-3
 
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3H-ii REV 21  5/08 LIST OF FIGURES
 
3H-1 Taut Wire Displacement Transducer Locations 
 
3H-2 Structural Integrity Test - Dome and Vertical Displacement Measurements 
 
3H-3 Structural Integrity Test - Radial Di splacement Measurements at Equipment Hatch
 
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3H-1 REV 21  5/08 APPENDIX 3H CONTAINMENT STRUCTURAL ACCEPTANCE TEST
 
3H.1 INTRODUCTION
 
The purpose of the containment structural acceptance test is to demonstrate that, when the containment is pressurized to the design loading, the deflections of the containment's structural elements and the cracks at the exterior surface concrete a re within the acceptable limits. This confirms that the design and construction of the containment are adequate to withstand such pressure loading. The structural
 
acceptance test will be performed in conjunction with the containment integrated leak rate test and will generally comply with Regulatory Guide 1.18 as discussed in appendix 3A. 
 
A complete test procedure will be prepared and subm itted to the NRC for review at least 90 days prior to conducting the structural acceptance test. 
 
3H.2 TEST DESCRIPTION
 
3H.2.1  GENERAL 
 
Prior to reactor fuel loading and operation, the inte grity of the containment is demonstrated by a pressure proof test. The pressure test permits verification that the structural response due to the induced load is consistent with the predicte d behavior. This is accomplished by measurements of the structure's deflections by the use of internally mounted taut wires. 
 
3H.2.2  TEST PRESSURE 
 
The pressure proof test is performed by subjecting the containment to a continuous increase in pressure from atmospheric pressure to 1.15 times the design p ressure. Deflection measurements are recorded at atmospheric pressure, at 5 psi pressure increments during pressuri zation and depressurization cycles, and at the completion of depressurization. At maximu m pressure level the pressure is held constant for at least 1 hour. Crack patterns are recorded at atmo spheric pressure prior to the test, at maximum pressure, and following depressurization. 
 
3H.2.3  DEFLECTION MEASUREMENTS 
 
The structural deflections are measured with taut wire extensometers. Each extensometer consists of an invar wire spanning selected points, with one end (the dead end) fixed in position and the live end attached to a spring-loaded frame incorporating a linear potentiometer. The entire system spans the distance to be measured. The springs used are th e "negator" type that apply an essentially constant force independent of extension. The springs selected apply a force of approximately 15 lb each, and they are
 
used in matched pairs with a back-to-back mounting to avoid eccentricity. Accuracy of the extensometer is +/-0.002 inch.
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3H-2 REV 21  5/08 Radial deflections are measured along six equally spaced meridians at the spring line, at midheight of the cylinder, and at a point above the base slab at a hei ght equal to three times the thickness of the wall at the location where the deflection is measured. The locations of these measurements are illustrated in figure 3H-1. 
 
Vertical deflections are measured at the apex, at the spring line of the dome, and at three intermediate points. The locations of these measu rements are shown in figure 3H-2. 
 
The radial and tangential deflections of the cont ainment wall adjacent to the equipment hatch are measured at 12 points as shown on figure 3H-3. 
 
3H.2.4  CRACK PATTERNS 
 
The patterns of cracks that exceed 0.01 inch in width at the exterior surface concrete are mapped near the base wall intersection, at midheight of the wall, at the springline of the dome, and around the equipment hatch. The crack patterns are mapped also at the intersection between a buttress and the wall, at the intersection between the top ring girder and the wall, and on the top shelf of the ring girder. At each
 
location, an area of at least 40 ft 2 is mapped. 
 
3H.2.5  STRAIN MEASUREMENTS 
 
The Farley containment is similar to those of Turkey Point Unit 3 (Docket No. 50-250), Palisades Plant Unit 1 (Docket No. 50-255), and Point Beach Nuclear Power Plant Units 1 and 2 (Docket Nos. 50-266
 
and 50-301, respectively). The containments fo r both Turkey Point Unit 3 and Palisades Unit 1 were completely instrumented. The Tu rkey Point instruments provided approx imately 400 strain measurements at 55 locations throughout the containment concrete and liner. In addition, about 55 taut wire
 
measurements of structural deformation have been made. The Palisades instrumentation was comparable. The data from the instrumenta tion permitted detailed comparison between design calculations and observed response. The basic struct ural design and the accuracy of the calculation procedures used by Bechtel have th erefore been verified by these tests. This verification is applicable to the Farley containment design. 
 
Since the detailed confirmation of the design techniqu es has been made, strain gage instrumentation of the Farley containment is not required and it is c oncluded that no additional confirmation of design techniques is necessary. 
 
3H.2.6  TESTING ENVIRONMENT 
 
The structural acceptance test will be scheduled for periods in which extremely inclement weather is not forecast. However, due to the state of the art of weather forecasting, and the time involved in the preparation and performance of the test, should snow, heavy rain, or strong wind occur during the test, it may be continued and the results considered va lid unless evidence indicates otherwise. 
 
The environmental conditions during the test are meas ured to permit the evaluation of their contribution to the response of the containment. Atmospheric temperature, pressure, and humidity inside and outside the containment are monitored continuously during the test. In addition, the temperature inside and FNP-FSAR-3H
 
3H-3 REV 21  5/08 outside the containment is measured at sufficiently long periods prior to the test to establish an average temperature of the wall for the evaluation of effects of temperature change on the deflection measurements. 
 
3H.2.7  CONDITIONS FOR REPEATING TEST 
 
The test will be repeated under the following conditions: 
 
A. If the structural response deviates at any time during the test, up to a value that may jeopardize the containment integrity, the containment will be depressurized and the cause(s) for the deviation of response determ ined. If repair to the containment is necessary the test will be repeated. 
 
B. If extremely inclement weather, such as snow, heavy rain, or strong wind, occurs during the test and the results of the test are found to be invalid, a retest will be performed. 
 
C. If any significant modifications or repairs are made to the containment following the test, the test will be repeated. 
 
3H.3 STRUCTURE RESPONSE
 
The numerical values of the predic ted structure response are established by the analytical techniques described in section 3.8.1 and appendix 3F. These va lues and the tolerances to be permitted in the acceptance test are developed as the result of the c ontainment structural analysis and will be determined prior to the test. 
 
3H.4 TEST REPORT
 
The following information will be included in the final test report: 
 
A. A description of the test procedure and the taut wire system. 
 
B. A comparison of the test measurements w ith the allowable limits (predicted response plus tolerance) for deflections and crack width. 
 
C. An evaluation of the estimated accuracy of the measurements. 
 
D. An evaluation of any deviations (i.e., t est results that exceed the allowable limits), the disposition of the deviations, and the need for corrective measures. 
 
E. A discussion of the calculated safety margin provided by the structure as deduced from the test results.
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REV 21  5/08
[TAUT WIRE DISPLACEMENT TRANSDUCER LOCATIONS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3H-1
]
REV 21  5/08
[STRUCTURAL INTEGRITY TEST - DOME & VERTICAL DISPLACEMENTS MEASUREMENTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3H-2
]
REV 21  5/08
[STRUCTURAL INTEGRITY TEST - RADIAL DISPLACEMENT MEASUREMENTS AT EQUIPMENT HATCH JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3H-3
]
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3I-i  REV 21  5/08 3I. LINER PLATE STABILITY LIST OF FIGURES
 
3I-1 Liner - Stiffener Weld Test Results
 
3I-2 Liner Plate Loading Conditions
 
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3I-1  REV 21  5/08 APPENDIX 3I LINER PLATE STABILITY
 
The stability of the liner plate was studied for the loading cases and deformations to which it
 
may reasonably be subjected. The critical loading cases considered include the loss-of-coolant
 
accident (LOCA) condition and the operating condition during the winter. 
 
Two separate solutions of the plate stability were studied: 
 
A. The plate as a compressed panel under biaxial  compression, assuming that the channel and angle stiffeners are rigid in their attachment to the prestressed
 
concrete containment and the liner. 
 
B. The plate as a compressed panel under biaxial  compression, assuming the panel to be a portion of a  large cylinder with a flexible stiffener system. 
 
Figure 3.8-1, Detail 2, illustrates the actual physical configuration of the stiffening system used
 
for the liner plate. The channels function as horizontal stiffeners and the angles as vertical
 
stiffeners. 
 
For the solution, an initially deflected form for the liner plate is expressed in terms of a Fourier
 
series of the form 
 
nsin m cos mn1 m 0 n===  where  defines the central angle in a plan view of the cylinder, from a point on the circumference where there is zero radial deflection to the point on the circumference where there is maximum radial deflection;  defines the radial deflection and  defines the unsupported length in the vertical direction. 
 
Under normal operating conditions, the overall structural stability of the liner plate is maintained. 
 
The most critical stress for the liner plate exists in the condition illustrated by figure 3I-2. In this
 
condition, Panel 1 and Panel 3 have outward initial curvature and Panel 2 has inward initial
 
curvature. When a load is applied parallel to the liner plate, Panels 1 and 3 bear against the
 
concrete and Panel 2 deforms inward. If the load is primarily from concrete shrinkage creep, prestress, and thermal effects, the membrane stress (N/t) in Panels 1 and 3 tends to relax to a value of (N-N/t) in Panel 2. The anchors between the panels with inward and outward curvature must restrain a force of N for static equilibrium. Due to inward deformation, flexural stress also exists in Panel 2 and the anchors are subjected to the moment (M).  (See
 
figure 3I-2.) 
 
The maximum compressive strains are caused by accident pressure, thermal loading, prestress, shrinkage, and creep. The maximum calculated strains do not exceed 0.0025 in./in., and the
 
liner plate always remains in a stable condition.
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3I-2  REV 21  5/08 The anchorage has the capability of resisting the full force (N) due to a theoretically fixed anchor, but in addition it has sufficient ductility to accept the 0.038-in. displacement without
 
failure. The above displacement results from a uniform membrane strain of 0.0025-in./in.
 
distributed over a 15-in. anchor spacing. Various patterns of welds attaching the angle anchors
 
to the liner plate have been tested for ductility and strength when subjected to a transverse shear load such as N and are shown in figure 3I-1. 
 
Also of concern is the nature of the state of stress and behavior at the point of attachment
 
between the stiffeners and the liner plate. Special tests (1) have been conducted on simulated models of the liner plate and vertical stiffener assembly to determine the shear capacity of the
 
angle anchorage. The results of these tests and the various weld configurations are shown in
 
figure 3I-1. Note in the test results that two different configurations of support were used for the
 
simulated continuous anchor. The case in which the 0-in. gap was used simulates the expected
 
condition that exists in the containment. The case with the 5/8-in. gap attempts to simulate the
 
condition that might exist at an isolated location if the concrete were not in continuous contact
 
with the anchor. Being guided in the proportioning of the liner plate stiffeners by the values of
 
shear transfer for the case of the 5/8-in. gap will, in general, result in a margin of safety for
 
progressive failure of anchors of approximately 2.7. The weld configuration shown in
 
figure 3.8-1, Detail 2, is adequate to transfer all loads that are considered in the design on the
 
containment between the liner plate and the stiffener-anchors. 
 
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3I-3  REV 21  5/08 REFERENCE 
: 1. Liner Plate Anchorage Tests for Job No. 6600 Arkansas  Nuclear One, Arkansas Power
& Light Company; Job No. 6292 Rancho Seco Nuclear Station - Unit 1 Sacramento
 
Municipal Utilities District; Job No. 6750 Calvert Cliffs - Units 1 and 2, Baltimore Gas &
 
Electric Company.... Prepared by  Bechtel Corporation, San Francisco, Calif.  (April 18, 1969).   
 
REV 21  5/08 LINER - STIFFENER WELD TEST RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3I-1
 
REV 21  5/08 LINER PLATE LOADING CONDITIONS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 3I-2
 
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3J-i REV 21  5/08 3J  MAIN STEAM LINE BREAK OUTSIDE CONTAINMENT PRESSURE AND TEMPERATURE ANALYSIS TABLE OF CONTENTS Page
 
3J.0 BACKGROUND............................................................................................................3J-1
 
3J.1 FNP APPLICABILITY...................................................................................................3J-1
 
3J.2 WESTINGHOUSE OWNERS GROUP EFFORTS.......................................................3J-2
 
3J.3 APPLICATION OF WOG BLOWDOWN DATA............................................................3J-3
 
FNP-FSAR-3J
 
3J-ii REV 21  5/08 LIST OF FIGURES 3J-1 HELB Outside Containment 0.05 ft 2 Break at 102-percent Power 30-min. Operator Action - Temperature vs Time
 
3J-2 HELB Outside Containment 0.2 ft 2 Break at 70-percent power - Temperature vs Time
 
3J-3 HELB Outside Containment 0.2 ft 2 Break at 102-percent Power - Temperature vs Time
 
3J-4 HELB Outside Containment 0.4 ft 2 Break at 102-percent Power - Temperature vs Time
 
3J-5 HELB Outside Containment 0.6 ft 2 Break at 102-percent Power - Temperature vs Time
 
3J-6 HELB Outside Containment 0.8 ft 2 Break at 102-percent Power - Temperature vs Time
 
3J-7 HELB Outside Containment 1.2 ft 2 Break at 102-percent Power - Temperature vs Time
 
3J-8 HELB Outside Containment 1.1 ft 2 Break at 102-percent Power - Temperature vs Time
 
3J-9 HELB Outside Containment 4.6 ft 2 Break at 102-percent Power - Temperature vs Time
 
3J-10 HELB Outside Containment Combined Temperature Profile
 
3J-11 HELB Outside Containment 0.05 ft 2 Break at 102-percent Power - Pressure vs Time
 
3J-12 HELB Outside Containment 0.2 ft 2 Break at 102-percent Power - Pressure vs Time
 
3J-13 HELB Outside Containment 1.1 ft 2 Break at 102-percent Power - Pressure vs Time
 
3J-14 HELB Outside Containment 4.6 ft 2 Break at 102-percent Power - Pressure vs Time
 
3J-15 HELB Outside Containment Combined Pressure Profile
 
3J-16 HELB Outside Containment Combined Temperature Profile for Model 54F Cases at 102% Power
 
3J-17 HELB Outside Containment Combined Temperature Profile for Model 54F Cases at 70% Power
 
FNP-FSAR-3J
 
3J-1  REV 21  5/08 APPENDIX 3J MAIN STEAM LINE BREAK OUTSIDE CONTAINMENT PRESSURE AND TEMPERATURE ANALYSIS  3J.0 BACKGROUND IE Information Notice 84-90, "Main Steam Line Br eak Effect on Environmental Qualification of Equipment," informed licensees of a concern with analyses of main steam line breaks (MSLB). 
 
IE Information Notice 84-90 stated that the assumption of large breaks bounding all others was
 
not true for temperature effects and that smaller breaks would actually result in higher
 
temperatures in the steam released from the br eak. The higher temperatures would result from steam remaining in the steam generator for longer periods than previously assumed due to the steam exiting the break at a slower rate. As steam generator water level decreases due to
 
break flow, the secondary water level could allow tube bundle uncovery. Consequently, the
 
temperature of the steam being generated would approach the reactor coolant system (RCS) temperature and become superheated before exiting the steam generator and the break in the
 
main steam line.
 
In response to this issue, analyses were performed for a spectrum of break sizes using a more
 
advanced computer code than was available at the time of the analyses presented in
 
appendix 3K. The new analyses supersede the appendix 3K analyses with respect to the main steam valve room pressure and temperature re sponse to postulated main steam line breaks.
 
3J.1 FNP APPLICABILITY The maximum blowdown from any steam generator would be equivalent to that produced by a
 
1.069 ft 2 break due to the integral flow restrictors on the outlet nozzle for each steam generator.
Since a MSLB downstream of the main steam isol ation valves (MSIV) could result in blowdown from all three steam generators prior to MSIV closure, the maximum equivalent break size is
 
3.207 ft 2.
The MSLB for Farley Nuclear Plant (FNP) is not postulated for main steam piping outside
 
containment up to and including the MSIVs. The basis for this position is discussed below. 
 
Paragraph 3.6.2.4 lists the specific location criteria for breakpoints in ASME Section III, Class 2
 
and 3 lines, with reference to appendix 3K. Appendix 3K describes the criteria for postulating
 
pipe ruptures or cracks in high-energy lines outside containment and the methods for evaluating
 
the effects of these breaks. Attachment A, part II of appendix 3K specifically addresses the
 
postulated break and leakage locations in the main steam lines outside of containment and
 
demonstrates that this piping conforms to Branch Technical Positions (BTP) APCSB 3-1 and
 
MEB 3-1 of Standard Review Plan Sections 3.6.1 and 3.6.2, respectively. This is consistent
 
with Section 3.6 to Supplement 1 of the NRC Safety Evaluation Report for the Joseph M. Farley
 
Nuclear Plant - Units 1 and 2.
 
The only postulated break upstream of the MSIVs is the 3-in. diameter branch line to the
 
turbine-driven auxiliary feedwater. This line is not part of the "no break zone" and a break must
 
be postulated in this line as part of the FNP licensing basis. Thus, the only break upstream of FNP-FSAR-3J
 
3J-2  REV 21  5/08 the MSIVs which must be considered for FNP is the 3-in. branch line to the turbine-driven auxiliary feedwater pump.
 
3J.2 WESTINGHOUSE OWNERS GROUP EFFORTS In response to the NRC concern, the Westinghouse Owners Group (WOG) formed the
 
High-Energy Line Break/Superheated Blowdown Outside Containment (HELB/SBOC) subgroup.
 
The WOG determined mass/energy release data corresponding to a full spectrum of breaks
 
(0.05 ft 2 to 4.6 ft 2 at 70-percent and 100-percent power). The results of the WOG analysis are presented in WCAP-10961 (reference 1).
 
In support of the Farley Nuclear Plant power uprating, the full spectrum of steamline breaks
 
outside containment with superheated steam blowdown was reanalyzed. Mass/energy release
 
data at 70-percent and 102-percent power were revised using FNP plant-specific assumptions
 
including the increased power level. The revised blowdown analysis is presented in
 
WCAP-14722 (reference 4) and supersedes the blowdown results documented in
 
WCAP-10961.
 
In support of the Farley Nuclear Plant steam generator replacement, a limited spectrum of
 
steam line breaks outside containment with superheated steam blowdown, based on the power uprating analysis, was reanalyzed. Mass/energy release data at 70-percent and 102-percent
 
powers were again calculated using FNP plant-specific assumptions including those associated
 
with the replacement steam generators. The revised blowdown analysis is presented in
 
WCAP-15097 (reference 5) and supersedes the blowdown results documented in
 
WCAP-14722.
 
MSIV closure time is a significant parameter in determination of the consequences of a
 
postulated MSLB in the main steam lines because, for breaks downstream of the MSIVs, main
 
steam line isolation terminates the blowdown. A dditionally, if MSIV closure occurs prior to tube bundle uncovery, the transient will not result in any superheated blowdown. For the power
 
uprating and steam generator replacement analyses, Westinghouse determined that the earliest
 
actuation of MSIV closure for Farley Nuclear Plant is produced by a low-steam pressure signal
 
for breaks outside containment. 
 
Farley Nuclear Plant pressure transmitters are not located in an area where they would be
 
subjected to the harsh environment during a steam line break. Thus, the Farley Nuclear Plant
 
pressure transmitters can be assumed to operate with normal error allowances, and MSIV
 
closure for Farley Nuclear Plant would occur much sooner than for a plant with pressure
 
transmitters located in a harsh environment with corresponding environmentally-induced errors.
Accordingly, Westinghouse provided the appropriate information in WCAP-10961 to determine
 
the specific MSIV closure time corresponding to various break sizes for FNP. The following
 
discussion addresses the specific low steam line pressure setpoint for FNP.
 
The FNP Technical Specification nominal trip setpoint for low-steam line pressure is 585 psig. 
 
FSAR paragraph 7.3.1.2 specifies a historical +
4-percent actuation signal accuracy for a range of 0 to 1200 psig. Therefore, a 48-psi inaccuracy was conservatively applied to the nominal trip
 
setpoint, which results in a safety analysis limit (SAL) of 537 psig for safety injection and main
 
steamline isolation by low steam line pressure. Setpoint uncertainty calculations for these FNP-FSAR-3J
 
3J-3  REV 21  5/08 ESFAS functions demonstrate adequate margin between the SAL and the corresponding nominal trip setpoint. In addition, the dynamic signal compensation specified in the Technical
 
Specifications is explicitly modeled in the sa fety analysis and conservatively implemented by plant procedures. Therefore, FNP can be assured that the MSIV ESFAS actuation signal will
 
close when steam pressure falls to 537 psig (551.7 psia). The power uprating and steam
 
generator replacement analyses provide sufficient information to demonstrate that the FNP MSIVs will close prior to tube bundle uncovery for all breaks of 0.6 ft 2 and larger. Therefore, superheated blowdown will not occur for 0.6 ft 2 and larger breaks.
 
The cases analyzed for Farley Nuclear Plant in the power uprating and steam generator
 
replacement analyses were the 3.2, 2.0, 1.4, 1.0, 0.9, 0.8, 0.7, 0.6, 0.5, 0.4, 0.3, 0.2, 0.1, and
 
0.05-ft 2 breaks at 70-percent and 102-percent power. Due to the integral exit nozzle flow restrictors on each steam generator for FNP, only breaks of 3.207 ft 2 and smaller apply to Farley Nuclear Plant. See section 3J.4 for the discussion of the Model 54F replacement steam
 
generators. 
 
As discussed in section 3J.1, the 3-in. diameter branch line to the turbine-driven auxiliary
 
feedwater pump is not considered a part of the "no break zone" and must be postulated to
 
break. Westinghouse analyzed the 0.05-ft 2 break and presented the results in WCAP-15097 (reference 5). Because MSIV closure is not automatically initiated for this break, the results of
 
the power uprating and steam generator replacement analyses for the time period analyzed (i.e., 0-1800 s) are applicable to a break either downstream or upstream of the MSIVs.
 
Due to the relatively low blowdown rate of the 0.05-ft 2 break, auxiliary feedwater flow is sufficient to delay tube bundle uncovery for nearly 1800 s. At this point, operator action to close
 
the MSIVs is assumed and the transient for the downstream break is terminated. However, the
 
break postulated upstream of the MSIV (a break in the branch line to the turbine-driven auxiliary feedwater pump) is not isolated by MSIV closure. Termination of this transient requires MSIV
 
closure and termination of auxiliary feedwater flow to the steam generator with the faulted line.
The steam blowdown through this break would then continue until steam generator dryout.
 
3J.3 APPLICATION OF WOG BLOWDOWN DATA The results of the WOG HELB/SBOC analysis are presented in WCAP-10961. Blowdown data
 
from WCAP-10961 was analyzed by Westinghouse to determine compartment temperatures using the COMPACT code. COMPACT is a multinode containment code developed by Westinghouse for analysis of containment and outside- containment compartment transients. 
 
This code models the mass transfer between an upper gaseous region and a lower sump region
 
within each node. The code also models natural circulation flow induced by large temperature
 
gradients, which promotes better compartment gas mixing and, hence, more uniform
 
temperature distribution within the compartments.
 
The full spectrum of breaks for Farley Nuclear Plant was analyzed (0.05 ft 2 to 4.6 ft 2 at 70-percent and 102-percent power) even though WCAP-10961 indicated that only the 0.2 ft 2 and smaller breaks may produce superheated steam. This approach was taken to ensure a
 
consistent methodology basis for all postulated st eam line breaks in the main steam valve room (MSVR). Westinghouse determined that the results from the analysis of nine breaks would
 
envelop the environmental conditions for all postu lated breaks. The breaks downstream of the FNP-FSAR-3J
 
3J-4  REV 21  5/08 MSIVs analyzed by Westinghouse were the 0.2-ft 2 break at 70-percent power and the 0.2, 0.4, 0.6, 0.8, 1.0, 1.1, and 4.6-ft 2 breaks at 102-percent power. Westinghouse also analyzed the 0.05-ft 2 break at 102-percent power upstream of the MSIVs. The results of these analyses are documented in WCAP-11652 (reference 2).
 
The break location was selected in the lower portion of the MSVR. The break compartment (Compartment 1) volume was selected to represent a volume occupied by high temperature
 
steam exiting the break in the lower portion of the MSVR. The selection of break location and
 
small break compartment volume ensures that the calculated gas temperatures in the MSVR would be conservative.
 
For all breaks downstream of the MSIVs, the mass and energy release data from WCAP-10961
 
were applied. For the 0.05-ft 2 break upstream of the MSIVs, the mass and energy releases were based on the 102-percent power Case 67 of Category 4 in WCAP-10961. These releases
 
are utilized until 1800 s, when it is assumed that the operator takes action to close the MSIVs
 
and isolate auxiliary feedwater to the faulted steam generator. The mass and energy releases
 
following the operator actions assume a conservative linear blowdown until steam generator
 
dryout and take no credit for continued cooldown of the RCS.
 
During winter months, plastic sheeting may be applied to the exterior of penthouse grating to
 
prevent the freezing of MSVR instruments. Application of the plastic sheeting is discussed in
 
appendix 3K. The plastic sheeting is installed such that the maximum pressure required to tear
 
the sheeting away is 1.25 psig. The impact of the installation of plastic sheeting on the pressure
 
and temperature transients was considered in the WCAP-11652 and WCAP-15560 analyses.
 
The results of the Westinghouse analyses demonstrate that the calculated environmental
 
temperatures in the break compartment do not exceed 325°F for a wide range of break sizes and power levels. Since no credit was taken for heat removal by concrete and steel structural
 
heat sinks, the results of the calculations are very conservative.
 
The results presented in WCAP-11652 demonstrated that for the spectrum of breaks analyzed
 
in this WCAP, the peak temperatures produced by superheat are not a problem for FNP since
 
the existing MSVR analysis peak temperat ure of 308°F and the existing environmental qualification temperature profile are greater than those produced by the 0.2-ft 2 and smaller breaks. However, the WCAP-11652 results indicate that the larger breaks, which do not
 
produce superheated blowdown, result in peak temperatures which exceed the previous temperature profiles. The most limiting case identified by this Westinghouse analysis, with
 
regard to peak temperature, is the 0.8-ft 2 break at 102-percent power, which yields a maximum temperature of slightly less than 325°F. Breaks larger than 0.8 ft 2 are less limiting due to rapid MSIV closure and the corresponding termination of blowdown. The results of this WCAP
 
determined that the thermal transient induced by breaks smaller than 0.8 ft 2 were less limiting due to the reduction in mass flowrate and by the early onset of natural circulation. 
 
The effects of long term steam releases, during postulated main steam line ruptures, on
 
outside-containment equipment environmental qualification are presented in WCAP-15560.
This analysis alters the assumptions that ma ximum superheated steam releases by maximizing the total mass and energy released over time into the MSVR. These analyses expand the
 
power-level/break-area spectrum used in previous superheated steam mass and energy analyses and revise plant parameters to maximize the total energy released through each FNP-FSAR-3J
 
3J-5  REV 21  5/08 postulated steam line rupture. In effect, the analyses presented in WCAP-15560 revised the inputs for the initial steam generator inventory and main feedwater flowrates to develop
 
bounding EQ requirements for the MSVR when used in combination with the prior superheated
 
steam releases. The results of the steam line break analysis presented in WCAP-15560
 
represent the revised basis for the bounding EQ temperature envelope.
 
The combined temperature profile, presented in WCAP-15560, for all of the analyzed cases is
 
shown in figure 3J-10. MSVR equipment which is required to be environmentally qualified has been reviewed against the temperature profile and found to be acceptable.
 
Although the pressure transient associated with the release of superheated steam was never
 
considered to be an issue for NRC Information Notice 84-90, the WOG blowdown data was
 
analyzed to determine the peak pressure which would result from the new analysis. 
 
Westinghouse analyzed the pressure transient associated with 4.6, 1.1, 0.2, and 0.05-ft 2 breaks of 102-percent power. These four breaks were determined by Westinghouse to envelope the
 
pressure transient associated with the spectrum of breaks applicable to FNP because they
 
include the two largest and two smallest flowrates. The analysis indicates the pressure
 
transient resulting from these breaks is slight due to the large venting area available to the
 
MSVR.
 
The composite pressure profile is shown on figure 3J-15. As indicated on the graph, MSVR
 
pressure does not exceed 16 psia for any MSLB. The only discernible pressure increase is the
 
1.25-psi pulse required to clear the sheeting from the grating. The new maximum pressure of
 
less than 16 psia is well below the peak pressure of 20.5 psia from the previous MSVR pressure
 
analysis. A comparison of the mass and energy releases for uprated conditions (reference 4) to
 
those of the previous evaluations (references 1 through 3) indicates that the blowdown for
 
uprated conditions remains bounded by the previous analyses. Accordingly, the new pressure
 
analysis did not impact MSVR equipment qualification or structural integrity.
 
3J.4 Discussion of Results for Model 54F Replacement Steam Generators The original design basis analyses for main steam line breaks outside the containment were
 
documented in WCAP-11652, Rev. 2 and in WCAP-14013. The spectrum of cases that was
 
presented in these documents was used as the basis for determining the break spectrum for the
 
Model 54F replacement steam generators (SG). Since maximizing the amount of superheat
 
was the primary consideration, the following nine cases were studied (reference 6) for the
 
impact of the Model 54F SGs on the main steam valve room (MSVR) post-accident temperature profile.
 
Case 1: 0.4 ft 2 break area at 102% power  Case 2: 0.3 ft 2 break area at 102% power  Case 3: 0.2 ft 2 break area at 102% power  Case 4: 0.1 ft 2 break area at 102% power  Case 5: 0.05 ft 2 break area at 102% power  Case 6: 0.3 ft 2 break area at 70% power  Case 7: 0.2 ft 2 break area at 70% power  Case 8: 0.1 ft 2 break area at 70% power  Case 9: 0.05 ft 2 break area at 70% power FNP-FSAR-3J
 
3J-6  REV 21  5/08 There are two cases (Case 5 and Case 9) that are upstream of the MSIV for the Farley units, which means that the break releases cannot be terminated by MSIV closure. These cases
 
model a very small break (0.05 ft
: 2) in which the releases continue for close to one hour, until the faulted SG is emptied. Because of the relatively low flow rates, these cases are among the
 
least limiting in terms of the peak compartment tem perature. However, the releases last for the longest time, and thus these cases define the compartment temperature envelope as it returns
 
to normal temperatures.
 
The case which yields the most limiting compartment temperature is Case 1. The peak temperature in the MSVR for this case is 320.11
°F. This is less than the maximum temperature of approximately 325
°F that occurs for the original Model 51 steam generators. Figure 3J-16 provides a comparison of Cases 1 through 5 at 102% power to the Equipment Temperature Envelope. It can be seen that all of these cases are within the temperature limit. Figure 3J-17
 
provides a similar comparison with Cases 6 through 9 at 70% power. This comparison also shows that the cases for the model 54F result in compartment temperatures that are within the temperature envelope.
 
Thus, the model 54F replacement steam gener ators do not impact the MSVR equipment qualification or the structural integrity.
 
FNP-FSAR-3J
 
3J-7  REV 21  5/08 REFERENCES
: 1. WCAP-10961 , Revision 1, "Steamline Break Mass/Energy Releases for Equipment Environmental Qualification Outside Containment," Proprietary, October 1985.
: 2. WCAP-11652 , Revision 2, "Joseph M. Farley Nuclear Station Units 1 and 2 Main Steam Valve Room Temperature Response to S uperheated Steam Releases," Proprietary, June 1988.
: 3. WCAP-14013 , "Joseph M. Farley Nuclear Station Units 1 and 2 Main Steam Valve Room Temperature Response to Superheated Steam," March, 1994.
: 4. WCAP-14722 (Proprietary), "Farley Nuclear Plant Units 1 and 2 Power Uprate Project NSSS Engineering Report," November 5, 1997.
: 5. WCAP-15097 (Proprietary), "Farley Nuclear Plant Units 1 and 2 Replacement Steam Generator Program NSSS Engineering Report," November 1998.
: 6. Westinghouse Letter, ALA-98-233, Rev. 1, Southern Nuclear Operating Company, Joseph M. Farley Nuclear Plant, Units 1 and 2, "Main Steam Valve Room Analyses for
 
Main Steam Line Breaks, Revision 1, "November 17, 1998.
: 7. WCAP-15560 , "Joseph M. Farley Nuclear Station Units 1 and 2 Main Steam Valve Room Analysis for Steam Line Breaks Outside Containment", February 2001.}}

Latest revision as of 16:13, 14 January 2025

Updated Final Safety Analysis Report, Revision 21, Chapter 3, Table of Contents Through Appendix 3J, Page 3J-7
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