Text

Revision 28 to Updated Final Safety Analysis Report, Chapter 3, Design of Structures, Components, Equipment and Systems, Part 1
	ML18312A070
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Site:	Farley
Issue date:	10/30/2018
From:	; Southern Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation
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FNP-FSAR-3

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

3-v REV 21 5/08 TABLE OF CONTENTS Page 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

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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)

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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

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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

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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 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.

FNP-FSAR-3 TABLE 3.2-3 (SHEET 1 OF 6)

LISTING OF P&IDs

System Drawing Number REV 23 5/11 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 23 5/11 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-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 23 5/11 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 23 5/11 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 FNP-FSAR-3 TABLE 3.2-3 (SHEET 5 OF 6)

System Drawing Number REV 23 5/11 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 23 5/11 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

FNP-FSAR-3

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

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.

FNP-FSAR-3

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

FNP-FSAR-3

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 ItemsSection III, Class 1, 2, and 3

1690 Stock Materials for Section III Construction

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.

FNP-FSAR-3

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

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 eliminated 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 components 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.

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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

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3.7-5 REV 21 5/08 b. Rectangular Base Motion Spring Constant Horizontal

()kGB L 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.

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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.

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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.

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3.7-8 REV 21 5/08 Simplified analytical models are used for analysis 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.

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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.

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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.

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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).

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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: KK N1K1NMK N1i S1j 2 i 2 TRR2R 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

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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 2KK K}1{K++= and

()[]2/12 KKK 1= dk 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.

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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.

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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.

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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 data 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 di stribution 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

FNP-FSAR-3 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 21 5/08 TABLE 3.7-6 COMPARISON OF TRANSLATIONAL AND TORSIONAL FREQUENCIES Horizontal

Translational Torsional Mode Fre q uenc y Fre q uenc y Structure NR (Hz)

(Hz)(a) Internal structures 1 16.21 26.72 2 41.51 63.31 A uxiliar y buildin g 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 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 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

FNP-FSAR-3 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 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

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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.

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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 hours0.00278 days 0.0667 hours 3.968254e-4 weeks 9.132e-5 months . 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.

FNP-FSAR-3

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 ASM E 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 ASM E 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 effects are considered in the 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 22 8/09 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-65 CVI Seismic Analysis Report (Dynamic/Seismic Analysis) Penetration Room Filtration System

Fans AMCA Test Code 300-67, 211A-65 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 (charc oal) 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-65 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-65 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 22 8/09 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 protection 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 radiation 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 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The power range detector has been tested in temperatures in excess of

175°F for a period of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months 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 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months . 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-3A

3A-i REV 23 5/11 3A CONFORMANCE WITH NRC REGULATORY GUIDES TABLE OF CONTENTS

Page Regulatory Guide 1.1 ....................................................................................................... 3 A-1.1-1 Regulatory Guide 1.2 ....................................................................................................... 3 A-1.2-1 Regulatory Guide 1.3 ....................................................................................................... 3 A-1.3-1 Regulatory Guide 1.4 ....................................................................................................... 3 A-1.4-1 Regulatory Guide 1.5 ....................................................................................................... 3 A-1.5-1 Regulatory Guide 1.6 ....................................................................................................... 3 A-1.6-1 Regulatory Guide 1.7 ....................................................................................................... 3 A-1.7-1 Regulatory Guide 1.8 ....................................................................................................... 3 A-1.8-1 Regulatory Guide 1.9 ....................................................................................................... 3 A-1.9-1 Regulatory Guide 1.10 ................................................................................................... 3A-1

.10-1 Regulatory Guide 1.11 ................................................................................................... 3A-1

.11-1 Regulatory Guide 1.12 ................................................................................................... 3A-1

.12-1 Regulatory Guide 1.13 ................................................................................................... 3A-1

.13-1 Regulatory Guide 1.14 ................................................................................................... 3A-1

.14-1 Regulatory Guide 1.15 ................................................................................................... 3A-1

.15-1 Regulatory Guide 1.16 ................................................................................................... 3A-1

.16-1 Regulatory Guide 1.17 ................................................................................................... 3A-1

.17-1 Regulatory Guide 1.18 ................................................................................................... 3A-1

.18-1 Regulatory Guide 1.19 ................................................................................................... 3A-1

.19-1 Regulatory Guide 1.20 ................................................................................................... 3A-1

.20-1 Regulatory Guide 1.21 ................................................................................................... 3A-1

.21-1 Regulatory Guide 1.22 ................................................................................................... 3A-1

.22-1 Regulatory Guide 1.23 ................................................................................................... 3A-1

.23-1 Regulatory Guide 1.24 ................................................................................................... 3A-1

.24-1 Regulatory Guide 1.25 ................................................................................................... 3A-1

.25-1 Regulatory Guide 1.26 ................................................................................................... 3A-1

.26-1 Regulatory Guide 1.27 ................................................................................................... 3A-1

.27-1 [HISTORICAL]

[Regulatory Guide 1.28] ........................................................................ 3A-1.28-1 Regulatory Guide 1.29 ................................................................................................... 3A-1

.29-1 Regulatory Guide 1.30 ................................................................................................... 3A-1

.30-1 Regulatory Guide 1.31 ................................................................................................... 3A-1

.31-1 Regulatory Guide 1.32 ................................................................................................... 3A-1

.32-1 Regulatory Guide 1.33 ................................................................................................... 3A-1

.33-1 Regulatory Guide 1.34 ................................................................................................... 3A-1

.34-1 [HISTORICAL] [Regulatory Guide 1.35] ........................................................................

3A-1.35-1 Regulatory Guide 1.36 ................................................................................................... 3A-1

.36-1 Regulatory Guide 1.37 ................................................................................................... 3A-1

.37-1 Regulatory Guide 1.38 ................................................................................................... 3A-1

.38-1 Regulatory Guide 1.39 ................................................................................................... 3A-1

.39-1 Regulatory Guide 1.40 ................................................................................................... 3A-1

.40-1 Regulatory Guide 1.41 ................................................................................................... 3A-1

.41-1 FNP-FSAR-3A

3A-ii REV 23 5/11 TABLE OF CONTENTS Page Regulatory Guide 1.42 ................................................................................................... 3A-1

.42-1 Regulatory Guide 1.43 ................................................................................................... 3A-1

.43-1 Regulatory Guide 1.44 ................................................................................................... 3A-1

.44-1 Regulatory Guide 1.45 .................................................................................................... 3A-1

.45-1 Regulatory Guide 1.46 .................................................................................................... 3A-1

.46-1 Regulatory Guide 1.47 .................................................................................................... 3A-1

.47-1 Regulatory Guide 1.48 .................................................................................................... 3A-1

.48-1 Regulatory Guide 1.49 .................................................................................................... 3A-1

.49-1 Regulatory Guide 1.50 .................................................................................................... 3A-1

.50-1 Regulatory Guide 1.51 .................................................................................................... 3A-1

.51-1 Regulatory Guide 1.52 .................................................................................................... 3A-1

.52-1 Regulatory Guide 1.53 .................................................................................................... 3A-1

.53-1 Regulatory Guide 1.54 .................................................................................................... 3A-1

.54-1 Regulatory Guide 1.55 .................................................................................................... 3A-1

.55-1 Regulatory Guide 1.56 .................................................................................................... 3A-1

.56-1 Regulatory Guide 1.57 .................................................................................................... 3A-1

.57-1 Regulatory Guide 1.58 .................................................................................................... 3A-1

.58-1 Regulatory Guide 1.59 .................................................................................................... 3A-1

.59-1 Regulatory Guide 1.60 .................................................................................................... 3A-1

.60-1 Regulatory Guide 1.61 .................................................................................................... 3A-1

.61-1 Regulatory Guide 1.62 .................................................................................................... 3A-1

.62-1 Regulatory Guide 1.63 .................................................................................................... 3A-1

.63-1 Regulatory Guide 1.64 .................................................................................................... 3A-1

.64-1 Regulatory Guide 1.65 .................................................................................................... 3A-1

.65-1 Regulatory Guide 1.66 .................................................................................................... 3A-1

.66-1 Regulatory Guide 1.67 .................................................................................................... 3A-1

.67-1 Regulatory Guide 1.68 .................................................................................................... 3A-1

.68-1 Regulatory Guide 1.69 .................................................................................................... 3A-1

.69-1 Regulatory Guide 1.70 .................................................................................................... 3A-1

.70-1 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

.71-1 Regulatory Guide 1.72 .................................................................................................... 3A-1

.72-1 Regulatory Guide 1.73 .................................................................................................... 3A-1

.73-1 Regulatory Guide 1.74 .................................................................................................... 3A-1

.74-1 Regulatory Guide 1.75 .................................................................................................... 3A-1

.75-1 Regulatory Guide 1.76 .................................................................................................... 3A-1

.76-1 Regulatory Guide 1.77 .................................................................................................... 3A-1

.77-1 Regulatory Guide 1.78 .................................................................................................... 3A-1

.78-1 FNP-FSAR-3A

3A-iii REV 23 5/11 TABLE OF CONTENTS

Page Regulatory Guide 1.79 .................................................................................................... 3A-1

.79-1 Regulatory Guide 1.80 .................................................................................................... 3A-1

.80-1 Regulatory Guide 1.81 .................................................................................................... 3A-1

.81-1 Regulatory Guide 1.82 .................................................................................................... 3A-1

.82-1 Regulatory Guide 1.83 .................................................................................................... 3A-1

.83-1 Regulatory Guide 1.84 .................................................................................................... 3A-1

.84-1 Regulatory Guide 1.85 .................................................................................................. 3A-1

.85-1 Regulatory Guide 1.86 .................................................................................................. 3A-1

.86-1 Regulatory Guide 1.87 .................................................................................................. 3A-1

.87-1 Regulatory Guide 1.88 .................................................................................................. 3A-1

.88-1 Regulatory Guide 1.95 .................................................................................................. 3A-1

.95-1 Regulatory Guide 1.99 .................................................................................................. 3A-1

.99-1 Regulatory Guide 1.108 .............................................................................................. 3A-1.10 8-1 Regulatory Guide 1.109 .............................................................................................. 3A-1.10 9-1 Regulatory Guide 1.111 ................................................................................................ 3A-1.11 1-1 Regulatory Guide 1.112 ................................................................................................ 3A-1.11 2-1 Regulatory Guide 1.113 ................................................................................................ 3A-1.11 3-1 Regulatory Guide 1.127 ................................................................................................ 3A-1.12 7-1 Regulatory Guide 1.155 .............................................................................................. 3A-1.15 5-1 Regulatory Guide 1.163 .............................................................................................. 3A-1.16 3-1 Regulatory Guide 1.182 .............................................................................................. 3A-1.18 2-1 Regulatory Guide 1.190 ................................................................................................ 3A-1.19 0-1 Regulatory Guide 1.194 ................................................................................................ 3A-1.19 4-1 Regulatory Guide 1.195 ................................................................................................ 3A-1.19 5-1 Regulatory Guide 1.196 .............................................................................................. 3A-1.19 6-1 Regulatory Guide 1.197 .............................................................................................. 3A-1.19 7-1

FNP-FSAR-3A

3A-1 REV 23 5/11 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 23 5/11 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 pumped fluid temperatures are less than the saturation temperature at the Technical Specifications (TS) minimum operating containment pressure

prior to the accident, NPSH is determined assuming no increase in containment

pressure from that present prior to postulated loss-of-coolant accidents.

2. If expected maximum pumped fluid temperatures exceed the saturation temperature at the TS minimum operating containment pressure prior to the

accident, NPSH is determined assuming the containment pressure is equal to the

pumped liquid vapor pressure (saturated conditions).

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 during recirculation

following a LBLOCA event which generates bounding debris head losses. 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 for sump temperatures above the saturation temperature at the minimum

operating containment pressure prior to the accident.

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 23 5/11 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 23 5/11 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-1.5-1 REV 23 5/11 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 23 5/11 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 23 5/11 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-1.9-1 REV 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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.

FNP-FSAR-3A

3A-1.13-1 REV 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 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 23 5/11

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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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-1.26-1 REV 23 5/11 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 23 5/11 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 23 5/11

[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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 assurance audits of the procedural development and maintenance program utilizing a representative sampling 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 23 5/11 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 23 5/11

[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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 guidelines 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 the 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, except for the containment air particulate and gas monitors, is not seismically qualified.

The containment air particulate and gas monitor (channels R-11 and R-12) will be qualified to function following a safe shutdown earthquake as described in paragraph 11.4.2.2.3. The skid monitor control panel will provide an alarm signal to the control room on increasing radiation levels and a local indication.

FNP-FSAR-3A

3A-1.46-1 REV 23 5/11 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 23 5/11 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.

FNP-FSAR-3A

3A-1.47-2 REV 23 5/11 10. Service water.

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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 initial 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 regulations 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 23 5/11 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 accordance 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 are 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 23 5/11 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-1.66-1 REV 23 5/11 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 23 5/11 Regulatory Guide 1.67 - INSTALLATION OF OVERPRESSURE PROTECTION DEVICES (Rev. 0, 10/73)

CONFORMANCE (a)

All structures, systems, and components of overpressure 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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 23 5/11 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-1.78-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.79-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.79-2 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.79-3 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.79-4 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.80-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.81-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.81-2 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.82-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.82-2 REV 23 5/11

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 sump water levels for LBLOCA event which generates bounding debris head losses were used in applicable NPSH calculations. The

NPSH calculations used a range of sump fluid temperatures between 120 ºF and

212 ºF. This temperature range covers the time period during the post-LOCA

cooldown that results in the minimum post-LOCA containment sump NPSH

margin.

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.1.3.3.8: 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 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. 3: The sumps are located on the lowest floor elevation in the containment exclusive of the reactor vessel cavity. Each sump intake 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.

FNP-FSAR-3A

3A-1.82-3 REV 23 5/11 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.

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.]

FNP-FSAR-3A

3A-1.83-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.84-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.85-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.86-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.87-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.88-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.95-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.99-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.108-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.109-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.111-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.112-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.113-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.127-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.155-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.163-1 REV 23 5/11 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.

FNP-FSAR-3A

3A-1.190-1 REV 23 5/11 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.

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 substantiate 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 the 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 eigenvectors 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 matrix 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

FNP-FSAR-3F

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.

FNP-FSAR-3F

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

FNP-FSAR-3F

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

FNP-FSAR-3F

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.

FNP-FSAR-3H

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

FNP-FSAR-3H

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

FNP-FSAR-3H

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 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . 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.

FNP-FSAR-3H

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 for 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 instrumentation 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 techniques 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 predicted 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.

]

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

]

FNP-FSAR-3I 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

FNP-FSAR-3I

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

nsinm cos mn1m0n=== 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.

FNP-FSAR-3I

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.

FNP-FSAR-3I

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

FNP-FSAR-3J

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 break. 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 temperature 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 temperature. 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.

ML18312A070

Text

3.0 DESIGN

3.1.1 Criterion

3.1.2 Criterion

3.1.4 Criterion

3.1.5 Criterion

3.1.6 Criterion

3.1.7 Criterion

3.1.8 Criterion

3.1.9 Criterion

3.2 CLASSIFICATION

3.2.1 Seismic

3.2.2 System

3.3.2 Tornado

3.4 WATER

3.4.1 Flood

3.4.2 Analysis

3.5 MISSILE

3.5.1 Missile

3.5.2 Missile

3.5.3 Selected

3.5.4 Barrier

3.5.5 Missile

3.6 PROTECTION

3.6.1 Systems

3.6.2 Design

3.6.3 Design

3.6.4 Dynamic

3.6.5 Protective

3.6.6 Structural

3.7 SEISMIC

3.7.1 Seismic

3.7.2 Seismic

3.7.3 Seismic

3.7.4 Seismic

3.7.5 Seismic

3.8 DESIGN

3.8.1 Concrete

3.8.2 Steel

3.8.3 Internal

3.8.4 Other

3.8.5 Foundations

3.8.6 Masonry

3.9 MECHANICAL

3.9.1 Dynamic

3.9.3 Components

3.9.4 Operability

3.8.3 Containment

SUMMARY

3.4.1 FLOOD

3.4.2 ANALYSIS

3.6.1 SYSTEMS

3.6.2 DESIGN

3.6.3 DESIGN

3.6.5 PROTECTIVE

3.6.6 STRUCTURAL

3.7.1 SEISMIC

3.7.2 SEISMIC

3.7.3 SEISMIC

3.7.4 SEISMIC

3.9.1 DYNAMIC

3.9.3 COMPONENTS

3.9.4 OPERABILITY

SUMMARY

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

1.1 Integrated

1.4 Recirculation

3.1 Accumulator

3.2 Testing

3.3 Initiating

3.4 Testing

3.5 System

References: