ML17212A070

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Final Safety Analysis Report, Rev. 30, Chapter 3, Design of Structures, Components, Equipment, and Systems
ML17212A070
Person / Time
Site: Millstone Dominion icon.png
Issue date: 06/29/2017
From:
Dominion Nuclear Connecticut
To:
Office of Nuclear Reactor Regulation
Shared Package
ML17212A038 List:
References
17-208
Download: ML17212A070 (835)
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Text

MPS-3 FSAR Millstone Power Station Unit 3 Safety Analysis Report Chapter 3

Table of Contents tion Title Page CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA ................... 3.1-1 1 Conformance with Single Failure Criterion................................................ 3.1-1 1.1 Single Failure Criterion .............................................................................. 3.1-1 1.2 Definitions of Terms Used in Single Failure Criterion............................... 3.1-1 1.3 Application of Single Failure Criterion ...................................................... 3.1-2 2 Criterion Conformance ............................................................................... 3.1-6 2.1 Quality Standards and Records (Criterion 1).............................................. 3.1-6 2.2 Design Bases for Protection against Natural Phenomena (Criterion 2)...... 3.1-6 2.3 Fire Protection (Criterion 3) ....................................................................... 3.1-8 2.4 Environmental and Missile Design Bases (Criterion 4) ............................. 3.1-9 2.5 Sharing of Structures, Systems, and Components (Criterion 5) ............... 3.1-10 2.6 (Criterion 6) .............................................................................................. 3.1-11 2.7 (Criterion 7) .............................................................................................. 3.1-11 2.8 (Criterion 8) .............................................................................................. 3.1-11 2.9 (Criterion 9) .............................................................................................. 3.1-11 2.10 Reactor Design (Criterion 10)................................................................... 3.1-11 2.11 Reactor Inherent Protection (Criterion 11) ............................................... 3.1-12 2.12 Suppression of Reactor Power Oscillations (Criterion 12)....................... 3.1-12 2.13 Instrumentation and Control (Criterion 13) .............................................. 3.1-13 2.14 Reactor Coolant Pressure Boundary (Criterion 14) .................................. 3.1-14 2.15 Reactor Coolant System Design (Criterion 15) ........................................ 3.1-14 2.16 Containment Design (Criterion 16) .......................................................... 3.1-15 2.17 Electric Power Systems (Criterion 17) ..................................................... 3.1-15 2.18 Inspection and Testing of Electric Power Systems (Criterion 18)............ 3.1-16 2.19 Control Room (Criterion 19) .................................................................... 3.1-17 2.20 Protection System Functions (Criterion 20) ............................................. 3.1-17 2.21 Protection System Reliability and Testability (Criterion 21) ................... 3.1-18 2.22 Protection System Independence (Criterion 22) ....................................... 3.1-19 2.23 Protection System Failure Modes (Criterion 23) ...................................... 3.1-19 2.24 Separation of Protection and Control Systems (Criterion 24) .................. 3.1-20 2.25 Protection System Requirements for Reactivity Control Malfunctions (Criterion 25) ............................................................................................ 3.1-20 2.26 Reactivity Control System Redundancy and Capability (Criterion 26) ... 3.1-21 2.27 Combined Reactivity Control System Capability (Criterion 27).............. 3.1-22 2.28 Reactivity Limits (Criterion 28) ............................................................... 3.1-22 2.29 Protection against Anticipated Operational Occurrences (Criterion 29) .. 3.1-23 2.30 Quality of Reactor Coolant Pressure Boundary (Criterion 30)................. 3.1-23 2.31 Fracture Prevention of Reactor Coolant Pressure Boundary (Criterion 31) . 3.1-3-i Rev. 30

Table of Contents (Continued) 24 2.32 Inspection of Reactor Coolant Pressure Boundary (Criterion 32) ............ 3.1-25 2.33 Reactor Coolant Makeup (Criterion 33) ................................................... 3.1-26 2.34 Residual Heat Removal (Criterion 34) ..................................................... 3.1-26 2.35 Emergency Core Cooling (Criterion 35) .................................................. 3.1-27 2.36 Inspection of Emergency Core Cooling System (Criterion 36)................ 3.1-28 2.37 Testing of Emergency Core Cooling System (Criterion 37) .................... 3.1-28 2.38 Containment Heat Removal (Criterion 38)............................................... 3.1-28 2.39 Inspection of Containment Heat Removal System (Criterion 39) ............ 3.1-29 2.40 Testing of Containment Heat Removal System (Criterion 40)................. 3.1-29 2.41 Containment Atmosphere Cleanup (Criterion 41).................................... 3.1-30 2.42 Inspection of Containment Atmosphere Cleanup Systems (Criterion 42) 3.1-31 2.43 Testing of Containment Atmosphere Cleanup Systems (Criterion 43) .... 3.1-31 2.44 Cooling Water (Criterion 44).................................................................... 3.1-31 2.45 Inspection of Cooling Water System (Criterion 45) ................................. 3.1-32 2.46 Testing of Cooling Water System (Criterion 46)...................................... 3.1-33 2.47 (Criterion 47) ............................................................................................ 3.1-33 2.48 (Criterion 48) ............................................................................................ 3.1-33 2.49 (Criterion 49) ............................................................................................ 3.1-33 2.50 Containment Design Basis (Criterion 50)................................................. 3.1-34 2.51 Fracture Prevention of Containment Pressure Boundary (Criterion 51) .. 3.1-34 2.52 Capability for Containment Leakage Rate Testing (Criterion 52) ........... 3.1-35 2.53 Provisions for Containment Testing and Inspection (Criterion 53).......... 3.1-35 2.54 Piping Systems Penetrating Containment (Criterion 54).......................... 3.1-36 2.55 Reactor Coolant Pressure Boundary Penetrating Containment (Criterion 55) ....

3.1-36 2.56 Primary Containment Isolation (Criterion 56).......................................... 3.1-37 2.57 Closed System Isolation Valves (Criterion 57) ........................................ 3.1-38 2.58 (Criterion 58) ............................................................................................ 3.1-38 2.59 (Criterion 59) ............................................................................................ 3.1-38 2.60 Control of Releases of Radioactive Materials to the Environment (Criterion 60) 3.1-39 2.61 Fuel Storage and Handling and Radioactive Control (Criterion 61) ........ 3.1-40 2.62 Prevention of Criticality in Fuel Storage and Handling (Criterion 62) .... 3.1-41 2.63 Monitoring Fuel and Waste Storage (Criterion 63) .................................. 3.1-41 2.64 Monitoring Radioactivity Releases (Criterion 64).................................... 3.1-42 3 Reference for Section 3.1.......................................................................... 3.1-43 CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS.. 3.2-1 1 Seismic Classification................................................................................. 3.2-1 3-ii Rev. 30

Table of Contents (Continued) 2 System Quality Group Classification ......................................................... 3.2-1 3 Quality Assurance Categories..................................................................... 3.2-6 4 Other Classification Systems ...................................................................... 3.2-6 5 Tabulation of Codes and Classifications .................................................... 3.2-7 WIND AND TORNADO LOADINGS ............................................................... 3.3-1 1 Wind Loadings............................................................................................ 3.3-1 1.1 Design Wind Velocity ................................................................................ 3.3-1 1.2 Determination of Applied Forces ............................................................... 3.3-1 2 Tornado Loadings ....................................................................................... 3.3-1 2.1 Applicable Design Parameters.................................................................... 3.3-1 2.2 Determination of Forces on Structures ....................................................... 3.3-2 2.3 Effect of Failure of Structures or Components not Designed for Tornado Loads 3.3-4 3 References for Section 3.3 .......................................................................... 3.3-4 WATER LEVEL (FLOOD) DESIGN ................................................................. 3.4-1 1 Flood Protection.......................................................................................... 3.4-1 1.1 Flood Protection Measures for Seismic Category I Structures................... 3.4-1 1.2 Permanent Dewatering System ................................................................... 3.4-3 2 Analytical and Test Procedures .................................................................. 3.4-3 3 Reference for Section 3.4............................................................................ 3.4-3 MISSILE PROTECTION .................................................................................... 3.5-1 1 Missile Selection and Description .............................................................. 3.5-1 1.1 Internally Generated Missiles (Outside Containment) ............................... 3.5-1 1.2 Internally Generated Missiles (Inside Containment) .................................. 3.5-3 1.2.1 Missile Selection and Description .............................................................. 3.5-3 1.2.2 Missile Protection Provided........................................................................ 3.5-6 1.3 Turbine Missiles ......................................................................................... 3.5-7 1.3.1 Turbine Placement and Orientation ............................................................ 3.5-7 1.3.2 Missiles Identification and Characteristics ................................................. 3.5-7 1.3.3 Target Description ...................................................................................... 3.5-7 1.3.4 Probability Analysis.................................................................................... 3.5-7 1.3.5 Turbine Overspeed Protection .................................................................... 3.5-8 1.3.6 Turbine Valve Testing ................................................................................ 3.5-8 1.4 Missiles Generated by Natural Phenomena ................................................ 3.5-8 1.5 Missiles Generated by Events Near the Site ............................................... 3.5-9 3-iii Rev. 30

Table of Contents (Continued) 1.6 Aircraft Hazards.......................................................................................... 3.5-9 2 Structures, Systems, and Components to be Protected from Externally Generat-ed Missiles .................................................................................................. 3.5-9 3 Barrier Design Procedures .......................................................................... 3.5-9 3.1 Concrete Barriers ........................................................................................ 3.5-9 3.2 Steel Barriers............................................................................................. 3.5-11 3.3 Design Evaluation..................................................................................... 3.5-11 3.4 Secondary Missiles ................................................................................... 3.5-12 4 References for Section 3.5 ........................................................................ 3.5-12 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURES OF PIPING .......................................................... 3.6-1 1 Postulated Piping Failures in Fluid Systems Inside and Outside of Containment 3.6-1 1.1 Design Bases............................................................................................... 3.6-1 1.1.1 Design Basis Protection Criteria................................................................. 3.6-1 1.1.2 Design Basis Pipe Break/Crack Criteria ..................................................... 3.6-2 1.1.3 Essential Systems, Components, and Structures ........................................ 3.6-2 1.1.4 Design Approach ........................................................................................ 3.6-2 1.2 Description.................................................................................................. 3.6-3 1.3 Safety Evaluation ........................................................................................ 3.6-4 1.3.1 Operability of Essential Systems and Components .................................... 3.6-4 1.3.2 Failure Mode and Effects............................................................................ 3.6-5 1.3.3 Pipe Break/Crack Analysis ......................................................................... 3.6-6 2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping .................................................................... 3.6-18 2.1 Criteria Used to Define Break and Crack Location and Configuration.... 3.6-18 2.1.1 Criteria for Inside Containment ................................................................ 3.6-19 2.1.2 Criteria for Outside Containment ............................................................. 3.6-20 2.1.3 Design Basis Break/Crack Types and Orientation ................................... 3.6-24 2.1.4 Conformance with Regulatory Guide 1.46 ............................................... 3.6-25 2.2 Analytical Methods to Define Forcing Functions and Response Models 3.6-26 2.2.1 Introduction............................................................................................... 3.6-26 2.2.2 Time Dependent Blowdown Force ........................................................... 3.6-27 2.2.3 Simplified Blowdown Analysis ................................................................ 3.6-33 2.2.4 Lumped-Parameter Dynamic Analysis ..................................................... 3.6-34 2.2.5 Energy Balance Analysis .......................................................................... 3.6-35 2.2.6 Local Pipe Indentation .............................................................................. 3.6-38 2.2.7 Concrete Barrier Impact............................................................................ 3.6-39 2.3 Dynamic Analysis Methods to Verify Integrity and Operability ............. 3.6-39 3-iv Rev. 30

Table of Contents (Continued) 2.3.1 Pipe Rupture Restraints ............................................................................ 3.6-41 2.4 Guard Pipe Assembly Design Criteria...................................................... 3.6-42 3 References for Section 3.6 ........................................................................ 3.6-42 SEISMIC DESIGN .............................................................................................. 3.7-1 B.1 Seismic Input ............................................................................................3.7B-2 B.1.1 Design Response Spectra..........................................................................3.7B-2 B.1.2 Design Time History.................................................................................3.7B-2 B.1.3 Critical Damping Values ..........................................................................3.7B-2 B.1.4 Supporting Media for Seismic Category I Structures ...............................3.7B-3 B.2 Seismic System Analysis ..........................................................................3.7B-3 B.2.1 Seismic Analysis Methods........................................................................3.7B-3 B.2.2 Natural Frequencies and Response Loads ................................................3.7B-4 B.2.2.1 Containment Summary .............................................................................3.7B-4 B.2.2.2 Main Steam Valve Building Summary .....................................................3.7B-5 B.2.2.3 Emergency Generator Enclosure Summary..............................................3.7B-5 B.2.3 Procedures Used for Analytical Modeling................................................3.7B-6 B.2.4 Soil-Structure Interaction..........................................................................3.7B-7 B.2.5 Development of Floor Response Spectra..................................................3.7B-7 B.2.6 Three Components of Earthquake Motion................................................3.7B-8 B.2.7 Combination of Modal Responses ............................................................3.7B-8 B.2.8 Interaction of Non-Category I Structures with Seismic Category I Structures ...

3.7B-8 B.2.9 Effects of Parameter Variations on Floor Response Spectra ....................3.7B-8 B.2.10 Use of Constant Vertical Static Factors ....................................................3.7B-8 B.2.11 Method Used to Account for Torsional Effects........................................3.7B-8 B.2.12 Comparison of Responses.........................................................................3.7B-9 B.2.13 Methods for Seismic Analysis of Category I Dams .................................3.7B-9 B.2.14 Determination of Seismic Category I Structure Overturning Moments ...3.7B-9 B.2.15 Analysis Procedure for Damping............................................................3.7B-10 B.3 Seismic Subsystem Analysis ..................................................................3.7B-12 B.3.1 Seismic Analysis Methods......................................................................3.7B-12 B.3.1.1 Equipment and Components ...................................................................3.7B-12 B.3.1.2 Piping Systems........................................................................................3.7B-19 B.3.2 Determination of Number of Earthquake Cycles ...................................3.7B-20 B.3.2.1 Equipment and Components ...................................................................3.7B-20 B.3.2.2 Piping Systems........................................................................................3.7B-20 B.3.3 Procedures Used for Modeling ...............................................................3.7B-21 B.3.3.1 Equipment and Components ...................................................................3.7B-21 B.3.3.2 Piping Systems........................................................................................3.7B-21 3-v Rev. 30

Table of Contents (Continued)

B.3.4 Basis for Selection of Frequencies..........................................................3.7B-21 B.3.4.1 Equipment and Components ...................................................................3.7B-21 B.3.4.2 Piping Systems........................................................................................3.7B-21 B.3.5 Use of Equivalent Static Load Method of Analysis ...............................3.7B-21 B.3.5.1 Equipment and Components ...................................................................3.7B-21 B.3.5.2 Piping Systems........................................................................................3.7B-25 B.3.6 Three Components of Earthquake Motion..............................................3.7B-26 B.3.6.1 Equipment and Components ...................................................................3.7B-26 B.3.6.2 Piping Systems........................................................................................3.7B-26 B.3.7 Combination of Modal Responses ..........................................................3.7B-26 B.3.8 Analytical Procedures for Piping Systems..............................................3.7B-28 B.3.8.1 Natural Frequencies and Mode Shapes...................................................3.7B-29 B.3.8.2 Dynamic Response .................................................................................3.7B-29 B.3.8.3 Response Spectrum Modal Analysis ......................................................3.7B-30 B.3.9 Multiply Supported Equipment and Components with Distinct Inputs..3.7B-31 B.3.10 Use of Constant Vertical Static Factors ..................................................3.7B-31 B.3.10.1 Equipment and Components ...................................................................3.7B-31 B.3.10.2 Piping Systems........................................................................................3.7B-31 B.3.11 Torsional Effects of Eccentric Masses....................................................3.7B-32 B.3.12 Buried Seismic Category I Piping Systems ............................................3.7B-32 B.3.12.1 Seismic Wave Effect in the Free Field ...................................................3.7B-32 B.3.12.2 Effects of Differential Movements Between Structure and Adjacent Soil Due to Seismic Motion .......................................................................................3.7B-36 B.3.12.3 Effects of Differential Movements Due to Structural Settlement...........3.7B-37 B.3.12.4 Accommodations for Buried Piping Structural Penetrations..................3.7B-37 B.3.13 Interaction of Other Systems (Piping and Equipment) with Seismic Category I Systems (Piping and Equipment)............................................................3.7B-37 B.3.14 Seismic Analysis for Reactor Internals...................................................3.7B-38 B.3.15 Analysis Procedure for Damping............................................................3.7B-38 N SEISMIC DESIGN ....................................................................................... 3.7N-110 N.1 Seismic Input ....................................................................................... 3.7N-110 N.1.1 Design Response Spectra..................................................................... 3.7N-110 N.1.2 Design Time History............................................................................ 3.7N-110 N.1.3 Critical Damping Values ..................................................................... 3.7N-110 N.1.4 Supporting Media for Seismic Category I Structures .......................... 3.7N-111 N.2 Seismic System Analysis ..................................................................... 3.7N-111 N.3 Seismic Subsystem Analysis ............................................................... 3.7N-111 N.3.1 Seismic Analysis Methods................................................................... 3.7N-111 N.3.1.1 Dynamic Analysis - Mathematical Model ........................................... 3.7N-111 3-vi Rev. 30

Table of Contents (Continued)

N.3.1.2 Modal Analysis .................................................................................... 3.7N-113 N.3.1.3 Response Spectrum Analysis............................................................... 3.7N-116 N.3.2 Determination of Number of Earthquake Cycles ................................ 3.7N-116 N.3.3 Procedure Used for Modeling.............................................................. 3.7N-116 N.3.4 Basis for Selection of Frequencies....................................................... 3.7N-116 N.3.5 Use of Equivalent Static Load Method of Analysis ............................ 3.7N-117 N.3.6 Three Components of Earthquake Motion........................................... 3.7N-117 N.3.7 Combination of Modal Responses ....................................................... 3.7N-117 N.3.8 Analytical Procedures for Piping ......................................................... 3.7N-119 N.3.9 Multiply Supported Equipment Components with Distinct Inputs ..... 3.7N-119 N.3.10 Use of Constant Vertical Static Factors ............................................... 3.7N-119 N.3.11 Torsional Effects of Eccentric Masses................................................. 3.7N-119 N.3.12 Buried Seismic Category I Piping Systems and Tunnels .................... 3.7N-119 N.3.13 Interaction of Other Piping with Seismic Category I Piping ............... 3.7N-119 N.3.14 Seismic Analyses for Reactor Internals ............................................... 3.7N-119 N.3.15 Analysis Procedure for Damping......................................................... 3.7N-120 N.4 Seismic Instrumentation ...................................................................... 3.7N-120 4.1 Comparison with Regulatory Guide 1.12 ............................................... 3.7-122 4.2 Location and Description of Instrumentation ......................................... 3.7-122 4.3 Control Room Operator Notification ...................................................... 3.7-123 4.4 Comparison of Measured and Predicted Responses ............................... 3.7-124 5 References for Section 3.7 ...................................................................... 3.7-124 DESIGN OF CATEGORY I STRUCTURES ..................................................... 3.8-1 1 Concrete Containment ................................................................................ 3.8-1 1.1 Description of the Containment .................................................................. 3.8-1 1.1.1 Base Foundation ......................................................................................... 3.8-2 1.1.2 Cylindrical Wall.......................................................................................... 3.8-2 1.1.3 Dome........................................................................................................... 3.8-3 1.1.4 Steel Liner and Penetrations ....................................................................... 3.8-3 1.1.5 Ring Girder ................................................................................................. 3.8-7 1.2 Applicable Codes, Standards, and Specifications....................................... 3.8-7 1.2.1 General........................................................................................................ 3.8-7 1.2.2 Structural Specifications ............................................................................. 3.8-8 1.2.3 Steel Liner and Penetrations ..................................................................... 3.8-10 1.3 Loads and Loading Combinations ............................................................ 3.8-11 1.3.1 Containment Mat, Shell, and Dome.......................................................... 3.8-11 1.3.2 Steel Liner and Penetrations ..................................................................... 3.8-14 1.4 Design and Analysis Procedures............................................................... 3.8-15 1.4.1 Containment Structure .............................................................................. 3.8-15 3-vii Rev. 30

Table of Contents (Continued) 1.4.2 Steel Liner and Penetrations ..................................................................... 3.8-19 1.5 Structural Acceptance Criteria.................................................................. 3.8-19 1.5.1 Containment Structure .............................................................................. 3.8-19 1.5.2 Steel Liner and Penetration....................................................................... 3.8-20 1.6 Materials, Quality Control, and Special Construction Techniques .......... 3.8-20 1.6.1 Concrete .................................................................................................... 3.8-20 1.6.2 Reinforcing Steel ...................................................................................... 3.8-23 1.6.3 Structural Steel.......................................................................................... 3.8-26 1.6.4 Waterproofing Membrane......................................................................... 3.8-26 1.6.5 Steel Liner and Penetrations ..................................................................... 3.8-27 1.6.6 Backfill Around Containment Structure ................................................... 3.8-28 1.7 Testing and Inservice Surveillance Requirements.................................... 3.8-28 1.7.1 Concrete Containment .............................................................................. 3.8-28 2 Steel Containment..................................................................................... 3.8-30 3 Concrete and Structural Steel Internal Structures of Steel or Concrete Contain-ments ......................................................................................................... 3.8-30 3.1 Description of Internal Structures............................................................. 3.8-30 3.2 Applicable Codes, Standards, and Specifications..................................... 3.8-31 3.3 Loads and Loading Combinations ............................................................ 3.8-31 3.4 Design and Analysis Procedures............................................................... 3.8-32 3.5 Structural Acceptance Criteria.................................................................. 3.8-32 3.6 Materials, Quality Control, and Special Construction Techniques .......... 3.8-33 3.7 Testing and Inservice Surveillance Requirements.................................... 3.8-34 4 Other Seismic Category I Structures (and Major Nonsafety Related Structures) 3.8-34 4.1 Description of the Structures .................................................................... 3.8-34 4.2 Applicable Codes, Standards, and Specifications..................................... 3.8-43 4.3 Loads and Loading Combinations ............................................................ 3.8-43 4.4 Design and Analysis Procedures............................................................... 3.8-44 4.5 Structural Acceptance Criteria.................................................................. 3.8-44 4.6 Materials, Operating Control, and Special Construction Techniques ...... 3.8-46 4.7 Testing and Inservice Surveillance Requirements.................................... 3.8-46 4.8 Masonry Walls .......................................................................................... 3.8-46 5 Foundations............................................................................................... 3.8-46 5.1 Description of the Foundations................................................................. 3.8-46 5.2 Applicable Codes, Standards, and Specifications..................................... 3.8-47 5.3 Loads and Loading Combinations ............................................................ 3.8-48 5.4 Design and Analysis Procedures............................................................... 3.8-48 5.5 Structural Acceptance Criteria.................................................................. 3.8-49 5.6 Materials, Quality Control, and Special Construction Techniques .......... 3.8-49 5.7 Testing and Inservice Surveillance Requirements.................................... 3.8-50 3-viii Rev. 30

Table of Contents (Continued) 6 References for Section 3.8 ........................................................................ 3.8-50 MECHANICAL SYSTEMS AND COMPONENTS .......................................... 3.9-1 B.1 Special Topics for Mechanical Components ............................................3.9B-2 B.1.1 Design Transients .....................................................................................3.9B-2 B.1.2 Computer Programs Used in Analysis......................................................3.9B-3 B.1.3 Experimental Stress Analysis ...................................................................3.9B-3 B.1.4 Consideration for the Evaluation of the Faulted Conditions ....................3.9B-3 B.1.4.1 Loading Conditions...................................................................................3.9B-3 B.1.4.2 Evaluation of Reactor Coolant Loop and Supports for Faulted Loading Condi-tion ............................................................................................................3.9B-4 B.1.4.3 Reactor Coolant Loop Models and Methods ............................................3.9B-5 B.1.4.4 Primary Component Supports Models and Methods ................................3.9B-8 B.1.4.5 Equipment and Components .....................................................................3.9B-9 B.2 Dynamic Testing and Analysis ...............................................................3.9B-10 B.2.1 Preoperational Vibration and Dynamic Effects Testing on Piping.........3.9B-10 B.2.2 Seismic Qualification Testing of Safety Related Mechanical Equipment..3.9B-11 B.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures...................................................................................3.9B-12 B.3.1 Loading Combinations, Design Transients, and Stress Limits ...............3.9B-12 B.3.1.1 ASME III Class 1 Components ..............................................................3.9B-12 B.3.1.2 ASME III Class 2 and 3 Components.....................................................3.9B-12 B.3.2 Pump and Valve Operability Assurance .................................................3.9B-13 B.3.2.1 Pump Operating Program .......................................................................3.9B-13 B.3.2.2 Valve Operability Program .....................................................................3.9B-16 B.3.3 Design and Installation Details for Mounting of Pressure Relief Devices .3.9B-18 B.3.3.1 Open Relief System ................................................................................3.9B-19 B.3.3.2 Closed Relief System..............................................................................3.9B-20 B.3.4 Component Supports...............................................................................3.9B-20 N MECHANICAL SYSTEMS AND COMPONENTS ..................................... 3.9N-27 N.1 Special Topics for Mechanical Components ......................................... 3.9N-27 N.1.1 Design Transients .................................................................................. 3.9N-27 N.1.2 Computer Programs Used in Analyses .................................................. 3.9N-42 N.1.3 Experimental Stress Analysis ................................................................ 3.9N-42 N.1.4 Considerations for the Evaluation of the Faulted Condition ................. 3.9N-43 N.1.4.1 Loading Conditions................................................................................ 3.9N-43 3-ix Rev. 30

Table of Contents (Continued)

N.1.4.2 Analysis of Primary Components .......................................................... 3.9N-43 N.1.4.3 Dynamic Analysis of Reactor Pressure Vessel for Postulated Loss-of-Coolant Accident ................................................................................................. 3.9N-44 N.1.4.4 Stress Criteria for Reactor Coolant System Components ...................... 3.9N-47 N.2 Dynamic Testing and Analysis .............................................................. 3.9N-47 N.2.1 Preoperational Vibration and Dynamic Effects Testing on Piping........ 3.9N-47 N.2.2 Seismic Qualification Testing of Safety Related Mechanical Equipment. 3.9N-47 N.2.3 Dynamic Response Analysis of Reactor Internals Under Operational Flow Transients and Steady State Conditions................................................. 3.9N-47 N.2.4 Preoperational Flow-Induced Vibration Testing of Reactor Internals... 3.9N-49 N.2.5 Dynamic System Analysis of the Reactor Internals Under Faulted Conditions..

3.9N-51 N.2.6 Correlations of Reactor Internals Vibration Tests with the Analytical Results...

3.9N-55 N.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures.................................................................................. 3.9N-55 N.3.1 Loading Combinations, Design Transients, and Stress Limits for Class 2 Components ........................................................................................... 3.9N-55 N.3.1.1 Design Loading Combinations .............................................................. 3.9N-55 N.3.1.2 Design Stress Limits .............................................................................. 3.9N-55 N.3.2 Pumps and Valve Operability Assurance .............................................. 3.9N-56 N.3.2.1 ASME Code Class Pumps ..................................................................... 3.9N-56 N.3.2.2 Valve Operability................................................................................... 3.9N-56 N.3.2.3 Pump Motor and Valve Operator Qualification .................................... 3.9N-58 N.3.3 Design and Installation Details for Mounting of Pressure-Relieving Devices ....

3.9N-58 N.3.4 Component Supports.............................................................................. 3.9N-58 N.3.4.1 Component Supports for Tanks and Heat Exchangers .......................... 3.9N-58 N.3.4.2 Component Supports for Pumps ............................................................ 3.9N-59 N.4 Control Rod Drive System (CRDS)....................................................... 3.9N-59 N.4.1 Descriptive Information of CRDS ......................................................... 3.9N-59 N.4.2 Applicable CRDS Design Specifications .............................................. 3.9N-64 N.4.3 Design Loads, Stress Limits, and Allowable Deformations .................. 3.9N-65 N.4.3.1 Pressure Vessel Assembly ..................................................................... 3.9N-65 N.4.3.2 Drive Rod Assembly.............................................................................. 3.9N-66 N.4.3.3 Latch Assembly and Coil Stack Assembly............................................ 3.9N-66 N.4.4 CRDS Performance Assurance Program ............................................... 3.9N-68 N.5 Reactor Vessel Internals ........................................................................ 3.9N-69 N.5.1 Design Arrangements ............................................................................ 3.9N-69 N.5.2 Design Loading Conditions ................................................................... 3.9N-74 3-x Rev. 30

Table of Contents (Continued)

N.5.3 Design Loading Categories.................................................................... 3.9N-75 N.5.4 Design Bases.......................................................................................... 3.9N-76 N.6 Inservice Testing of Pumps and Valves................................................. 3.9N-77 7 Inservice Testing of Pumps and Valves.................................................... 3.9-90 7.1 Inservice Testing of Pumps....................................................................... 3.9-90 7.2 Inservice Testing of Valves ...................................................................... 3.9-90 SEISMIC QUALIFICATION OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT ................................................................. 3.10-1 B.1 Seismic Qualification Criteria ................................................................3.10B-2 B.2 Methods and Procedures for Qualifying Electrical Equipment and Instrumenta-tion ..........................................................................................................3.10B-2 B.3 Methods and Procedures of Analysis or Testing of Supports of Electrical Equip-ment and Instrumentation .......................................................................3.10B-3 0B.4 Replacement Items..................................................................................3.10B-3 B.5 Operating License Review......................................................................3.10B-4 N SEISMIC QUALIFICATION OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT .............................................................. 3.10N-5 N.1 Seismic Qualification Criteria ............................................................... 3.10N-5 N.1.1 Qualification Standards.......................................................................... 3.10N-5 N.1.2 Performance Requirements for Seismic Qualification .......................... 3.10N-5 N.1.3 Acceptance Criteria................................................................................ 3.10N-6 N.2 Methods and Procedures for Qualifying Electrical Equipment and Instrumenta-tion ......................................................................................................... 3.10N-6 N.2.1 Seismic Qualification by Type Test....................................................... 3.10N-6 N.2.2 Seismic Qualification by Analysis......................................................... 3.10N-7 N.3 Method and Procedures for Qualifying Supports of Electrical Equipment and Instrumentation ...................................................................................... 3.10N-8 N.4 Operating License Review..................................................................... 3.10N-8

.1 Replacement Items.................................................................................... 3.10-9

.2 References for Section 3.10 ...................................................................... 3.10-9 ENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRICAL EQUIPMENT .................................................................................................... 3.11-1 B.1 Equipment Identification and Environmental Conditions ......................3.11B-3 B.1.1 Environmental Conditions ......................................................................3.11B-3 B.1.2 Equipment Identification ........................................................................3.11B-4 3-xi Rev. 30

Table of Contents (Continued)

B.2 Qualification Tests and Analyses ...........................................................3.11B-4 B.2.1 Regulatory Guides ..................................................................................3.11B-4 B.2.2 Safety-Related (Class 1E) Equipment and Component Qualifications ..3.11B-5 B.3 Qualification Test Results.......................................................................3.11B-8 B.3.1 Nuclear Steam Supply System Equipment Qualification Program ........3.11B-8 B.3.2 Qualification of Safety-Related Equipment (BOP) not covered by Section 3.11.

3.11B-8 B.4 Loss of Ventilation..................................................................................3.11B-8 B.5 Chemical and Radiation Environment ....................................................3.11B-9 B.5.1 Radiation Environment ...........................................................................3.11B-9 B.5.2 Chemical Environment .........................................................................3.11B-10 N ENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRICAL EQUIPMENT ............................................................................................... 3.11N-11 N.1 Equipment Identification and Environmental Conditions ................... 3.11N-11 N.2 Qualification Tests and Analysis ......................................................... 3.11N-11 N.2.1 Environmental Qualification Criteria .................................................. 3.11N-11 N.2.2 Performance Requirements for Environmental Qualification ............. 3.11N-11 N.2.3 Methods and Procedures for Environmental Qualification ................. 3.11N-12 N.3 Qualification Test Results.................................................................... 3.11N-12 N.4 Loss of Ventilation............................................................................... 3.11N-12 N.5 Estimated Chemical and Radiation Environment ................................ 3.11N-12

.6 References for Section 3.11 .................................................................... 3.11-13 PENDIX 3A -COMPUTER PROGRAMS FOR DYNAMIC AND STATIC ANALYSIS OF SEISMIC CATEGORY I STRUCTURES, EQUIPMENT, AND COMPO-NENTS..................................................................................................... 3A.1-0 1 STRUCTURES ........................................................................................ 3A.1-1 1.1 STRUDL II .............................................................................................. 3A.1-1 1.2 SHELL 1 .................................................................................................. 3A.1-2 1.3 STRUDL-SW........................................................................................... 3A.1-3 1.4 ASAAS (Asymmetric Stress Analysis of Axisymmetric Solids) ............ 3A.1-3 1.5 TAC2D (A General Purpose Two-Dimensional Heat Transfer Computer Code) 3A.1-5 1.6 Time History Program ............................................................................. 3A.1-7 1.7 PLAXLY.................................................................................................. 3A.1-8 1.8 MAT5 (Circular Mat with Axisymmetric Loading) ................................ 3A.1-8 1.9 ANSYS .................................................................................................... 3A.1-8 1.10 MEMBRANE (Membrane Stress Analysis)............................................ 3A.1-9 1.11 SBMMI (Single Barrier Mass Missile Impact)........................................ 3A.1-9 1.12 GHOSH-WILSON................................................................................. 3A.1-10 3-xii Rev. 30

Table of Contents (Continued) 1.13 References for Appendix 3A.1 .............................................................. 3A.1-11 2 EQUIPMENT AND COMPONENTS..................................................... 3A.2-1 2.1 ASAAS (Asymmetric Stress Analysis of Axisymmetric Solids) ............ 3A.2-2 2.2 Limita 25 - 2D Nonlinear Transient Dynamic Analysis.......................... 3A.2-3 2.3 MISSILE .................................................................................................. 3A.2-5 2.4 SLOSH..................................................................................................... 3A.2-6 2.5 LION (ME-112) ....................................................................................... 3A.2-7 2.6 LIMITA 3 ................................................................................................ 3A.2-7 2.7 TAC2D (A General Purpose, Two-Dimensional Heat Transfer Computer Code) 3A.2-9 2.8 SHELL 1 .................................................................................................. 3A.2-9 2.9 Vessel Penetration Analysis..................................................................... 3A.2-9 2.10 DINASAW (Dynamic Inelastic Nonlinear Analysis by Stone and Webster) .....

3A.2-10 2.11 LIMITA 2 .............................................................................................. 3A.2-11 2.12 STARDYNE .......................................................................................... 3A.2-14 2.13 ASYMPR (ME-171) .............................................................................. 3A.2-15 2.14 LIDOP (ME-184)................................................................................... 3A.2-16 2.15 Dynamic Load Factors (DLF ME-185) ................................................. 3A.2-17 2.16 References for Appendix 3A.2 .............................................................. 3A.2-19 3 PIPING SYSTEMS.................................................................................. 3A.3-1 3.1 NUPIPE II................................................................................................ 3A.3-1 3.2 PITRUST ................................................................................................. 3A.3-3 3.3 PILUG...................................................................................................... 3A.3-3 3.4 SAVAL .................................................................................................... 3A.3-4 3.5 STEHAM ................................................................................................. 3A.3-5 3.6 WATHAM ............................................................................................... 3A.3-5 3.7 PSPECTRA.............................................................................................. 3A.3-6 3.8 STRUDL-SW........................................................................................... 3A.3-7 3.9 STRUDL-II (ICES).................................................................................. 3A.3-9 3.10 APEN ..................................................................................................... 3A.3-10 3.11 PITRIFE................................................................................................. 3A.3-11 3.12 PITAB .................................................................................................... 3A.3-12 3.13 CHPLOT ................................................................................................ 3A.3-12 3.14 LOADCOMB......................................................................................... 3A.3-13 3.15 BEARST ................................................................................................ 3A.3-14 3.16 ANCCOMB ........................................................................................... 3A.3-14 3.17 BENDCORD ......................................................................................... 3A.3-16 3.18 NUPIPE-SWPC ..................................................................................... 3A.3-16 3.19 PC-PREPS ............................................................................................. 3A.3-17 3.20 PILUG-PC ............................................................................................. 3A.3-17 3-xiii Rev. 30

Table of Contents (Continued) 3.21 References for Appendix 3A.3 .............................................................. 3A.3-18 PENDIX 3B -ENVIRONMENTAL DESIGN CONDITIONS .................................. 3B.B-0 3-xiv Rev. 30

List of Tables mber Title 1 List Of QA Category I and Seismic Category I Structures, Systems, and Components 2 Safety Classes 2 and 3 Instrument Tubing Requirements 3 Instrument Tubing Examination and Testing Program 4 Comparison of Proposed Tubing Examination and Testing with ASME III Requirements 1 Structural Panels Subject to Tornado Pressure Drop 1 Safety Related Structures, Systems, and Components Outside Containment Required for Safe Reactor Shutdown*

2 Safety Related Structures, Systems, and Components Inside Containment Required for Safe Reactor Shutdown*

3 Summary of Control Rod Drive Mechanism Missile Analysis 4 Valve-Missile Characteristics*

5 Piping Temperature Element Assembly - Missile Characteristics 6 Characteristics of Other Missiles Postulated within Reactor Containment 7 Turbine-Target Distances and Impact Areas 8 Deleted by FSARCR 02-MP3-13 9 Deleted by FSARCR 02-MP3-13 10 Deleted by FSARCR 02-MP3-13 11 Deleted by FSARCR 02-MP3-13 12 Deleted by FSARCR 02-MP3-13 13 Postulated Tornado-Generated External Missiles 14 Barrier Deflection and Ductility Ratios for Tornado-Borne Missiles Plus 360 MPH Tornado Wind

-15 Barrier Deflection and Ductility Ratio for a Beam-Column Plus 360 MPH Tornado Wind 1 High-Energy Systems Remote from Essential Structures, Systems, and Components 3-xv Rev. 30

mber Title 2 High-energy Systems in Proximity to Essential Structures, Systems, and Components 3 Moderate-Energy Systems in Remote from Essential Structures, Systems, and Components 4 Moderate-Energy Systems Proximate to Essential Structures, Systems, and Components 5 Essential Structures, Systems, and Components Required for Safe Reactor Shutdown 6 Postulated Breaks Main Steam System 7 Pipe Whip Effects Main Steam System 8 Jet Impingement Effects - Main Steam System 9 Postulated Breaks Main Feedwater System

-10 Pipe Whip Effects - Main Feedwater System 11 Jet Impingement Effects - Main Steam System 12 Postulated Breaks - Reactor Coolant System - Loop Stop Valve Bypass Piping, Excess Letdown Piping, Loop Fill Piping, Loop Drain Piping. Letdown Line, and Normal Charging

-13 Pipe Whip Effects - Reactor Coolant System - Loop Stop Valve Bypass Piping, Excess Letdown Piping, Loop Fill Piping, Loop Drain Piping, Letdown and Normal Charging 14 JET Impingement Effects - Reactor Coolant System - Loop Stop Valve Bypass Piping, Excess Letdown Piping, Loop Drain Piping, Loop Fill Piping, Letdown Line, and Normal Charging 15 Postulated Breaks - Reactor Coolant System - Pressurizer Cubicle Piping

-16 Pipe Whip Effects - Reactor Coolant System - Pressurizer Cubicle Piping 17 Jet Impingement Effects - Reactor Coolant System - Pressurizer Cubicle Piping 18 Postulated Pipe Breaks - Reactor Coolant System - Low Pressure Safety Injection, High Pressure Safety Injection, and Residual Heat Removal Piping 19 Pipe Whip Effects - Reactor Coolant System - Low Pressure Safety Injection, High Pressure Safety Injection, and Residual Heat Removal Piping 20 Jet Impingement Effects - Low Pressure Safety Injection, High Pressure Safety Injection, and Residual Heat Removal Piping 3-xvi Rev. 30

mber Title 21 Postulated Breaks - Chemical Volume Control System - Normal Charging 22 Pipe Whip Effects - Chemical Volume Control System - Normal Charging 23 Jet Impingement Effects - Chemical Volume Control System - Normal Charging 24 Postulated Breaks - Chemical Volume Control System - Letdown Line

-25 Pipe Whip Effects - Chemical Volume Control System - Letdown Line

-26 Jet Impingement Effects - Chemical and Volume Control System - Letdown Line Piping 27 Postulated Breaks - Seal Water Injection System 28 Pipe Whip Effects - Chemical Volume Control System - Seal Water Injection Line 29 Jet Impingement Effects - Chemical Volume Control System - Seal Water Injection Line 30 Postulated Breaks - Steam Generator Blowdown System 31 Pipe Whip Effects - Steam Generator Blowdown System

-32 Jet Impingement Effects - Steam Generator Blowdown System 33 Energy Absorbing Capacity of a 4 Inch Schedule 80 Pipe 34 Postulated Breaks Auxiliary Feedwater System

-35 Pipe Whip Effects Auxiliary Feedwater System 36 Jet Impingement Effects Auxiliary Feedwater Systems 1 Seismic Instrumentation B-1 Damping Factors B-2 Methods of Seismic Analysis Used for Seismic Category I Structures B-3 Containment and Internal Structures Significant Modal Frequencies and Participation Factors Uncracked Model B-4 Containment and Internal Structures Significant Modal Frequencies and Participation Factors Cracked Model B-5 Containment and Internal Structures Significant Mode Shapes (Normalized Eigenvectors) Uncracked Model B-6 Containment and Internal Structures Significant Mode Shapes (Normalized Eigenvectors) Cracked Model 3-xvii Rev. 30

mber Title B-7 Containment and Internal Structures SRSS Accelerations and Displacements for Safe Shutdown Earthquake Uncracked Model, Response Spectrum Analysis B-8 Containment and Internal Structures SRSS Accelerations and Displacements for Safe Shutdown Earthquake Cracked Model, Response Spectrum Analysis B-9 Containment and Internal Structures Degrees of Freedom B-10 Containment and Internal Structure Enveloped Accelerations and Displacements Horizontal-motion

  • B-11 Containment and Internal Structure Enveloped Accelerations and Displacements Vertical-Motion
  • B-12 Main Steam Valve Building Significant Modal Frequencies and Participation Factors B-13 Main Steam Valve Building Significant Mode Shapes (Eigenvectors)

B-14 Main Steam Valve Building CSM* Accelerations for Safe Shutdown Earthquake from Response Spectrum Analysis B-15 Main Steam Valve Building CSM* Displacements for Safe Shutdown Earthquake from Response Spectrum Analysis B-16 Main Steam Valve Building ABS

  • of Safe Shutdown Earthquake Responses B-17 Main Steam Valve Building Degrees of Freedom B-18 Emergency Generator Enclosure Accelerations and Displacements Horizontal-Motion*

B-19 Emergency Generator Enclosure Accelerations and Displacements Vertical -

Motion B-20 Comparison of Response Spectra and Time History Analysis Results Containment and Internal Structures (Uncracked Properties)

B-21 Comparison of Response Spectra and Time History Analysis Results Containment and Internal Structures (Cracked Properties)

B-22 Comparison of Response Spectra and Time History Analysis Results Main Steam Valve Building SSE Accelerations*

B-23 1 G Flat Response B-24 Modal Density, n*

B-25 Amplified Response Dynamic Factor Study B-26 Piping System Seismic Design and Analysis Criteria 3-xviii Rev. 30

mber Title N-1 Damping Values Used for Seismic Analysis for Westinghouse Supplied Equipment 1 Loading Conditions - Liner Plate and Access Openings 2 Loading Conditions, Penetrations 3 Loads and Loading Combinations 4 Load Combinations for ASME III Class 2 Penetrations except Quench, Recirculation, and Safety Injection Piping 5 Load Combinations for ASME III Class 2 Penetrations for the Quench Spray, Recirculation Spray, and Safety Injection Systems 6 Load Combinations for ASME III Class MC Sleeved Penetrations 7 Nomenclature for Tables 3.8-4 through 3.8-6 B-1 List of Input Documents Describing Design Transient for Fatigue Analysis of RCP and Associated Class 1 Piping B-2 List of MPS-3 Documents Describing Components Requiring Inelastic Analysis B-3 Preoperational Tests (1)

B-4 Omitted B-5 Stress Limits for ASME Section III Class 1 (NB) Seismic Category I Components (Elastic Analysis)

B-6 Comparison of Class 1 Requirements Regulatory Guide 1.48 vs. Tables 3.9B-5 and 3.9B-10 B-7 Stress Limits for ASME Section III Class 2 and 3 Components (Elastic Analysis)

B-8 Comparison of Classes 2 and 3 Requirements Regulatory Guide 1.48 vs Table 3.9B-7 B-9 Stress Limits for ASME Section III Class 1, 2, and 3 Component Supports

  • B-9A Load Combinations for ASME Section III Class 1, 2, and 3 Component Supports B-10 ASME III Class 1 Stress and Fatigue Analysis Requirements per NB3650 B-11 ASME III Class 2 and 3* Stress Analysis Requirements per NC3650 and ND3650 B-12 Loadings Applicable to Piping Systems B-13 Active Pumps and Valves B-14 Omitted B-15 Postulated Primary Loop Breaks 3-xix Rev. 30

mber Title B-16 Steam Generator and Reactor Coolant Pump Support Snubber Embedment Loads(1)

(KIPS)

B-17 Steam Generator and Reactor Coolant Pump Support Column Embedment Loads(1)(KIPS)

B-18 Factors of Safety for Primary Members of Steam Generators and Reactor Coolant Pump Supports B-19 Minimum Design Margins for Pressurizer Support B-20 Design Margins for the Primary Members of the Pressurizer Safety Valve Support B-21 Design Margins for Primary Members of Reactor Pressure Vessel Support B-22 Design Margins for Primary Loop Bumper Support Structure N-1 Summary of Reactor Coolant System Design Transients N-2 Loading Combinations for Reactor Coolant System Components N-3 Allowable Stresses for Reactor Coolant System Components N-4 Design Loading Combinations for ASME Code Class 2 and 3 Components and Supports N-5 Stress Criteria for Safety Code Class 2 (1) and Class 3 Tanks N-6 Stress Criteria for ASME Code Class 2 Tanks(1)

N-7 Stress Criteria for ASME Code Class 2 and Class 3 Inactive Pumps N-8 Stress Criteria for Safety Related ASME Code Class 2 and Class 3 Active and Inactive Valves N-9 Stress Criteria for ASME Code Class 2 and 3 Piping N-10 Design Criteria for Active Pumps N-11 Active Pumps N-12 Active Valves N-13 Maximum Deflections Allowed for Reactor Internal Support Structures N-1 Seismic Category I Instrumentation and Electrical Equipment in Westinghouse NSSS Scope of Supply N-1 Safety-Related Equipment in Westinghouse NSSS Scope of Supply (Historical, not subject to future updating)

.1.2-1 Thin-Wall Cylinder, Pertinent Parameters 3-xx Rev. 30

mber Title

.1.2-2 Exact and Computer Stresses for Thin-wall Cylinder

.1.4-1 Infinitely Long Solid Cylinder, Pertinent Parameters

.1.5-1 Input Thermal Parameter Functions for TAC2D Sample Problem

.1.6-1 Peak Acceleration and Displacement

.1.6-2 Horizontal Amplified Response Spectra

.1.10-1 Pertinent Parameters of a Cylindrical Shell

.1.10-2 Summary of Results

.1.11-1 Test Problem Data

.1.11-2 A Comparison of Hand and Computer Program Results

.2.1-1 Infinitely Long Solid Cylinder, Pertinent Parameters

.2.4-1 Comparison of Results of SLOSH vs AEC Analysis*

.2.6-1 Comparison of Experimental Data with Analytical Data Using LIMITA 3 3-1 Comparison of Support Reaction Due to Thermal, Anchor Movement, and External Force Loading 3-2 Comparison of Deflections and Rotations Due to Thermal, Anchor Movement, and External Force Loading 3-3 Comparison of Stress Due To Thermal, Anchor Movement, and External Force Loading 3-4 Comparison of Internal Forces Due To Deadweight Analysis 3-5 Comparison of Deflections and Rotation Due to Deadweight Analysis 3-6 Comparison of Stresses Due to Deadweight Analysis 3-7 Comparison of Natural Frequencies 3-8 Comparison of Natural frequencies 3-9 Comparison of Class 1 Pipe Stress Analysis 3-10 Individual Pair Usage Factor for Point No. 30 3-11 Comparison of PITRUST With Franklin Institute Program CYLNOZ and Hand Calculation 3-12 Comparison of PITRUST With Reference 8 Results 3-13 Comparison of PILUG Computer Program Output with Hand Calculations 3-xxi Rev. 30

mber Title 3-14 Summary of Comparison of SAVAL Computer Output with Hand Calculation As-Designed Condition 3-15 Summary of Comparison of SAVAL Computer Output with Hand Calculation Reinforced Condition (1 1/4 Pad) 3-16 Nodal Force Comparison 3-17 Input Data for WATHAM 3-18 Comparison of NODAL Force Calculation at Time = 2.34 Seconds 3-19 Comparison of PITRIFE Computer Program Output With STRUDL-II Output 3-20 Comparison of PITRIFE Computer Program Output With Hand Calculations 3-21 BENDCORD Program--Verification Problem 3-22 Pipe 1, Segment 1 Force Versus Time for Run Number S2807011 3-xxii Rev. 30

List of Figures mber Title 1 Line Designation, Safety Class Boundaries 1 Dynamic Load Factor Curve for Tornado Pressure Drop 1 Circulating and Service Water Pumphouse 1 Turbine Placement and Orientation for Three-Unit Site 2 Turbine Placement and Orientation Profile

-3 Deleted

-4 Deleted

-5 Deleted 6 Missile Barrier Beam Column 1 Fluid Systems Subjected to Break and Crack Analysis 2 Relative Locations of Safety Related Equipment, Elevation 3 feet 8 inches 3 Relative Locations of Safety Related Equipment, Elevation 24 feet 6 inches 4 Relative Locations of Safety Related Equipment, Elevation 51 feet 4 inches 5 Relative Locations of Safety Related Equipment, Section 1-1 6 Relative Locations of Safety Related Equipment, Section 2-2 7 Relative Locations of Safety Related Equipment, Sections 3-3, 4-4

-8 Main Steam System Notes to Figure 3.6-8: Data for Pipe Rupture Restraints - Main Steam System 9 Main Steam System, (Continued) 10 Feedwater System Postulated Pipe Rupture Locations 11 Feedwater System Pipe Rupture Restraints Notes to Figure 3.6-11: Data for Pipe Rupture Restraints - Main Feedwater System

-12 Reactor Coolant System - Primary Coolant Piping Notes to Figure 3.6-12: Data for Pipe Rupture Restraints - Reactor Coolant System

- Primary Coolant Piping

-13 Reactor Coolant System - Low Pressure Safety Injection and Residual Heat Removal Piping Notes to Figure 3.6-13: Data for Pipe Rupture Restraints Reactor Coolant System, Low Pressure Safety Injection, High Pressure Safety Injection, and Residual Heat 3-xxiii Rev. 30

NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.

ure Title Removal Piping

-14 Reactor Coolant System - Pressurizer Cubicle Piping Notes to Figure 3.6-14: Data for Pipe Rupture Restraints Reactor Coolant System Pressurizer Cubicle Piping

-15 Chemical and Volume Control System - Charging Line Piping Notes to Figure 3.6-15: Data for Pipe Rupture Restraints Chemical Volume Control System - Normal Charging 16 Chemical and Volume Control System - Letdown Line Piping Notes to Figure 3.6-16: Data for Pipe Rupture Restraints Chemical Volume Control System - Letdown Line

-17 Chemical and Volume Control System - Seal Water Injection and Piping 18 Steam Generator Blowdown System Notes to Figure 3.6-18: Data for Pipe Rupture Restraints Steam Generator Blowdown System 19 Pipe Rupture Analysis Flow Chart 20 Variation of Steady State Steam Blowdown vs Friction 21 Pipe Restraint Intermediate Structure System, Mathematical Model

-22 Laminated Strap Restraint for Small Lines 23 Feedwater System Short Loop Strap near the Steam Generator, NET Shear Force at Node 10 24 Feedwater System Short Loop Strap near the Steam Generator, NET Shear Force at Node 14 25 Feedwater System Short Loop Strap near the Steam Generator, NET Moment Force at Node 14 26 Force Acting on the Strap Node 8 27 Feedwater System Short Loop Strap near the Steam Generator, Energy 28 Energy Balance Analysis Model 29 Pipe Crush Bumper 30 Pipe Crush Bumper 31 Omni Directional Pipe Rupture Restraint 32 Auxiliary Feedwater System 33 Reactor Coolant System Small Bore Piping in Steam Generator Cubicles 3-xxiv Rev. 30

NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.

ure Title B-1 Horizontal Design Response Spectra 0.17g SSE B-2 Horizontal Design Response Spectra 0.09g 1/2 OBE B-3 Horizontal Time History Response Spectra 0.5 Percent Damping Value B-4 Horizontal Time History Response Spectra 1.0 Percent Damping Value B-5 Horizontal Time History Response Spectra 2.0 Percent Damping Value B-6 Horizontal Time History Response Spectra 5.0 Percent Damping Value B-7 Horizontal Time History Response Spectra 7.0 Percent Damping Value B-8 Horizontal Time History Response Spectra 10 Percent Damping Value B-9 Dynamic Model of the Containment Structure B-10 Dynamic Model of the Main Steam Valve Building B-11 Dynamic Model of the Emergency Generator Enclosure N-S View B-12 Dynamic Model of the Emergency Generator Enclosure E-W View B-13 Reactor Containment Mat Elevation -37 feet 3 inches Amplified Response Spectra N-S Excitation SSE B-14 Reactor Containment Mat Elevation -37 feet 3 inches Amplified Response Spectra E-W Excitation SSE B-15 Reactor Containment Mat Elevation -37 feet 3 inches Amplified Response Spectra Vertical Excitation SSE B-16 Reactor Containment Steam Generators Support Slab Elevation 3 feet 0 inches Amplified Response Spectra N-S Excitation SSE B-17 Reactor Containment Steam Generators Support Slab Elevation 3 feet 0 inches Amplified Response Spectra E-W Excitation SSE B-18 Reactor Containment Steam Generators Support Slab Elevation 3 feet 0 inches Amplified Response Spectra Vertical Excitation SSE B-19 Reactor Containment Top of Primary Shield Wall Elevation 24 feet 6 inches Amplified Response Spectra N-S Excitation SSE B-20 Reactor Containment Top of Primary Shield Wall El 24 feet 6 inches Amplified Response Spectra E-W Excitation SSE B-21 1 Reactor Containment Top of Primary Shield Wall Elevation 24 feet 6 inches Amplified Response Spectra Vertical Excitation SSE B-22 Reactor Containment Operation Floor Elevation 50 feet 10 inches Amplified 3-xxv Rev. 30

NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.

ure Title Response Spectra N-S Excitation SSE B-23 Reactor Containment Operating Floor Elevation 50 feet 10 inches Amplified Response Spectra E-W Excitation SSE B-24 Reactor Containment Operating Floor Elevation 50 feet 10 inches Amplified Response Spectra Vertical Excitation B-25 Reactor Containment Top of Crane Wall Elevation 109 feet 1 inch Amplified Response Spectra N-S Excitation SSE B-26 Reactor Containment Top of Crane Wall Elevation 109 feet 1 inch Amplified Response Spectra E-W Excitation SSE B-27 Reactor Containment Top of Crane Wall Elevation 109 feet 1 inch Amplified Response Spectra Vertical Excitation SSE B-28 Reactor Containment Springline Elevation 104 feet 0 inches Amplified Response Spectra N-S Excitation SSE B-29 Reactor Containment Springline Elevation 104 feet 0 inches Amplified Response Spectra E-W Excitation SSE B-30 Reactor Containment Springline Elevation 104 feet 0 inches Amplified Response Spectra Vertical Excitation B-31 Reactor Containment Dome Apex Elevation 163 feet 4 inches Amplified Response Spectra N-S Excitation SSE B-32 Reactor Containment Dome Apex Elevation 163 feet 4 inches Amplified Response Spectra E-W Excitation SSE B-33 Reactor Containment Dome Apex Elevation 163 feet 4 inches Amplified Response Spectra Vertical Excitation SSE B-34 Main Steam Valve Building Mat Elevation 9 feet -0 inches Amplified Response Spectra N-S Excitation SSE B-35 Main Steam Valve Building Mat Elevation 9 feet 0 inches Amplified Response Spectra E-W Excitation SSE B-36 Main Steam Valve Building Mat Elevation 9 feet 0 inches Amplified Response Spectra Vertical Excitation SSE B-37 Main Steam Valve Building Floor Slab Elevation 41 feet 0 inches Amplified Response Spectra N-S Excitation SSE B-38 Main Stem Valve Building Floor Slab Elevation 41 feet 0 inches Amplified Response Spectra E-W Excitation SSE 3-xxvi Rev. 30

NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.

ure Title B-39 Main Steam Valve Building Floor Slab Elevation 41 feet 0 inches Amplified Response Spectra Vertical Excitation SSE B-40 Main Steam Valve Building Roof Slab Elevation 85 feet 4 inches Amplified Response Spectra N-S Excitation SSE B-41 Main Steam Valve Building Roof Slab Elevation 85 feet 4 inches Amplified Response Spectra E-W Excitation SSE B-42 Main Steam Valve Building Roof Slab Elevation 85 feet 4 inches Amplified Response Spectra Vertical Excitation SSE B-43 Emergency Generator Enclosure Transfer Function of Base of Structure Elevation 9 ft-0 in for Horizontal Acceleration N-S Excitation SSE B-44 Emergency Generator Enclosure Transfer Function at Floor Slab Elevation 51 feet 0 inches for Horizontal Acceleration N-S Excitation SSE B-45 Emergency Generator Enclosure Transfer Function at Roof Slab Elevation 66 feet 0 inches for Horizontal Acceleration N-S Excitation SSE B-46 Emergency Generator Enclosure Transfer Function at Base of Structure Elevation 9 feet 0 inches for Vertical Acceleration Vertical Excitation SSE B-47 Emergency Generator Enclosure Transfer Function at Floor Slab Elevation 51 feet 0 inches for Vertical Acceleration Vertical Excitation SSE B-48 Emergency Generator Enclosure Transfer Function at Roof Slab Elevation 66 feet 0 inches for Vertical Acceleration Vertical Excitation SSE B-49 Emergency Generator Enclosure Transfer Function at Base of Structure Elevation 9 feet 0 inches for Horizontal Acceleration E-W Excitation SSE B-50 Emergency Generator Enclosure Transfer Function at Floor Slab Elevation 51 feet 0 inches for Horizontal Acceleration E-W Excitation SSE B-51 Emergency Generator Enclosure Transfer Function at Roof Slab Elevation 66 feet 0 inches for Horizontal Acceleration E-W Excitation SSE B-52 Emergency Generator Enclosure Floor Slab Elevation 24 feet 6 inches Amplified Response Spectra N-S Excitation SSE B-53 Emergency Generator Enclosure Floor Slab Elevation 24 feet 6 inches Amplified Response Spectra E-W Excitation SSE B-54 Emergency Generator Enclosure Floor Slab Elevation 24 feet 6 inches Amplified Response Spectra Vertical Excitation SSE B-55 Emergency Generator Enclosure Floor Slab Elevation 51 feet 0 inches Amplified 3-xxvii Rev. 30

NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.

ure Title Response Spectra N-S Excitation SSE B-56 Emergency Generator Enclosure Floor Slab Elevation 51 ft-0 in Amplified Response Spectra E-W Excitation SSE B-57 Emergency Generator Enclosure Floor Slab Elevation 51 feet 0 inches Amplified Response Spectra Vertical Excitation SSE B-58 Emergency Generator Enclosure Roof Slab Elevation 66 feet 0 inches Amplified Response Spectra N-S Excitation SSE B-59 Emergency Generator Enclosure Roof Slab Elevation 66 feet 0 inches Amplified Response Spectra E-W Excitation SSE B-60 Emergency Generator Enclosure Roof Slab Elevation 66 feet 0 inches Amplified Response Spectra Vertical Excitation SSE B-61 Emergency Generator Enclosure Diesel Generator Mat Elevation 24 feet 0 inches Amplified Response Spectra N-S Excitation SSE B-62 Emergency Generator Enclosure Diesel Generator Mat Elevation 24 feet 6 inches Amplified Response Spectra Vertical Excitation SSE B-63 Emergency Generator Enclosure Horizontal Acceleration Time History at Bedrock N-S Excitation SSE B-64 Emergency Generator Enclosure Horizontal Acceleration Time History at Base of Structure N-S Excitation SSE B-65 Typical Mathematical Model of a Piping System B-66 Representation of Family of Peak Responses Curves within Broadened Resonant Peak B-67 Hypothetical vs Actual Response of Multiple Modes within Broadened Response Peak B-68 Justification of Static Load Factor B-69 Typical Amplified Response Spectra B-70 Model Beams B-71 Damping Value for Seismic Analysis of Piping N-1 Multi-Degree of Freedom System

-1 Containment Structure Foundation Detail 2 Typical Detail of Dome Cylinder Junction 3 Knuckle Plate 3-xxviii Rev. 30

NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.

ure Title

-4 Mat Reinforcement with Radial Shear Bar Assembly

-5 Reinforcement Around Main Steam Penetrations

-6 Reinforcing Details Sections through Ring Beam of Personnel Access Lock 7 Reinforcing Details Personnel Access Lock Opening 8 Typical Reinforcing Details Sections through Ring Beam of Equipment Access Hatch 9 Reinforcing Details Equipment Access Hatch Opening

-10 Base Plate Details Containment Enclosure Structure 11 Ring at Apex of Containment Structure Dome 12 Elevation Dome Maintenance Truss

-13 Liner Knuckle Plate 14 Basic Liner Dimensions 15 Typical Piping Penetrations

-16 Typical Sleeved Piping Penetration 17 Multiple Pipe Penetration Assembly 18 Unsleeved Piping Penetration 19 Electrical Penetration 20 Fuel Transfer Tube 21 Personnel Hatch 22 Equipment Hatch

-23 Containment Structure Foundation Detail with Ring Girder 24 Detail of Ring Girder at ESF Building

-25 Containment Structure Mat Moment and Shear Diagrams

-26 Containment Structure Mat Moment and Shear Diagrams

-27 Containment Structure Mat Moment and Shear Diagrams

-28 Containment Structure Mat Moment and Shear Diagrams

-29 Containment Structure Mat Moment and Shear Diagrams

-30 Containment Structure Mat Moment and Shear Diagrams 3-xxix Rev. 30

NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.

ure Title

-31 Containment Structure Mat Moment and Shear Diagrams 32 Load Plot Nomenclature

-33 Containment Structure Wall and Dome Force Resultants

-34 Containment Structure Wall and Dome Force Resultants

-35 Containment Structure Wall and Dome Force Resultants

-36 Containment Structure Wall and Dome Force Resultants

-37 Containment Structure Wall and Dome Force Resultants

-38 Containment Structure Wall and Dome Force Resultants

-39 Containment Structure Wall and Dome Force Resultants

-40 Containment Structure Wall and Dome Force Resultants

-41 Containment Structure Wall and Dome Force Resultants

-42 Containment Structure Wall and Dome Force Resultants

-43 Containment Structure Wall and Dome Force Resultants

-44 Containment Structure Wall and Dome Force Resultants

-45 Containment Structure Wall and Dome Force Resultants

-46 Containment Structure Wall and Dome Force Resultants

-47 Containment Structure Wall and Dome Force Resultants

-48 Containment Structure Wall and Dome Force Resultants

-49 Containment Structure Wall and Dome Force Resultants

-50 Containment Structure Wall and Dome Force Resultants

-51 Containment Structure Wall and Dome Force Resultants

-52 Containment Structure Wall and Dome Force Resultants

-53 Containment Structure Wall and Dome Force Resultants

-54 Containment Structure Wall and Dome Force Resultants 55 Containment Structure Wall and Dome 56 Details of Wall Diagonals at Containment Mat 57 Details of Waterproof Membrane 58 Displacement Profile for Containment Shell Under Pressure Test (Predicted) 3-xxx Rev. 30

NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.

ure Title 59 Containment Structure, Plan Views 60 Containment Structure, Elevation Views

-61 Containment Enclosure Building 62 Auxiliary Building 63 Fuel Building

-64 Control Room Area

-65 Cable Tunnel 66 Emergency Generator Enclosure Notes to Figure 3.8-66: Data for Pipe Rupture Restraints Steam Generator Blowdown System 67 Engineered Safety Features Building

-68 Main Steam Valve Building 69 Circulating and Service Water Pumphouse 70 Recombiner Building 71 3.8-71 Circulating Discharge Structure and Tunnel 72 Service Building

-73 Turbine Building 74 Solid and Liquid Waste Disposal Building 75 Warehouse No. 5 76 Auxiliary Boiler Enclosure 77 Arrangement of Horizontal Shear Keys

-78 Typical Shear Key Detail 79 Fuel Pool Temperature Transients 80 Fuel Pool--Fuel Building Temperature Conditions 81 Fuel Pool Temperature Transients--for Higher Enrichment Fuel

-82 Fuel Pool Temperature Transients--Full Core Offload

-83 Millstone Stack Elevations and Sections

-84 Millstone Stack - Mathematical Model B-1 Reactor Coolant PPG 3-xxxi Rev. 30

NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.

ure Title B-2 Dynamic Model - Typical RCS Loop B-3 Dynamic Model - Center Section B-4 X2 View - Rotated Axes B-5 Snubber Orientation N-1 Reactor Pressure Vessel and Internals System Model N-2 Deleted by PKG FSC 07-MP3-039 N-3 RPV Support Model for Westinghouse Internals Response Analysis N-4 Control Rod Drive Mechanism N-5 Control Rod Drive Mechanism Schematic N-6 Nominal Latch Clearance at Minimum and Maximum Clearance N-7 Control Rod Drive Mechanism Latch Clearance Thermal Effect N-8 Lower Core Support Assembly (Core Barrel Assembly)

N-9 Upper Core Support Assembly N-10 Plan View of Upper Core Support Assembly N-11 Neutron Shield Panel Design N-12 Neutron Shield Panel Design Details 1.2-1 One Hundred-Element Idealization of Thin-Wall Cylinder 1.4-1 Element Plot 1.4-2 Harmonic Axisymmetric Plain Strain 1.5-1 TAC2D Sample Problem Thermal Model 1.5-2 Transient Temperatures in a Right Circular Cylinder-Comparison of TAC2D Results with Series Solution 1.6-1 Structural Model 1.7-1 Finite Element Grid for PLAXLY-FLUSH Comparison 1.7-2 Horizontal Amplified Response Spectra at Top Mass in Beam Structure 2.3-1 Target for Low Trajectory Missile 2.11-1 Cantilever Pipe Used in MIT Experiment 3-1 Mathematical Model for Flexibility Analysis Verification 3-xxxii Rev. 30

NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.

ure Title 3-2 NUPIPE Program Force Time-History Verification 3-3 Mathematical Model for Class 1 Stress Verification 3-4 Test Problem - SAVAL 3-5 Sudden Discharge of a Gas from a Pipeline through a Nozzle 3-6 Sudden Discharge of a Gas from a Pipeline through a Nozzle 3-7 Comparison of Pressure Response at the Closed End 3-8 Comparison of Pressure Response at the Open End 3-9 Comparison of Pressure Responses by STEHAM and Experiment 3-10 Hydraulic Network for Verification Problem 3-11 Hydraulic Network for WATHAM Verification 3-12 Head versus Time Plot for Junction J 3-13 Head versus Time Plot at Valve 3-14 Design Input to a Problem 3-15 Pipe 1 Segment 1 Force-Time History 3-xxxiii Rev. 30

CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA s section evaluates the design bases of Millstone 3 as measured against the NRC General ign Criteria for Nuclear Power Plants, Appendix A to 10 CFR 50, as amended through ober 27, 1978. The General Design Criteria (GDC) consist of the single failure definition and ndividual criteria, and are intended to establish the basic requirements for the principal design eria of nuclear power plants.

General Design Criteria were not written specifically for the pressurized water reactor; rather, were intended to guide the design of all water-cooled nuclear power plants. As a result, the eria are generic in nature and subject to interpretation. For this reason, there are some cases re conformance to a particular criterion is not directly assessable. In these cases, the formance of plant design to the interpretation of the criteria is discussed. For the single failure nition and each of the 55 criteria, a specific assessment of the plant design is made and rences are included, where necessary, to identify where detailed design information pertinent ach criterion is treated in this FSAR. For the purpose of this report, the terms important to ty, safety related, and safety systems are synonymous.

ed on the contents herein, the Applicants conclude that Millstone 3 satisfies and complies with GDC, with the exceptions as noted.

1 CONFORMANCE WITH SINGLE FAILURE CRITERION 1.1 Single Failure Criterion endix A to 10 CFR 50, as amended through October 27, 1978, defines single failure criterion ollows:

A single failure means an occurrence which results in the loss of capability of a component to perform its intended safety functions. Multiple failures resulting from a single occurrence are considered to be a single failure. Fluid and electric systems are considered to be designed against an assumed single failure if neither (1) a single failure of any active component (assuming passive components function properly) nor (2) a single failure of a passive component (assuming active components function properly) results in a loss of the capability of the system to perform its safety functions.

1.2 Definitions of Terms Used in Single Failure Criterion Active Component Failure active component is one in which mechanical movement must occur to complete the ponents intended function. An active component failure is failure of the component to plete its intended function upon demand.

3.1-1 Rev. 30

h spurious action.

mples of active component failures include the failure of a powered, manual, or check valve ept ECCS check valves within the NSSS vendor original scope of design) to move to its ect position; the failure of an electrical breaker or relay to respond; and the failure of a pump, or emergency diesel generator to start.

Passive Component Failure assive component is one in which mechanical movement need not occur for the component to orm its intended function. A passive component failure is the structural failure of the ponent or the blockage of a process flowpath by the passive component so that the component s not perform its intended function. For fluid pressure boundary components, a passive ponent failure is a break of, or crack in, the pressure boundary.

sive components include piping, cables, and valve bodies.

Short Term short term is defined as the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the start of an incident.

Long Term long term is defined as the period following the short term (i.e., greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />), during ch time the system safety function is still required as defined in the individual equipment lification documentation described in Section 3.11.

Related Service Systems ated service systems are those systems which provide the services necessary to a fluid system nable that system to complete its intended safety function.

mples of related service systems for the emergency core cooling system (ECCS) include the ty injection pumps cooling system, charging pumps cooling system, service water system, tric power supply system, protection systems, and the ECCS equipment area ventilation ems.

1.3 Application of Single Failure Criterion incidents of moderate frequency (Chapter 15) that can result in automatic reactor or turbine when Millstone 3 is generating power for offsite transmission, Millstone 3 is designed to ntain capability for cold safe shutdown assuming a single failure.

3.1-2 Rev. 30

ch requires their function, without loss of safety function to Millstone 3. Examples of systems his category are: the ECCS, the containment depressurization systems, the supplementary leak ection and release systems, the auxiliary feedwater system, and the cooling water systems d in combination with these systems.

single failure considered is a random failure in addition to:

1. The initiating event for which the system is required,
2. Any failures which are direct consequences of the initiating event, and
3. Loss of offsite electric power if the initiating event results in a trip of either the turbine generator or reactor protection system.

fluid systems, the single failure is limited to an active component failure during the short

, and assuming no prior failure during the short term, the single failure is either an active or a sive component failure during the long term.

h energy piping ruptures and moderate energy leakage cracks are considered separately as le postulated initial events occurring during normal plant operation. A single active failure is med in systems used to mitigate the consequences of the piping failure and to shut down the tor. The single active failure is assumed to occur in addition to the postulated piping failure any direct consequences of the piping failure.

ere the initial postulated piping failure is assumed to occur in one of two or more redundant ns of a dual-purpose moderate-energy essential system (i.e., one required to operate during mal plant conditions as well as to shut down the reactor and mitigate the consequences of the ng failure), single failures of components in the other train or trains of that system are not med, provided that the system is designed to Seismic Category I standards, is powered from h offsite and onsite sources, and is constructed, operated, and inspected to QA Category I and he testing and inservice inspection standards appropriate for nuclear safety systems. Examples ystems that qualify as dual-purpose essential systems are the service water system, the reactor t component cooling system, and the residual heat removal systems.

ngle active failure is not assumed in a dual-purpose safe shutdown system only if:

the initial failure event occurs from a source external to the dual-purpose safe shutdown system; the event results in the loss of function of one of the two redundant trains of the shutdown system; the event does not result in a reactor or turbine trip; and 3.1-3 Rev. 30

er these circumstances, continued operation of the plant is controlled by plant Technical cifications.

ingle active failure is not assumed in a safety-related system if the initial failure event is tulated in one of two redundant trains of that system whose function is not required during mal plant operation or to achieve cold shutdown if the failure does not require an automatic ective action to mitigate the consequences of the failure. Under these circumstances, continue ration of the plant is controlled by plant Technical Specifications.

electrical and protection systems, the single failure can be either an active or passive failure ng the short or long term. This is in accordance with IEEE-308 and Regulatory Guide 1.53 ction 7.1.2).

the purpose of single failure criterion, the system subject to the criterion includes the safety em itself and its related service systems, such as reactor plant component cooling, service er, electric power, etc., required for the safety system to perform its safety function. Even ugh each system indicated above is designed to the single failure criterion, only one single ure and its consequences are assumed to take place in the aggregate of safety systems and ted service systems in the plant.

safety related systems are designed such that their failure does not cause safety related ems to lose their safety function.

he proper active function of a component is demonstrated despite any reasonable postulated dition, that component is considered especially qualified for service and exempt from active ponent failure. Examples of such component function exemptions include opening of code ty valves and ECCS swing check valves within the NSSS vendor original scope of design.

s exemption from single active failure consideration applies only to the ECCS system design osophy employed by the NSSS vendor. As active components (described in FSAR Section

, ECCS check valves are subject to stringent design criteria and their operational readiness is odically verified in accordance with applicable plant technical specifications and the MP3 rvice Testing Program.

he case of ECCS check valves, the differential pressures required to open the check valve in forward direction, and the reverse flow velocity required to close the check valve, are very ll in comparison to the differential pressures and reverse flow velocities which will be erated if the check valve does not change position immediately as required. Therefore, failure he check valve to open or close is likely to be momentary, and not of sufficient duration to act the safety function of the system. In the design of the ECCS, therefore, the failure of a ty-related check valve to open when required to pass flow in the forward direction, or the ure of a safety-related check valve to seat to prohibit flow in the reverse direction is not sidered to be a credible event.

3.1-4 Rev. 30

des. As an example, review of systems involving piping, heat exchangers, valves, flange ts, and system interface barriers results in a definition of a design leak rate for passive ponent failure evaluation based on maximum flow through a failed packing, a mechanical

, or a piping crack (having an opening in size equal to half the pipe diameter in length and half pipe wall thickness in width for non-ECCS systems). Passive component failure is considered system functional design adequacy and effects from flooding only.

an exception, for ECCS systems containing recirculated sump fluid during the long term post CA mode of operation, only limited passive failures, with leakage rates up to 50 gpm, are tulated to occur in piping components or as pump mechanical seal failures. In some areas of ESF building and pipe tunnel, these long term failures are precluded by conservative piping em design (refer to Section 3.6 for passive failure exclusion design criteria).

ans are provided to detect and isolate limited passive failures of up to 50 gpm, from ECCS ponents, within approximately 30 minutes in the ESF building and in approximately 60 utes in the Auxiliary building (refer to Section 6.3 for ECCS failure modes and effects lysis and a description of leakage detection and isolation design features).

roper passive functions of a component can be demonstrated despite any reasonable postulated ditions, that component is considered especially qualified for service and exempt from passive ponent failure. One example of such a case is not assuming a passive component failure in twork.

ieu of a review of all the components in a system, complete severance of a line is assumed as passive component failure.

sive component failure or leakage in excess of limits defined in the Technical Specifications apter 16) is not considered in the primary reactor containment vessel boundary and the tainment isolation systems.

servative piping design criteria have been applied in conjunction with dual isolation valves ted outside containment as an acceptable alternative to General Design Criterion 56 for ain containment penetrations described in Section 6.2. Passive failures are not assumed in the outside containment (between the containment and the outboard isolation valve), provided the stresses are low, augmented inservice inspection is performed, and protection from tulated missiles is assured.

design, the mass of fluid discharged through the crack or break prior to effective isolation is sidered. Leakage duration is conservatively determined consistent with leakage detection, tion, and isolation mechanisms. As a guide, a nominal 30-minute leakage duration is sidered conservative for manual operator action outside of the control room.

3.1-5 Rev. 30

us indications.

2 CRITERION CONFORMANCE 2.1 Quality Standards and Records (Criterion 1)

Criterion uctures, systems, and components important to safety shall be designed, fabricated, erected, tested to quality standards commensurate with the importance of the safety functions to be ormed. Where generally recognized codes and standards are used, they shall be identified and luated to determine their applicability, adequacy, and sufficiency and shall be supplemented or dified as necessary to assure a quality product in keeping with the required safety function. A lity assurance program shall be established and implemented in order to provide adequate rance that these structures, systems, and components will satisfactorily perform their ctions. Appropriate records of the design, fabrication, erection, and testing of structures, ems, and components important to safety shall be maintained by, or under the control of, the lear power unit licensee throughout the life of the unit.

Design Conformance ctures, systems, and components important to safety are designed, fabricated, erected, and ed to quality standards commensurate with the importance of the safety functions to be ormed.

lity standards applicable to safety related structures, systems, and components are generally tained in codes such as the ASME Boiler and Pressure Vessel Code. The applicability of these es is specifically identified throughout this report and is summarized in Section 3.2.5.

pter 17 provides direct reference to the Quality Assurance Program established to provide rance that safety related structures, systems, and components satisfactorily perform their nded safety functions. The procedures for generating and maintaining appropriate design, ication, erection, and testing records are contained within the referenced documents.

2.2 Design Bases for Protection against Natural Phenomena (Criterion 2)

Criterion uctures, systems, and components important to safety shall be designed to withstand the cts of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and hes without loss of capability to perform their safety functions. The design bases for these ctures, systems, and components shall reflect: (1) appropriate consideration of the most severe he natural phenomena that have been historically reported for the site and surrounding area, h sufficient margin for the limited accuracy, quantity, and period of time in which the historical have been accumulated, (2) appropriate combinations of the effects of normal and accident 3.1-6 Rev. 30

Design Conformance se features of plant facilities that are essential to the prevention of accidents that could affect public health and safety or to the mitigation of accident consequences are designed to:

1. Quality standards that reflect the importance of the function to be performed.

Approved design codes are used when appropriate to the nuclear application.

2. Performance standards that enable the facility to withstand, without loss of the capability to protect the public, the additional forces imposed by the most severe earthquake, flooding condition, wind, ice, or other natural phenomena for the site, and credible combinations of the effects of normal and accident conditions with the effects of the natural phenomena.

tures of the facility essential to accident prevention and mitigation of accident consequences, ch are designed to withstand the effects of natural phenomena, are:

1. The reactor coolant pressure boundary and containment barriers
2. The controls and emergency cooling systems whose functions are to maintain the integrity of these barriers
3. Systems which depressurize the containment following a loss of coolant accident (LOCA)
4. Power supply and essential services
5. Reactivity systems, monitoring systems, and fuel systems
6. The components used to store and cool spent reactor fuel piping, components, and supporting structures of the reactor and safety related systems are gned to withstand a specified seismic disturbance and credible combinations of effects of mal and accident conditions coincident with the effects of natural phenomena. Plant design eria specify that there is to be no loss of function of such equipment in the event of the safe tdown earthquake (SSE) ground acceleration acting in the horizontal and vertical directions ultaneously. The dynamic response of Seismic Category I structures to ground acceleration, ed on an envelope of characteristics of the site foundation soils and on the critical damping of foundation and structures, is included in the design analysis.

ign of structures for protection against natural phenomena is described in Section 3.8. Safety ted structures have sufficient capacity to accept a combination of normal operating loads, 3.1-7 Rev. 30

emergency onsite power sources are not subject to interruption due to earthquake, windstorm, ds, or to disturbances on the external power transmission system.

er cabling, motors, and other equipment required for operation of the engineered safety ures are suitably protected against the effects of the design basis accident (DBA) and from ere external weather conditions, as applicable.

t design criteria which ensure protection against natural phenomena are described in tion 3.2 (Classification of Structures, Systems, and Components), Section 3.3 (Wind and nado Loadings), Section 3.4 (Water Level Design), and Section 3.7 (Seismic Design).

2.3 Fire Protection (Criterion 3)

Criterion uctures, systems, and components important to safety shall be designed and located to imize, consistent with other safety requirements, the probability and effect of fires and losions. Noncombustible and heat-resistant materials shall be used wherever practical ughout the unit, particularly in locations such as the containment and control room. Fire ction and fighting systems of appropriate capacity and capability shall be provided and gned to minimize the adverse effects of fires on structures, systems, and components ortant to safety. Fire fighting systems shall be designed to assure that their rupture or vertent operation does not significantly impair the safety capability of these structures, ems, and components.

Design Conformance design of Millstone 3 minimizes the probability and effect of fires and explosions on ctures, systems, and components important to safety. Noncombustible and heat-resistant erials are used wherever practical throughout the unit. Fire detection and fire suppression ems of sufficient capacity and capability minimize the adverse effects of fires on structures, ems, and components important to safety. Fire suppression systems are designed to assure that ure or inadvertent operation does not significantly impair the safety capability of these ctures, systems, and components.

tion 9.5.1 and Millstone 3 Fire Protection Evaluation Report describe the fire protection em in detail.

3.1-8 Rev. 30

Criterion uctures, systems, and components important to safety shall be designed to accommodate the cts of and to be compatible with the environmental conditions associated with normal ration, maintenance, testing, and postulated accidents, including loss-of-coolant accidents.

se structures, systems, and components shall be appropriately protected against dynamic cts, including the effects of missiles, pipe whipping, and discharging fluids, that may result m equipment failures and from events and conditions outside the nuclear power unit.

Design Conformance ctures, systems, and components important to safety are designed to accommodate the effects nd to be compatible with the environmental conditions associated with normal operating, ntenance, testing, and postulated accidents including LOCAs. These items are either ected from accident conditions or designed to withstand, without failure, exposure to the bination of temperature, pressure, humidity, radiation, and dynamic effects expected during required operational period.

sical separation, physical protection, pipe restraints, and redundancy are included in the gn of safety related systems to ensure that each such system performs its intended safety ction.

letter from B. J. Youngblood (NRC) to J. F. Opeka (NNECO) dated June 5, 1985, Millstone 3 granted an exemption for a period of two cycles of operation from those portions of General ign Criterion 4 which require protection of structures, systems, and components from the amic effects associated with postulated breaks in the reactor coolant system primary loop ng.

ederal Register, Volume 51, No. 70, dated April 11, 1986, the NRC published a final rule difying General Design Criterion 4 to allow use of leak-before-break technology for excluding m the design basis the dynamic effects of postulated ruptures in primary coolant loop piping in surized water reactors. This rule obviates the need for the above exemption.

ctures, systems, and components important to safety are classified as QA Category I and are gned in accordance with the codes and classifications indicated in Section 3.2.5.

pter 3 provides the details of the environmental activities and dynamic effects to which the ctures, systems, and components important to safety are designed.

3.1-9 Rev. 30

Criterion uctures, systems, and components important to safety shall not be shared among nuclear er units unless it can be shown that such sharing will not significantly impair their ability to orm their safety functions, including, in the event of an accident in one unit, an orderly tdown and cooldown of the remaining units.

Design Conformance ystem important to safety and shared by Millstone 1, 2, and 3 is the radioactive gaseous waste em discharge stack described in Section 11.3. The sharing of the discharge stack causes no ty impairment of the three units.

following equipment may be shared and utilized by Millstone Unit 2 to meet its GDC 17 uirements for an alternate offsite source to relieve one of its emergency diesel generators and ply power to minimum post-accident loads:

1. Main Transformer 15G-3X
2. Normal Station Service Transformer 15G-3SA
3. Reserve Station Service Transformer 15G-23SA sharing of this equipment does not impair its ability to perform its safety function. The sformers are adequately sized and have sufficient capacity to meet maximum postulated Unit ading requirements while supplying Unit 2 GDC 17 minimum loads.

er facilities and systems not important to safety within the definitions of GDC 5, but which are ed by the three units are:

1. Environmental monitoring systems
2. Machine shops
3. Offsite transmission lines and switchyard
4. Office buildings
5. Roadway access
6. Railroad access
7. General warehouses 3.1-10 Rev. 30
9. Potable (Waterford) water
10. Warehousing facility houses the Millstone 2 condensate polishing system, the Millstone 2 and 3 (removed from service) condensate demineralizer radioactive liquid waste systems. The Unit 2 condensate polishing solid waste system is designed to process radioactive waste from Millstone 2 and 3.
11. Alternate AC (SBO) Diesel Generator (shared by Units 2 and 3)
12. Security System
13. Auxiliary Steam System 2.6 (Criterion 6) erion 6 has not been promulgated by the NRC.

2.7 (Criterion 7) erion 7 has not been promulgated by the NRC.

2.8 (Criterion 8) erion 8 has not been promulgated by the NRC.

2.9 (Criterion 9) erion 9 has not been promulgated by the NRC.

2.10 Reactor Design (Criterion 10)

Criterion e reactor core and associated coolant, control, and protection systems shall be designed with ropriate margin to assure that specified acceptable fuel design limits are not exceeded during condition of normal operation, including the effects of anticipated operational occurrences.

Design Conformance reactor core and associated coolant, control, and protection systems are designed with quate margins to:

1. Assure that fuel damage is not expected during normal core operation and operational transients (Condition I) or any transient conditions arising from 3.1-11 Rev. 30

plant cleanup system and are consistent with plant design bases.

2. Ensure return of the reactor to a safe state following infrequent incident (Condition III) events with only a small fraction of fuel rods damaged, although sufficient fuel damage might occur to preclude immediate resumption of operation.
3. Assure that the core is intact with acceptable heat transfer geometry following transients arising from occurrences of limiting faults (Condition IV).

e that fuel damage as used under Item 1 is defined as penetration of the fission product barrier

, the fuel rod clad). Also note that ANSI N18.2-73 expands the definitions of the four ditions enumerated in Items 1 through 3.

pter 4 discusses the design bases and design evaluation of reactor components. Chapter 7 s the details of the control and protection systems instrumentation design and logic. This rmation supports the accident analyses of Chapter 15 which show that the acceptable fuel gn limits are not exceeded for Condition I and II occurrences.

2.11 Reactor Inherent Protection (Criterion 11)

Criterion e reactor core and associated coolant systems shall be designed so that in the power operating ge the net effect of the prompt inherent nuclear feedback characteristics tends to compensate a rapid increase in reactivity.

Design Conformance mpt compensatory reactivity feedback effects are assured when the reactor is critical by the ative fuel temperature effect (Doppler effect) and by ensuring that the moderator temperature fficient is maintained within the limits provided in Technical Specification 3/4.1.1.3. Section 2.3 discusses these reactivity coefficients.

2.12 Suppression of Reactor Power Oscillations (Criterion 12)

Criterion e reactor core and associated coolant, control, and protection systems shall be designed to re that power oscillations which can result in conditions exceeding specified acceptable fuel gn limits are not possible or can be reliably and readily detected and suppressed.

3.1-12 Rev. 30

al power oscillations of the fundamental mode are inherently stable by the negative power fficient of reactivity.

illations, due to xenon spatial effects, in the radial, diametral, and azimuthal overtone modes heavily damped due to the inherent design and due to the negative power coefficient of tivity.

illations, due to xenon spatial effects, in the axial first overtone mode may occur. Assurance fuel design limits are not exceeded by xenon axial oscillations is provided by reactor trip ctions using the measured axial power imbalance as an input.

illations, due to xenon spatial effects, in axial modes higher than the first overtone are heavily ped due to the inherent design and due to the negative Doppler coefficient of reactivity.

tion 4.3 discusses xenon and samarium stability control.

2.13 Instrumentation and Control (Criterion 13)

Criterion trumentation shall be provided to monitor variables and systems over their anticipated ranges normal operation, for anticipated operational occurrences, and for accident conditions as ropriate to assure adequate safety, including those variables and systems that can affect the ion process, the integrity of the reactor core, the reactor coolant pressure boundary, and the tainment and its associated systems. Appropriate controls shall be provided to maintain these ables and systems within prescribed operating ranges.

Design Conformance rumentation and controls are provided to monitor and control neutron flux, control rod ition, temperatures, pressures, flows, and levels as necessary to assure that adequate plant ty can be maintained. Instrumentation is provided in the reactor coolant system, steam and er conversion system, the containment, engineered safety features systems, and other iliaries. Parameters that must be provided for operator use under normal operating and dent conditions are indicated in proximity with the controls for maintaining the indicated meter in the proper range.

quantity and types of process instrumentation provided ensures safe and orderly operation of ystems over the full design range of the plant. These systems are described in Chapters 6, 7, 8, 1, and 12.

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Criterion e reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to e an extremely low probability of abnormal leakage, of rapidly propagating failure, and of ss rupture.

Design Conformance reactor coolant system boundary is designed to accommodate the system pressures and peratures attained under all expected modes of plant operation, including all anticipated sients, and to maintain the stresses within applicable stress limits (Section 3.9). Reactor lant pressure boundary materials, selection, and fabrication techniques ensure a low bability of gross rupture or abnormal leakage.

ddition to the loads imposed on the system under normal operating conditions, consideration lso given to abnormal loading conditions, such as seismic and pipe rupture, as discussed in tions 3.6 and 3.7. The system is protected from overpressure by means of pressure relieving ices as required by applicable codes (Section 5.2.2).

reactor coolant system boundary has provisions for inspection, testing, and surveillance of cal areas to assess the structural and leaktight integrity (Section 5.2). For the reactor vessel ction 5.3), a material surveillance program conforming to applicable codes is provided.

2.15 Reactor Coolant System Design (Criterion 15)

Criterion e reactor coolant system and associated auxiliary, control, and protection systems shall be gned with sufficient margin to assure that the design conditions of the reactor coolant pressure ndary are not exceeded during any condition of normal operation, including anticipated rational occurrences.

Design Conformance design pressure and temperature for each component in the reactor coolant and associated iliary, control, and protection systems are selected to be above the maximum coolant pressure temperature under all normal and anticipated transient load conditions.

itionally, reactor coolant pressure boundary components achieve a large margin of safety by use of proven ASME materials and design codes, use of proven fabrication techniques, destructive shop testing, and integrated hydrostatic testing of assembled components. Chapter scusses the reactor coolant system design.

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Criterion actor containment and associated systems shall be provided to establish an essentially

-tight barrier against the uncontrolled release of radioactivity to the environment and to assure the containment design conditions important to safety are not exceeded for as long as tulated accident conditions require.

Design Conformance teel-lined reinforced concrete containment structure, maintained at subatmospheric pressure, loses the entire reactor coolant system with an essentially leaktight barrier, as described in tion 6.2.1. The containment structure and the engineered safety features are designed to hstand internal and external environmental conditions that may reasonably be expected during life of the unit and to ensure that the short and long term conditions following a LOCA do not eed the design values. Following a design basis accident (DBA), the containment heat removal ems reduce the containment pressure, as described in Sections 6.1.2 and 6.2.2. Most of the age from the containment structure is collected and processed through the supplementary leak ection and release system, described in Section 6.2.3. This process reduces the amount of oactivity released to the environment.

2.17 Electric Power Systems (Criterion 17)

Criterion onsite electric power system and an offsite electric power system shall be provided to permit ctioning of structures, systems, and components important to safety. The safety function for h system (assuming the other system is not functioning) shall be to provide sufficient capacity capability to assure that (1) specified acceptable fuel design limits and design conditions of reactor coolant pressure boundary are not exceeded as a result of anticipated operational urrences and (2) the core is cooled and containment integrity and other vital functions are ntained in the event of postulated accidents.

e onsite electric power supplies, including the batteries, and the onsite electric distribution em, shall have sufficient independence, redundancy, and testability to perform their safety ctions assuming a single failure.

ectric power from the transmission network to the onsite electric distribution system shall be plied by two physically independent circuits (not necessarily on separate rights of way) gned and located so as to minimize to the extent practical the likelihood of their simultaneous ure under operating and postulated accident and environmental conditions. A switchyard mon to both circuits is acceptable. Each of these circuits shall be designed to be available in icient time following a loss of all onsite alternating current power supplies and the other ite electric power circuit, to assure that specified acceptable fuel design limits and design ditions of the reactor coolant pressure boundary are not exceeded. One of these circuits shall 3.1-15 Rev. 30

ovisions shall be included to minimize the probability of losing electric power from any of the aining supplies as a result of, or coincident with, the loss of power generated by the nuclear er unit, the loss of power from the transmission network, or the loss of power from the onsite tric power supplies.

Design Conformance o connections to the offsite power system are provided. The preferred offsite connection is a kfeed through the main and normal station service transformers with the generator breaker

n. The alternate offsite connection is through the reserve station service transformers. Each ite source has 100 percent capacity for all emergency and normal loads during all phases of ration plus the capacity to supply Millstone Unit 2 GDC 17 requirements through the NSST or ST as an alternate offsite source for minimum post-accident loads.

o onsite power systems are provided. Each system has an emergency diesel generator. Each el generator has 100 percent capacity for the emergency loads in the event of the postulated dents or required for reactor cooldown.

design of the electrical system (Chapter 8) conforms to Criterion 17.

2.18 Inspection and Testing of Electric Power Systems (Criterion 18)

Criterion ectric power systems important to safety shall be designed to permit appropriate periodic ection and testing of important areas and features, such as wiring, insulation, connections, and tchboards, to assess the continuity of the systems and the conditions of their components. The ems shall be designed with a capability to test periodically (1) the operability and functional ormance of the components of the systems, such as onsite power sources, relays, switches, buses, and (2) the operability of the systems as a whole and, under conditions as close to gn as practical, the full operation sequence that brings the systems into operation, including ration of applicable portions of the protection systems, and the transfer of power among the lear power unit, the offsite power system, and the onsite power system.

Design Conformance engineered safety features power supply buses, the power supply buses for equipment uired for cooldown, and associated emergency generators, all of which comprise the onsite er system, are arranged for periodic testing of each system independently. During refueling tdowns, tests are conducted to prove the operability of the automatic starting and load uencing capability of the emergency generators. Full load testing of each emergency generator erformed periodically. These tests prove the operability of the electric power systems under ditions as close to design as practical to assess the continuity of these systems and condition of 3.1-16 Rev. 30

2.19 Control Room (Criterion 19)

Criterion control room shall be provided from which actions can be taken to operate the nuclear power safely under normal conditions and to maintain it in a safe condition under accident ditions, including loss-of-coolant accidents. Adequate radiation protection shall be provided ermit access and occupancy of the control room under accident conditions without personnel iving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the y, for the duration of the accident.

uipment at appropriate locations outside the control room shall be provided (1) with a design ability for prompt hot shutdown of the reactor, including necessary instrumentation and trols to maintain the unit in a safe condition during hot shutdown, and (2) with a potential ability for subsequent cold shutdown of the reactor through the use of suitable procedures.

Design Conformance control room provided is equipped to operate the unit safely under normal and accident ditions. Its shielding and ventilation design permits continuous occupancy of the control room the duration of a DBA without the dose to personnel exceeding 5 rem whole body. Based on CFR 50.67, the applicable dose criterion was modified to 5 rem TEDE.

e auxiliary shutdown panel located in the west switchgear room has equipment, controls, and rumentation to accomplish, in conjunction with controls and indication located on the adjacent 0V switchgear, a prompt hot shutdown and the capability for subsequent cold shutdown of the tor through the use of suitable procedures. The panel is physically located outside the control

m. Thus, the uninhabitability of the control room would have no effect on the availability of auxiliary shutdown panel and adjacent controls (Section 7.4.1.3).

design of the control building (Section 3.8.4), which houses the control room and the iliary shutdown panel area, conforms to Criterion 19. Section 9.4.1 describes the control ding ventilation system. Control room habitability is discussed in Section 6.4. Fire protection ems are discussed in Section 9.5.1 2.20 Protection System Functions (Criterion 20)

Criterion e protection system shall be designed (1) to initiate automatically the operation of appropriate ems including the reactivity control systems, to assure that specified acceptable fuel design ts are not exceeded as a result of anticipated operational occurrences and (2) to sense accident ditions and to initiate the operation of systems and components important to safety.

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ully automatic protection system with appropriate redundant channels is provided to cope with sients where insufficient time is available for manual corrective action. The design basis for protection systems is in accordance with IEEE Standard 279-1971 and IEEE Standard

-1972. The reactor protection system automatically initiates a reactor trip when any variable eeds the normal operating range. Setpoints are designed to provide an envelope of safe rating conditions with adequate margin for uncertainties to ensure that fuel design limits are exceeded.

ctor trip is initiated by removing power to the rod drive mechanisms of all the full length rod ter control assemblies. This causes the rods to insert by gravity rapidly reducing the reactor er output. The response and adequacy of the protection system have been verified by analysis nticipated transients.

engineered safety features (ESF) actuation system automatically initiates emergency core ling, and other safeguards functions, by sensing accident conditions using redundant analog nnels measuring diverse variables. Manual actuation of safeguards may be performed where le time is available for operator action. The ESF actuation system automatically trips the tor on manual or automatic safety injection signal (SIS) generation.

2.21 Protection System Reliability and Testability (Criterion 21)

Criterion e protection system shall be designed for high functional reliability and inservice testability mensurate with the safety functions to be performed. Redundancy and independence gned into the protection system shall be sufficient to assure that (1) no single failure results in of the protection function and (2) removal from service of any component or channel does result in loss of the required minimum redundancy unless the acceptable reliability of ration of the protection system can be otherwise demonstrated. The protection system shall be gned to permit periodic testing of its functioning when the reactor is in operation, including a ability to test channels independently to determine failures and losses of redundancy that may e occurred.

Design Conformance protection system is designed for high functional reliability and inservice testability.

mpliance with this criterion is discussed in detail in Sections 7.2.2.2.3 and 7.3.2.2.5.

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Criterion e protection system shall be designed to assure that the effects of natural phenomena, and of mal operating, maintenance, testing, and postulated accident conditions on redundant channels not result in loss of the protection function, or shall be demonstrated to be acceptable on some r defined basis. Design techniques, such as functional diversity or diversity in component gn and principles of operation, shall be used to the extent practical to prevent loss of the ection function.

Design Conformance tection system components are designed and arranged so that the environment accompanying emergency situation in which the components are required to function does not result in loss he safety function. Various means are used to accomplish this. Functional diversity has been gned into the system. The extent of this functional diversity has been evaluated for a wide ety of postulated accidents. Diverse protection functions automatically terminate an accident ore intolerable consequences could occur.

tion 7.1.2.1.8 provides details of ESF system diversity.

omatic reactor trips are based upon process parameters and neutron flux measurements. Trips process parameters include reactor coolant loop temperature measurements, pressurizer sure and level measurements, and reactor coolant pump underspeed trip. Trips may also be ated manually or by SIS. Section 7.2 describes all the trips and provides further details.

h quality components, conservative design and applicable quality control, inspection, bration, and tests are utilized to guard against common-mode failure. Sections 3.10 and 3.11 vide details concerning qualification testing. Qualification testings is performed on the various ty systems to demonstrate functional operation at normal and post-accident conditions of perature, humidity, pressure, and radiation for specified periods, if required. Typical ection system equipment is subjected to type tests under simulated seismic condition using servatively large accelerations and applicable frequencies. The test results indicate no loss of protection function.

2.23 Protection System Failure Modes (Criterion 23)

Criterion e protection system shall be designed to fail into a safe state or into a state demonstrated to be eptable on some other defined basis if conditions such as disconnection of the system, loss of rgy (e.g., electric power, instrument air), or postulated adverse environments (e.g., extreme t or cold, fire, pressure, steam, water, and radiation) are experienced.

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protection system is designed with due consideration of the most probable failure modes of components under various perturbations of the environment and energy sources. Sections 7.2 7.3 discuss this protection system.

2.24 Separation of Protection and Control Systems (Criterion 24)

Criterion e protection system shall be separated from control systems to the extent that failure of any le control system component or channel, or failure or removal from service of any single ection system component or channel which is common to the control and protection systems es intact a system satisfying all reliability, redundancy, and independence requirements of the ection system. Interconnection of the protection and control systems shall be limited so as to re that safety is not significantly impaired.

Design Conformance protection system is separate and distinct from the control systems. Control systems may be endent on the protection system in that control signals are derived from protection system surements, where applicable. These signals are transferred to the control system by isolation ices which are classified as protection components. The adequacy of system isolation is fied by testing under conditions of postulated credible faults. The failure of any single control em component or channel, or failure or removal from service of any single protection system ponent or channel which is common to the control and protection systems, leaves intact a em which satisfies the requirements of the protection system. Distinction between channel train is made in this discussion. The removal of a train from service is allowed only during ing of the train. Chapter 7 gives further details.

2.25 Protection System Requirements for Reactivity Control Malfunctions (Criterion 25)

Criterion e protection system shall be designed to assure that specified acceptable fuel design limits are exceeded for any single malfunction of the reactivity control systems, such as accidental hdrawal (not ejection or dropout) of control rods.

Design Conformance protection system is designed to limit reactivity transients so that fuel design limits are not eeded. Reactor shutdown by full length rod insertion is completely independent of the normal trol function since the trip breakers interrupt power to the rod mechanisms regardless of ting control signals. Thus, in the postulated accidental withdrawal (assumed to be initiated by ntrol malfunction), flux, temperature, pressure, level, and flow signals would be generated pendently. Any of these signals (trip demands) would operate the breakers to trip the reactor.

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rator has ample time to determine the cause of dilution, terminate the source of dilution, and ate boration before the shutdown margin is lost. The analyses show that acceptable fuel age limits are not exceeded even in the event of a single malfunction of either system.

2.26 Reactivity Control System Redundancy and Capability (Criterion 26)

Criterion o independent reactivity control systems of different design principles shall be provided. One he systems shall use control rods, preferably including a positive means for inserting the rods shall be capable of reliability controlling reactivity changes to assure that under conditions of mal operation, including anticipated operational occurrences, and with appropriate margin for functions such as stuck rods, specified acceptable fuel design limits are not exceeded. The ond reactivity control system shall be capable of reliably controlling the rate of reactivity nges resulting from planned, normal power changes (including xenon burnout) to assure eptable fuel design limits are not exceeded. One of the systems shall be capable of holding the tor core subcritical under cold conditions.

Design Conformance o reactivity control systems are provided. These are rod cluster control assemblies (RCCAs) chemical shim (boric acid). The RCCAs are inserted into the core by the force of gravity.

ing operation the shutdown rod banks are fully withdrawn. The rod control system matically maintains a programmed average reactor temperature compensating for reactivity cts associated with scheduled and transient load reductions. The rod control system cannot matically withdraw control rods. Operator action is required to restore the plant to ilibrium conditions following load increases. The shutdown rod banks along with the control ks are designed to shut down the reactor with adequate margin under conditions of normal ration and anticipated operational occurrences, thereby ensuring that specified fuel design ts are not exceeded. The most restrictive period in core life is assumed in all analyses and the t reactive rod cluster is assumed to be in the fully withdrawn position.

chemical and volume control system maintains the reactor in the cold shutdown state ependent of the position of the control rods and can compensate for xenon burnout transients.

pter 4 presents details of the construction of the RCCAs and Chapter 7 discusses the ration. Chapter 9 describes the means of controlling the boric acid concentration. Chapter 15 udes performance analyses under accident conditions.

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Criterion e reactivity control systems shall be designed to have a combined capability, in conjunction h poison addition by the emergency core cooling system, of reliably controlling reactivity nges to assure that under postulated accident conditions and with appropriate margin for stuck s the capability to cool the core is maintained.

Design Conformance facility is provided with means of making and holding the core subcritical under any cipated conditions and with appropriate margin for contingencies. Chapter 4 and 9 discuss e means in detail. Combined use of the rod cluster control system and the chemical shim trol system permits the necessary shutdown margin to be maintained during long term xenon ay and plant cooldown. The single highest worth control cluster is assumed to be stuck full-out n trip for this determination. Chapter 15 describes accident assumptions in detail.

2.28 Reactivity Limits (Criterion 28)

Criterion e reactivity control systems shall be designed with appropriate limits on the potential amount rate of reactivity increase to assure that the effects of postulated reactivity accidents can her (1) result in damage to the reactor coolant pressure boundary greater than limited local ding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel rnals to impair significantly the capability to cool the core. These postulated reactivity dents shall include consideration of rod ejection (unless prevented by positive means), rod pout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water ition.

Design Conformance maximum reactivity worth of control rods and the maximum rates of reactivity insertion loying control rods are limited to values that prevent rupture of the reactor coolant system ndary or disruptions of the core or vessel internals to a degree that could impair the ctiveness of emergency core cooling.

maximum positive reactivity insertion rates for the withdrawal of rod cluster control mblies (RCCAs) and the dilution of the boric acid in the reactor coolant system are limited by physical design characteristics of the RCCAs and of the chemical and volume control system.

hnical Specifications on shutdown margin and on RCCA insertion limits and bank overlaps as ctions of power provide additional assurance that the consequences of the postulated accidents no more severe than those presented in the analyses of Chapter 15. Reactivity insertion rates, tion, and withdrawal limits are also discussed in Section 4.3. The capability of the chemical 3.1-22 Rev. 30

urance of core cooling capability following Condition IV accidents, such as rod ejections, m line break, etc., is given by keeping the reactor coolant pressure boundary stresses within ted condition limits as specified by applicable ASME Codes. Structural deformations are cked also and limited to values that do not jeopardize the operation of necessary safety ures.

2.29 Protection against Anticipated Operational Occurrences (Criterion 29)

Criterion e protection and reactivity control systems shall be designed to assure an extremely high bability of accomplishing their safety functions in the event of anticipated operational urrences.

Design Conformance protection and reactivity control systems are designed to assure extremely high probability of orming their required safety functions in any anticipated operational occurrences. Equipment d in these systems is designed, constructed, operated, and maintained with a high level of ability. Chapter 7 covers details of system design. Also refer to the discussions of GDC-20 ugh 28.

2.30 Quality of Reactor Coolant Pressure Boundary (Criterion 30)

Criterion mponents which are part of the reactor coolant pressure boundary shall be designed, icated, erected, and tested to the highest quality standards practical. Means shall be provided detecting and, to the extent practical, identifying the location of the source of reactor coolant age.

Design Conformance ctor coolant pressure boundary components are designed, fabricated, inspected, and tested in formance with ASME Nuclear Power Plant Components Code,Section III. All components classified according to ANSI N18.2-73 and N18.2a-75 and are accorded the quality measures ropriate to the classification. The design bases and evaluations of reactor coolant pressure ndary components are discussed in Chapter 5.

kage is detected by an increase in the amount of makeup water required to maintain a normal l in the pressurizer. The reactor vessel closure joint is provided with a temperature monitored off between double gaskets. Leakage into the reactor containment is drained to the reactor ding sump where it is monitored.

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tor coolant in the system at the pressurizer, volume control tank and coolant drain collection s make available an accurate indication of integrated leakage.

tion 5.2.5 discusses the reactor coolant pressure boundary leakage detection system.

2.31 Fracture Prevention of Reactor Coolant Pressure Boundary (Criterion 31)

Criterion e reactor coolant pressure boundary shall be designed with sufficient margin to assure that n stressed under operating, maintenance, testing, and postulated accident conditions (1) the ndary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is imized. The design shall reflect consideration of service temperatures and other conditions of boundary material under operating, maintenance, testing, and postulated accident conditions the uncertainties in determining (1) material properties, (2) the effects of irradiation on erial properties, (3) residual, steady state and transient stresses, and (4) size of flaws.

Design Conformance se control is maintained over material selection and fabrication for the reactor coolant system ssure that the boundary behaves in a non-brittle manner. The reactor coolant system materials ch are exposed to the coolant are corrosion resistant stainless steel or Inconel. The NIL tility reference temperature (RTNDT) of the reactor vessel structural steel is established by rpy V-notch and drop weight tests in accordance with 10 CFR 50, Appendix G.

part of the reactor vessel specification, certain requirements which are not specified by the licable ASME Codes are performed as follows:

1. Ultrasonic Testing - In addition to code requirements, a 100-percent volumetric ultrasonic test of reactor vessel plate for shear wave and a post-hydro test ultrasonic map of all full penetration ferritic pressure boundary welds in the pressure vessel are performed. Cladding bond ultrasonic inspection to more restrictive requirements than those specified in the code are also required to preclude interpretation problems during inservice inspection.
2. Radiation Surveillance Program - In the surveillance programs, the evaluation of the radiation damage is based on pre-irradiation and post-irradiation testing of Charpy V-notch and tensile specimens. These programs are directed toward evaluation of the effect of radiation on the fracture toughness of reactor vessel steels based on the reference transition temperature approach and the fracture mechanics approach, and are in accordance with ASTM-E-185-82, Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, and the requirements of 10 CFR 50, Appendix H.

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over the life of the plant.

fabrication and quality control techniques used in the fabrication of the reactor coolant em are consistent with those used for the reactor vessel. The inspections of reactor vessel, surizer, piping, pumps, and steam generator are governed by ASME Code requirements.

fer to Chapter 5.)

owable pressure-temperature relationships for plant heatup and cooldown rates are calculated g methods derived from the ASME Code,Section III, Appendix G, Protection Against Non-tile Failure. The approach specifies that allowed stress intensity factors for all vessel rating conditions may not exceed the reference stress intensity factor (KIR) for the metal perature at any time. Operating specifications include conservative margins for predicted nges in the material reference temperatures (RTNDT) due to irradiation.

2.32 Inspection of Reactor Coolant Pressure Boundary (Criterion 32)

Criterion mponents which are part of the reactor coolant pressure boundary shall be designed to permit periodic inspection and testing of important areas and features to assess their structural and

-tight integrity, and (2) an appropriate material surveillance program for the reactor pressure sel.

Design Conformance design of the reactor coolant pressure boundary provides the capability for accessibility ng service life to the entire internal surfaces of the reactor vessel, certain external zones of the sel including the nozzle to reactor coolant piping welds and the top and bottom heads, and rnal surfaces of the reactor coolant piping except for the area of pipe within the primary lding concrete. The inspection capability complements the leakage detection systems in ssing the pressure boundary components integrity. The reactor coolant pressure boundary is odically inspected under the provisions of ASME Boiler and Pressure Vessel Code, Section Section 5.2.4 gives details of the Inservice Inspection Program.

nitoring of changes in the fracture toughness properties of the reactor vessel core region plates ing, weldments, and associated heat treated zones are performed in accordance with 10 CFR Appendix H, Reactor Vessel Material Surveillance Program Requirements. Samples of tor vessel plate materials are retained and catalogued in case future engineering development ws the need for further testing.

material properties surveillance program includes not only the conventional tensile and act tests, but also fracture mechanics specimens. The observed shifts in RTNDT of the core on materials with irradiation are used to confirm the allowable limits calculated for all rational transients. Section 5.3 gives further details.

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Criterion system to supply reactor coolant makeup for protection against small breaks in the reactor lant pressure boundary shall be provided. The system safety function shall be to assure that cified acceptable fuel design limits are not exceeded as a result of reactor coolant loss due to age from the reactor coolant pressure boundary and rupture of small piping or other small ponents which are part of the boundary. The system shall be designed to assure that for onsite tric power system operation (assuming offsite power is not available) and for offsite electric er system operation (assuming onsite power is not available) the system safety function can ccomplished using the piping, pumps, and valves used to maintain coolant inventory during mal reactor operation.

Design Conformance chemical and volume control system provides a means of reactor coolant makeup and stment of the boric acid concentration. Makeup is added automatically if the level in the ume control tank falls below the normal operating range. The high-pressure centrifugal rging pumps provided are capable of supplying the required makeup and reactor coolant seal ction flow when power is available from either onsite or offsite electric power systems. These ps also serve as high head safety injection pumps. Functional reliability is assured by vision of standby components assuring a safe response to probable modes of failure. Sections and 9.3 include details of system design and Chapter 8, details of the electric power system.

2.34 Residual Heat Removal (Criterion 34)

Criterion system to remove residual heat shall be provided. The system safety function shall be to sfer fission product decay heat and other residual heat from the reactor core at a rate such that cified acceptable fuel design limits and the design conditions of the reactor coolant pressure ndary are not exceeded.

itable redundancy in components and features, and suitable interconnections, leak detection, isolation capabilities shall be provided to assure that for onsite electric power system ration (assuming offsite power is not available) and for offsite electric power system operation uming onsite power is not available) the system safety function can be accomplished, ming a single failure.

Design Conformance residual heat removal system, in conjunction with the steam and power conversion system, is gned to transfer the fission product decay heat and other residual heat from the reactor core hin acceptable limits. The transfer of the heat removal function from the steam and power 3.1-26 Rev. 30

able redundancy at temperatures below approximately 350°F is accomplished with the two dual heat removal pumps (located in separate compartments with means available for draining monitoring of leakage), the two heat exchangers and the associated piping, cabling, and tric power sources. The residual heat removal system is able to operate on either onsite or ite electrical power system.

able redundancy at temperatures above approximately 350°F is provided by the steam erators and associated piping system.

tion 5.4.7 and Chapter 10 give details of the system design.

2.35 Emergency Core Cooling (Criterion 35)

Criterion system to provide abundant emergency core cooling shall be provided. The system safety ction shall be to transfer heat from the reactor core following any loss of reactor coolant at a such that (1) fuel and clad damage that could interfere with continued effective core cooling revented and (2) clad metal-water reaction is limited to negligible amounts.

itable redundancy in components and features, and suitable interconnections, leak detection, ation, and containment capabilities shall be provided to assure that for onsite electric power em operation (assuming offsite power is not available) and for offsite electric power system ration (assuming onsite power is not available) the system safety function can be omplished, assuming a single failure.

Design Conformance emergency core cooling system is provided to cope with any loss-of-coolant accident in the t design basis. Abundant cooling water is available in an emergency to transfer heat from the at a rate that clad metal-water reaction is limited to less than one percent. Adequate design visions are made to assure performance of the required safety functions even with a single ure.

tion 6.3 includes details of the capability of the systems. Chapter 15 includes an evaluation of adequacy of the system functions. Performance evaluations are conducted in accordance with CFR 50.46 and Appendix K to 10 CFR 50.

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Criterion e emergency core cooling system shall be designed to permit appropriate periodic inspection mportant components, such as spray rings in the reactor pressure vessel, water injection zles, and piping to assure the integrity and capability of the system.

Design Conformance ign provisions facilitate access to the critical parts of the injection nozzles, pipes, and valves visual inspection and for nondestructive inspection where such techniques are desirable and ropriate. The design is in accordance with ASME,Section XI requirements.

components outside the containment are accessible for leaktightness inspection during ration of the reactor.

2.37 Testing of Emergency Core Cooling System (Criterion 37)

Criterion e emergency core cooling system shall be designed to permit appropriate periodic pressure functional testing to assure (1) the structural and leaktight integrity of its components, (2) the rability and performance of the active components of the system, and (3) the operability of the em as a whole and, under conditions as close to design as practical, the performance of the full rational sequence that brings the system into operation, including operation of applicable ions of the protection system, the transfer between normal and emergency power sources, and operation of the associated cooling water system.

Design Conformance ive components of the emergency core cooling system can be actuated from the emergency er source at any time during unit operation to demonstrate operability.

ts are performed during refueling shutdowns to demonstrate proper automatic operation of the rgency core cooling system. An integrated system test is performed. Sections 6.3 and 7.3 cribe the above tests.

2.38 Containment Heat Removal (Criterion 38)

Criterion system to remove heat from the reactor containment shall be provided. The system safety ction shall be to reduce rapidly, consistent with the functioning of other associated systems, containment pressure and temperature following any loss-of-coolant accident and maintain m at acceptably low levels.

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em operation (assuming offsite power is not available) and for offsite electric power system ration (assuming onsite power is not available) the system safety function can be omplished, assuming a single failure.

Design Conformance t is removed from the containment structure following a LOCA by the containment ressurization systems, which consist of the quench spray system and the containment rculation system (Section 6.2.2). The quench spray system, consisting of two 100-percent acity subsystems, transfers water from the refueling water storage tank to two parallel

-degree spray headers. The quench spray system transfers heat from the containment osphere to water on the containment structure floor. The containment recirculation system, ch consists of two 100-percent capacity subsystems (each consisting of two pumps, two lers, and a 360-degree spray header), transfers heat from the water collected in the tainment structure sump to the service water system (Section 9.2.1) via the containment rculation coolers. The quench spray pumps and the containment recirculation pumps and lers are located in the engineered safety features building (Section 3.8).

containment depressurization systems are designed so that no single active failure in the short or no single active or passive failure in the long term impairs their ability to perform their ty function. Redundant components are isolated, physically and electrically. Each subsystem onnected to a separate electrical bus which can be connected to either offsite or onsite power.

2.39 Inspection of Containment Heat Removal System (Criterion 39)

Criterion e containment heat removal system shall be designed to permit appropriate periodic ection of important components, such as the torus, sumps, spray nozzles, and piping to assure integrity and capability of the system.

Design Conformance design of the containment depressurization systems permits appropriate periodic inspection he important components, as described in Section 6.2.2.4.

2.40 Testing of Containment Heat Removal System (Criterion 40)

Criterion e containment heat removal system shall be designed to permit appropriate periodic pressure functional testing to assure (1) the structural and leaktight integrity of its components, (2) the rability and performance of the active components of the system, and (3) the operability of the em as a whole, and, under conditions as close to the design as practical, the performance of 3.1-29 Rev. 30

rces, and the operation of the associated cooling water system.

Design Conformance design of the containment depressurization systems permits periodic pressure and functional ing, as described in Section 6.2.2.4.

2.41 Containment Atmosphere Cleanup (Criterion 41)

Criterion stems to control fission products, hydrogen, oxygen, and other substances which may be ased into the reactor containment shall be provided as necessary to reduce, consistent with the ctioning of other associated systems, the concentration and quality of fission products released he environment following postulated accidents, and to control the concentration of hydrogen xygen and other substances in the containment atmosphere following postulated accidents to re that containment integrity is maintained.

ch system shall have suitable redundancy in components and features, and suitable rconnections, leak detection, isolation, and containment capabilities to assure that for onsite tric power system operation (assuming offsite power is not available) and for offsite electric er system operation (assuming onsite power is not available) its safety function can be omplished, assuming a single failure.

Design Conformance supplementary leak collection and release system (SLCRS) (Section 6.2.3.2) collects oactive leakage from the containment to the containment enclosure and contiguous areas owing a LOCA.

quench spray system (Section 6.2.2 and 6.5.2) sprays borated water into the containment osphere to reduce the containment pressure and remove airborne iodine. The pH in the tainment sumps is controlled by the dissolution of trisodium phosphate (stored in baskets) in sump water.

recirculation spray system (Sections 6.2.2 and 6.5.2) sprays containment sump water into the tainment atmosphere to reduce containment pressure and remove airborne iodine.

se systems are sufficiently redundant to perform their safety function assuming a single active ure in the short term or a single active or passive failure in the long term and are operable with er onsite or offsite power.

3.1-30 Rev. 30

Criterion e containment atmosphere cleanup systems shall be designed to permit appropriate periodic ection of important components, such as filter frames, ducts, and piping to assure the integrity capability of the systems.

Design Conformance design of the supplementary leak collection and release system permit appropriate periodic ection of the important components, as described in Section 6.5.1.4.

2.43 Testing of Containment Atmosphere Cleanup Systems (Criterion 43)

Criterion e containment atmosphere cleanup systems shall be designed to permit appropriate periodic sure and functional testing to assure (1) the structural and leaktight integrity of its ponents, (2) the operability and performance of the active components of the systems such as

, filters, dampers, pumps, and valves and (3) the operability of the systems as a whole and er conditions as close to design as practical, the performance of the full operational sequence brings the systems into operation, including operation of applicable portions of the protection em, the transfer between normal and emergency power sources, and the operation of ciated systems.

Design Conformance design of the supplementary leak collection and release system permits periodic pressure and ctional testing of components, as described in Section 6.5.1.4.

2.44 Cooling Water (Criterion 44)

Criterion system to transfer heat from structures, systems, and components important to safety, to an mate heat sink shall be provided. The system safety function shall be to transfer the combined t load of these structures, systems, and components under normal operating and accident ditions.

itable redundancy in components and features, and suitable interconnections, leak detection, isolation capabilities shall be provided to assure that for onsite electric power system ration (assuming offsite power is not available) and for offsite electric power system operation uming onsite power is not available) the system safety function can be accomplished, ming a single failure.

3.1-31 Rev. 30

reactor plant component cooling water system, the charging pump cooling system, spent fuel l cooling and purification system and the safety injection pump cooling system, transfer heat m systems containing reactor coolant to the service water system. Together, these systems sfer heat to the ultimate heat sink from structures, systems, and components important to ty during normal and accident conditions.

se systems are designed with suitable redundancy in components, with leak protection, and h the capability to isolate redundant components. The systems are designed to satisfy the ling water requirements assuming a single failure and either a loss of onsite or offsite power.

igns of these systems are described in FSAR sections, as follows:

Title Section No.

AC Power Supply System 8.3.1 Service Water System 9.2.1 Reactor Plant Component Cooling Water System 9.2.2.1 Charging Pumps Cooling System 9.2.2.4 Safety Injection Pumps Cooling System 9.2.2.5 Ultimate Heat Sink 9.2.5 Spent Fuel Pool Cooling and Purification 9.1.3 2.45 Inspection of Cooling Water System (Criterion 45)

Criterion e cooling water system shall be designed to permit appropriate periodic inspection of ortant components, such as heat exchangers and piping, to assure the integrity and capability he system.

Design Conformance service water system (Section 9.2.1), the reactor plant component cooling water system ction 9.2.2.1), the charging pumps cooling system (Section 9.2.2.4), the safety injection pumps ling system (Section 9.2.2.5), and the spent fuel pool cooling and purification systems ction 9.1.3) are designed to permit appropriate periodic inspection in order to ensure the grity of the components and the systems as a whole. In addition, as the service water, reactor t component cooling water, charging pumps cooling, and spent fuel pool cooling and fication systems function almost continuously during normal unit operation, their capability integrity are continuously demonstrated. The safety injection pumps cooling system is rated periodically to assure its capability and integrity.

3.1-32 Rev. 30

Criterion e cooling water system shall be designed to permit appropriate periodic pressure and ctional testing to assure (1) the structural and leaktight integrity of its components, (2) the rability and the performance of the active components of the system, and (3) the operability of system as a whole and, under conditions as close to design as practical, the performance of the operational sequence that brings the system into operation for reactor shutdown and for

-of-coolant accidents, including operation of applicable portions of the protection system and transfer between normal and emergency power sources.

Design Conformance service water system (Section 9.2.1), reactor plant component cooling water system ction 9.2.2.1), charging pumps cooling system (Section 9.2.2.4), safety injection pumps ling system (Section 9.2.2.5), and the spent fuel pool cooling and purification system ction 9.1.3) are designed to permit periodic pressure and functional testing. With the exception he safety injection pumps cooling system, these systems operate during normal operation and tdown; thus, the structural and leaktight integrity of the system components, the operability performance of most of the active components, and the operability of the system as a whole continuously demonstrated. The active components that cannot be tested during normal em operation are tested during shutdown.

safety injection pumps cooling system, which is not normally in service, is periodically tested ssure structural and leaktight integrity of its components, the operability and performance of ctive components, and the operability of the system as a whole.

performance of the full operational sequence for the safety related portions of the above ems that brings each system into operation for reactor shutdown, LOCA, or loss of unit power valuated periodically in conjunction with the applicable portions of the protection system.

nsfer between normal and emergency power sources is discussed in Section 8.3.

2.47 (Criterion 47) erion 47 has not been promulgated by the NRC.

2.48 (Criterion 48) erion 48 has not been promulgated by the NRC.

2.49 (Criterion 49) erion 49 has not been promulgated by the NRC.

3.1-33 Rev. 30

Criterion e reactor containment structure, including access openings, penetrations, and the containment t removal system shall be designed so that the containment structure and its internal partments can accommodate without exceeding the design leakage rate and, with sufficient gin, the calculated pressure and temperature conditions resulting from any loss-of-coolant dent. This margin shall reflect consideration of (1) the effects of potential energy sources ch have not been included in the determination of the peak conditions, such as energy in steam erators and as required by Paragraph 50.44 energy from metal-water and other chemical tions that may result from degradation but not total failure of emergency core cooling ctioning, (2) the limited experience and experimental data available for defining accident nomena and containment responses, and (3) the conservatism of the calculational model and ut parameters.

Design Conformance containment structure is designed with a leakage rate shown in Table 1.3-3. The containment esigned to withstand, by a sufficient margin, loads above those that are conservatively ulated to result from a DBA as discussed in Section 6.2.1.

2.51 Fracture Prevention of Containment Pressure Boundary (Criterion 51)

Criterion e reactor containment boundary shall be designed with sufficient margin to assure that under rating, maintenance, testing, and postulated accident conditions (1) its ferritic materials ave in a nonbrittle manner and (2) the probability of rapidly propagating fracture is imized. The design shall reflect consideration of service temperatures and other conditions of containment boundary material during operation, maintenance, testing, and postulated dent conditions, and the uncertainties in determining (1) material properties, (2) residual, dy state, and transient stresses, and (3) size of flaws.

Design Conformance itic materials for the containment structure boundary are specified so that, when the liner, ipment latch, personnel lock, penetrations, and fluid system components, including valves uired to isolate the system are exposed to postulated accident, test, operating and normal ditions, the corresponding and resultant stress levels are below the maximum stress level mitted by the crack arrest temperature (CAT) curve of NRL Report 6900 for each applicable espondence temperature condition.

3.1-34 Rev. 30

ure 23 of NRL Report 6900 shows the fracture analysis diagram (FAD), which plots stress vs perature in excess of nil ductility transition temperature (NDTT). The containment structure r is designed so that no stress exceeds the CAT curve shown in this FAD. This approach is y conservative and ensures that flaws of any size are not propagated to a rapid (i.e., brittle) ture.

ertainties in the determination of NDTT are minimized by using the drop weight test, ASTM 8, for material 5/8 inch or thicker. Charpy V-notch tests are required for all material which m part of the containment structure boundary.

2.52 Capability for Containment Leakage Rate Testing (Criterion 52)

Criterion e reactor containment and other equipment which may be subjected to containment test ditions shall be designed so that periodic integrated leakage rate testing can be conducted at tainment design pressure.

Design Conformance design of the containment structure and related equipment, which are subjected to the tainment structure test conditions, as described in Section 6.2.6, allows for conducting odic integrated leakage rate testing of the containment structure.

2.53 Provisions for Containment Testing and Inspection (Criterion 53)

Criterion e reactor containment shall be designed to permit (1) appropriate periodic inspection of all ortant areas, such as penetrations, (2) an appropriate surveillance program, and (3) periodic ing at containment design pressure of the leaktightness of penetrations which have resilient s and expansion bellows.

Design Conformance reactor containment is designed to permit (1) appropriate periodic inspection of all important s, such as penetrations, (2) an appropriate surveillance program, and (3) periodic testing at tainment design pressure of the leaktightness of penetrations which have resilient seals and ansion bellows, as discussed in Section 6.2.6 (Containment Leakage Testing).

3.1-35 Rev. 30

Criterion ping systems penetrating primary reactor containment shall be provided with leak detection, ation, and containment capabilities having redundancy, reliability, and performance abilities which reflect the importance to safety of isolating these piping systems. Such piping ems shall be designed with a capability to test periodically the operability of the isolation es and associated apparatus and to determine if valve leakage is within acceptable limits.

Design Conformance piping systems penetrating the containment structure are designed to minimize leakage.

tainment isolation valves provide the capability to seal most penetrations redundantly; tion 6.2.4 describes the few exceptions in detail. Pressure taps provide the capability to orm a Type C (10 CFR 50 Appendix J) test to measure containment isolation valve leakage s, as outlined in Section 6.2.4.

2.55 Reactor Coolant Pressure Boundary Penetrating Containment (Criterion 55)

Criterion ch line that is part of the reactor coolant pressure boundary and that penetrates primary reactor tainment shall be provided with containment isolation valves as follows, unless it can be onstrated that the containment isolation provisions for a specific class of lines, such as rument lines, are acceptable on some other defined basis:

1. One locked closed isolation valve inside and one locked closed isolation valve outside containment; or
2. One automatic isolation valve inside and one locked closed isolation valve outside containment; or
3. One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or
4. One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.

lation valves outside containment shall be located as close to containment as practical and n loss of actuating power, automatic isolation valves shall be designed to take the position that vides greater safety.

3.1-36 Rev. 30

quate safety. Determination of the appropriateness of these requirements, such as higher lity in design, fabrication, and testing, additional provisions for inspection, protection against e severe natural phenomena, and additional isolation valves and containment, shall include sideration of the population density, use characteristics, and physical characteristics of the site irons.

Design Conformance lines that are part of the reactor coolant pressure boundary and that penetrate the containment cture are provided with containment isolation valves in accordance with the above criterion, escribed in Section 6.2.4. Valves outside the containment structure are located as close as tical to the containment structure. All containment penetrations and isolation valves are ected against possible environmental effects including missiles. The isolation valves are ject to periodic Type C tests (10 CFR 50, Appendix J), as outlined in Section 6.2.4, and, matic valves, upon loss of actuating power, take the position that provides the greatest safety.

re are no additional requirements for the mechanical design of those lines that are part of the tor coolant pressure boundary and that penetrate the containment structure beyond those uired by the applicable standards and codes.

2.56 Primary Containment Isolation (Criterion 56)

Criterion ch line that connects directly to the containment atmosphere and penetrates primary reactor tainment shall be provided with containment isolation valves as follows, unless it can be onstrated that the containment isolation provisions for a specific class of lines, such as rument lines, are acceptable on some other defined basis:

1. One locked closed isolation valve inside and one locked closed isolation valve outside containment; or
2. One automatic isolation valve inside and one locked closed isolation valve outside containment; or
3. One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or
4. One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.

3.1-37 Rev. 30

vides greater safety.

Design Conformance lines that connect to the containment atmosphere are provided with containment isolation es in accordance with the above criterion except for those lines identified otherwise in tion 6.2.4. Valves outside the containment structure are located as close as possible to the tainment structure. The isolation valves are subject to periodic Type C tests (10 CFR 50, endix J) as outlined in Section 6.2.4 and, upon loss of actuating power, take the position that vides the greatest safety.

rument line penetrations are in accordance with Regulatory Guide 1.11 (Section 6.2.4) and e a remote manual isolation valve outside of the containment structure.

2.57 Closed System Isolation Valves (Criterion 57)

Criterion ch line that penetrates primary reactor containment and is neither part of the reactor coolant sure boundary nor connected directly to the containment atmosphere shall have at least one tainment isolation valve which shall be either automatic, or locked closed, or capable of ote manual operation. This valve shall be outside containment and located as close to the tainment as practical. A simple check valve may not be used as the automatic isolation valve.

Design Conformance lines connected to closed systems are provided with at least one containment isolation valve ted as close as possible to the outside of the containment structure as described in tion 6.2.4. The isolation valves are subject to periodic tests (10 CFR 50, Appendix J) as ined in Section 6.2.4. These automatic isolation valves, upon loss of actuating power, take the ition that provides the greatest safety.

2.58 (Criterion 58) erion 58 has not been promulgated by the NRC.

2.59 (Criterion 59) erion 59 has not been promulgated by the NRC.

3.1-38 Rev. 30

Criterion e nuclear power unit design shall include means to control suitably the release of radioactive erials in gaseous and liquid effluents and to handle radioactive solid wastes produced during mal reactor operation, including anticipated operational occurrences.

ficient holdup capacity shall be provided for retention of gaseous and liquid effluents taining radioactive materials, particularly where unfavorable site environmental conditions be expected to impose unusual operational limitations upon the release of such effluents to the ironment.

Design Conformance ll cases, the design for radioactivity control is based on:

1. The requirements of 10 CFR 20, 10 CFR 50, and Appendix I to 10 CFR 50 for normal operations and for any transient situation that might reasonably be anticipated to occur.
2. 10 CFR 50.67 dose level guidelines for potential accidents of extremely low probability of occurrence release paths, including ventilation and process streams, are monitored and controlled as cribed in Section 11.5.

activity level of the radioactive gaseous waste effluents subsequent release through the

-foot Millstone stack are monitored (Section 11.3.2.4). Under conditions of concurrent fuel ure and steam generator tube leakage, some radioactive gas would be present and suitably trolled in the steam jet air ejector discharge in the condenser air removal system (Section 4.2) and in the flow from the steam packing exhauster fan in the turbine generator gland seal exhaust system (Section 10.4.3). The steam jet air ejector discharge is directed to the lstone stack while the seal steam packing exhauster fan discharges through the condensate shing enclosure roof.

trol of liquid waste effluents (Sections 11.2 and 11.5) is maintained by batch processing of all ids, sampling before discharge, and a controlled rate of release. Liquid effluents are monitored radioactivity and rate of flow. Radioactive liquid waste system capacities are sufficient to dle any expected transient in the processing of liquid waste.

d wastes are prepared for offsite disposal by either compaction or solidification ction 11.4). Solid waste is prepared for shipment by placement in properly labeled containers meet applicable NRC and Department of Transportation dose rate requirements as detailed in CFR 71, 49 CFR 170-178, and Section 11.4.

3.1-39 Rev. 30

Criterion e fuel storage and handling, radioactive waste, and other systems which may contain oactivity shall be designed to assure adequate safety under normal and postulated accident ditions. These systems shall be designed (1) with a capability to permit appropriate periodic ection and testing of components important to safety, (2) with suitable shielding for radiation ection, (3) with appropriate containment, confinement, and filtering systems, (4) with a dual heat removal capability having reliability and testability that reflects the importance to ty of decay heat and other residual heat removal, and (5) to prevent significant reduction in storage coolant inventory under accident conditions.

Design Conformance ety related components in the fuel storage and handling system (Section 9.1.3.4) are designed llow periodic inspection and testing to ensure proper operation. Performance of components ortant to safety in the radioactive liquid and gaseous waste systems is verified by extensive cess fluid analysis and continuous radiation monitoring of gaseous effluents, respectively.

new and spent fuel storage areas are designed to meet the requirements of 10 CFR 20 in viding radiation shielding for operating personnel during new and spent fuel transfer and age. The fuel transfer canal and spent fuel pool wall thickness are sufficient to shield adjacent k areas to meet the requirements of 10 CFR 20 for personnel access during actual fuel transfer.

te storage and processing facilities in the auxiliary building and the waste disposal building shielded to meet the requirements of 10 CFR 20 for operating personnel. Periodic surveys by ation protection personnel and continuously operated radiation monitors located in areas cted to afford maximum personnel protection (Section 12.1) ensure that radiation design ls are not exceeded during the operating lifetime of the unit.

w and spent fuel handling systems are designed to preclude gross mechanical failures which ld lead to significant radioactivity releases. Floor and equipment drains are provided to collect age which might occur from valve stem leakoffs, pump seals, and other equipment, and to sfer the leakage to one of the building sumps for eventual processing by the liquid waste em.

iation gases and particulates released from components are collected by the reactor plant ted vents system. Uncontrolled leakage of radioactive gases and particulates which may leak m spent fuel, radioactive waste, or components containing radioactive fluids is collected and ted by the respective building ventilation filtration system (Section 9.4) or supplementary age collection and release system (Section 6.5.1). All discharges from these systems are nitored for radioactivity.

ay heat from spent fuel is dissipated in the water of the spent fuel pool and subsequently oved by the cooling portion of the fuel pool cooling and purification system (Section 9.1.3).

undancy of fuel pool cooling and purification system components ensures reliability in 3.1-40 Rev. 30

cated and alarmed. Special tests are not required because at least one pump and heat hanger are normally in operation when spent fuel is stored in the spent fuel pool.

piping connected to the spent fuel pool is designed so that an acceptable water level is ntained in the event of a pipe rupture. Instrumentation to annunciate spent fuel pool water l changes above or below preset levels is provided on the fuel pool control panel in the main trol room. Redundancy of makeup water sources ensures adequate supply and availability of eup to the spent fuel pool even under loss of normal electrical power.

2.62 Prevention of Criticality in Fuel Storage and Handling (Criterion 62)

Criterion iticality in the fuel storage and handling system shall be prevented by physical systems or cesses, preferably by use of geometrically safe configurations.

Design Conformance icality is prevented in the new fuel storage racks by a combination of geometry and poison erial as described in Section 9.1.1 and 4.3.2.6.

icality is prevented in the spent fuel storage area by the physical separation of fuel assemblies, ts on the enrichment, burnup and decay times of the fuel, and the use of fixed neutron poisons egion 1 and 2. Soluble boron in the spent fuel pool water is credited for certain accident ditions. Sections 9.1.1 and 9.1.2 discuss criticality prevention in more detail.

2.63 Monitoring Fuel and Waste Storage (Criterion 63)

Criterion propriate systems shall be provided in fuel storage and radioactive waste systems and ciated handling areas (1) to detect conditions that may result in loss of residual heat removal ability and excessive radiation levels and (2) to initiate appropriate safety actions.

Design Conformance spent fuel pool water temperature is continuously monitored. Safety related temperature cators are provided in the control room where alarms are also provided should the water perature increase above a preset level. Spent fuel pool temperature is also indicated and med at the fuel pool cooling and purification panel (FP) in the fuel building.

ety related spent fuel pool low water level indicating lights are displayed in the control room.

itionally, alarms are provided in the control room and at the FP in the fuel building should the l increase or decrease beyond preset levels. The spent fuel pool cooling system cooler outlet 3.1-41 Rev. 30

m is actuated at the FP and the fuel pool cooling system trouble alarm is actuated in the trol room.

the event of high temperature, low flow, or abnormal spent fuel pool level indication, inistrative procedures provide for checking the operating status and integrity of the spent fuel l and support cooling systems, inspecting for spent fuel pool leakage, and ensuring that ective measures are taken to restore all system parameters to normal.

fuel pool demineralizer can be aligned to either the SFP or the RWST. The specific ductivities of the SFP and the RWST (the fuel pool demineralizer influent sources) are nitored and recorded weekly. The fuel pool demineralizer effluent specific conductivity is nitored and recorded monthly. In the event of abnormal conductivity levels in the SFP, RWST, he effluent of the fuel pool demineralizer, administrative procedures provide for taking itional samples to assist in determining the source and/or cause of the abnormal conductivity ls and for ensuring that corrective measures are taken.

iation levels in the area of spent fuel storage are continuously monitored by radiation ctors located around the periphery of the storage areas. Other continuously operating ation detectors are located in the fuel and waste disposal buildings in areas best suited for ting operating personnel of high local radiation levels. Radiation levels in excess of the preset es for either the fuel building or waste storage areas initiate alarms, both locally and in the trol room.

he event of a high airborne gross activity alarm, the fuel building ventilation system exhaust ction 9.4.2) is remote manually diverted to its exhaust filtration system. In the waste disposal ding, the ventilation system exhaust (Section 9.4.5) can be remote manually diverted to the iliary building filtration system upon receipt of a high airborne activity alarm.

2.64 Monitoring Radioactivity Releases (Criterion 64)

Criterion eans shall be provided for monitoring the reactor containment atmosphere, spaces containing ponents for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the t environs for radioactivity that may be released from normal operations, including cipated operational occurrences, and from postulated accidents.

Design Conformance containment atmosphere is monitored during normal and transient operations of the reactor t by the containment structure particulate and gas monitor located in the upper level of the iliary building (Section 12.3.4) or by grab sampling. Normal unit effluent discharge paths are nitored during normal plant operation by the ventilation particulate samples and gas monitors he auxiliary building and engineered safety buildings (Section 11.5). After a postulated 3.1-42 Rev. 30

he containment structure including the areas that contain loss-of-coolant accident fluids. In ition, the service water outlet from each pair of containment recirculation coolers is monitored nsure that any leakage of radioactive fluids into the service water system is detected ction 11.5). Radioactivity levels in the environs are controlled during normal and accident ditions by the various radiation monitoring systems (Sections 11.5 and 12.3.4) and monitored he collection of samples as part of the offsite radiological monitoring program.

3 REFERENCE FOR SECTION 3.1 3.1-43 Rev. 30

1 SEISMIC CLASSIFICATION mic Category I structures, systems, and components are those necessary to ensure:

1. The integrity of the reactor coolant pressure boundary
2. The capability to shut down the reactor and maintain it in a safe shutdown condition, or
3. The capability to prevent or mitigate the consequences of accidents which could result in potential off site exposures comparable to the guideline exposure of 10 CFR 100.

mic Category I structures, systems, and components are designed to remain functional during fe shutdown earthquake (SSE). Strain limits in excess of yield are allowed provided safety ctions are maintained. Seismic Category I structures, systems, and components are also gned to be well within the elastic limit for a vibratory motion at 50 percent of the SSE. This uirement is called the operating basis earthquake (OBE).

SSE and the OBE are described in Section 2.5. The seismic design of Seismic Category I ems and components is described in Sections 3.7 and 3.10 and of structures in Section 3.8.

mic Category I structures, systems, and components are listed in Table 3.2-1.

seismic classification of structures, systems, and components complies with Regulatory de 1.29.

ctures, systems, and components designated Seismic Category I in accordance with NRC ulatory Guide 1.29 are listed in Table 3.2-1.

2 SYSTEM QUALITY GROUP CLASSIFICATION containment structure and safety related fluid systems, listed in Table 3.2-1, are classified ording to the classes listed below. Supports are in the same safety class as the components for ch they provide support if failure of the support could cause a loss of a safety function ciated with the supported component.

Safety Class 1 ety Class 1 (SC-1) applies to reactor coolant pressure boundary components whose failure ng normal reactor operation would prevent orderly reactor shutdown and cooldown. The tor coolant pressure boundary is defined in Paragraph 50.2 of 10 CFR 50 as being all those sure-containing components ... such as pressure vessels, piping, pumps, and valves, which 3.2-1 Rev. 30

2. Connected to the reactor coolant system up to and including any and all of the following:
a. The outermost containment isolation valve in system piping which penetrates the primary reactor containment.
b. The second of two valves normally closed during normal reactor operation in system piping which does not penetrate the primary reactor containment.
c. The reactor coolant system safety and relief valves.

ording to Section 50.55a of 10 CFR 50, components which are connected to the reactor lant system and are part of the reactor coolant pressure boundary may be downgraded in safety s provided that:

1. In the event of postulated failure of the component during normal reactor operation, the reactor can be shut down and cooled down in an orderly manner, assuming makeup is provided by the reactor coolant makeup system only, or
2. The component is or can be isolated from the reactor coolant system by two valves (both closed, both open, or one closed and the other open). Each open valve must be capable of automatic actuation and, assuming the other valve is open, its closure time must be such that, in the event of postulated failure of the component during normal reactor operation, each valve remains operable and the reactor can be shut down and cooled down in an orderly manner, assuming makeup is provided by the reactor coolant makeup system only.

ut down and cooled down in an orderly manner is taken as an action where the pressurizer s not empty, assuming normal makeup is available.

Safety Class 2 ety Class 2 applies to:

1. Pressure containing components of the reactor coolant pressure boundary not covered in Safety Class 1
2. The reactor containment, including those valves and components of systems used to affect isolation of the reactor containment
3. Components necessary to the system safety function of the following:
a. Residual heat removal system 3.2-2 Rev. 30
c. Reactor containment cooling spray systems
d. Emergency core cooling system including injection and recirculation modes
e. Containment air purification and cleanup systems used to clean up the containment atmosphere and, thereby, to reduce the radioactivity present in leakage from the containment structure, regardless of whether such systems are located inside or outside the containment structure
f. Portions of the main steam and feedwater systems extending from and including the secondary side of the steam generator up to the first restraint beyond the main steam isolation and feedwater isolation valves in the main steam valve building.
g. Portions of the auxiliary feedwater system
h. Hydrogen recombiner system Safety Class 3 ety Class 3 applies to:
1. Components necessary to the safety system function of the following:
a. Portions of the reactor auxiliary systems that provide boric acid for the reactor coolant letdown and makeup loop
b. Portions of the auxiliary feedwater system
c. Portions of the reactor plant component cooling and service water systems that transfer heat from components whose heat removal capability serves a safety function or are required for orderly reactor shutdown. This generally includes portions of the reactor plant component cooling and service water systems serving the emergency core cooling, containment heat removal, and residual heat removal systems; also cooling systems for the control room and safety related electrical equipment. If a heat transfer component is classified as Safety Class 2 or 3 just to maintain a flow path or because its failure would cause uncontrollable release of gaseous radioactivity, then the cooling water lines to and from the component do not have to be assigned a safety class.
d. Fuel pool cooling system 3.2-3 Rev. 30

IEEE-387-1977)

f. Portions of the main steam system that supply steam to the turbine drive of the steam generator auxiliary feedwater pump
g. Air purification and cleanup systems used to clean up the atmosphere after leakage from the containment structure and other air purification and cleanup systems used after accidents
2. Portions of any system whose failure would result in calculated potential exposure comparable to the guideline exposure of 10 CFR 100. The designation of SC 3 applies to portions of the following systems, the failure of which would result in uncontrollable release to the environment of significant gaseous radioactivity normally held up, and meets the intent of 10 CFR 100:
a. Reactor coolant auxiliary systems that form the reactor coolant letdown and makeup loop not covered by Safety Class 2
b. Portions of the radioactive waste processing and handling systems Nonnuclear Safety Class s class applies to portions of the unit not covered in Safety Classes 1, 2, or 3. This class udes most of the steam and power conversion systems, radioactive liquid waste system, and ions of the boron recovery system containing degassed liquid.

Quality Group Classification System quality group classification system and its relation to industrial codes conform with ulatory Guide 1.26 (Section 1.8) with the following exceptions:

1. The safety class terminology of ANSI N18.2-1973 and ANSI N18.2a-1975 is used instead of the quality group terminology. Thus, the terms Safety Class 1, Safety Class 2, Safety Class 3, and non-nuclear safety (NNS) are used instead of Quality Groups A, B, C, and D, respectively.
2. Regarding Regulatory Guide 1.26, Positions C.1.e and C.2.c, one safety valve designed, manufactured, and tested in accordance with ASME III, Division 1 (i.e.,

a code safety valve) is considered acceptable as the boundary between the reactor coolant pressure boundary and lower safety class or NNS line.

lstone 3 has constructed components in safety-related systems to the ASME Boiler and ssure Vessel Code,Section III, Nuclear Power Plant Components, Division I, as follows:

3.2-4 Rev. 30

2. Quality Group B (Section III, Class 2) components comply with the requirements of Subsection NA-2130 of the code
3. Quality Group C (Section III, Class 3) components comply with the requirements of Subsection NA-2130 of the code except for:
a. Rubber expansion joints 3SWP*EJ6A, B, C, and D, which are procured in accordance with Appendix B requirements or commercially dedicated.
b. Valves SWP*V673, SWP*V674, SWP*V23, SWP*V22, SWP*V55, and SWP*V56 which are not N stamped but were procured per the requirements of Generic Letter 89-09.
c. The service water system supply and return piping for the post accident sample cooler, which is designed to ANSI B31.1 requirements and is seismically qualified.
d. The service water cubicle sump drain lines which are designed to ANSI 31.1 requirements and are seismically qualified.

boundary of jurisdiction of ASME Code Section III, Classes 2 and 3 process piping extends nd includes the root valve. The appropriate safety class extends from the root valve to the sing instrument. Seismic Category I supports are employed for Safety Classes 2 and 3 rument tubing. The tubing used is one-half inch or less in diameter. The requirements for ety Classes 2 and 3 instrument tubing are listed in Tables 3.2-2, 3.2-3 and 3.2-4.

safety classes of safety-related fluid systems are given in Table 3.2-1. In addition, the safety s boundaries are shown on the various piping and instrumentation diagrams (P&IDs) located ughout this safety analysis report. The following line designations are used on P&IDs to cate these boundaries:

ety Class 1, SC-1 (Quality Group A) = line designator, - 1 ety Class 2, SC-2 (Quality Group B) = line designator, - 2 ety Class 3, SC-3 (Quality Group C) = line designator, - 3 nuclear Safety Class, NNS (Quality Group D) = line designator - 4 QA classification process for the identification of structures, systems, and components as lear safety-related and non safety-related is controlled via an Engineering Design cification. The methodology for system identifications, system interfaces, line designators, 3.2-5 Rev. 30

ety class boundaries on FSAR figures (P&IDs) are extended to the first piping restraint beyond indicated boundary.

3 QUALITY ASSURANCE CATEGORIES le 3.2-1 lists Millstone 3 structures, systems, and components which are classified QA egory I.

4 OTHER CLASSIFICATION SYSTEMS Tornado Design Classification tornado design classification conforms to Regulatory Guide 1.117 (Table 1.8-1). Table 3.2 1 the tornado protection criteria for Category I structures, systems, and components.

tion 3.8.4.1 describes those structures which are not safety-related but whose failure could uce to an unacceptable safety level the functional capability of any plant feature included in items listed in the appendix of Regulatory Guide 1.117.

ASME Code Classes ME Code Classes 1, 2, 3, and MC are defined in the ASME Boiler and Pressure Vessel Code, tion III, Nuclear Power Plant Components, 1971. With regard to pumps, valves, piping, tanks, pressure vessels, there is a direct one-for-one correlation between Code Classes 1, 2, and 3, SC 1, 2, and 3 (Section 3.2.2). Exceptions to this are noted on the engineering document. The es for the concrete portions of the containment structure, classified as SC 2, are specified in tion 3.8. Metal containment systems, such as the personnel access lock, are classified as Code ss MC. Paragraphs NE 1100 and NE 1140 of ASME Section III delineate the portions of the tainment structure classified as Code Class MC.

code classes of structures, systems, and components are given in Table 3.2 1.

Engineered Safety Features ineered safety features (ESF) are those systems used to directly mitigate the consequences of ajor loss-of-coolant accident, up to and including the design basis accident, which is the ble ended rupture displacement of the largest pipe in the reactor coolant pressure boundary.

following types of systems are classified as ESF:

containment systems, including containment structure and containment enclosure building; ESF actuation systems; 3.2-6 Rev. 30

containment heat removal systems; containment combustible gas control systems; containment isolation systems; and supplementary leak collection and release system.

ould be noted that systems supporting the above (e.g., cooling systems and electrical systems) not classified as ESF, but are safety-related (QA Category I).

IEEE Classification Systems E Std-308-1971, IEEE Standard Criteria for Class 1E Electric Systems for Nuclear Power erating Stations, delineates three classes of structures and equipment (Classes I, II, and III) one class of electric systems (Class 1E). Class 1E electric systems provide the electric power d to shut down the reactor and limit the release of radioactive material following a design basis nt (i.e., postulated events used in the design to establish the performance requirements of the ctures and systems). Class II and III electric equipment may be supplied from Class 1E tric systems.

5 TABULATION OF CODES AND CLASSIFICATIONS s subsection provides a concise compilation of the safety classes, codes, and design sifications of the structures, systems, and components in Table 3.2-1 that are QA Category I.

Category I structures, systems, and components are defined in Table 3.2-1. Seismic Category uctures, systems, and components are defined in Section 3.2.1 and are designed in accordance h the seismic design criteria of Sections 3.7, 3.8, and 3.10. Because the definitions of Seismic egory I and QA Category I are different, the various structures, systems, and components ng in one category do not necessarily fall in the other. QA Category I items which are not mic Category I are pointed out in the notes column of Table 3.2-1.

safety class, as defined in Section 3.2.2, is indicated for each QA Category I structure, em, and component to which it applies. The codes applicable to the QA Category I ponents are also presented; in addition, Table 3.2-1 indicates which structures, systems, and ponents are designed for tornado resistance and are protected from tornado effects.

3.2-7 Rev. 30

(Symbols and references are defined at the end of this Table)

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes STRUCTURES Containment Structure Reinforced Concrete Substructure 2 ACI 318-71 N/A CS P Reinforced Concrete Superstructure 2 ACI 318-71 N/A CS D Containment Enclosure Building N/A ACI 318-71 N/A CS D Containment Enclosure Building N/A AISC Steel N/A CEB N/A Construction Manual, 7th Edition Reinforced Concrete Interior Shields 2 ACI 318-71 N/A CS P and Walls Containment Structure Liner 2 ASME III MC CS P (Note 2)

Piping and Duct Penetrations 2 ASME III 2 CS P (Note 2)

Electrical Penetrations 2 IEEE-317 N/A CS P Personnel Access Lock 2 ASME III MC CS P (Note 2)

Equipment Hatch 2 ASME III MC CS P (Note 2) (Note 3)

Containment Dome Closure 2 ASME III MC CS D (Note 2)

Containment Sump Strainer 2 AISC 6th N/A CS P Edition 3.2-8 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Circulating Water Discharge Tunnel N/A ACI 318-71 N/A OY D Demineralizer Water Storage Tank N/A ACI 318-71 N/A OY D Enclosure Cable Tunnel from Auxiliary Building N/A ACI 318-71 N/A SB D to Control Building Main Steam Valve Building N/A ACI 318-71 N/A MSV D Engineered Safety Features Building N/A ACI 318-71 N/A ESB D Auxiliary Building Reinforced Concrete Structure N/A ACI 318-71 N/A AB D MCC and Rod Control Area N/A ACI 318-71 N/A AB P Hydrogen Recombiner Building N/A ACI 318-71 N/A HRB D Fuel Building Reinforced Concrete Structure N/A ACI 318-71 N/A FB D (Note) Capital new fuel pool, spent fu pool, pipe tunnel below grade, and the fuel pool cooler and pump area only will be Seismi Category I and protected from tornado missiles.

3.2-9 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Steel Roof Structure N/A ACI 318-71 N/A FB D (Note) Not missile protected. Designe to withstand tornado winds on (Section 3.8.4.3).

Spent Fuel Pool Liner (including N/A ASME VIII N/A FB P Gates)

Spent Fuel Pool Racks 3 ASME III 3 FB P Control Building N/A ACI 318-71 N/A CB D Emergency Generator Enclosures N/A ACI 318-71 N/A EGE D Circulating and Service Water N/A ACI 318-71 N/A CSP D (Note) Service water pump cubicles Pumphouse only.

West Retaining Wall N/A ACI 318-71 N/A CSP D Emergency Generator Fuel Oil Transfer N/A ACI 318-71 N/A OY D Pump Vault SYSTEMS Reactor Coolant System Reactor Vessel 1 ASME III 1 CS P 3.2-10 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Reactor Internals, Core Supports 1 N/A N/A CS P Reactor core supports and internals are designed to the intent of Subsection NG of ASME III.

Reactor Head Vent Piping and Valves 1 ASME III 1 CS P Fuel Assemblies 1 N/A N/A CS P Full Length Control Rod Drive 1 ASME III 1 CS P Mechanism Housing Part Length Control Rod Drive 1 ASME III 1 CS P Mechanism Housing CRDM Pressure Vessel 1 ASME III 1 CS P CRDM Latch Assembly NNS N/A N/A CS P CRDM Drive Rod Assembly NNS N/A N/A CS P Control Rods 1 N/A N/A CS P Steam Generator Tube Side 1 ASME III 1 CS P Shell Side 2 ASME III 2 CS P Reactor Coolant Stop Valves 1 ASME III 1 CS P Pressurizer 1 ASME III 1 CS P 3.2-11 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Pressurizer Heaters (Groups A&B) 1 No Code N/A CS P Pressurizer Spray Valves and Piping 1 ASME III 1 CS P Reactor Coolant Hot and Cold Leg 1 ASME III 1 CS P Piping, Fittings, and Fabrication Surge Pipe, Fittings, and Fabrication 1 ASME III 1 CS P Loop Bypass Line 1 ASME III 1 CS P Reactor Coolant Thermowell 1 ASME III 1 CS P Reactor Coolant Thermowell Boss 1 ASME III 1 CS P Safety Valves 1 ASME III 1 CS P Relief Valves (PORV) 1 ASME III 1 CS P Actuators are qualified to IEEE-323-74.

Power Operated Block Valves 1 ASME III 1 CS P Actuators are qualified to IEEE-323-74.

Valves to Reactor Coolant System 1 ASME III 1 CS P Boundary Control Rod Drive Mechanism Head 1 ASME III 1 CS P Adapter Plugs Reactor Coolant Pump Reactor Coolant Pump Casing 1 ASME III 1 CS P 3.2-12 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Main Flange 1 ASME III 1 CS P Thermal Barrier 1 ASME III 1 CS P No. 1 Seal Housing 1 ASME III 1 CS P No. 2 Seal Housing 2 ASME III 1 CS P No. 2 Seal Housing is permitte to be Code Class 2; however, i is supplied as Code Class 1 by Westinghouse.

No. 3 Seal Housing 2 ASME III 2 CS P Pressure Retaining Bolting 1 ASME III 1 CS P Reactor coolant pump seal bolting is NSS.

Reactor Coolant Pump Seals N/A N/A N/A CS P Special requirements are included in the specifications.

Reactor Coolant Pump Motor RCP motor is not safety-relate Shaft Coupling 2 NEMA MG1 N/A CS P Spool Piece 2 NEMA MG1 N/A CS P Armature 2 No Code N/A CS P Flywheel 2 No Code N/A CS P Motor Bolting 2 No Code N/A CS P Upper Oil Cooler 3.2-13 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Tube Side-Component Cooling 3 ASME III 3 CS Water Shell Side-oil 2 ASME III 2 CS P Lower Oil Cooler Tube Side-Component Cooling 3 ASME III 3 CS P Water Shell Side-oil 2 ASME III 2 CS P Air-Water Coolers 3 ASME III 3 CS P Piping and Valves*** See Note CS P All piping, valves, and other pressure retaining equipment which are inside or part of the reactor coolant pressure boundary (Section 3.2.2) are SC-1 and ASME III. Other equipment outside the bounda is ASME III 2 or NNS.

Instrumentation and Controls required N/A IEEE-279-71 N/A CS / CR / P to perform safety function in QA IRR / MRC Category I portions of system***

Inadequate Core Cooling N/A IEEE-323-74 N/A CS / CR P Instrumentation 3.2-14 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes IEEE-344-75 Subcooled/Superheat Margin Monitor Core Exit Thermocouple Heated Junction Thermocouple Core Exit Thermocouple Pressure Boundary/Reactor Internal 1 ASME III 1 CS P Modifications for Inadequate Core Cooling Instrumentation Supports for QA Category I Same as component being supported.

Components*

Chemical and Volume Control System Regenerative Heat Exchanger**

Tube Side 2 ASME III 2 CS P Shell Side 2 ASME III 2 CS P Letdown Heat Exchanger**

Tube Side 2 ASME III 2 AB P Shell Side 3 ASME III 3 AB P Mixed Bed Demineralizer*** 3 ASME III 3 AB P 3.2-15 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Cation Bed Demineralizer** 3 ASME III 3 AB P Reactor Coolant Filter** 2 ASME III 2 AB P Volume Control Tank** 2 ASME III 2 AB P Centrifugal Charging Pump** 2 ASME III 2 AB P Seal Water Injection Filter** 2 ASME III 2 AB P Letdown Orifices** 2 ASME III 2 CS P Excess Letdown Heat Exchanger**

Tube Side 2 ASME III 2 CS P Shell Side 3 ASME III 3 CS P Seal Water Return Filter** 2 ASME III 2 AB P Seal Water Heat Exchanger**

Tube Side 2 ASME III 2 AB P Shell Side 3 ASME III 3 AB P Letdown Filter 3 ASME III 3 AB P Boric Acid Tanks

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Boric Acid Tank Orifice* 3 ASME III 3 AB P Instrumentation and Controls See Note AB / CS P Pressure retaining portions of inline instruments are the sam as connecting piping.

Moderating Heat Exchanger**

Tube Side 3 ASME III 3 AB P Shell Side 3 ASME III 3 AB P Letdown Chiller Heat Exchanger**

Tube Side 3 ASME III 3 AB P Shell Side NNS ASME VIII N/A AB P Letdown Reheat Heat Exchanger**

Tube Side 2 ASME III 2 AB P Shell Side 3 ASME III 3 AB P Thermal Regeneration 3 ASME III 3 AB P Demineralizer**

Piping and Valves inside RCPB* 1 ASME III 1 CS P Figures 9.2-5 (P&ID 105), 9.3 No. 1 Seal Water Injection Lines (P&ID 103), and 9.3-8 (P&ID 104) delineate SC boundaries.

3.2-17 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Piping and Valves outside RCPB* See Note Same references as inside RCPB.

Letdown and charging lines 2 ASME III 2 AB / CS P Charging pump alternate minimum flow lines downstream of isolation valve 3CHS*MV8512 A/B to the RWST are ANS Safety Class Seal water injection and return lines 2 ASME III 2 AB / CS P Mixed-bed and cation-bed lines 3 ASME III 3 AB P Boric acid lines 3 ASME III 3 AB P Thermal regeneration lines 3 ASME III 3 AB P Emergency boration 2 ASME III 2 AB P Supports for QA Category I Same as component being supported.

Components*

Residual Heat Removal System Residual Heat Removal Pump** 2 ASME III 2 ESB P Residual Heat Exchanger**

Tube Side 2 ASME III 2 ESB P 3.2-18 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Shell Side 3 ASME III 3 ESB P Piping and Valves inside RCPB* 1 ASME III 1 CS / ESB P Figure 5.4-5 delineates SC boundaries.

Piping and Valves outside RCPB* 2 ASME III 2 CS / ESB P Figure 5.4-5 (P&ID 112) delineates SC boundaries.

Valve interlocks for over-pressurization N/A IEEE-279-71 N/A CS / ESB / P Section 7.6.2.2 for criteria protection IRR / ESR / comments.

CR Supports for QA Category I Same as components being supported.

Emergency Core Cooling System Accumulators** 2 ASME III 2 CS P Safety Injection Pumps** 2 ASME III 2 ESB P Piping and Valves inside RCPB* 1 ASME III 1 CS P Figures 5.4-5 (P&ID 112), 6.3 (P&ID 113), and 9.2-4 (P&ID 114) delineate SC boundaries.

Piping and Valves outside RCPB* Figures same as inside RCPB.

High and low pressure safety 2 ASME III 2 CS / ESB P injection lines Supports for QA Category I Same as component being supported.

Components

  • 3.2-19 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Hydrogen Recombiner System*

Hydrogen Recombiner Blower 2 ASME III 2 HRB P Preheater Coil 2 ASME III 2 HRB P Air Cooler 2 ASME III 2 HRB P Thermal Recombiner 2 ASME III 2 HRB P Piping and Valves 2 ASME III 2 CS / HRB P Figure 6.2-36 (P&ID 115) delineates SC boundaries.

Instrumentation and Controls required N/A IEEE-279-71 N/A HRB P to perform safety function Supports for QA Category I Same as component being supported.

Components

  • Quench Spray System
  • Refueling Water Storage Tank (RWST) 2 ASME III 2 OY N/A Quench Spray Pumps 2 ASME III 2 ESB N/A Piping and Valves, excluding RWST 2 ASME III 2 ESB / CS N/A Figure 6.2-36 (P&ID 115) recirculation lines and test piping delineates SC boundaries.

(NNS) 3.2-20 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Instrumentation and Controls required N/A IEEE-279-71 N/A OY / ESB / P to perform safety function in QA IRR / ESR /

IEEE-323-74 Category I portions of system CR IEEE-336-71 Supports for QA Category I Same as component being supported.

Components*

Containment Recirculation System*

Containment Recirculation Pumps 2 ASME III 2 ESB P Containment Recirculation Coolers Tube Side 3 ASME III 3 ESB P Shell Side 2 ASME III 2 ESB P Piping and Valves, excluding test 2 ASME III 2 ESB / CS P Figure 5.4-5 (P&ID 112) piping delineates SC boundaries.

Instrumentation and Controls required N/A IEEE-279-71 N/A CS / E / P to perform safety function in QA IRR / ESR /

IEEE-323-74 Category I portions of system ESB / CR IEEE-336-71 Supports for QA Category I Same as component being supported.

Components 3.2-21 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Emergency Generator Fuel Oil System*

Emergency Generator Fuel Oil Day 3 ASME III 3 EGE P Tanks Emergency Generator Fuel Oil Transfer 3 ASME III 3 OY P Located underground in Pumps emergency generator fuel oil transfer pump vault.

Emergency Generator Fuel Oil Storage 3 ASME III 3 OY P Underground vault.

Tanks Piping and Valves 3 ASME III 3 OY / EGE P Figure 9.5-2 (P&ID 117) delineates SC boundaries.

Instrumentation and Controls required N/A IEEE-279-71 N/A EGE / OY / P to perform safety function. IRR /

IEEE-323-74 ESR / CR IEEE-336-71 Supports for QA Category I Same as component being supported.

Components*

Emergency Diesel Engine Air Start System*

Air Receiver Tanks 3 ASME III 3 EGE P Piping and Valves 3 ASME III 3 EGE P 3.2-22 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Instrumentation and Controls required N/A IEEE-279-71 N/A EGE / P to perform safety function OY / IRR /

IEEE-323-74 ESR / CR IEEE-336-71 Supports for QA Category I Same as component being supported.

Components*

Emergency Diesel Engine Jacket Water Cooling System*

Fresh Water Expansion Tank 3 ASME III 3 EGE P Piping and Valves 3 ASME III 3 EGE P Instrumentation and Controls required N/A IEEE-279-71 N/A EGE P to perform safety function IEEE-323-74 IEEE-336-71 Supports for QA Category I Same as component being supported.

Components*

Emergency Diesel Engine Exhaust and Combustion Air System*

Air Piping 3 ASME III 3 EGE P Supports for QA Category I Same as component being supported.

Components*

3.2-23 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Emergency Diesel Engine Lube Oil System*

Piping 3 Mfgr Std See Note EGE P Equivalent to ANSI B31.1. Se NRC Question Q430.73.

Valves 3 ASME III 3 EGE P Lube Oil Heat Exchanger 3 ASME III 3 EGE P 3EGO*E3A,B Lube Oil Strainers 3 ASME III See 3 EGE P Additional strainer to be Note purchased/installed ASME III, 3EGO*STR1A,B or if not available, ASME VII 3EGO*STR2A,B Lube Oil Filter 3 ASME III 3 EGE P 3EGO*FLT1A Heater 3 ASME VIII N/A EGE P 3EGO*H1A,B Lube Oil and Pre-Lube Pump 3 Mfgr Std N/A EGE P ASME material 3EGO*P3A,B 3EGO*P4A,B Rocker Arm Lube Oil System*

3.2-24 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Rocker Arm Lube Oil and Prelube 3 Mfgr Std N/A EGE P Pump 3EGO*P2A,B 3EGO*P1A,B Piping 3 Mfgr Std See Note EGE P Equivalent to ANSI B31.1. Se NRC Question Q430.73.

Valves 3 ANSI B16.5 N/A EGE P Filter 3 Mfgr Std N/A EGE P 3EGO*FLT2A,B Fuel Pool Cooling and Purification System*

Fuel Pool Cooling Pumps 3 ASME III 3 FB P Fuel Pool Coolers 3 ASME III 3 FB P Piping and Valves required for cooling 3 ASME III 3 FB P Figure 9.1-6 (P&ID 111) delineates SC boundaries.

Service Water Piping for Emergency 3 ASME III 3 FB P Figure 9.1-6 (P&ID 111)

Makeup to Fuel Pool delineates SC boundaries.

Instrumentation and Controls required N/A IEEE-279-71 N/A FB / CR / P Pressure retaining portions of to perform safety function in QA ESR inline instruments are the sam IEEE-323-74 Category I portions of system as connecting piping.

IEEE-336-71 3.2-25 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Supports for QA Category I Same as component being supported.

Components Containment Isolation System*

Containment isolation valves and 2 ASME III 2 CS / FB / P Individual fluid system figures associated piping for all system MSV / AB / delineate SC boundaries.

penetrating containment structure. ESB / HRB Instrumentation and Controls required N/A IEEE-279-71 N/A CS / FB / P to perform safety function MSV / AB /

IEEE-323-74 CR / IRR /

IEEE-336-71 ESB / HRB Supports Same as component being supported.

Service Water System*

Service Water Pumps 3 ASME III 3 CSP P Service Water Strainer 3 ASME III 3 CSP P Piping and Valves supplying cooling 3 ASME III 3 AB / ESB / P Figure 9.2-1 (P&ID 133) water to QA Category I equipment EGE / OY delineates SC boundaries.

(Buried) / CR

/ CSP Instrumentation and Controls required N/A IEEE-279-71 N/A CSP / ESB / P to perform safety function in QA EGE / CR /

IEEE-323-74 IRR / ESR Category I portions of system.

IEEE-336-71 3.2-26 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Supports for QA Category I Same as component being supported.

Components Reactor Plant Component Cooling Water Subsystem*

Component Cooling Pumps 3 ASME III 3 AB P Component Cooling Surge Tank 3 ASME III 3 AB P Component Cooling Heat Exchangers 3 ASME III 3 AB P Piping and Valves supplying cooling 3 ASME III 3 AB / CS / P Figure 9.2-2 (P&ID 121) water to QA Category I equipment ESB / FB delineates SC boundaries.

Instrumentation and Controls required N/A IEEE-279-71 N/A AB / CS / P to perform safety function in QA CR / IRR /

IEEE-323-74 ESB /

Category I portions of system IEEE-336-71 ESR / FB Supports for QA Category I Same as component being supported.

Components Neutron Shield Tank Cooling System*

Neutron Shield Tank 1 N/A N/A CS P Classified as SC-1 because portions support reactor vessel not because of shielding function.

Charging Pumps Cooling System*

3.2-27 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Charging Pumps Cooling Pumps 3 ASME III 3 AB P Charging Pumps Coolers 3 ASME III 3 AB P Charging Pumps Surge Tank 3 ASME III 3 AB P Piping and Valves 3 ASME III 3 AB P Instrumentation and Controls required N/A IEEE-279-71 N/A AB / CR / P to perform safety function IRR IEEE-336-71 IEEE-323-74 Supports for QA Category I Same as component being supported.

Components*

Safety Injection Pumps Cooling System*

Safety Injection Pumps Coolers 3 ASME III 3 ESB P Safety and Injection Pumps Cooling 3 ASME III 3 ESB P Pumps Safety Injection Pumps Surge Tank 3 ASME III 3 ESB P Piping and Valves 3 ASME III 3 ESB P Instrumentation and Controls required N/A IEEE-279-71 N/A ESB / CR / P to perform safety function IRR IEEE-323-74 IEEE-336-71 3.2-28 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Supports for QA Category I Same as component being supported.

Components*

Fuel Handling System (Note 3)

Reactor Vessel Head Lifting Device 1 & NNS N/A N/A CS P Portions that furnish support t Control Rod Drive Mechanism are SC-1.

Refueling Machine NNS CMMA N/A CS P Spec #70 Spent Fuel Shipping Cask Trolley NNS CMMA N/A FB P Spec #70 Spent Fuel Assembly Handling Tool NNS N/A N/A FB P Spent Fuel Pit Bridge and Hoist 3 N/A N/A FB P Fuel Transfer Tube and Flange 2 ASME III N/A FB / CS P Portions of Containment Boundary Fuel Basket 3 N/A N/A FB / CS P Protects fuel during transportation from damage.

Drive Mechanism and Controls NNS N/A N/A FB / CS P Supports for QA Category I Same as component being supported.

Components*

Main Steam System 3.2-29 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Main Steam Piping and Valves from 2 ASME III 2 CS / ESB / P Figure 10.3-1 (P&ID 123) steam generators up to and including MSV delineates SC boundaries.

main steam isolation trip valves and isolation valves in main steam supply lines to auxiliary feed pump turbines*

Main Steam Piping and Valves from 3 ASME III 3 ESB P Figure 10.3-1 (P&ID 123) isolation valves to the steam generator delineates SC boundaries.

auxiliary feed water pump turbine

Main Steam Piping from main steam 3 ASME III 3 MSV P Figure 10.3-1 (P&ID 123) isolation valves to turbine building delineates SC boundaries.

Main Steam Flow Restrictors 2 ASME III 2 CS P Instrumentation and Controls required N/A IEEE-279-71 N/A MSV / CR / P to perform safety function in QA IRR IEEE-323-74 Category I portions of system***

IEEE-336-71 Supports for QA Category I Same as component being supported.

Components*

Auxiliary Feedwater System*

3.2-30 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Demineralized Water Storage Tank 3 ASME III 3 OY P (DWST)

Steam Generator Auxiliary Feedwater 3 ASME III 3 ESB P Pumps (Turbine- and Motor-Driven)

Piping and Valves supplying auxiliary 3 ASME III 3 OY / ESB P Figure 10.4-6 (P&ID 130) feedwater from DWST to steam delineates SC boundaries.

generator auxiliary feedwater isolation valves Piping and Valves from steam 2 ASME III 2 ESB / CS P Figure 10.4-6 (P&ID 130).

generator auxiliary feedwater isolation valves to steam generator feedwater lines Instrumentation and Controls required N/A IEEE-279-71 N/A ESB / CR / P to perform safety function in QA IRR / OY /

IEEE-323-74 Category I portions of system CS IEEE-336-71 Supports for QA Category I Same as component being supported.

Components Feedwater System 3.2-31 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Steam Generator Feedwater Piping and 2 ASME III 2 MSV / CS P Figure 10.4-6 (P&ID 130)

Valves inside containment structure up delineates SC boundaries.

to and including the first restraint beyond isolation valve outside containment structure

  • Feedwater Piping and Valves from the 3 ASME III 3 MSV P Figure 10.4-6 (P&ID 130) first restraint beyond isolation valve delineates SC boundaries.

outside containment to turbine building wall Instrumentation and Controls required N/A IEEE-279-71 N/A MSV / CR / P to perform safety function in QA CS IEEE-323-74 Category I portions of system***

IEEE-336-71 Supports for QA Category I Same as component being supported.

Components*

Steam Generator Blowdown System Piping and Valves from steam 2 ASME III 2 CS / MSV P Figure 10.3-1 (P&ID 123) generator to containment isolation delineates SC boundaries.

valves Reactor Plant Sampling Systems*

3.2-32 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Piping and Valves for reactor coolant 2 ASME III 2 CS / AB P Figure 9.3-2 (P&ID 144) loop, pressurizer, and safety injection delineates SC boundaries.

accumulator sampling up to and including first isolation valve outside containment structure Instrumentation and Controls in QA N/A IEEE-279-71 N/A CS / AB/ P Category I portions of system CR / IRR IEEE-323-74 IEEE-336-71 Sample lines originating from safety- 2 or 3 ASME III 2 or 3 AB / ESB P Figure 9.3-2 (P&ID 144) related components, up to and delineates SC boundaries.

including remotely operated sample selection valve or second manual isolation valves Supports for QA Category I Same as component being supported.

Components*

Reactor Plant Aerated Drains Instrumentation and Control (sump N/A IEEE-279-71 N/A AB / ESB P level indication) required to provide IEEE-336-71 leak detection IEEE-323-74 3.2-33 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Porous Concrete Groundwater Sump, 3 ANSI B31.1 N/A ESB P Figure 9.3-6 (P&ID 106D)

Piping & Containment Recirculation delineates SC boundaries Cubicle Sumps Containment Monitoring System Instrumentation and Controls required N/A IEEE-279-71 N/A CS P to monitor hydrogen concentration IEEE-336-71 IEEE-323-74 Heating, Ventilation, and Air Conditioning System*

ESF Building All air-conditioning systems 3 SMACNA ASME III ESB P ASME III, Class 3 for service ARI, AMCA water side of refrigeration condenser.

Emergency Ventilation System 3 SMACNA, N/A ESB P for mechanical equipment room AMCA and auxiliary feedwater pump room Emergency Generator Enclosure Ventilation (except normal 3 SMACNA, N/A EGE P exhaust fan) AMCA 3.2-34 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Control Building All air-conditioning systems, 3 SMACNA, ASME III CB P ASME III, Class 3 for chilled including control building AMCA water and service water chilled water system pressure-containing components.

Battery rooms ventilation and 3 SMACNA, N/A CB P chiller room ventilation AMCA (excluding the control room kitchenette and toilet exhausts)

Control Room Pressurization Control room post-accident do analyses do not credit this Piping 3 ANSI B31.1 N/A CB P system or its components.

Tanks 3 ASME VIII N/A CB P Valves 3 ANSI B31.1 N/A CB P Control Room Emergency Ventilation 3 Note A N/A CB P See ESF Filter Systems, below Auxiliary Building Charging pumps and service 3 SMACNA, N/A AB P water pumps cubicles AMCA ventilation (excluding auxiliary building exhaust filtration system) 3.2-35 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Auxiliary building exhaust 3 Note A AB P See ESF Filter Systems, below filtration MCC and rod control area air- 3 SMACNA, ASME III AB P ASME III, Class 3 for service conditioning AMCA water coils.

Hydrogen Recombiner Building Hydrogen recombiner cubicles 3 SMACNA, N/A HRB P ventilation AMCA Main Steam Valve Building Main Stream Valve building 3 SMACNA, N/A MSV P ventilation (except normal AMCA exhaust fans)

Yard Structures Service Water Pumphouse 3 SMACNA, N/A CSP P ventilation AMCA Fuel Building Fuel Building Filtration System 3 Note A AB P A. See ESF Filter Systems, below. This system is not credited for post-accident analyses.

Supplementary Leak Collection and 3 Note A AB (Note B) B. Only portions inside tornad Release System (SLCRS) protected buildings.

3.2-36 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Instrumentation and Controls required N/A IEEE-279-71 N/A ESB / EGE P Only portions required to to perform safety function for QA / CB / CSP mitigate effects of LOCA will IEEE-336-71 Category I ventilation systems / AB Seismic Category I.

Supports for QA Category I Same as component being supported.

Components*

Fuel Building Fuel Building Exhaust Filtration 3 See Note See ESF Filter Systems, below This system is not credited for post-accident analyses.

ESF Filter Systems and SLCRS Fans 3 AMCA N/A Note C Note C C. For location and tornado criteria, see individual system listing.

Motors 3 NEMA, N/A Note C Note C IEEE-323, 344, 334 Filter Trains 3 ANSI N509 N/A Note C Note C Duct 3 SMACNA N/A Note C Note C Dampers 3 AMCA N/A Note C Note C 3.2-37 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Filter segment drainage up to and 3 ANSI B31.1 N/A AB Note C Applicable only to Auxiliary including the isolation valves Building Ventilation Filters, Fu Building Ventilation Filters an SLCRs.

Electrical Systems*

Emergency Generators, including N/A IEEE-308-74 N/A EGE P Auxiliaries IEEE-323-74 Unit Batteries and Chargers N/A IEEE-308-74 N/A BR P IEEE-323-74 Vital Bus and Inverters N/A IEEE-308-74 N/A ESR P IEEE-323-74 Emergency Unit Substations N/A IEEE-308-74 N/A ESR / P MAC IEEE-323-74 Emergency Station Service Switchgear N/A IEEE-308-74 N/A ESR P IEEE-323-74 Emergency Motor Control Centers N/A IEEE-308-74 N/A ESR / MRC P

/ EGE / CSP IEEE-323-74 / ESB Electric Motors-Pumps N/A IEEE-334-74 N/A Misc P (Note 1) 3.2-38 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Control Panelboards N/A IEEE-308-74 N/A Main control board and panels with CR P safety related functions Diesel engine driven emergency RCR P generators panel Radiation monitor panel CR P Auxiliary shutdown panel ESR P Air-conditioning control panel CR P Hydrogen recombiner control panel HRB P Trays, Conduits, and Ducts Carry N/A N/A N/A CS / CSP / P Safety Related Wiring AB / CB /

FB / EGE /

ESB / HRB OY / MSV /

MRC Class 1E AC Instrumentation, Control N/A IEEE-323-74 1E CS /CSP AB P and Power Cables - Essential Buses / ESF / CB /

IEEE-383-74 FB / HRB /

(orange/purple trains)

OY / MSV/

MRC / EGE 3.2-39 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Class 1E DC Power Cables N/A IEEE-323-74 1E CS/ CSP / P AB / ESF /

IEEE-383-74 CB / FB /

MSV / EGE

/ HRB Class 1E dc Switchgear, Distribution N/A IEEE-323-74 1E CB / CSP P Panels, and Protective Relays IEEE-344-75 Emergency Lighting Battery Pack N/A Mfgr Std See Note CS / AB P Seismic Category I only.

Supports CSP / ESF /

FB / HRB /

MSV / CB /

EGE Reactor Trip System ***

All portions of reactor trip system N/A IEEE-279-71 N/A CS / CR / P Includes instrumentation and which must operate to safely shut down IRR control components from and reactor to hot subcritical condition turbine impulse pressure transmitters.

Incore Instrumentation System***

Instrumentation and Conduit Tubes 1 ASME III 1 (Note) CS P Design Basis allowed use of Subsection NC 3600 Bottom Mounted Instrumentation 2 Mfgr. Std. N/A CS P Thimble Tubes 3.2-40 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes Engineered Safety Features Actuation N/A IEEE-279-71 N/A CS / CR P System ***

Miscellaneous

  • Containment Structure Polar Cran N/A N/A N/A CS P Designed for earthquake in unloaded condition.

LEGEND General Symbols Location Symbols N/A - Not applicable AB - Auxiliary building SC - Safety class BR - Battery room (in control building)

NNS - Nonnuclear safety CB - Control building QA Category - Quality assurance category CR - Control room (in control building)

RCPB - Reactor coolant pressure boundary CS - Containment structure

  • - SWEC Scope of Supply CEB - Containment enclosure building
      • - Scope of Supply shared between WNES and SWEC EGE - Emergency generator enclosure ESB - Engineered safety features building Tornado Criteria Symbols ESR - Emergency switchgear room (in control building) 3.2-41 Rev

ANS Safety Code Tornado Class Code (Note 1) Class Location Criterion Notes P - Protected from tornado effects by a structure or FB - Fuel building because below grade D -Designed to withstand tornado effects HRB -Hydrogen recombiner building IRR -Instrument rack room (in control building)

MRC -MCC and rod control area (in auxiliary building)

MSV -Main steam valve building OY -Outside, yard SB -Service building TB -Turbine building WDB -Waste disposal building NOTES:

Note 1. The mechanical system components satisfy the codes and addenda (ASME Section III, Division 1) in effect at the time of component order.

Note 2. There was no applicable code for the design of concrete containment structure liners at the construction of the Millstone 3 liner However, ASME Sections III and VIII, 1971 Edition, were used as guides. See Section 3.8.1.2.3.

Note 3. Protection p is provided by the tornado missile shield blocks. Based on a probabilistic analysis, installation of the tornado mis shield blocks is not required during Modes 5 and 6.

Note 4. This FSAR table identifies safety-related pumps for a given system. Unless otherwise indicated, motors for these safety-related pumps are also safety-related and included under the same safety class.

3.2-42 Rev

ASME III (All Ref. Summer 1973) QA Category I Program Organization Required Same Training Required Same Design Specification Pressure boundary integrity for SSE and Same, except for code references to dead load, thermal certification in specification (1)

Engineering, Design, and Document Category I Same Control Procurement Control ASME-approved suppliers ASME III design and materials, Category I supplier, no N-stamp required Receiving, Inspection, Identification, Physical inspection and review of Same Storage, and Handling Control documentation; ANSI storage and material identification Fabrication and Installation Control Control Drawing Package, FQC, and ANI Same, except no mandatory holdpoints; no review and established holdpoints; material third party documentation review; no traceability individual packages per drawing; normal Category I IR System (Table 3.2-3 and 3.2-4)

Field Welding and Brazing Control ASME III Procedures - Weld data package Same (Table 3.2-4) each weld; ASME IX welders (2)

Bolted and Other Mechanical Joints Data sheet for special bolted joints No special bolted joints; mechanical fittings installed to MFG requirements; documented on inspection report Tubing Supports Design to AISC, 7th Edition (Refer to NRC Same Question 210.36 for Details) 3.2-43 Rev

ASME III (All Ref. Summer 1973) QA Category I Program Heat Treatment and Special Operations and Not Applicable Same Repairs Fabrication and Installation Inspection FQC, ANI, ASME acceptance; material Same, except limited third party traceability required of selected surveillance; Category I material marking components to specific point of installation or exclusive purchase of Category I material (Tables 3.2-3 and 3.2-4)

Nondestructive Testing Dye penetrant for Class 2, visual for Same Class 3; traceability (2)

Nonconformances N&D Same Control of Measuring and Test Equipment Required Same Authorized Nuclear Inspector and Code N-stamp Not Applicable Certification Quality Assurance Audit Program SWEC, ASME, ANI, NUSCO SWEC, NUSCO Company Quality Assurance and Control SWEC QA Program Manual SWEC QA Program Manual Manual SWEC ASME III Control Manual Final Documentation As-built data package Documented on inspection reports; FQC FQC-ANI Certification acceptance Certificate Holder (Installation Not Applicable Same Subcontractor)

Pressure Testing 1.25 times design pressure Same FQC/ANI to witness SWEC's Responsibilities when Owner's Prepare code data forms, N5, N3, ANI Not Applicable Designee witness N-stamp 3.2-44 Rev

ASME III (All Ref. Summer 1973) QA Category I Program SWEC Operations under ASMI Section XI Governs repair of components Not Applicable NOTES:

1. For tubing which serves a nonsafety-related function attached to a Category I process pipe, functional capability is not required; therefore, the allowable stress values for ASME equation 9 utilize the faulted allowable of 2.4 Sh for all loading conditions. The faulte allowable for this tubing will ensure that the pressure boundary is maintained, thereby protecting the safety-related function of the proc piping.
2. Compression fittings are used exclusively for tubing installation except where transition from pipe to tubing is required. The pipe-to-tubing transition is controlled via piping specification requirements as part of the ASME III piping system 3.2-45 Rev

Safety Classes 2 and 3 Socket and Butt Welds

1. One hundred percent visual inspection by the construction department prior to release to Field Quality Control (FQC) - (document via construction checklist).
2. One hundred percent FQC inspection using ASME III NDE procedures (1)

(document via IR).

3. Third party inspector (ANI) witness NDE, percentage as determined by ANI.
4. In-process surveillance inspections performed by FQC - (document via IR).
5. One hundred percent pressure tested per ASME III pressure test requirements with 100 percent visual inspection of welds - (document via Pressure Test Report).

Safety Classes 2 and 3 Compression Fittings

1. One hundred percent of fitting make-up by construction prior to release to FQC using vendors recommended practices and inspection tools (document via construction checklist).
2. One hundred percent of fitting make-up by FQC using vendor's recommended practices and inspection tools (document via IR).
3. In-process surveillance inspection performed by FQC (document via IR).
4. One hundred percent pressure tested per QA Category I pressure test requirements with 100 percent visual inspection of fittings (document via Pressure Test Report).

TE:

or Class 2, LP is required, for Class 3, visual inspection is required.

3.2-46 Rev. 30

TESTING WITH ASME III REQUIREMENTS ASME Program Classes 2 and 3 QA Category I Exam/Testing Program One hundred percent visual inspection. 1. Same One hundred percent LP inspection (1) 2. Same (1)

Surveillance by ASME and ANI; 3. Surveillance by a third party inspector; approximately 10 percent in process approximately 10 percent in-process activities activities - includes welding and weld hydros Hydro - 100 percent inspection by FQC 4. Pressure test (Hydro or Pneumatic) -100 and ANI at 1.25 times design pressure percent inspection by FQC at 1.25 times design pressure; ANI to witness.

Surveillance inspection performed by 5. Same FQC; i.e., In process Welding, Weld material Control, Material Control All inspection performed, with the 6. Same exception of Item 5, are documented in the weld data packages; i.e., Weld Data Sheets Welders and procedures to be qualified to 7. Same ASME IX TE:

or Class 3, LP is not required by ASME III.

3.2-47 Rev. 30

1 WIND LOADINGS s section discusses the design wind load on Seismic Category I structures.

1.1 Design Wind Velocity maximum wind experienced in the vicinity of the Millstone site was associated with a 1960 icane. According to the preliminary site evaluation for Millstone 1 (TRC 1965), the Montauk nt Coast Guard Station on the tip of Long Island, across Long Island Sound from Millstone nt, recorded a wind speed of 115 mph with gusts to 140 mph. This wind speed is greater than 1,000 year mean recurrence interval wind speed for the New Haven area (Hollister 1970).

ocity profiles are as given in ANSI A58.1 - 1972 (ANSI 1972).

Seismic Category I structures are designed to remain elastic under the 115 mph basic wind ed. The gust factors are included in the determination of effective velocity pressures as cribed in ANSI A58.1. Because the site is considered to be flat, open country, the 1/7 power for vertical velocity distribution is used.

1.2 Determination of Applied Forces SI A58.1-1972 (ANSI 1972) is used to develop wind pressures and distributions on structures.

effective external velocity pressures for structures, qf, and for portions thereof, qp, at various hts above ground are in accordance with Tables 5 and 6 of ANSI A58.1 for exposure C.

ctive velocity pressures for calculating internal pressure (qm) are in accordance with ANSI

.1 Table 12. External pressure coefficients (Cp) are in accordance with ANSI A58.1 Tables 7 10 and internal pressure coefficients (Cpi) are in accordance with ANSI A58.1 Table 11.

tep function of pressure with height is used. The specified resultant design wind pressure at a n height is applied over a height zone defined by one-half the difference in adjacent heights which the design wind pressures are specified. The resultant design wind pressure acts normal he surface of the structure being considered.

2 TORNADO LOADINGS mic Category I structures requiring tornado design are listed in Table 3.2-1.

2.1 Applicable Design Parameters nado design parameters are in conformance with Regulatory Guide 1.76 (Section 1.8) as cribed below:

Maximum wind speed 360 mph Rotational speed 290 mph 3.3-1 Rev. 30

Radius of maximum rotational speed 150 ft Pressure drop 3.0 psi Rate of pressure drop 2.0 psi/sec maximum wind speed is the sum of the maximum rotational speed component and the imum translational speed component.

le 3.5-13 lists the tornado generated missiles and their characteristics.

2.2 Determination of Forces on Structures al tornado load is determined as a result of tornado wind (Ww), differential pressure (Wp), and sile (Wm) loadings.

Tornado Wind Load (Ww) nado wind load is determined as follows:

Ww = q Cp re:

q = External effective velocity pressure Cp = External wind pressure coefficient for structure being considered external effective velocity pressure (q) is determined in accordance with ASCE Paper No.

9 (ASCE 1961) as follows:

q = 0.00256 v2 psf re:

v = Applicable tornado wind speed in mph external wind pressure coefficient (Cp) defines the pressure acting on the surface of the cture. The external pressure coefficients for rectangular shape structures are in accordance h ANSI A58.1. The coefficients for cylindrical shape structures are based on Table 4(f) of CE Paper No. 3269 (ASCE 1961). The coefficients for the containment structure dome are ed on Wind Stresses in Domes (Gondikas and Salvadori 1960). Coefficients for structural steel pes are based on ASCE Paper No. 3269 (ASCE 1961).

Tornado Differential Pressure Load (Wp) 3.3-2 Rev. 30

pressure drop).

the fuel building, the section of metal roof at elevation 55 feet-3 inches and between column s G.5 and H is capable of venting the building. For this reason, the design is based on two ditions. The first condition considers the structure to be vented through the area covered by metal roof. Those sections of walls and floors of interior cubicles which are affected by the ting of the building through the roof area are designed for the full 3 psi pressure drop. For the ond condition, it is assumed that this roof does not vent the building. Therefore, the exterior ls and roofs of the building are designed for the full pressure drop (3 psi).

the emergency generator enclosure, the diesel generator muffler cubicle above elevation 51

-0 inch is capable of venting. Therefore, the design of this building is also based on two ditions. The first condition considers the diesel generator muffler cubicle above elevation to vented by the large openings in the exterior walls of the cubicle. The portions of walls and rs of interior cubicles which are affected by venting of the muffler cubicle are designed for the 3 psi pressure drop. For the second condition, it is assumed that this cubicle does not vent and exterior walls and the roof of the structure affected are designed for the full 3 psi pressure p.

ynamic load factor of 1.0 is applied to the pressure drop since all wall and floor panels subject ressure drop have frequencies greater than 4 Hz as shown in Table 3.3-1. Figure 3.3-1 presents pressure drop time history and the calculated dynamic load factor curve for the pressure drop.

mparing the two, it is evident that the Millstone 3 structural elements experience no dynamic lification during the pressure drop condition.

Tornado Missile Load (Wm) tion 3.5.3 describes the methods of analysis to determine the impact effects of tornado siles.

Total Tornado Load (Wt) total tornado load to be applied to Seismic Category I structures is determined from the most erse combination of individual tornado loadings. The loading combinations considered are:

W = Ww W = Wp W = Wm W = Ww + 0.5 Wp W = Ww + W m 3.3-3 Rev. 30

total tornado load is then combined with other loads as specified in Sections 3.8.1.3, 3.8.3.3, 4.3, and 3.8.5.3.

2.3 Effect of Failure of Structures or Components not Designed for Tornado Loads ctures which do not require Category I design are either located so that structural failure does affect the ability of safety related structures or systems to perform their intended design ction, or designed so that they do not collapse under tornado wind load. The metal siding and ing of the service, turbine, waste disposal, containment enclosure buildings, and portions of fuel building are assumed to blow off under tornado wind load. The resulting siding and ing missiles are less severe than the missiles described in Section 3.5. The structural steel ing of these structures is designed to withstand tornado wind loads so as not to compromise integrity of any safety related structure, system, or component.

3 REFERENCES FOR SECTION 3.3 1 American National Standards Institute (ANSI) 1972. A58.1 1972, Building Code Requirements for Minimum Design Loads in Buildings and Other Structures. New York, N.Y.

2 American Society of Civil Engineers (ASCE) 1961. Wind Forces on Structures. In:

Transactions of the American Society of Civil Engineers, Paper No. 3269, Vol 126, Part II.

3 Gondikas, P. and Salvadori, M.C. 1960. Wind Stresses in Domes. Engineering Mechanics Division, American Society of Civil Engineers (ASCE).

4 Hollister, S. C. 1970. The Engineering Interpretation of Weather Bureau Records for Wind Loading on Structures. National Bureau of Standards (U.S.), Building Sciences Service, 30.

5 TRC Service Corporation 1965. Millstone Unit No. 1, Preliminary Site Evaluation.

3.3-4 Rev. 30

FUNDAMENTAL THICKNESS FREQUENCY PANEL SIZE (FT) (FT) (HZ) REMARKS SVB North Wall 36x73 2 6.5 Simple Supports

x. Bldg. Roof 49.5x63 2 4.6 Simple Supports
x. Bldg. Roof 126.5x26.5 2 9.0 Simple Supports Beam el Bldg. Roof 59.5x66.5 2 4.0 Simple Supports ntrol Bldg. 57.5x102 2 4.2 Partial Fixity of 3.3-5 Rev. 30

s section discusses the flood and the highest groundwater level design for Seismic Category I ctures and components. Internal flooding has been addressed as part of equipment lification requirements and is described in Section 3.11.

1 FLOOD PROTECTION 1.1 Flood Protection Measures for Seismic Category I Structures design basis flood (maximum combination of storm surge and wave runup) established for lstone 3 is elevation +23.8 feet msl and the maximum still water level is elevation +19.7 feet ction 2.4.5). All unit safety-related structures and equipment, except the circulating and ice water pumphouse, are protected from flooding by the site grade of elevation +24 feet msl.

h pair of service water pumps and pump motors is located at elevation +14.5 feet msl inside vidual watertight cubicles in the seismically designed pumphouse (Figure 3.4 1). The walls of e cubicles are watertight up to elevation +25.5 feet msl protecting the pump motors and ciated electrical equipment from wave action and probable maximum hurricane (PMH) surge.

accesses to safety-related structures and facilities are at an elevation of 24 feet-6 inches above nominal site grade elevation of 24 feet-0 inch and are consequently protected from flooding to groundwater, storm surge, and direct rainfall, except for the doors that are discussed in tion 2.4.2.3. The two access openings to the service water cubicles inside the pumphouse ch are below the flood protection level of +23.8 feet msl are fitted with watertight steel doors able of withstanding the maximum hydrostatic load occurring at their respective location.

phouse roof ventilators are weatherproof and located above the maximum accumulation of w resulting from the occurrence of the 100-year accumulated snow depth of 52 inches.

ipment access openings on the pumphouse roof over the service water cubicles are fitted with ertight covers. During normal plant operation, the service water cubicles have open drain lines alled in the cubicle sump to enable the service water pump seal water leak off to drain directly the intake structure pump bay. During severe weather or flooding conditions, the drain lines isolated and the service water cubicle sumps are drained using sump pumps 3PBS-P1A and S-P1B.

ndations of safety-related structures are constructed of reinforced concrete. All subgrade ts between walls and slabs are sealed with waterstops cast in concrete.

storm drain system uses catch basins and underground conduits and/or drainage ditches to vey runoff to Niantic Bay. Roof and site storm drain systems are designed for a maximum ipitation of 6.5 inch per hour. The probable maximum precipitation of 70.4 inch per hour for a inute period (Hansen, et al., 1982) would result in temporary flooding of the site area in the nity of the safety-related buildings as described in Section 2.4.2.3.

3.4-1 Rev. 30

effects of local intense precipitation can be found in Section 2.4.2.3.

seaward wall of the intake structure is constructed of reinforced concrete, designed to hstand the forces of a standing wave, or clapotis, with a maximum crest elevation of +41.2 feet

. Further discussion on the effects of PMH storm surge and wave action, including the ltant pressure distribution on the intake wall, can be found in Section 2.4.5.

mbinations of the maximum surge level with a coincident wave and the maximum wave height h a coincident surge for three different speed PMH were examined to determine the maximum ft pressure on the pumphouse floor. The most critical combination is the maximum wave ht of 16.2 feet and a surge level of elevation 19.7 feet msl associated with the slow speed H.

calculated maximum uplift pressure on the pumphouse floor due to the most critical bination of wave action and storm surge during PMH conditions is 863 lb/sq ft. The floor is gned to withstand more than this pressure, precluding the possibility of failure.

water level fluctuations within the pumphouse, resulting from storm surge and wave action, dampened by the energy lost in passage through the restricted openings in the trash racks, eling screens, and operating deck. Internal water level fluctuations are further attenuated ause water must enter the structure through a submerged opening (elevation -7 to -30 feet msl) ugh which the pressure response factor is less than unity.

discharge outfall structure is also designed to withstand maximum wave forces induced by most critical combination of wave action and storm surge during PMH conditions. The imum horizontal pressure is determined for the combination of a maximum wave height of 1 feet and a maximum surge level of 19.7 feet msl. The maximum vertical pressure on the er outfall structure is obtained for a minimum water submergence at the surge level of ation 9.46 feet msl and a coincident breaking wave of 14.7 feet. The calculated maximum zontal pressure and the maximum vertical pressure on the discharge outfall structure for the ve-mentioned combination events are 2,325 lb/sq ft and 585 lb/sq ft, respectively.

tion 2.5.5.1 discusses shoreline protection in the vicinity of the pumphouse.

od protection complies with Regulatory Guide 1.102, Flood Protection for Nuclear Power nts, as follows:

1. C1 Flood protection is accomplished by the unit's location on a Dry Site, with the exception of the circulating and service water pumphouse which utilizes Incorporated Barriers.
2. C2 Not applicable Refer to Section 1.8 for clarification to Position C1.

3.4-2 Rev. 30

re is no safety-related dewatering system for lowering groundwater levels for Millstone 3.

s system is not applicable. Removal of water which bypasses the membrane installed below Containment Structure is collected in the Engineered Safety Features Building porous crete groundwater sump and is removed by the Underdrain System (see Section 9.3.3).

2 ANALYTICAL AND TEST PROCEDURES und levels of all Category I structures, except for the circulating and service water pumphouse the discharge structure, are located above the design basis flood (DBF) level. This level is ed on the maximum combination of storm surge due to the PMH and associated wave run-up ction 3.4.1). Structures located above this level are designed for the hydrostatic effects of ft and lateral water pressure resulting from the DBF or normal groundwater, whichever is e severe. Groundwater levels are based on piezometric readings taken at the site ure 2.5.4-37).

circulating water discharge structure and discharge tunnel and the circulating and service er pumphouse are located below the DBF level.

circulating water discharge structure and discharge tunnel are designed for the hydrostatic dynamic effects of the DBF as described in Section 3.4.1.

circulating and service water pumphouse is designed laterally for a standing wave and for ft on the operating floor due to confined wave action within the pumphouse (Section 3.4.1).

ndation loadings used in the design reflect saturated soil conditions, where applicable.

design wind loading described in Section 3.3.1 is applied concurrently with the hydrostatic dynamic effects of the DBF (Sections 3.8.1.3, 3.8.3.3 and 3.8.4.3). Tornado loading is not lied concurrently with the DBF.

3 REFERENCE FOR SECTION 3.4 1 Hansen, E.M., Schreiner, L.C., and Miller, J.F, 1982. Application of Probable Maximum Precipitation Estimate - U.S. East of the 105th Meridian. Hydrometeorological Report No. 52, National Weather Service, NOAA, U.S. Department of Commerce, Washington, D.C.

3.4-3 Rev. 30

1 MISSILE SELECTION AND DESCRIPTION tems and components located both inside and outside the containment have been examined to tify and classify potential missiles. Two broad categories of systems and components are ewed to determine the potential for generating missiles; pressurized components and high ed rotating machinery. Only designs where a single failure could lead to missile ejection are sidered. The basic approach to ensure missile protection of systems and components both de and outside of containment involves the following considerations:

1. examination of systems in order to identify and classify potential missile sources;
2. evaluation of the design adequacy of equipment to preclude generation of missiles; and
3. evaluation of the effects of the generation of missiles where the potential exists and provisions for protection against them.

objective is to ensure design adequacy against generation of missiles and means of protecting ntial structures, systems, and components should a missile be generated.

1.1 Internally Generated Missiles (Outside Containment) design bases consider missiles generated outside the containment but internal to the plant site.

se shall not cause damage that may affect the safe shutdown or cause radiation release during rating conditions, and postulated accident conditions associated with the effects of missile mation. Table 3.5-1 identifies the safety-related structures, systems, and components outside containment required for safe shutdown of the reactor under all conditions of plant operation.

ves in high energy fluid systems are evaluated as potential missile sources. Valves are typically gned with parts which are removable for maintenance. It is these removable parts which ent the most significant potential for missile producing failures. Valves provided with k-seated stems are not considered credible sources of missiles. This design feature effectively inates the possibility of ejecting valve stems even if the stem threads fail. Valve bonnets are sidered credible sources for missiles in cases where the bonnet is bolted to the body. The net and the connection bolts are postulated to be ejected. Valve bonnet missiles are not sidered credible where the bonnet is welded to the body, the bonnet is integral with the body, he bonnet bolts are torqued in a controlled manner. Credit may be taken for air and motor rators which interfere with the ejection of valve stems and bonnets. The Applicants review of es located outside containment has indicated that all the valves in the high energy systems can ategorized as follows.

1. Valves that are isolated or enclosed from safety related equipment. These valves do not impose any danger.

3.5-1 Rev. 30

determined, then failure is admitted and the consequences of such a failure are assessed. If an unacceptable interaction occurs, then the missile source is subject to reorientation if feasible, the target is subject to relocation if feasible, or a barrier is provided to prevent interaction. An unacceptable interaction is an interaction with an essential structure, system, or component.

trifugal pumps and fans located outside the containment in areas containing safety related ponents have been evaluated for missiles caused by overspeed or failure. The maximum oad speed of these centrifugal pumps and fans is equivalent to the maximum operating speed heir motors. Consequently, no overspeed is expected and missiles associated with overspeed ditions in centrifugal pumps and fans outside the containment are not postulated.

s are further evaluated for missile generation under normal operating speeds due to fatigue ure or manufacturing defects. Fan fragments are postulated only where a credible single failure hanism results in fragmentation. Such fragments have been shown either to lack sufficient rgy to penetrate the fan housing or to result in acceptable interactions with essential targets. In assessment, the fragments are assumed to be unimpeded by any flexible connections between fan housing and attached ducting.

auxiliary feedwater pump turbine is equipped with redundant overspeed detection devices a regularly tested turbine trip valve, as such overspeed in the turbine is considered credible y up to the trip setting at 10 percent over rated speed. At this speed there exist substantial gins between the energy available in the fragments generated from the turbine-driven pump the energy required to escape the pump casing; therefore, missiles are not postulated from the p component. Similarly, fragments generated from the turbine component lack the etrating geometry and the energy to perforate its casing. As such, neither the pump nor the ine driver is considered a source of internally generated missiles and no essential systems or ponents can be adversely affected. Nonetheless, the auxiliary feedwater pump turbine is ted and oriented within a concrete cubicle to prevent any generated missiles from affecting r safety systems, such as the motor-driven auxiliary feedwater pumps, in adjacent cubicles.

motor-generator that provides power to the control rod drive mechanisms (CRDM) is located ide the containment. The flywheel on this component has been evaluated as a potential sile. The fabrication specifications of the motor-generator-set-flywheel control the material to t ASTM-A533-70, Grade B, Class I with inspections per MIL-I-45208A and flame cutting machining operations governed to prevent flaws in the material. Nondestructive testing sisting of nilductility (ASTM-E-208), Charpy V-notch (ASTM-A593), ultrasonic TM-A577 and A578), and magnetic particle (ASME Section III, NB2545) is performed on h flywheel material lot. In addition to these requirements, stress calculations are performed sistent with guidelines of ASME Section III, Appendix A, to show the combined stresses due entrifugal forces and the shaft interference fit shall not exceed one-third of the yield strength ormal operating speed (1,800 rpm) and likewise, shall not exceed two-thirds of the yield ngth at 25 percent overspeed. However, no overspeed is expected for the following reason: the heel weighs approximately 1,300 pounds and has dimensions of 35.36 inches in diameter and 3.5-2 Rev. 30

or is insufficient for overspeed. Therefore, there are no credible missiles from the CRDM or-generator flywheel.

luation of missiles being generated from the emergency generator enclosure concluded that e is no need to evaluate missile generation. Safety related emergency generators are located in ructure designed for tornado missile protection; consequently, missiles from the diesel engines considered unable to penetrate this structure. The essential diesel generator systems are undant and separated so that a missile generated by one diesel engine will not affect the other.

rs are offset from the generators axes precluding the possibility of missiles exiting from the rway of the structure.

1.2 Internally Generated Missiles (Inside Containment) design bases are such that missiles generated within the reactor containment will not cause of function in any redundant engineered safety feature nor radiation release or damage the tainment boundary.

ddition, a missile accident which is not caused by a LOCA shall not initiate a LOCA.

le 3.5-2 identifies the structures, systems, and components inside the containment whose ure could lead to offsite radiological consequences or which are required for safe plant tdown to a cold condition assuming an additional single failure.

ipment inside the containment has been evaluated for potential missile generation. As a result his review, the following information concerns potential missile sources and systems which uire protection from internally generated missiles inside the containment.

1.2.1 Missile Selection and Description ure of the reactor vessel, steam generators, pressurizer, and reactor coolant pump casings ing to missile generation are not considered credible because of the combination of material racteristics, inspections, quality control during fabrication, erection, and operation, servative design, and prudent operation as applied to the particular component.

reactor coolant pump flywheel is not considered a source of missiles for the reasons ussed in Section 5.4.1. Nuts and bolts are of negligible concern because of the small amount tored elastic energy.

trifugal pumps, fans, and air compressors (centrifugal and axial) located inside the tainment have been evaluated for missiles associated with overspeed failure. The maximum oad speed of these centrifugal pumps, fans, and air compressors is equivalent to the operating ed of their motors. Therefore, no overspeed is expected and missiles associated with overspeed ditions in centrifugal pumps, fans, or air compressors within the containment are not tulated.

3.5-3 Rev. 30

hanism results in fragmentation. Such fragments have been shown either to lack sufficient rgy to penetrate the fan housing or to result in acceptable interactions with essential targets. In assessment, the fragments are assumed to be unimpeded by any flexible connections between fan housing and attached ducting.

following nuclear steam supply system components are considered to have a potential for sile generation inside the reactor containment:

1. control rod drive mechanism housing plug, drive shaft, and the drive shaft and drive mechanism latched together;
2. valves;
3. temperature and pressure sensor assemblies; and
4. pressurizer heaters.

ss failure of a control rod mechanism housing, sufficient to allow a control rod to be rapidly ted from the core, is not considered credible for the following reasons.

1. Control rod drive mechanisms are shop tested at 4,100 +/- 75 psi.
2. Control rod drive mechanism housings are individually hydrotested to 3,107 psi after they are installed on the reactor vessel to the head adapters and checked again during the hydrotest of the completed reactor coolant system.
3. Control rod drive mechanism housings are made of Type 304 stainless steel. This material exhibits excellent notch toughness at all temperatures that are encountered.

wever, it is postulated that the top plug on the control rod drive mechanism could become e and would be forced upward by the water jet. The following sequence of events is assumed.

1. The drive shaft and control rod cluster are forced out of the core by the differential pressure of 2,500 psi across the drive shaft.
2. The drive shaft and control rod cluster, latched together, are assumed fully inserted when the accident starts.
3. After travelling approximately 12 feet, the rod cluster control spider hits the underside of the upper support plate.
4. Upon impact, the flexure arms in the coupling join the drive shaft and the control cluster fracture freeing the drive shaft from the control rod cluster.

3.5-4 Rev. 30

mechanism missile shield provided.

control rod drive mechanism (CRDM) missiles are summarized in Table 3.5-3. The missile cities have been calculated by balancing the forces due to the water jet. No spreading of the er jet has been assumed. CRDM missile impact velocity, kinetic energy, and penetration are sidered in the design and layout of the missile shield.

ve stems in motor-operated (MOV) or air-operated (AOV) valves are not considered credible rces of missiles. All the isolation valves installed in the reactor coolant system have stems with ack seat. This effectively eliminates the possibility of ejecting valve stems even if the stem ads fail.

ves within the reactor coolant pressure boundary have been reviewed to identify potential siles. This review identified no credible failures that could result in missile generation, except valve missiles in the region where the pressurizer extends above the operating floor. Valves in region are the pressurizer safety valves, the motor-operated isolation valves in the relief line, air-operated relief valves, and the air-operated spray valves. Although failure of these valves ot considered credible, failure of the valve bonnet body bolts is, nevertheless, postulated and integrity of the containment liner and safety related equipment from the resultant bonnet sile is assured by the pressurizer cubicle walls and roof which act as a missile barrier.

missile characteristics of the valves in the region where the pressurizer extends above the rating deck are given in Table 3.5-4.

only credible sources of jet-propelled missiles from the reactor coolant piping and piping ems connected to the reactor coolant system are the temperature and pressure sensor mblies. The resistance temperature sensor assemblies are of two types: with well and thout well. Two rupture locations have been postulated: around the weld (or thread) between temperature element assembly and the boss of the without well element, and the weld (or ad) between the well and the boss for the with well element.

missile characteristics of the piping temperature sensor assemblies are given in Table 3.5-5. A degree half-angle expansion water or steam jet has been assumed. The missile characteristics he piping pressure element assemblies are less severe than those of Table 3.5-5.

perature and pressure sensors are installed on the reactor coolant pumps close to the radial ring assembly. A hole is drilled in the gasket and sealed on the internal end of a steel plate. In luating missile potential, it is assumed that this plate breaks and the pipe plug on the external of the hole becomes a missile.

missile characteristics of the reactor coolant pump temperature sensor, the instrumentation l of the pressurizer, and the pressurizer heaters are given in Table 3.5-6. A 10-degree ansion jet has been assumed.

3.5-5 Rev. 30

initial flight direction of these missiles has been determined and only low kinetic energies are sible. Only fragile components would be damaged by this missile category. Protection was lemented by keeping such components as electrical equipment and cables out of the primary ht path.

ssurizer heaters could loosen and become jet-propelled missiles. The integrity of safety related ipment from postulated heater vessels is assured by the pressurizer cubicle floor and walls.

valves have been designed against bonnet-to-body connection failure and subsequent bonnet tion by means of:

1. Compliance with the ASME Code,Section III
2. Control of load during the bonnet-to-body connection stud tightening process proper stud torquing procedures limit the stress of the studs to the allowable limits blished in the ASME Code. This stress level is far below the material yield. The valves are rotested per the ASME Code,Section III. The bodies and bonnets are volumetrically and ace tested to verify soundness. Critical valves are also designated for inservice inspection to ME Code,Section XI, requirements (Section 6.6).

-ASME III valves were examined for potential missiles. Provision for protection against them undertaken based on the following:

1. Valve stems are provided with back seats. This prevents stem ejection should the threads fail.
2. Valves are oriented to prevent postulated missiles from impacting critical targets.
3. Probability of failure is low (P 10-4 per year), so a low impact probability may be argued against generation of missiles.
4. Barriers erected specifically for protection against internal missiles.

siles generated due to a seismic event are addressed in Section 3.7B.3.13.

1.2.2 Missile Protection Provided ety related structures, systems, and components whose safety function might be impaired, are ected from postulated missiles by:

1. Locating the systems or components in individual missile proof structures 3.5-6 Rev. 30
3. Providing special localized protective shields or barriers.

ability of structures or barriers to withstand the effects of potential internally generated siles is discussed in Section 3.5.3.

1.3 Turbine Missiles 1.3.1 Turbine Placement and Orientation ure 3.5-1 shows the turbine placement and orientation for the three unit site. This figure also cates the +/-25-degree missile ejection zone with respect to the low pressure turbine wheels for h turbine unit within reach of plant structures. Figure 3.5-2 shows an elevation view of lstone 3 structures within the +/-25-degree missile ejection zone.

1.3.2 Missiles Identification and Characteristics ounding value of 1.0E-2 per year for the probability of unacceptable damage in the event that rbine missile is generated by wheel failure, is used.

1.3.3 Target Description vy zones on Figure 3.5-1 identify target areas for low trajectory missiles. Table 3.5-7 provides licable elevations over which those wall areas are considered as targets for each heavy line.

as other than those shown by the heavy lines have been excluded as targets because they are er outside the ejection zone for low trajectory missiles or they meet the criteria that the bability of the missile strike damaging its target is zero, P3 = 0.

get areas for high trajectory missiles are the roof areas of all Category I structures. These ets are shown as shaded areas on Figure 3.5-1.

1.3.4 Probability Analysis ulatory Guide 1.115 Revision 1, Protection Against Low-Trajectory Turbine Missiles states the NRC staff considers a hazard due to low trajectory turbine generated missiles of less than E-7 per year an acceptable risk rate for the loss of an essential system for a single event. The hodology is based on maintenance and inspection activities which minimize the potential for sile generation. Using the missile damage probability of 1.0E-2 (for unfavorable ntation), the probability of turbine failure, P1, must be shown to be less than 1.0E-5 in order eet overall R.G. 1.115 acceptable risk rate.

turbine system maintenance program is established based on manufacturers calculations of sile probabilities. The program ensures that the 1.0E-5 probability is maintained. Higher babilities of turbine failure are acceptable for short durations (for example, to support on-line 3.5-7 Rev. 30

1.3.5 Turbine Overspeed Protection turbine control system is an electrohydraulic control (EHC) system that includes both digital analog circuitry, electronic servo hardware, and hydraulic valve actuators.

EHC provides a normal overspeed protection system and an emergency overspeed protection em to limit turbine overspeed. These two systems are essentially separate and independent.

normal overspeed protection system is part of the turbine load and speed control system and esigned to limit turbine overspeed without a turbine trip under all load conditions. The rgency overspeed protection system is part of the emergency trip system and is designed to the turbine if the turbine speed exceeds 110 percent of rated speed (Section 10.2).

1.3.6 Turbine Valve Testing main turbine generator control, main stop, intercept, and reheat stop valves are routinely ed (Section 10.2.3.6).

trol and main stop valves are tested one at a time, and as each test is completed, the valve is rned to its original position before the next valve is tested. Intercept and reheat stop valves are rlocked so that a pair of these valves in one crossover pipe is tested together. For this test, one of pipe is tested and the valves returned to the open position before the next pair is tested.

1.4 Missiles Generated by Natural Phenomena only credible missiles generated by natural phenomena are those generated by a design basis ado. Those Category I structures designed to withstand the effects of tornado missiles and the ems and components thus protected are identified in Tables 3.5-1 and 3.5-2.

inimum of 2 feet thickness of reinforced concrete having a minimum strength of 3,000 psi (28 compressive strength) was used for walls, roofs, and floors designated as missile protection.

minimum reinforcing steel each way in each face of any square foot of wall or slab providing sile protection is 1.85 square inches.

tulated missiles generated by the design basis tornado (Section 3.3.2) are listed in le 3.5-13, which includes all parameters necessary to determine missile penetration. These siles are considered capable of striking in any orientation.

tilation openings in the various facility buildings housing essential shutdown equipment are ected by reinforced concrete labyrinths.

re are no other design basis missiles resulting from flood or any other natural phenomena cribed in Section 2.2.3.

3.5-8 Rev. 30

missiles of any significance are expected to be generated by events near the site due to ances from nearby transportation routes. The possibility of missiles from a Providence &

rcester (P&W) railroad tank car explosion was considered; however, it was determined that h a tank car explosion would not generate significant missiles at the plant site. For a detailed cription, see Section 2.2.3.

1.6 Aircraft Hazards tudy of the probability of aircraft which use the nearby airport and airways colliding with the ty related structures of the Millstone site has been conducted. Due to the conservatism of the lysis, the values derived by the analytical model are believed to be substantially higher than true probability. The study concludes that the aircraft accident probability would be less than x 10-7 per year for a number of years since no increase in air traffic is projected in the vicinity he site.

2 STRUCTURES, SYSTEMS, AND COMPONENTS TO BE PROTECTED FROM EXTERNALLY GENERATED MISSILES siles to be considered in this section are identified in Section 3.5.1.

discussed in Section 3.5.1, all plant systems and components must be protected whose failure lead to unacceptable offsite radiological consequences or which are required to shut down the tor and maintain it in a safe condition, assuming an additional single failure. Safety related ctures, systems, and components that are required for a safe shutdown of the reactor are tified in Tables 3.5-1 and 3.5-2. All components containing radioactive fluids are protected or within the results of the design basis offsite radiological analyses which are presented in tions 15.7.1, 15.7.2, and 15.7.3.

3 BARRIER DESIGN PROCEDURES sile barriers are designed to withstand the effects of missiles described in Section 3.5.1 hout compromising plant safety.

st internally generated missiles have low to moderate energy. Consequently most barriers ted specifically for protection against internal missiles are rather light and usually protect y a limited area. Small steel barriers, rather than concrete structures are generally most able. There are a few exceptions, such as the control rod drive missiles, where sufficient rgy exists to make a concrete barrier more practical.

3.1 Concrete Barriers sile barrier design requirements include:

3.5-9 Rev. 30

2. Scabbing particles are not generated or are limited to energy levels which still permit safe shutdown of the plant
3. Structural response to missile impact permits the barrier and its supports to safely carry other loads during and after impact. In addition, the deflection of the barrier does not impair the safety or safety function of a Category I system.

concrete barriers, the requirement to stop the missile is fulfilled by showing that perforation he barrier, or the final barrier in a multiple barrier design, does not occur. This is established ed either on test data (Nusbaum et al., 1976; Stephenson 1976; and Rotz 1975), or an empirical luation of these test data.

ermination of whether scabbing occurs is also based on procedures set forth in Stone &

bster Engineering Corporation Topical Report, SWEC-7703, Appendix B (SWEC 1977).

overall structural response to missile impact is based on dynamic time history calculations for tic-plastic behavior of the missile and the barrier. The general method for calculating overall ctural response is described in Introduction to Structural Dynamics (Biggs 1964). The mate load capacity of concrete barriers is determined by yield line theory. The barriers are gned so that the calculated ductility ratio of the barriers for any load combination is less than maximum allowable ductility ratio.

tility ratio is defined as the ratio of maximum acceptable displacement (Xm) to the lacement at the effective yield point (Xy) of the structure.

analytical procedures for analyzing concrete barriers for overall structural response to tornado siles refer to Stone & Webster Engineering Corporation Topical Report SWEC 7703, endix C (SWEC 1977).

he concrete barrier is required to carry loads during and after missile impact, the maximum wable ductility ratio is limited to a factor of 10. In particular, for beam-column members, re the compressive load is equal to or less than one-third of that which produces balanced in conditions for the cross section, the allowable ductility ratio is 10.

er to Table 3.5-14 for examples of maximum barrier elastic deflection, maximum barrier ection, and ductility ratio as a function of missile and barrier span for 24 inch concrete iers.

an example of maximum barrier elastic deflection, maximum barrier deflection, and ductility o for a beam-column, refer to Table 3.5-15.

concrete barrier is not required to carry other loads during and after impact, the maximum wable rebar elongation is limited to 5 percent.

3.5-10 Rev. 30

l barriers are designed with consideration given to local perforation, overall response, and port loads.

thickness of the barrier plate required to prevent perforation is determined from the Stanford ation (Gwaltney 1968). The BRL formula for steel perforation is used if the Stanford equation ot applicable. However, impacts near a support are most limiting with respect to perforation where these are postulated, the Stanford equation is used with the artificial, but conservative, mption that the support spacing equals the missile diameter.

preferred design is a single plate supported at the corners and, if necessary, reinforced by ing around the periphery to channels or other rolled shapes. The overall response is evaluated sidering a central impact and equating the missile energy to the energy absorbing capacity of barrier plate. Plastic limit theory and an allowable of 50% ultimate uniform strain are used to rmine this capacity.

load delivered to the supports is dynamic and equal to the plastic limit load of the plate. Since impact may occur near one support, each support is sized accordingly. When the resulting load o high for founding structure, soft supports are used. In this case the supports, whether elastic lastic, are designed to accept the available energy:

m E = E o ---------------

M+m re Eo = Initial missile energy, m = Missile mass, and M = Barrier mass participating in missile momentum transfer 3.3 Design Evaluation y one missile is postulated at a time and is assumed to strike the barrier end-on. The missiles e two effects on structures, walls, or any other barrier: local effects and overall response. The l effects include penetration, perforation, and spalling or scabbing. The overall response udes the flexural and shear effects.

nforced concrete external roofs and walls of Seismic Category I structures form barriers inst tornado-generated missiles. Buried underground, safety related duct runs and piping are protected.

al Effect: The estimate of missile penetration in concrete barrier is based on the modified y formula (Amirikian 1950). Sufficient thickness of concrete is provided to prethickness 3.5-11 Rev. 30

rall Barrier Response: The overall response of structural barriers to missile impact is rmined by time history analysis of the slab missile system. The ductility ratio so determined t be less than the allowable ductility ratio. This ductility ratio is a function of the controlling re of the structural behavior. For beams, walls, and slabs where flexural controls design, the missible ductility ratios are given in Section 3.5.3.1 and are based on Stone & Webster ineering Corporation Topical Report, SWEC 7703 (SWEC 1977). Flexural strength is rmined from an ultimate strength theory with the limitations on ductility.

example of a missile shield that has been designed using these procedures is the control rod e mechanism missile shield.

shield is provided over the control rod drive mechanisms to block any missiles which might ssociated with a fracture of the pressure housing of any mechanism. This shield is constructed einforced concrete with a steel facing plate, and is located above, and as near as possible, to housing. This limits the velocity of the ejected missile and prevents missiles from bypassing he shield is designed using procedures given in ORNL-NSIC-5 (Greenstreet et al., 1965).

3.4 Secondary Missiles crete impacts produce secondary missiles from the front face of the barrier. These missiles e considerably less energy than the primary missile. Concrete secondary missiles were only sidered for a postulated primary missile which has more than 4 ft-kips of kinetic energy. Our ew showed that scabbing does not damage safety related equipments.

ondary missiles from impacts are limited to components that may be torn loose by the primary sile. The energy dissipated in creating secondary missiles cause the remaining energy in the ary missile to be negligible. Only concrete walls were found to exhibit this phenomena of ing parts or scabbings. Nevertheless, all secondary missiles were found passive.

4 REFERENCES FOR SECTION 3.5 1 Amirikian, A. 1950. Design of a Protective Structure. NAVDOCKS, P-51, Bureau of Yards and Docks, Dept. of the Navy.

2 Biggs, J.M. 1964. Introduction to Structural Dynamics. McGraw-Hill, New York, New York.

3 Bush, S.H., Probability of Damage to Nuclear Components Due to Turbine Failures, Nuclear Safety. Vol. 14, No. 3, May - June, 1973.

4 General Electric Co., March 1973. Hypothetical Turbine Missiles - Probability of Occurrence, Memo Report. Turbine Department.

3.5-12 Rev. 30

Oak Ridge National Laboratory and Bechtel Corp., US AEC Technical Information Center, TD-4500.

6 Gwaltney, R.C. 1968. Missile Generation and Protection in Light Water-Cooled Power Reactor Plants. ORNL-NSIC-22.

7 Nusbaum, M.S., Welch, R.E., and Foxx, C.E. 1976. One Quarter Size Reinforced Concrete Missile Barrier Tests. ITT Research Institute Report.

8 Rotz, J.V. 1975. Results of Missile Impact Tests on Reinforced Concrete Panels. Second ASCE Specialty Conference on Structural Design of Nuclear Plant Facilities, New Orleans, La.

9 Sihweil, I. 1976. Interim Position on Ductility Ratios of Reinforced Concrete Structural Elements. Memorandum, USNRC, (SEB:306).

10 Stephenson, A.E. 1946. Full-Scale Tornado - Missile Impact Tests. Electric Power Research Institute, Interim Report EPRI NP-148.

11 Stevenson, J.D. 1974. 2nd Annual Short Course on Structural Design, Analysis and Testing of Nuclear Plant Equipment and Structures. Case Western Reserve University, Cleveland, Ohio.

12 Stone & Webster Engineering Corporation (SWEC) 1977. Missile Barrier Interaction, Topical Report.

3.5-13 Rev. 30

OUTSIDE CONTAINMENT REQUIRED FOR SAFE REACTOR SHUTDOWN*

Structures Location** FSAR Section FSAR Figure ntainment penetrations AB, EF, MS, 3.8.1 3.8-18, 3.8-19 HR ntainment hatches (personnel and AB, HR 3.8.1 3.8-21, 3.8-22 uipment) outside containment ntainment reinforced concrete external CS 3.8.1 3.8-1 thru erstructure 3.8-14 and 3.8-19 ble tunnel from containment structure to CB, AB, SB 3.8.4 3.8-62 thru ntrol building 3.8-67 and 8.3-1 ilding reinforced concrete walls AB, EF, FB, 3.8.4 3.8-62 thru rounding safety related systems and SB 3.8-67 mponents in steam and feedwater valve Areas MS 3.8.4 3.8-62 thru 3.8-67 gineered safety features areas EF 3.8.4 3.8-62 thru 3.8-67 ntrol building CB 3.8.4 8.3-1 rvice water pumphouse SP 3.8.4 3.8-69 and 3.4-1 esel generator building DG 3.8.4 3.8-68, 3.8-69 esel generator fuel oil pump House OY 3.8.4, 3.8.5 3.8-68, 3.8-69 stems emical and Volume Control System 9.3.4 9.3-7 and 9.3-8 arging pumps AB al water injection filter AB ric acid tanks AB ric acid transfer pumps AB ing, valves, and portions of the system AB uired for charging, seal water injection d boration 3.5-14 Rev. 30

Structures Location** FSAR Section FSAR Figure trumentation, cable, and controls required AB, CR, SB perform a safety function in the above rtions of system sidual Heat Removal System 5.4 5.4-4 sidual heat removal pumps EF sidual heat removal exchanger EF Tube side Shell side ing and valves required to cool and EF, AB intain the RCS in a cold shutdown ndition trumentation, cable, and controls required EF, AB, SB, perform a safety function in the above CR rtions of system ergency Core Cooling System 6.3 6.3-1 thru 6.3-4 gh-head safety injection pumps (charging AB mps) w-head safety injection pumps (residual EF at removal pumps) ing and valves required for injection to EF, AB RCS trumentation, cable and controls required EF, AB, SB, perform a safety function in the above CR rtions of system ench Spray System 6.2.2 6.2-36 fueling water storage tank, quench spray EF, OY mps, piping and valves required to form intended safety functions trumentation, cables and controls required EF, OY, CR perform a safety function in the above rtions of system ergency Generator Fuel Oil 9.5.4 9.5-2 rage and Transfer Systems 3.5-15 Rev. 30

Structures Location** FSAR Section FSAR Figure ergency generator fuel oil day Tanks DG ergency generator fuel oil transfer pumps DG fuel oil transfer Pump house ergency generator fuel oil storage tanks OY (Underground) ing and valves required to perform a OY/DG ety function trumentation, cable, and controls required DB / OY IR /

perform a safety function in the above ER rtions of system ergency Generator Cooling Water System 9.5.5 9.5-3 ergency generator cooling water DG changer oling water electric immersion heater DG ssure housing ergency generator cooling water DG pansion tank tor driven circulating water pump DG ing and valves required to perform a DG ety function trumentation and controls required to DG form a safety function in the above rtions of system ergency Generator Starting System 9.5.6 9.5-3 ergency generator air starting storage DG ks ing and valves required to perform a DG ety function trumentation, cable, and controls required DG perform a safety function in the above rtions of system ergency Generator Lubrication System 9.5.7 9.5-3 3.5-16 Rev. 30

Structures Location** FSAR Section FSAR Figure ergency generator lubrication oil cooler DG ergency generator lubrication oil filter DG ing and valves required to perform a DG ety function trumentation and controls required to DG form a safety function in the above rtions of system ergency Generator Air Intake and 9.5.8 9.5-4 haust System r filter DG r intake silencer DG haust silencer DG ing required to perform a safety function DG rvice Water System 9.2.1 9.2-1 rvice water pumps SP ing and valves supplying cooling water to AB, EF, OY ety related equipment (buried) trumentation, cable, and controls required EF, DG, CB perform a safety function in the above rtions of system actor Plant Component Cooling 9.2.2.1 9.2-2 ter System actor plant component cooling pumps AB actor plant component cooling surge tank AB actor plant components cooling heat AB changers ing and valves supplying cooling water to AB, SB ety related equipment required for tdown trumentation, cable and controls required AB, CB perform a safety function in the above rtions of system 3.5-17 Rev. 30

Structures Location** FSAR Section FSAR Figure am Generator Blowdown System 10.4.8 10.3-1 ing and valves from containment up to MS d including the blowdown isolation valves trumentation, cabling, and controls MS uired to trip close the isolation valves in Steam System 10.3 10.3-1,2,3,4 in steam piping and valves from MS ntainment up to and including main steam lation stop valves and all piping in the ve house in steam piping and valves from main MS am lines to steam generator auxiliary dwater pump turbine trumentation, cable, and controls required MS / CR / IR trip close the MS stop valves and bypass ves am generator atmospheric relief valves MS n be operated manually) xiliary Feedwater System 10.4.9 10.4-9 mineralized water storage tank OY xiliary feedwater pumps (turbine- and EF tor-driven) ing and valves supplying auxiliary OY, EF dwater from ST to containment netration trumentation, cable and controls required perform a safety function in the above rtions of system edwater System 10.4.7 10.4-8 edwater piping and valves from MS ntainment up to and including the dwater isolation valves 3.5-18 Rev. 30

Structures Location** FSAR Section FSAR Figure trumentation, cable, and controls required trip close the feedwater isolation valves d to trip the main feedwater pumps r Conditioning, Heating, Cooling 9.4 d Ventilation Systems ergency generator enclosure ventilation EGE 9.4.6 9.4-3 gineered safety features area unit cooler EF / AB 9.4.3 9.4-2 d ductwork ntrol building heating and ventilation, CB 9.4.1 9.4-1 luding control building chilled water tem rvice water pumphouse ventilation SP 9.4.8 9.4-3 trumentation and control required to EF / EGE form a safety related function for the / CB / AB ove portions of system hting System 9.5.3 None ergency lighting AB/CR ctrical Systems 8.3 8.3-1 thru 8.3-3 ergency generators EGE tteries, chargers, and battery switchboards ER al bus panels and inverters ER sential unit substations ER sential metalclad switchgear ER sential motor control centers ER ntrol panelboards: CR Main control board and panels ergency generator panel ER xiliary shutdown panel ER ntrol room air-conditioning control panel CR 3.5-19 Rev. 30

Structures Location** FSAR Section FSAR Figure bling and raceway supports for safety All locations ated equipment required for safe shutdown outside containment tors for safety related equipment required Same as safe shutdown components trumentation and cables required for the All locations ety related portion of the above electrical outside tems containment actor Trip Systems portions of the reactor trip and safeguards CR / IR 7.2, 7.4 7.2-1 thru uation channels including sensors, 7.2-4 cuitry, and processing equipment required control and safety shut down the reactor to cold shutdown condition (the protection cuits used to trip the reactor on dervoltage, underfrequency, and turbine are excluded) ntrols for defeating automatic safety CR / IR 7.3, 7.4 None ection actuation during a cooldown and pressurization ication of the following plant parameters 7.4, 7.5, 7.7 7.7-1, 7.7-4, uld be available to the operator: 7.7-5, 7.7-6 ication of plant parameters required for CR/IR e shutdown (see Table 7.5-1) ntainment Isolation System ntainment isolation valves and associated AB, EF, MS 6.2.4 6.2-47 ing for all systems penetrating ntainment structure trumentation, cable and controls required AB perform a safety function in the above rtions of system TES:

The applicable seismic category and quality group classification for the equipment listed in this table is delineated in Table 3.2-1.

Location Symbols 3.5-20 Rev. 30

CR - Control room (in control building)

CS - Containment structure EGE - Emergency generator enclosure EF - Engineered safety features building ER - Emergency switchgear room (in control building)

IR - Instrument rack room (in control building)

MS - Main steam valve enclosures SP - Service water pumphouse OY - Outside, yard AT - Auxiliary feedwater storage tank HR - Hydrogen recombiner building SB - Service building FB - Fuel building 3.5-21 Rev. 30

INSIDE CONTAINMENT REQUIRED FOR SAFE REACTOR SHUTDOWN*

uctures Location ** FSAR Section ntainment Structure 3.8.1, 3.8.3 Reinforced concrete interior substructure CS Reinforced concrete interior shields and walls CS Containment structure liner CS Piping and duct penetrations CS Electrical penetration/assemblies CS Personnel access lock CS Equipment hatch CS stems actor Coolant System 5.1, 5.2, 5.3, 5.4 Reactor vessel CS Control rod drive mechanisms CS Steam generators CS Pressurizer CS Piping, fitting and valves within RCPB CS Reactor coolant thermowell CS Reactor coolant thermowell boss CS Safety valves CS Relief valves CS Control rod drive mechanism head adapter plugs CS Isolation valves for reactor coolant system branch CS line Reactor coolant pump CS Reactor coolant pump casing CS Main flange Thermal barrier stems Location ** FSAR Section No. 1 seal housing 3.5-22 Rev. 30

No. 2 seal housing Bolting Impeller Diffuser Reactor coolant pump motor CS Rotor Flywheel Shaft Shaft coupling Bearings Upper oil cooler CS

- Tube side - ccw

- Shell side - oil Lower oil cooler CS

- Tube side - ccw

- Shell side - oil Instrumentation cable and controls required to CS perform a safety function in the above portion of system emical and Volume Control System 9.3.4 Piping, valves, and portions of the system required CS for charging, seal water injection, and boration Instrumentation, cable, and controls required to CS perform a safety function in the above portion of system sidual Heat Removal System 5.4 Piping and valves required to cool and maintain the CS RCS in a cold shutdown condition Instrumentation, cable, and controls required to CS perform a safety related function in the above portions of system 6.3 3.5-23 Rev. 30

ESF pumps CS Piping and valves required for injection to the RCS CS stems Location** FSAR Section Instrumentation, cable and controls required to CS perform a safety related function in the above portions of system ntainment Isolation System 6.2.4 Containment isolation valves and associated piping CS for all systems penetrating containment structure Instrumentation, cable and controls required to CS perform a safety related function in the above portions of system actor Plant Component Cooling Water System 9.2.2.1 Piping and valves supplying cooling water to CS safety related equipment required for shutdown Instrumentation, cable and controls required to CS perform a safety related function in the above portions of system xiliary Feedwater System 10.4.9 Piping and valves supplying auxiliary feedwater to CS the steam generators edwater System 10.4.7 Piping and valves supplying auxiliary feedwater to CS the steam generator

-Conditioning, Heating, Cooling, and Ventilation 9.4.7 stems Containment recirculation unit coolers and CS associated equipment excluding chillers Instrumentation, cable, and controls required to CS perform a safety related function for the above air-conditioning, heating, cooling and ventilation systems ctrical Systems 8.3 3.5-24 Rev. 30

Cabling and raceways for safety related equipment CS required for safe shutdown Motors for safety related equipment required for CS safe shutdown stems Location ** FSAR Section actor Trip System See Table 3.5-1 Those portions of the reactor trip system listed in CS Table 3.5-1 that are located inside containment TES:

The applicable seismic category and quality group classification for the equipment is listed in Table 3.2-1.

Location Symbols CS - Containment Structure 3.5-25 Rev. 30

ANALYSIS Missile Impact Kinetic Typical Examples of Weight Velocity Energy Penetration Postulated Missiles (lb) (ft/sec) (ft-lb) (in) Assumptions Mechanism housing cap 11 90 1,380 0.05 Plug becomes loose and is accelerated by the water jet Mechanism top cap and 133 150 46,757 0.08 Drive shaft further drive rod assembly pushes the plug into impacting on the same the shield missile shield spot Drive shaft latched to 1,500 12.1 1,490 0.057 -

mechanism 3.5-26 Rev. 30

Weight to Impact Flow Thrust Impact Area Weight Discharge Area Area Ratio Velocity Missile Description (lb) Area (in2) (in2) (in2) (psi) (fps) fety valve bonnet 350 2.86 80 24 14.6 110 inches x 6 inches) or nches x 6 inches nch motor-operated isolation 400 5.5 113 28 14.1 135 ve bonnet us motor and stem) nch solenoid-operated relief 75 1.8 20 20 3.75 115 ve bonnet (plus stem) nch air-operated spray valve 200 9.3 50 50 4 190 TE:

ypical values based upon manufacturer design data 3.5-27 Rev. 30

TABLE 3.5-5 PIPING TEMPERATURE ELEMENT ASSEMBLY -

MISSILE CHARACTERISTICS For a tear around the weld between the boss and the pipe:

Characteristics Without Well With Well w Discharge Area (in2) 0.11 0.60 rust Area (in2) 7.1 9.6 ssile Weight (lb) 11.0 15.2 ea of Impact (in2) 3.14 3.14 ssile Weight pact Area (psi) 3.5 4.84 locity (ft/sec) 20.0 120.0 For a tear at the junction between the temperature element assembly and the boss for the without well element and at the junction between the boss and the well for the with well element.

Characteristics Without Well With Well scharge Area (in2) 0.11 0.60 rust Area (in2) 3.14 3.14 ssile Weight (lb) 4.0 6.1 ea of Impact (in2) 3.14 3.14 ssile Weight pact Area (psi) 1.27 1.94 locity (ft/sec) 75.0 120.0 3.5-28 Rev. 30

POSTULATED WITHIN REACTOR CONTAINMENT Reactor Coolant Pump Temperature Instrument Well of Pressurizer Characteristics Sensor Pressurizer Heaters ight (lb) 0.25 5.5 15 scharge Area (in2) 0.50 0.442 0.80 rust Area (in2) 0.50 1.35 2.4 pact Area (in2) 0.50 1.35 2.4 ssile Weight pact Area (psi) 0.5 4.1 6.25 locity (ft/sec) 260 100 55 3.5-29 Rev. 30

High Trajectory Targets Target Area (sq ft) R * (ft) y ** (ft)

1. Control Building 11,730 216 15.5
2. Emergency Generator Enclosure 3,999 303 -13.0
3. Fuel Oil Storage Tanks 1,736 301 -44.5
4. Auxiliary Building 15,975 265 24.5
5. Main Steam Valve Building 1,856 171 14.9
6. Fuel Building 9,701 346 37.0
7. Engineered Safeguards Building 8,241 333 24.0
8. Demineralizer Water Storage Tank 1,133 423 11.0
9. Refueling Water Storage Tank 2,641 406 87.0
10. Hydrogen Recombiner Buildings 720 307 -17.7
11. Containment Structure 8,721 256 108.3
12. Service Water Pumps 2,750 570 -45.0 Target Area Low Trajectory Targets (sq ft) R * (ft) y ** min (ft) y ** max (ft)
1. Auxiliary Building Wall *** 0 234 14.9 24.5
2. Main Steam Valve Building Wall 1,560 157 -12.0 14.9
3. Refueling Water Storage Tank Wall 816 383 -17.7 18.0
4. Hydrogen Recombiner Building Wall 819 295 -45.0 -17.7
5. Containment Wall - Section 1 2,657 188 14.9 86.9
6. Containment Wall - Section 2 2,372 188 -45.0 86.9 3.5-30 Rev

= average distance from turbine

= height above spin axis of turbine Auxiliary building wall can be hit by missiles from only one wheel 3.5-31 Rev. 30

TABLE 3.5-8 DELETED BY FSARCR 02-MP3-13 3.5-32 Rev. 30

TABLE 3.5-9 DELETED BY FSARCR 02-MP3-13 3.5-33 Rev. 30

TABLE 3.5-10 DELETED BY FSARCR 02-MP3-13 3.5-34 Rev. 30

TABLE 3.5-11 DELETED BY FSARCR 02-MP3-13 3.5-35 Rev. 30

TABLE 3.5-12 DELETED BY FSARCR 02-MP3-13 3.5-36 Rev. 30

Horizontal Weight Velocity **

Missile Description * (lbs) (mph) 4 x 12 in plank, 12 ft long at 50 lb/ft3 200 210 Utility Pole 13.5 inch diameter, 35 feet long at 43 lb/ft3 1,496 120 1 inch solid steel rod, 3 feet long 8 130 3 inch schedule 40 pipe, 10 feet long 78 140 6 inch schedule 40 pipe 15 feet long 285 120 12 inch schedule 40 pipe, 15 feet long 743 110 Automobile frontal area 20 ft2 4,000 50 TES:

Missiles 1 through 6 are considered at all elevations. Maximum trajectory height of Missile 7 is 25 feet above grade.

Vertical velocities of 70 percent of horizontal velocities are used, except for Missile 3 which has the same speed in all directions.

3.5-37 Rev. 30

TORNADO WIND Maximum Elastic Barrier Mass Velocity Momentum Barrier Barrier Deflection Deflection (Xm) Barr Missile Description (slugs) (fps) (lb sec) Span (ft) (Xy) (ft) (ft) Ducti 12 inch schedule 40 pipe 23.07 161.33 3721.96 5. 0.0038 0.0069 1.7 15 feet long 743 lb 10. 0.0154 0.0266 1.7

20. 0.0615 0.0398 0.6
30. 0.1383 0.0489 0.3 4,000 lb auto 124.2 73.33 9108. 5. 0.0038 0.0295 7.6
10. 0.0154 0.0202 1.3
20. 0.0615 0.0414 0.6
30. 0.1383 0.0843 0.6 13.5 inch diameter utility 46.46 176 8177 5. 0.0038 0.0031 0.8 pole 35 feet long 1,496 lb
10. 0.0154 0.0122 0.7
20. 0.0615 0.0499 0.8
30. 0.1383 0.0871 0.6 Barrier Information:

Reinforcement: #11 at 10 inches GR 40 each way, each face Concrete: 3,000 psi Barrier Depth: 24 inches 3.5-38 Rev

WIND Maxim Barrier Maximum Maxim Elastic Barrier Barrier Barrier Allowa Barrier Missile Mass Velocity Momentum Deflection Deflection Ductility Compression Ductili Description Description (slug) (fps) (lb sec) (ft) (ft) Ratio Load (#) Ratio Refer to 13.5 inch diameter 46.46 176 8177 0.03 0.03 10. Barrier dead 10.

Figure 3.5-6 utility pole 35 ft load only long 3.5-39 Rev

MPS3 UFSAR 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURES OF PIPING This section addresses the subject of a postulated pipe break/crack inside and outside the containment. It also presents the results of analyses initiated in response to NRC Regulatory Guide 1.46 for inside containment and the Giambusso letter of December 18, 1972, for outside containment. The methods of evaluation, however, reflect the approach and methodology contained in the Branch Technical Positions ASB 3-1 and MEB 3-1 as qualified in Sections 3.6.1 and 3.6.2.

3.6.1 POSTULATED PIPING FAILURES IN FLUID SYSTEMS INSIDE AND OUTSIDE OF CONTAINMENT This section describes the design criteria and bases for protecting essential equipment from the effects of piping failures inside and outside of containment. These criteria and bases used in the design ensure that:

1. Functions of essential equipment necessary to maintain the capability of a safe plant shutdown during any piping failures are preserved.
2. The plant can be safely shut down to a cold shut down condition without offsite power.
3. Offsite doses in excess of applicable guidelines do not occur.

As used in this section, essential equipment is defined as the structures, systems, portions of systems, and components required to mitigate the effects of the postulated pipe rupture, to shut down the plant safely to a cold condition without offsite power, and to maintain the cold shutdown condition assuming a concurrent single active failure.

3.6.1.1 Design Bases 3.6.1.1.1 Design Basis Protection Criteria Within the plant there are high and moderate energy piping systems which are postulated to fail.

Failures of these systems are evaluated to determine their deleterious effects on essential equipment and those essential systems, components, and structures that are susceptible to the effects of these failures are provided protection. The criteria for protection against pipe rupture inside the containment are contained in NRC Regulatory Guide 1.46, as described in Section 1.8, the criteria for protection against pipe rupture outside containment conform to the NRC Technical Branch Position ASB 3-1. Section 3.6.1.1.4 discusses design features recommended for protection.

3.6-1 Rev. 30

MPS3 UFSAR 3.6.1.1.2 Design Basis Pipe Break/Crack Criteria The design basis pipe breaks are postulated to occur in high energy systems (or portions of systems) in accordance with the break criteria (Section 3.6.2.1). A high energy system is defined as a fluid system which operates during normal plant operating conditions and has a maximum operating pressure exceeding 275 psig or a maximum operating temperature exceeding 200°F.

Figure 3.6-1 presents these boundaries graphically, including the differences between inside and outside the containment.

Maximum operating temperature and pressure are the maximum temperature and pressure in the fluid systems during occurrences which are expected during normal plant conditions. Normal plant conditions are defined as startup, operation at power, hot standby, or reactor cooldown to cold shutdown conditions.

Through-wall leakage cracks are postulated to occur in moderate energy systems (located outside containment only) in accordance with Sections 3.6.2.1.2 and 3.6.2.1.3.

A moderate energy system is defined as a fluid system or portion of a fluid system which is pressurized during normal plant operating conditions but which has both a maximum operating pressure equal to 275 psig or less and a maximum operating temperature equal to 200°F or less.

3.6.1.1.3 Essential Systems, Components, and Structures Table 3.6-5 lists all essential equipment required to shut down the plant safely and to mitigate the consequences of postulated piping failures. Figures 3.6-2 through 3.6-7 show the relative locations of essential equipment within the plant. Thus, a determination can be made of which high and moderate energy systems are remote from or proximate to essential systems, components, and structures.

3.6.1.1.4 Design Approach The following subsections describe the design features provided to protect essential systems, components, and structures and to mitigate the consequence of piping failures.

3.6.1.1.4.1 Separation A primary objective in the piping layout and plant arrangement is to satisfy the separation criteria, so that the effects of postulated pipe breaks at any location are isolated or physically remote from essential structures, systems, and components.

Redundant essential equipment is located in separate cubicles so that a pipe failure in one cubicle cannot affect the backup equipment. Inside each redundant cubicle, high energy piping is located with as much separation as possible from the essential equipment so that the effects of pipe whip or jet impingement are minimized.

3.6-2 Rev. 30

MPS3 UFSAR Whenever possible, safety related electrical and control system equipment is located in areas which do not have piping (e.g., the emergency switchgear area). Cable tray runs and instrumentation process lines which cannot be separated from piping runs are located as far as practical from postulated piping failure locations. Every attempt is made to physically separate redundant cable runs and instrumentation lines. For all safety components where physical separation is impossible, one or more of the design features described below are provided to ensure that essential equipment remains operable.

3.6.1.1.4.2 Enclosures Where remote physical separation is not practical, protective enclosures are used. For example, each charging pump is located in a separate cubicle with concrete walls between redundant pumps. Redundant electrical and control systems in the area of high energy piping are located as far as possible from postulated pipe break locations so that potential damage is minimized.

Section 3.8.4 discusses the concrete design of these enclosures.

3.6.1.1.4.3 Restraints Where neither of the above design approaches is feasible, protection is provided by pipe restraints to limit pipe movement following a rupture to restrict the area affected by the failure. Placement of restraints is based on postulated pipe break locations. The design of the restraints for these locations limits pipe whip to acceptable levels and reduces jet impingement thereby preventing damage to essential equipment. Section 3.6.1.3.3 describes the design of pipe rupture restraints.

3.6.1.2 Description Every attempt, whenever practical, is made to satisfy the separation criteria by locating high and moderate energy systems physically remote from essential equipment. Where plant arrangement and/or piping layout does not permit complete physical separation, protective enclosures are provided. When these methods are not practical, restraints are installed to protect essential equipment.

Table 3.6-1 lists all high energy systems remote from essential systems, components, and structures. A major portion of these systems is in the turbine building where no essential equipment is located, thus satisfying the physical separation criterion. Pipe breaks are only postulated for the main steam line in this area due to its proximity to the control building.

Table 3.6-2 summarizes all high energy systems that are located in proximity to essential systems, components, and structures. The majority of essential systems, and components are found in the Engineered Safety Features, Auxiliary Building, and the Containment Structure. Pipe breaks and through wall leakage cracks are postulated in systems that are proximate to these essential systems, and components.

Table 3.6-3 contains all moderate energy systems outside the containment separated from essential systems, components, and structures by plant arrangement and/or piping layout.

Therefore, these systems are not evaluated for pipe rupture.

3.6-3 Rev. 30

MPS3 UFSAR Table 3.6-4 includes all moderate energy systems located outside containment in proximity to essential systems, components, and structures. Through-wall cracks are postulated in these systems or portions of systems to evaluate the effects of fluid spray and flooding. Jet impingement effects from cracks are not evaluated.

Each system listed in Table 3.6-2 is referenced by pipe break location drawing (figure number), as applicable, showing all postulated pipe breaks. Section 3.6.1.3.3 provides a summary of all postulated pipe breaks. It encompasses their detrimental effects on essential equipment along with single active failure criteria, loss of offsite power, seismic events, and all other design considerations.

3.6.1.3 Safety Evaluation 3.6.1.3.1 Operability of Essential Systems and Components The operability of an essential system or a component following a postulated piping failure may be dependent on one or more of the following assumptions in addition to consideration of the effects of the failure.

Single Failure Criterion A single active component failure is assumed to occur in essential systems used to mitigate consequences of the postulated piping failure and to shut down the reactor. The single active component failure is assumed to occur in addition to the postulated piping failure and any direct consequences of the piping failure, such as unit trip and loss of offsite power. Section 3.1.1 defines this failure criterion and its applications.

Loss of Offsite Power Offsite power is assumed unavailable if a trip of the turbine generator system or reactor protection system is a direct consequence of the postulated piping failure. However, a single failure of one emergency generator or one Class 1E bus can be assumed as the single failure if this assumption is the most limiting.

Seismic Event Credit for mitigating the consequences of a postulated event may be taken only for those systems and components designed to Seismic Category I requirements, except for the High Energy Line Break (HELB) isolation systems for the auxiliary steam and the hot water heating systems.

Although the electrical sensing and isolation devices are Category I and located in a Category I building, the valves are not Category I as they are installed in a non-Category I piping system located in a non-Category I structure. The location of the isolation valves provides for maximum isolation capability outside the area for which isolation is intended.

All available systems, including those actuated by operator actions, are used to mitigate the consequences of a postulated event. Judging the availability of systems includes consideration of 3.6-4 Rev. 30

MPS3 UFSAR the postulated failure and its direct consequences (e.g., unit trip and loss of offsite power) and the assumed single active component failure plus its direct consequences. The feasibility of the operator to take action is judged on the availability of ample time and adequate access to equipment for performing the proposed actions. Regulatory Guide 1.62 provides guidance in evaluating the feasibility of operator action.

3.6.1.3.2 Failure Mode and Effects An analysis of breaks in high energy systems, cracks in moderate energy systems, and the consequent failure modes and effects (e.g., environmental, pipe whip, and jet impingement) must include consideration of their sources and targets. The source comprises the pipe which is postulated to fail and the resulting effects of the failure. The target comprises structures, systems, and components considered essential for shutting down the plant safely, maintaining the safe shutdown, and mitigating the effects of the postulated pipe failure.

Interactions between sources and targets are analyzed individually to determine how each affects essential equipment in the area of the source. The interactions analyzed are pipe whip, jet impingement, and environmental effects.

Pipe Whip Section 3.6.2.1 describes the criteria for determining break locations in piping systems.

Section 3.6.2.2 describes the pipe whip analysis for defining forcing functions and pipe responses.

The effects of a particular piping failure are evaluated for each target required to mitigate the consequences of the failure. This evaluation is based on the location of the break and the forces generated by the whipping pipe. The targets impacted by the pipe are either designed to withstand these forces or are protected by rupture restraints (Section 3.6.1.3.3).

Jet Impingement The blowdown forces are calculated (Section 3.6.2.2), and the location and direction of the resulting jet of fluid determined, for each piping failure which could affect an essential target.

Essential targets impacted by the fluid must either be evaluated for effects of these jet forces or shielded as discussed in Section 3.6.2.2.

Environmental Effects The environmental effects of postulated pipe breaks are established as design basis environmental conditions for essential equipment affected by these pipe breaks and are included as environmental qualification criteria for the essential equipment. Section 3.11 describes the environmental qualification of essential equipment. The following environmental effects are considered:

1. Fluid spray - Essential equipment located proximate to high and moderate energy fluid systems are either designed to withstand the most severe environmental conditions resulting from fluid spray without loss of safety function or the most 3.6-5 Rev. 30

MPS3 UFSAR severe environmental conditions resulting from fluid spray which are humidity of 100 percent, maximum expected temperature due to the ruptured fluid system, and water spraying essential equipment.

2. Flooding - Compartments and areas containing essential equipment are examined for flooding potential. Flooding height is based on the leak rate from the piping failure and the time required to detect and isolate the leak. Compartment floor and wall structural integrity is maintained when subjected to the hydrostatic head resulting from the calculated flood level.
3. Pressure buildup in compartments - The calculation of pressure and temperature buildup in compartments is based on the maximum operating fluid conditions in the pipe which fails, the available vent area in the cubicle, and the volume within the cubicle. Pipe failure locations are determined as described in Section 3.6.2.

The results of the pressure calculations indicate the required pressure differential design or the vent area required for building compartments. Maximum calculated temperatures are used to determine essential equipment qualification requirements.

3.6.1.3.3 Pipe Break/Crack Analysis Each pipe break and crack postulated in the high and moderate energy systems, listed in Tables 3.6-2 and 3.6-4, have been systematically analyzed for their detrimental effects on safe shutdown targets. The following paragraphs describe in detail the interactions between the targets and the postulated breaks and cracks in each of the piping systems identified herein:

1. Main Steam System
2. Feedwater System
3. Reactor Coolant System
4. Chemical and Volume Control System
5. Steam Generator Blowdown System
6. Auxiliary Feedwater System Main Steam System Break Locations Pipe breaks in the main steam piping inside the containment are postulated at terminal ends and intermediate locations defined in Section 3.6.2.1.1. The main steam lines from steam generators 3RCS*SG1A and 3RCS*SG1D up to the containment penetrations are mirror-images of each other; therefore, pipe breaks are postulated at identical locations for both lines. Similarly, the other pair of main steam lines originating from steam generators 3RCS*SG1B and 3RCS*SG1C 3.6-6 Rev. 30

MPS3 UFSAR are symmetrical and the pipe break locations in the individual line are identical. Only one line of each pair is evaluated for the effects of the postulated piping failure; however, the results are also applicable to the other identical lines.

The portion of main steam piping between the containment penetration (inboard side) and the first restraint outboard of the isolation valve is considered a break exclusion zone. The stress distribution along this section of the main steam piping satisfies the break exclusion requirement (ASME Code,Section III, NC-3600; also Section 3.6.2):

Eq. (9) + Eq. (10) 0.8(1.2Sh + SA) (3.6.1)

The remainder of the main steam piping is physically located remote from essential systems, components, and structures except for a relatively short section near the control building.

However, breaks are postulated in accordance with Section 3.6.2.1.2 to ensure that protection is provided to preserve the integrity of the containment penetrations and the operability of the isolation valves in the event that pipe rupture should occur at any of the postulated locations.

Consideration is given to potential damage to the control building wall which may adversely affect the safety related equipment in the control room. Figures 3.6-8 and 3.6-9 show all postulated pipe breaks inside and outside of the containment. Table 3.6-6 lists the postulated break locations.

No single active failure has been assumed concurrent with breaks which are postulated in the break exclusion zone since the break is postulated to evaluate environmental effects only.

Separation As mentioned above, the main steam piping outside the containment, extending from the main steam valve building wall (Col. F, Figure 3.6-8) to the turbine building satisfy the physical separation criterion. Pipe breaks, however, are postulated in this region to assure that occurrence of such breaks do not jeopardize the integrity of the containment penetrations and operability of the isolation valves and essential equipment located in the control building.

Pipe Whip Effects Table 3.6-7 summarizes the results of pipe whip analysis inside and outside of containment. To prevent whipping of the main steam piping inside the containment, whip restraints are provided to protect the crane wall and the adjoining containment wall. Similarly, no whipping is permitted for the main steam piping outside containment to prevent damage to containment penetrations, isolation valves, and control building wall by placing restraints at specified locations shown on Figure 3.6-8. Details of these restraints are discussed in the next section.

Pipe Restraints The pipe rupture restraints on the Main Steam System are shown on Figure 3.6-8 along with postulated pipe rupture locations. Notes to Figure 3.6-8 provide location, type, and function for each pipe rupture restraint.

3.6-7 Rev. 30

MPS3 UFSAR Jet Impingement Effects Jet impingement effects are addressed in Table 3.6-8.

Environmental Effects Refer to Section 3.11B for information.

Conclusion Adequate protection and/or separation is provided for essential systems and components required to mitigate the consequences of main steam piping ruptures.

Essential systems, structures, and components remain functional subsequent to a postulated main steam line break.

Feedwater System Break Locations Pipe breaks in the feedwater piping inside the containment are postulated at terminal ends and intermediate locations in accordance with Section 3.6.2.1.1.

Pipe breaks are excluded in the main steam valve building between the containment penetration (inboard side) and the first restraint on the 20-inch feedwater lines and between the tap into the 20-inch line and the first restraint on the 8-inch bypass line, inasmuch as the stress distribution along this portion of the feedwater lines fully met the break exclusion requirement (Equation 3.6-1).

The portion of feedwater lines upstream of the break exclusion zone are postulated at terminal ends and intermediate locations in accordance with Section 3.6.2.1.2. Figures 3.6-10 and 3.6-11 show all the terminal and intermediate breaks inside and outside of containment. Table 3.6-9 lists the postulated break locations.

Separation The feedwater piping upstream of the break exclusion zone, extending from the main steam valve building wall (Col. F) to the turbine building fully meet the physical separation criterion, thereby precluding potential damage to essential equipment. Pipe breaks, however, are postulated to evaluate the effects of these breaks on the break exclusion zone to ensure that the integrity of the containment penetrations and the operability of the isolation valves are not lost in the event that this piping should fail.

Pipe Whip Effects 3.6-8 Rev. 30

MPS3 UFSAR Table 3.6-10 summarizes the results of pipe whip analysis inside and outside containment.

Whipping of feedwater piping inside the containment is not allowed except in the area where the whipping pipe cannot potentially damage an essential system, component, or structure. Such a case is confined only to the terminal breaks at the steam generator feedwater nozzles. The relatively short section of the feedwater piping in the turbine building is separated from the control building wall by approximately 40 feet at their nearest approach. A whipping feedwater pipe can not impact the wall.

Pipe Restraints The pipe rupture restraints on the Main Feedwater System are shown on Figure 3.6-11. The postulated pipe rupture locations are shown on Figure 3.6-10. Notes to Figure 3.6-11 provide location, type, and function for each pipe rupture restraint.

Jet Impingement Effects Jet impingement effects are addressed in Table 3.6-11.

Environmental Effects Refer to Section 3.11 for information.

Conclusion Adequate protection and/or separation is provided for essential systems and components required to mitigate the consequences of piping ruptures.

Reactor Coolant System Determination of the design basis break locations in the reactor coolant system are based on the criteria defined in Regulatory Guide 1.46. It takes into account certain modifications or exceptions specified in Section 1.8 and qualified in Section 3.6.2.1. For clarity of analysis, discussion of breaks and their consequential effects fall into three separate areas: 1. primary coolant piping (including Loop Stop Valve Bypass Piping, Excess Letdown Piping, Loop Drain Piping, Letdown Line, and Normal Charging Line) 2. pressurizer cubicle piping, and 3. safety injection system.

1. Primary Coolant Piping Break Locations Figures 3.6-12 and 3.6-33 show all pipe breaks in the primary coolant piping at postulated locations defined in Section 3.6.2.1. Each break location and orientation is determined on the basis of stress and fatigue analysis. Table 3.6-12 summarizes the postulated break locations. Only one loop (loop A) of the four identical loops is analyzed; however, results of analysis are applicable to all loops by virtue of their symmetry.

3.6-9 Rev. 30

MPS3 UFSAR The mechanistic effects of pipe breaks in the RCS hot leg, cold leg, and crossover leg need not be addressed due to the GDC4 exemption for primary loop pipe breaks.

Separation Protection criteria dictate physical separation of each of the reactor coolant loops in order to assure the continued integrity and operability of essential systems, components, and structures required to safely shut down the plant in the event of a loss of coolant accident, main steam, or feedwater break. This protection requirement is achieved by isolating each reactor coolant loop within a cubicle. Section 3.8-1 discusses the design of the cubicle.

Pipe Whip Effects Table 3.6-13 summarizes the results of pipe whip analysis of the reactor coolant loop bypass piping. Whipping of the ruptured coolant bypass piping originating at any postulated break locations jeopardizes the integrity and operability of the safe-shutdown targets identified in the table. To meet the relevant protection criterion whip restraints are installed at locations where whipping is prevented. Details of these restraints are as follows.

Pipe Restraints The pipe rupture restraints on the Reactor Coolant System Primary Coolant Piping are shown on Figure 3.6-12. The postulated pipe rupture locations are shown on Figure 3.6-12 for the Loop Stop Valve Bypass lines. Notes to Figure 3.6-12 provide location, type, and function for each pipe rupture restraint.

The pipe rupture restraints originally installed to prevent primary loop pipe whip have been deactivated due to the GDC4 exemption.

Jet Impingement Effects Jet impingement effects are addressed in Table 3.6-14.

Environmental Effects Refer to Section 3.11 for information.

Conclusion Adequate protection and/or separation is provided for essential systems and components required to mitigate the consequences of RCS loop (including Loop Stop Valve Bypass Piping, Excess Letdown Piping, Loop Drain Piping, Loop Fill Piping, Letdown Line, and Normal Charging Line) piping ruptures.

2. Pressurizer Cubicle Piping 3.6-10 Rev. 30

MPS3 UFSAR Break Locations Figure 3.6-14 shows the design basis breaks postulated in accordance with the rules specified in Section 3.6.2.1. Table 3.6-15 summarizes the postulated break locations.

Separation The pressurizer tank located in proximity to the reactor coolant loop is enclosed in a cubicle to satisfy the protection criterion. The pressurizer surge line, however, penetrates the cubicle wall as it connects to the hot leg of the reactor coolant loop (Loop B). Terminal and intermediate breaks are postulated in the surge line to assess their effects (pipe whip, jet impingement and environment) on the integrity of the pressurizer tank, pressurizer cubicle, and the reactor coolant system components located within the immediate vicinity of the postulated breaks.

Pipe Whip Effects Results of pipe whip analysis are summarized in Table 3.6-16.

Pipe Restraints The pipe rupture restraints on the Reactor Coolant System Pressurizer Cubicle Piping are shown on Figure 3.6-14. The postulated pipe rupture locations are shown on Figure 3.6-14.

Notes to Figure 3.6-14 provide location, type, and function for each pipe rupture restraint.

Jet Impingement Effects Jet impingement effects are addressed in Table 3.6-17.

Environmental Effects Refer to Section 3.11 for information.

Conclusion Adequate protection and/or separation is provided for essential systems and components required to mitigate the consequences of the Reactor Coolant System - Pressurizer Cubicle Piping ruptures.

3. Safety Injection System The safety injection system consists of the following major components:
1. Four accumulators which discharge borated water into each cold leg of the reactor coolant system.
2. Two charging pumps which supply borated water for cooling to each cold leg of the reactor coolant system (SI mode of CVCS).

3.6-11 Rev. 30

MPS3 UFSAR

3. Two safety injection pumps which supply borated water to each cold and hot leg of the reactor coolant system.
4. Two residual heat removal pumps which supply borated water for core cooling to each cold leg and two hot legs when the reactor coolant system pressure is low.

All of these components interconnect with the reactor coolant system which operate following a loss-of-coolant accident. Where the safety injection lines connect to the reactor coolant piping, the pressure boundary consists of the piping and valves leading to these connecting lines up to and including the second valve in each line. Pipe break postulation is confined only within the defined pressure boundary of the reactor coolant system with the remainder of the connecting systems excluded from the effects of pipe rupture.

Break Locations Figure 3.6-13 shows all postulated break locations within the reactor coolant pressure boundary except for the accumulator piping where break postulation is extended beyond the pressure boundary up to the accumulator tank discharge nozzle. Note that each pair of the accumulator tanks is located 180 degrees apart and that the main piping runs from each pair are symmetrical to each other. Only one piping run is analyzed for pipe breaks and their consequential effects; however, the results are applicable also to the remaining identical runs. Table 3.6-18 lists the postulated break locations.

Separation The safety injection accumulator tanks are conveniently located at the basement level of the containment structure, physically remote from high energy systems. Protection of each accumulator tank and its associated piping and valves from the effects of failure in the safety injection piping is further enhanced by judicious pipe routing, resulting in adequate physical separation.

Pipe Whip Effects Table 3.6-19 summarizes the pipe whip analysis.

Pipe Restraints The pipe rupture restraints on the Safety Injection System are shown on Figure 3.6-13. The postulated pipe rupture locations are shown on Figure 3.6-13. Notes to Figure 3.6-13 provide location, type, and function for each pipe rupture restraint.

Jet Impingement Effects Jet impingement effects are addressed in Table 3.6-20.

Environmental Effects 3.6-12 Rev. 30

MPS3 UFSAR Refer to Section 3.11 for information.

Conclusion Adequate protection and/or separation is provided for essential systems and components required to mitigate the consequences of safety injection system piping ruptures.

Chemical and Volume Control System The chemical and volume control system consists of charging, letdown, and seal water system; chemical control, purification, and makeup system; and boron thermal regeneration system.

Only the charging, letdown, and seal water system is systematically analyzed for pipe rupture, since it qualifies as a high energy system. Although the remaining systems function in conjunction with the charging and letdown lines, they are moderate-energy systems by definition.

Summaries for postulated pipe breaks and dynamic effects for the charging, letdown, and seal water subsystems of the CHS are given in Tables 3.6-21 through 3.6-29.

a. CHS Charging Lines:

Break Locations Figure 3.6-15 illustrates the design basis break locations.

Separation The charging lines inside and outside the containment fully meet the separation criteria, except the relatively short section that connects to the reactor coolant system. Separation of the charging lines from the essential equipment is achieved as follows:

Each charging pump is isolated in a cubicle so that a piping failure in one cubicle does not affect the other pumps and their associated piping and valves.

A portion of the charging line inside the containment is enclosed within the regenerative heat exchanger cubicle and the continuing piping run in the annulus area is physically routed away from safety related essential equipment. Similarly, the charging lines in the auxiliary building are routed in an area which precludes potential damage to essential equipment.

Pipe Whip Effects Since the separation criterion is fully met in the regenerative heat exchanger cubicle, the annulus area, and in the auxiliary building, prevention of pipe whip to mitigate its consequential effects in these areas is unnecessary. Table 3.6-22 provides the results of the pipe whip effects for the postulated pipe breaks on the CHS - Charging Line.

3.6-13 Rev. 30

MPS3 UFSAR Pipe Restraints The pipe rupture restraints on the CVCS charging line are shown on Figure 3.6-15 along with the postulated pipe rupture locations. Notes to Figure 3.6-15 provide location, type, and function for each pipe rupture restraint.

Jet Impingement Effects Jet impingement effects are addressed in Table 3.6-23.

Environmental Effects Refer to Section 3.11 for information.

b. CHS Letdown Line:

Break Locations Pipe breaks in the CHS letdown line are postulated at terminal ends and intermediate locations (Figure 3.6-16). However, breaks are excluded in the portion of letdown piping from the containment penetration to and including the containment isolation valves to ensure that Letdown Line breaks may be isolated assuming the most limiting single active failure as discussed in Chapter 15, Section 15.6.2. The stresses within this portion of the letdown line satisfy the break exclusion requirement (Equation 3.6-1).

Separation The letdown line inside and outside the containment is located with as much separation as possible from essential equipment by routing it in the regenerative heat exchanger cubicle, continuing in the annulus area as it penetrates the containment into the auxiliary building.

Pipe Whip Effects The effects of pipe whip have been mitigated by the installation of rupture restraints. Table 3.6-25 provides the results of pipe whip effects for the CHS-Letdown Line.

Pipe Restraints The pipe rupture restraints on the CHS letdown line are shown on Figure 3.6-16 along with the postulated pipe rupture locations. Notes to Figure 3.6-16 provide location, type, and function for each pipe rupture restraint.

Jet Impingement Effects Jet impingement effects are addressed in Table 3.6-26.

3.6-14 Rev. 30

MPS3 UFSAR Environmental Effects Refer to Section 3.11 for information.

Conclusion Adequate protection and/or separation is provided for essential systems and components required to mitigate the consequences of CHS charging and letdown piping ruptures.

c. CHS Seal Water Injection Line Break Locations Pipe breaks in the CHS seal water injection lines are postulated at terminal ends and intermediate locations and these postulated pipe break locations are shown on Figure 3.6-17 and listed in Table 3.6-27.

Separation The seal water injection line, inside and outside containment, is located with as much separation as possible from essential equipment by routing it in the annulus area as it penetrates the containment to the auxiliary building.

Pipe Whip Effects Table 3.6-28 indicates pipe whip effects. Pipe whip of the seal water injection lines, subsequent to a postulated pipe break, is not sustained since these lines connect the charging pumps to the reactor coolant pumps. The flow in the seal water injection line is limited.

Pipe Restraints Based upon the pipe whip effects noted above, pipe whip restraints are not required.

Jet Impingement Effects Jet impingement effects are addressed in Table 3.6-29.

Environmental Effects Refer to Section 3.11 for information.

Conclusion Essential systems, structures, and components remain functional subsequent to a postulated pipe rupture of the CHS seal water injection system.

3.6-15 Rev. 30

MPS3 UFSAR Steam Generator Blowdown System Break Locations Pipe breaks in the steam generator blowdown piping are postulated at terminal ends and intermediate locations shown on Figure 3.6-18.

The break exclusion criteria is invoked on portions of the steam generator blowdown piping from the containment penetration (outboard side) up to the isolation valves to ensure the integrity of a minimum of two intact and functioning Steam Generators. The stresses along this section of the steam generator blowdown piping fully meet the break exclusion requirements (Equation 3.6-1).

Pipe breaks are postulated at terminal and intermediate locations in the steam generator piping extending from the break exclusion zone through the main steam valve building to the turbine building. Table 3.6-30 summarizes these breaks.

Separation Every attempt has been made to locate the steam generator blowdown piping in areas physically remote from essential equipment to minimize the consequential effects of pipe rupture.

Pipe Whip Effects Table 3.6-31 summarizes the pipe whip effects resulting from the postulated piping failure.

Pipe Restraints The pipe rupture restraints on the steam generator blowdown system are shown on Figure 3.6-18 along with the postulated pipe rupture locations and data for these rupture restraints.

Jet Impingement Effects Jet impingement effects are addressed in Table 3.6-32.

Environmental Effects Refer to Section 3.11 for information.

Conclusion Adequate protection and/or separation is provided for essential systems and components required to mitigate the consequences of steam generator blowdown piping ruptures.

Auxiliary Feedwater System Break Locations 3.6-16 Rev. 30

MPS3 UFSAR Based upon application of the criteria specified in Section 3.6.2.1.1, pipe breaks on the auxiliary feedwater system inside the containment are postulated at terminal ends and intermediate locations.

Based upon application of the criteria specified in Section 3.6.2.1.2, pipe breaks in the auxiliary feedwater system outside the containment are postulated at terminal ends and at intermediate locations.

Auxiliary feedwater system pipe break locations and piping geometry are shown on Figure 3.6-32. Postulated pipe break locations for the Auxiliary Feedwater System are listed in Table 3.6-34.

Separation The Auxiliary Feedwater System satisfies the physical separation criterion Inside the Containment. Two Auxiliary Feedwater piping runs are routed along the exterior Crane Wall and the other two piping runs are routed along the interior Crane Wall at a higher elevation.

Inside the Engineered Safety Features (ESF) Building, the Auxiliary Feedwater System piping from each Motor Driven Pump is separated from the piping of the other Motor Driven Pump by the ESF Building walls. This prevents damage to a pair of intact Auxiliary Feedwater System pipes subsequent to a pipe break on any other Auxiliary Feedwater pipe.

Portions of the Turbine Driven Auxiliary Feedwater Pump discharge piping are high energy since this piping is pressurized due to operation of the Motor Driven Pumps. The Turbine Driven Auxiliary Feedwater Pump is separated from the Motor Driven Pumps by the ESF Building walls.

Turbine Driven Auxiliary Feedwater Pump discharge piping does not meet the separation criterion in one area of the ESF Building. Four piping runs associated with the Turbine Driven Auxiliary Feedwater Pump discharge piping by necessity are routed in a common area. These four piping runs are of the same nominal pipe size and wall thickness where the routing is common.

Based upon the criteria in Section 3.6.2.1, a break in one line does not affect the structural integrity of the adjacent lines.

Pipe Whip Effects Pipe whip of the auxiliary feedwater system is not sustained since backflow from each main feedwater line is prevented by dual check valves installed in the system and flow from each auxiliary feedwater pump is limited. Table 3.6-35 provides the results of pipe whip effects for the Auxiliary Feedwater System.

Pipe Restraints Pipe whip restraints are not required for postulated pipe breaks on the Auxiliary Feedwater System.

Jet Impingement Effects 3.6-17 Rev. 30

MPS3 UFSAR Jet impingement effects are addressed in Table 3.6-36.

In general, jet impingement forces, subsequent to postulated pipe breaks on the auxiliary feedwater system, are small due to the lack of upstream energy reservoirs to sustain pressure in the steady state.

Environmental Effects Refer to Section 3.11B for information.

Conclusion Essential systems, structures, and components remain functional subsequent to a postulated pipe rupture of the auxiliary feedwater system.

3.6.2 DETERMINATION OF BREAK LOCATIONS AND DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING This section describes the design bases used for defining postulated pipe break and crack locations in high- and moderate-energy piping systems inside and outside of the containment, the methods of analysis used to evaluate the jet reaction forces at the break locations, and the jet impingement effects and loading effects on adjacent essential systems, components, and structures.

3.6.2.1 Criteria Used to Define Break and Crack Location and Configuration Pipe breaks and cracks are postulated in those high- and moderate-energy piping systems located in proximity to essential systems, components, and structures required for the safe shutdown of the plant. All postulated breaks and cracks are evaluated for potential damage to essential systems, components, and structures due to pipe whip, jet impingement, and environmental effects. If the damage is unacceptable, protective measures are provided either by rerouting of piping, relocation of essential equipment, or providing enclosures. Where this is not feasible, pipe whip restraints and/or jet impingement shields are installed to protect the essential systems, components, and structures.

An unrestrained whipping pipe is considered capable of causing circumferential and longitudinal breaks, individually, in impacted pipes of smaller nominal pipe size and developing through-wall cracks in equal or larger nominal pipe sizes with thinner wall thicknesses. The impact into pipes of equal or larger nominal pipe sizes and wall thicknesses is considered inconsequential and is not evaluated. In all cases, the effects of jet impingement are less severe than the corresponding pipe whip impact. (In the limiting case of no clearance between the broken pipe and target pipe, the impact force equals the total thrust force and is localized whereas the jet force is less, being moderated by a shape factor, and is distributed. As the gap increases, the whip impact is enhanced but the jet intensity diminishes and becomes more diffuse.) Therefore, jet impingement on pipes of equal and larger nominal pipe size and wall thickness also is considered inconsequential and is 3.6-18 Rev. 30

MPS3 UFSAR not evaluated. In all other cases, the jet interaction is either avoided or the target pipe and its supports are evaluated for the jet load using Service Level D allowables.

Design basis break and crack locations, type, and orientation are postulated in accordance with the following sections.

3.6.2.1.1 Criteria for Inside Containment Break Locations - ASME Section III Code Class 1 Piping Breaks in ASME Section III Code Class 1 high-energy piping are postulated to occur at the following locations in each piping run or branch run:

1. At terminal ends of the pressurized portions of the runs (terminal ends are extremities of piping runs that connect to structures, components (e.g., vessels, pumps, valves) or pipe anchors. A branch connection to a main piping run is a terminal end of the branch run, except where the branch run is classified as part of a main run in the stress analysis and is shown to have a significant effect on the main run behavior.
2. At intermediate locations between terminal ends where either of the following criteria are exceeded:
a. At any intermediate locations between terminal ends where the maximum stress intensity ranges, for normal and upset plant conditions, and for a 1/2 safe shutdown earthquake (OBE) event transient exceed 2.4Sm (the design stress intensity as specified in Section III of the ASME Boiler and Pressure Vessel Code), calculated by Equation 10 and either Equation 12 or Equation 13 in Paragraph NB-3653 of the ASME Code,Section III, or
b. At any intermediate locations between terminal ends where the cumulative usage factor U (the cumulative usage factor as specified in Section III of the ASME Boiler and Pressure Vessel Code), derived from the piping fatigue analysis under the loadings associated with OBE and operational plant conditions, exceeds 0.1.

Break Locations - ASME Section III Code Classes 2 and 3 Piping Breaks in ASME Section III, Code Classes 2 and 3, high energy piping systems are not postulated at locations delineated in Section 3.6.2.1.2, Item 2. The portions of piping within the break exclusion zone are designed to meet the requirements of ASME Section III, Subarticle NE-1120 and the additional criteria specified in Section 3.6.2.1.2, Items 2a through 2f.

Breaks in ASME Section III Code Classes 2 and 3 high-energy piping are postulated to occur at the following locations in each piping run or branch run:

3.6-19 Rev. 30

MPS3 UFSAR

1. At terminal ends of the pressurized portions of the runs, and
2. At intermediate locations selected by either of the following criteria:
a. At each pipe fitting (e.g., elbow, tee, cross, flange, and nonstandard fitting),

welded attachment, and valve.

b. At each location where the stress ranges associated with normal and upset plant conditions and a OBE event, calculated by Equations 9 and 10, Paragraph NC-3652 of the ASME Code,Section III, exceed 0.8 (1.2Sh + SA).

SA is the stress calculated by the rules of NC 3600 and ND 3600 for Classes 2 and 3 components, respectively, of ASME Code Section III, 1971 edition up to and including the summer 1973 addenda. SA is the allowable stress range for expansion stress calculated by the rules of NC 3600 of ASME Code Section III, 1971 edition up to and including the summer 1973 addenda.

3.6.2.1.2 Criteria for Outside Containment High-Energy Piping Systems

1. Piping Systems Separated from Essential Structures, Systems and Components - A primary objective in the piping layout and plant arrangement is to have adequate separation, so that the effects of postulated pipe breaks at any location are isolated or physically remote from essential structures, systems, and components. Pipe breaks are not postulated in these separated high-energy piping systems.
2. Piping Systems in Containment Penetration Areas Break Exclusion Zone - Breaks are not postulated in certain high energy piping systems in the containment penetration areas. Breaks are not postulated in portions of high energy piping between the first restraint outboard of the containment isolation valve and the pipe to flued head weld on the inboard side of the containment penetration for the 30 Inch Main Steam System and the Main Feedwater System. For the Letdown Line, the 3 Inch Main Steam System in the Engineered Safety Features (ESF) Building, and the Steam Generator Blowdown System, breaks are not postulated between the outboard containment isolation valve and the pipe to flued head weld on the outboard side of the containment penetration. Portions of this piping outboard of the isolation valve, for which failure could affect the leaktight integrity of the containment structure, are provided with pipe whip restraints capable of resisting bending and torsional moments produced by the postulated piping failure outboard of the first restraint beyond the containment isolation valves. These restraints are located as close as practical to the containment isolation valves.

3.6-20 Rev. 30

MPS3 UFSAR The restraints are designed to withstand the loadings imposed by a postulated pipe rupture so that isolation valve structural integrity, operability, and leak tight integrity of the associated containment penetration is ensured.

The portions of piping within the break exclusion zone are designed to meet the requirements of ASME Code Section III, Subarticle NE-1120 and the following additional design requirements:

a. The following design stress and fatigue limits should not be exceeded for Class 2 piping:
  • The maximum stress ranges as calculated by Equations 9 and 10 in Paragraph NC-3652, ASME Code,Section III, considering normal and upset plant conditions (i.e., sustained loads, thermal expansion and a 1/2 SSE event) do not exceed 0.8 (1.2Sh + SA).
  • The maximum stresses as calculated by Equation 9 in Paragraph NC-3652 under the loadings resulting from a postulated piping failure beyond these portions of piping do not exceed 1.8Sh. This stress limit is applied between the containment isolation valve and the containment penetration. The portion of piping within the pipe break exclusion zone between the containment isolation valve and the first pipe rupture restraint must not develop a plastic hinge subsequent to postulated piping failures beyond the pipe break exclusion zone.
b. Welded attachments, for pipe supports or other purposes, to these portions of piping are permitted when a detailed stress analysis demonstrates compliance with the limits described in (a).
c. The number of circumferential and longitudinal piping welds and branch connections are minimized.
d. The length of these portions of piping is reduced to the minimum length practicable.
e. The design of pipe anchors or restraints (e.g., connections to containment penetrations and pipe whip restraints) does not require welding directly to the outer surface of the piping (e.g., flued integrally forged pipe fittings are used) except where such welds are capable of 100-percent volumetric inservice inspection. This criterion is also applicable to the portion of piping between the containment and the inside containment isolation valves.
f. For these portions of high-energy piping, inservice examination is performed in accordance with the requirements specified in ASME Code,Section XI, with the exception that examination of circumferential and 3.6-21 Rev. 30

MPS3 UFSAR longitudinal pipe welds between and including the boundaries of the pipe break exclusion zones is performed in accordance with the risk-informed methodology established in WCAP-14572, Revision 1-NP-A, Addendum

1. The examination includes the valve and pipe pressure boundaries.

Details of containment penetration, identification of pipe welds, access for inservice inspection, and points of fixity and discontinuity are provided in Section 6.6.

3. Piping Systems not designated as Break Exclusion Zone Breaks in ASME Section III Code Classes 2 and 3 high-energy piping are postulated at the following locations in each piping and branch run outboard of break exclusion zones (except those portions of piping systems identified in Section 3.6.2.1.2, Items 1 and 2):
a. At terminal ends of the pressurized portions of the runs.
b. At the extremity of the break exclusion zone (if applicable), and
c. At intermediate locations selected by either of the following criteria:
  • At each pipe fitting (e.g., elbow, tee, cross, and nonstandard fitting) welded attachment and valve or, if the run contains no fittings, at one location at each extreme of the run (a terminal end, if located within a protective structure, may substitute for one intermediate break), or
  • At each location where the stresses exceed 0.8 (1.2 Sh + SA).

Breaks in nonnuclear safety class high-energy piping are postulated at the following locations in each piping run or branch run:

1. At terminal ends of the pressurized portions of the runs, and
2. At intermediate locations selected by either of the following criteria:
a. At each pipe fitting, welded attachment, and valve, or
b. At each location where the stress ranges associated with normal and upset plant conditions and a 1/2 SSE event, calculated by Equations 9 and 10, paragraph NC-3652 of ASME Section III exceed 0.8 (1.2 Sh + SA).

Moderate-Energy Piping Systems For the purpose of satisfying the separation provisions of plant arrangement, a review of the piping layout and plant arrangement drawings is conducted to show that the effects of 3.6-22 Rev. 30

MPS3 UFSAR through-wall leakage cracks at any location are isolated or physically remote from essential systems, components, and structures.

For certain Moderate Energy Systems, leakage cracks are not postulated in those portions of ASME Section III Code Class 2 piping between the outboard containment penetration weld and the outboard containment isolation valve. A Crack Exclusion Zone is defined providing the piping is designed to the requirements of ASME Section III Subsection NE-1120 and that the maximum stress range (as calculated by Equations 9 and 10, Paragraph NC-3652 of Section III of the ASME Code) does not exceed 0.4 (1.2Sh + SA).

Crack Exclusion Zones are invoked for the Hydrogen Recombiner (HCS) System Suction piping, the Containment Atmosphere Monitoring (CMS) System Pump Suction piping, and the Containment Vacuum (CVS) System Pump Suction piping. The Crack Exclusion Zones are designated to ensure containment integrity post DBA since postulating through wall leakage cracks in this region of piping does not maintain containment integrity. As discussed in Section 6.2.4.2, these systems take exception to GDC-56 by placing two isolation valves in series outside containment. For these systems, Augmented Inservice Inspection (ISI) is invoked to support the GDC exception.

Through-wall leakage cracks are postulated in moderate-energy piping except where exempted by Moderate-Energy Piping Systems under Section 3.6.2.1.2, or where the maximum stress range in these portions of ASME Section III Code Class 2 or 3 piping is less than 0.4 (1.2Sh + SA). For nonnuclear, nonseismic piping systems, throughwall leakage cracks are postulated at any location that has the worst consequences for essential structures, systems, or components. Only environmental effects due to fluid spray and flooding are considered.

Cracks are not postulated in moderate-energy piping location in an area in which a break in high-energy piping occurs. Where a postulated leakage crack in the moderate-energy piping results in more limiting environmental conditions than the break in proximate high-energy piping, the provisions of the previous paragraph are applied.

Through-wall leakage cracks instead of breaks are postulated in the piping of those fluid systems that qualify as high-energy systems for only short operational periods, but qualify as moderate-energy systems for the major operational period.

An operational period is considered short if the fraction of time that the system operates within the pressure-temperature conditions specified for high-energy systems is less than 2 percent of the time that the system operates as a moderate-energy system (e.g., systems such as the reactor residual heat removal systems qualify as moderate-energy systems); however, systems such as auxiliary feedwater systems operated during reactor startup, hot standby, or shutdown, qualify as high-energy systems.

3.6-23 Rev. 30

MPS3 UFSAR 3.6.2.1.3 Design Basis Break/Crack Types and Orientation Circumferential Pipe Breaks The following circumferential breaks are postulated in high-energy piping at the locations specified in Sections 3.6.2.1.1 and 3.6.2.1.2:

1. Circumferential breaks are postulated in high-energy piping runs and branch runs exceeding a nominal pipe size of 1 inch. When the maximum stress range or usage factor exceeds the limits specified for break postulation, and if it is determined by detailed stress analysis, that the maximum stress range in the circumferential direction is at least 1.5 times that in the axial direction, then only longitudinal breaks are postulated.
2. Where break locations are selected at pipe fittings without the benefit of stress calculations, breaks are postulated at the piping weld to each fitting, valve, or welded attachment. If detailed stress analyses or tests are performed, the maximum stressed location in the fitting may be selected instead of the pipe-to-fitting weld.
3. Circumferential breaks are assumed to result in pipe severance and separation amounting to a one-diameter lateral displacement of the ruptured piping sections unless physically limited by piping restraints, structural members, or piping stiffness as may be demonstrated by inelastic analysis.
4. The dynamic force of the jet discharge at the break location is based on the effective cross-sectional flow area of the pipe and on a calculated fluid pressure as modified by a thrust coefficient. Limited pipe displacement at the break location, line restrictions, flow limiters, positive pump-controlled flow, and the absence of energy reservoirs are taken into account as applicable, in the reduction of jet discharge.
5. Pipe whipping is assumed to occur in the plane defined by the piping geometry and configuration, and is assumed to cause pipe movement in the direction of the jet reaction.

Longitudinal Pipe Breaks The following longitudinal breaks are postulated in high-energy piping at the locations of each circumferential break specified under Circumferential Pipe Breaks in this section, except as noted:

1. Longitudinal breaks in piping runs and branch runs are postulated in nominal pipe sizes 4 inches and larger. However, when the maximum stress range or usage factor exceeds the limits specified for break postulation and if it is determined by detailed stress analysis that the maximum stress range in the axial direction is at 3.6-24 Rev. 30

MPS3 UFSAR least 1.5 times that in the circumferential direction, then only a circumferential break is postulated.

2. Longitudinal breaks are not postulated at terminal ends.
3. Longitudinal breaks are assumed to result in an axial split without pipe severance.

Splits are located (but not concurrently) at two diametrically opposed points on the piping circumference such that a jet reaction causing out-of-plane bending of the piping configuration results. Alternately, a single split may be assumed at the section of highest stress as determined by detailed stress analysis.

4. The dynamic force of the fluid jet discharge is based on a circular break area equal to the effective cross-sectional flow area of the pipe at the break location, and on a calculated fluid pressure modified by an analytically or experimentally determined thrust coefficient as determined for a circumferential break at the same location.

Line restrictions, flow limiters, positive pump-controlled flow, and the absence of energy reservoirs are taken into account, as applicable, in the reduction of jet discharge.

5. Pipe movement is assumed to occur in the directions defined by the stiffness of the piping configuration and jet reaction forces, unless limited by structural members or piping restraints.

Through-Wall Leakage Cracks (outside of containment only)

The following through-wall leakage cracks are postulated in moderate-energy piping at the locations specified under Moderate-Energy Piping Systems in Section 3.6.2.1.2, item 3.

1. Cracks are postulated in moderate-energy piping runs and branch runs exceeding a nominal pipe size of 1 inch.
2. Fluid flow from a crack is based on a circular opening of area equal to that of a rectangle one-half pipe diameter in length and one-half pipe wall thickness in width.
3. The flow from the crack is assumed to result in an environment that wets all unprotected components within the compartment, with consequent flooding in the compartment and communicating compartments. Flooding effects are determined on the basis of a conservatively estimated time period required to effect corrective actions.

3.6.2.1.4 Conformance with Regulatory Guide 1.46 Refer to Section 1.8 for this information.

3.6-25 Rev. 30

MPS3 UFSAR 3.6.2.2 Analytical Methods to Define Forcing Functions and Response Models 3.6.2.2.1 Introduction Pipe rupture analyses include calculations to determine the fluid forces generated by blowdown of pressurized lines, complemented by dynamic or energy-balance methods to determine pipe motion and impact effects. Restraints for lines 6 inches and less in diameter are usually qualified on a generic basis using an energy balance method. However, restraints for larger lines are normally engineered individually for each system, using standard design concepts. The response of unrestrained lines is analyzed by either inelastic dynamic analysis or energy balance analysis.

Figure 3.6-19 provides a flowchart for the pipe rupture analysis.

Criteria for the pipe rupture response analysis include:

1. An analysis of the pipe run or branch is performed for each postulated longitudinal and circumferential rupture or, alternatively, for a worst case. Worst cases are selected on the basis of gap, fluid force, and piping system stiffness.
2. The loading condition of a pipe run or branch prior to postulated rupture in terms of internal pressure, temperature, and stress state is that condition associated with reactor operation at 100-percent power.
3. For a circumferential rupture, pipe whip dynamic analyses are only performed for that end (or ends) of the pipe or branch that is (are) connected to a contained fluid energy reservoir having sufficient capacity to develop a jet stream.
4. Dynamic analytical methods, used for calculating the piping or piping/restraint system response to the jet thrust developed after a postulated rupture, account for the effects of the following:
a. Mass, inertia, and stiffness properties of the system.
b. Impact and rebound (if any) as permitted by gaps between piping and restraint.
c. Elastic and inelastic deformation of piping and/or restraint.
d. Support boundary conditions.
5. An allowable design strain limit of 0.5 ultimate uniform strain of the restraints is used for energy-absorbing components. For compressive energy absorbing components, the following deformation limits are used:
a. The design limit for pipe crush bumpers shown on Figure 3.6-30 and metallic honeycomb is 80 percent of energy absorbing capacity.

3.6-26 Rev. 30

MPS3 UFSAR

b. The design limit for pipes crushed uniformly along their length is the lesser of:
1. one half of the pipe diameter, or
2. the maximum flattening limits as prescribed by ASTM A530.

The above deformation limits assure that the area under the essentially flat portion of the materials force-deflection curve is not exceeded.

6. A 10-percent increase of minimum specified yield strength (SY) may be used to account for strain rate effects in inelastic nonlinear analyses. Alternatively, experimental data may be used to determine the strain rate parameters for use in nonlinear codes which monitor strain rate.

3.6.2.2.2 Time Dependent Blowdown Force Blowdown force calculations are based on methods suggested by Moody (1973) and include consideration of the transient pressures, velocities, and other thermodynamic properties of the fluid. To provide the time history of pressure, velocity, etc., the method of characteristics is used to solve the continuity and momentum equations simultaneously. A general description of the method can be found in gas dynamics textbooks (De Haller 1945, Rudinger 1969, Owzarek 1968). For these one dimensional fluid mechanics analyses, the pipe run is treated as an equivalent section of straight pipe. The calculated momentum and pressure forces are applied at changes in direction or cross section of the piping to provide time-dependent loads for pipe dynamic analysis.

The transient forces result from wave propagation and fluid momentum. It is assumed that pipe bends and elbows neither attenuate the traveling pressure waves nor cause reflections.

Immediately following the rupture of a pipe, a decompression wave travels from the break at the speed of sound relative to the fluid. The fluid ahead of and behind the wave is at different thermodynamic states. This initial blowdown condition is maintained until a return signal from the pressure reservoir reaches the break. At this time, repeated wave reflections between the reservoir and break prevail until a steady state flow condition is established. Boundary conditions that govern the flow at the break are considered.

Fluid momentum changes will result in dynamic forces being exerted on pipe segments. The forcing function is calculated using Shapiro Volume I Equation 4.20 as follows:

2 m* u Ru F = ( P - P a )A + ------- = ( P - P a ) + ------------ A (3.6.2.2-1) g 144g where:

P = Local static pressure (psia) 3.6-27 Rev. 30

MPS3 UFSAR Pa = Ambient pressure (psia)

R = Fluid density (lbm/ft3) u = Velocity of blowdown fluid (fps)

A = Local flow cross-sectional area (in2) ft - lb m g = Dimensional Constant 32.2 --------------------- lb f - sec m* = Mass flow rate (lb /sec)

During the blowdown process, the local static pressure, mass flow rate, and other thermodynamic properties change with time; therefore, the forcing function varies with time.

3.6.2.2.2.1 Subcooled Nonflashing Waterline Blowdown When a pipe rupture occurs, the blowdown flow rate and properties must go from an initial set of conditions to the final or steady state condition. A decompression wave travels upstream toward the pressure source. The initial blowdown velocity can be calculated by applying the momentum equation across the wave:

RCu = gP (3.6.2.2-2)

(u-o) = 144g(Po-Pa)/RC (3.6.2.2-3) where:

C = The speed of sound (fps)

Po = The initial stagnation pressure (psia)

Therefore, the initial blowdown thrust for non-flashing water is:

2 g ( P o - P a ) ( 144 )

Ru F = ----------------- A = ---------------------------------------

- Po A (3.6.2.2-4) g ( 144 ) RC 2

For transient flow analysis, the one dimensional equations of conservation of momentum and continuity are solved simultaneously to obtain transient pressures and velocities, which are used to calculate the transient blowdown thrust. The solutions are subject to the following boundary conditions:

U (,t) O at t = 0 P (,t) Po at t = 0 3.6-28 Rev. 30

MPS3 UFSAR 2

Ru ( , t )

P ( , t ) = P o - ---------------------- at = 0 2g ( 144 )

The initial blowdown flow remains constant until the decompression wave, which is reflected from the pressure source, reaches the break end. It is then reflected again, causing a change of blowdown flow. These repeated wave transmissions and reflections continue until the steady state flow is established.

Steady State Flow with Friction For steady state flow with friction, the blowdown forcing function calculation becomes:

2 ( Po - Pa ) 1 F = -------------------------- ----------------- P o A (3.6.2.2-5)

Po fL 1 + -------e D

which is derived by applying Bernoullis equation across the pipe and by using the expression for the forcing function calculation, where:

Le = Total equivalent length of pipe friction f = Friction factor (Reynolds number and pipe surface roughness dependent), obtained from Moodys chart, (1961).

D = Pipe inside diameter. The friction parameter is calculated by considering all the possible losses on the line.

The friction parameter fLe/D is calculated by considering all the possible losses on the line.

Transient Flow With Friction With friction losses taken into consideration, transient pressures and velocities of a subcooled nonflashing water line blowdown can be obtained by simultaneously solving the continuity and momentum equations. The finite difference approximation using the method of characteristics is used as a principle for numerical solution of these two governing equations (Streeter 1967).

The computations proceed in the following manner. A grid is chosen in such a way that = Ct, where is a space increment, t a time increment, and C the propagation speed of the decompression wave. Starting from the initial conditions along the pipe, the pressure and velocity at t = t + t, and at any interior points of the pipe, can be calculated by using two characteristic equations. Whenever a boundary point is reached, the corresponding characteristic equation and boundary condition are used. The process proceeds until a steady state is reached.

3.6-29 Rev. 30

MPS3 UFSAR The friction losses are expressed in terms of pressure drop of the system. To accurately model the system with friction losses, a smaller must be used. This method can be applied for flow with or without friction losses.

The transient pressures, velocities, etc., are then used to calculate the blowdown forces using the equation described previously.

Side Thrust for Longitudinal Split For a longitudinal split, the blowdown flow comes from either upstream, downstream, or both pipe directions. Because of its geometry, the split is considered as an ideal nozzle with a discharge coefficient of unity.

A longitudinal break will cause a reaction force of 1 Po A for an extremely short time interval, which allows decompression waves to move out of the break region into the adjacent upstream and downstream pipe sections. Then the reaction force drops to a value corresponding to the blowdown thrust until the reflected waves arrive from each direction. The steady state blowdown thrust is:

2 ( P o - P a )A F = -----------------------------

- (3.6.2.2-6) fL e 1 + -------

D where:

A ~ The break opening area of the split 3.6.2.2.2.2 Steamline Blowdown Transient Flow without Friction Steam is treated as an ideal, single-phase gas with a constant specific heat ratio, k, of 1.3. Except for the case of steady state blowdown flow, the flow is assumed to be isentropic with negligible pipe friction. The characteristic method (Jonssen et al., 1973, Hartree 1952), which is a finite difference approximation using the principle of characteristics, is used as a basis for the numerical solution of the continuity and momentum equations. The transient pressure, mass flow rate, and other thermodynamic properties are then used to calculate the transient-state forcing function.

Immediately following the break, a decompression wave travels into the pipe toward the pressure reservoir. The fluid in front of the wave is at a state:

U1 = 0 C1 = C 3.6-30 Rev. 30

MPS3 UFSAR where:

U1 = Velocity of fluid Co = Speed of sound in fluid The fluid state at the exit is at the sonic condition (Shapiro 1953 Volume 1 Equation 4.6a).

Ue C 2 0.5

= -----e- = ------------- = 0.9325 (3.6.2.2-7)

Co Co K+1 The blowdown force can be calculated using Shapiro Volume I Equation 4.20 as:

2 P Re Ce F = -----e- + ---------------------- - P P o gP o ( 144 ) oA (3.6.2.2-8) 2 P Re Ce 2 Ro C o

= -----e- + ------ ------ ----------------------- P oA P o R o C o gP o ( 144 )

where:

2 KgP o ( 144 )

C o = ---------------------------

R The pressure ratio across the wave is:

k -

P Te k C e 2k 2 k-1

e- = ----- ----------- = ------ ----------- = ------------ = 0.55 (3.6.2.2-9)

Po T o k - 1 C o k - 1 k+1 where:

T = Temperature The density ratio is:

R Pe 1 C e 2k 1 2 ( 1/k - 1 )

e- = ------ --- = ------ ----------- --- = ------------- = 0.628 (3.6.2.2-10)

Ro Po k Co k - 1 k K+1 Therefore, the blowdown force can be reformulated as:

3.6-31 Rev. 30

MPS3 UFSAR 2 ( 2K/K-1 )

F = (1 + K ) ------------- P o A = 0.685 P o A (3.6.2.2-11)

K + 1 This blowdown force is constant until a return signal from the pressure source reaches the break.

When the wave reaches the reservoir, it is reflected as a compression wave. The boundary condition at the reservoir lies on the steady state ellipse. From, Shapiro Volume II Equation 24.36.

Ci 2 K - 1 Ui 2 C o + 2 C o = 1 (3.6.2.2-12) which is the energy equation applying across the vessel-pipe inlet. The boundary condition for this case is from Shapiro Volume II Page 963:

2 Ui T o = T i + -------------------------- - (3.6.2.2-13) 2C p g ( 778 )

where:

Btu Cp = The constant pressure specific heat of a fluid --------------------

lb m - °F i = The state at the inlet to the pipe If the steady state is reached, the flow in the pipe is uniform and, if the pressure in the pressure vessel remains high, then the boundary condition at the break always lies on the sonic line. For example:

U* C

= ------*- (3.6.2.2-14)

Co Co Then from the critical flow condition from Shapiro Volume I Equation 4.6a:

U* C 2 - = 0.9325

= ------*- = ----------- (3.6.2.2-15)

Co Co k+1 where:

  • = The critical flow condition Then, the steady state blowdown force is:

3.6-32 Rev. 30

MPS3 UFSAR P

  • R* ( U* ) 2 F = ------- + ----------------------- P o A P o P o g ( 144 )

(3.6.2.2-16) 2 ( k/k-1 )

= ( 1 + k ) ------------ P o A = 1.255 P o A k+1 Steady State with Friction For steady state flow with friction losses, the analysis is based on the theory of compressible flow with friction (Shapiro 1953). The pipe friction is the chief factor bringing about the change of fluid properties in the flow. A curve which describes the variation of steady state steam blowdown force versus friction parameter fLe/D is shown on Figure 3.6-20 (Moody 1973).

Transient Flow with Friction Using a method similar to the transient flow analysis for a non-flashing waterline, a hybrid method of characteristics has been adopted from Jonsson et al. (1973) to solve the one dimensional governing equations of mass, momentum, and energy simultaneously for pipe with a constant cross-sectional area with friction effects taken into consideration. The governing equations are first transformed into a system of characteristic equations. Then the finite difference approximation is used to integrate the fluid variables which represent the pressure, velocity, and entropy along the characteristic lines and the path line. The equations are solved using computer program ME-143 One-Dimensional Unsteady Flow of a Compressible Fluid with Friction (UFLOW). These transient pressures, velocities, etc., are used to calculate the blowdown forcing functions using the equation described previously.

3.6.2.2.3 Simplified Blowdown Analysis A conservative steady state forcing function may be used for calculations based on the energy balance method. The function has a magnitude of T = KPA (3.6.2.2-17) where:

P = System pressure prior to pipe break A = Pipe break area K = Thrust coefficient (theoretical maximum)

K values are 1.26 for saturated steam, water, and steam/water mixtures and 2.00 for nonflashing subcooled water.

3.6-33 Rev. 30

MPS3 UFSAR An amplification factor between 1.1 and 1.2 has been demonstrated by testing. To account for rebound, it is applied to the above force. Alternatively, the maximum fluid force during the energy input phase, as determined by the detailed methods of Section 3.6.2.2.2, may be used.

3.6.2.2.4 Lumped-Parameter Dynamic Analysis The piping system is modeled mathematically as a series of beam elements connected at nodes.

Distributed mass of the pipe and contained fluid is modeled as a lumped mass located at the nodal points. Beam elements have the stiffness properties of the pipe in the elastic range and approximate the plastic behavior above yield.

Before a rupture, pipe is stressed by internal pressure and is in static equilibrium. When initial conditions have an effect on the parameters being calculated, such as stresses in break exclusion regions or loads on attached components, this effect is considered.

As a circumferential break propagates, the load-carrying metal area of the pipe decreases so that a force unbalance results. The force initially transmitted across the break is assumed to drop linearly to zero in 1 millisecond. After the break, the forces exerted on the pipe by the fluid are determined by the time-dependent blowdown force derived (Section 3.6.2.2.2). Similarly, for a longitudinal split, the crack propagation speed limits the rate at which the split opens, so a 1-millisecond force rise time is assumed. Other break opening times may be used if justified.

Subsequent to a postulated rupture, the inelastic system response is analyzed by the use of an elastic-plastic lumped-mass beam element computer code such as DINASAW or LIMITA (Appendix Sections 3A.2.6, 3A.2.10, and 3A.2.11. The analysis considers the free motion of the pipe through a gap, if one exists, using the appropriate initial conditions and the fluid blowdown forces as calculated in Section 3.6.2.2.2. The mathematical model includes the restraint or barrier, and sometimes a member simulating the local crush resistance of the pipe. Rebound effects are considered by automatically connecting and disconnecting that member for impact and rebound, respectively.

3.6.2.2.4.1 Sample Pipe Rupture Dynamic Analysis Pipe rupture restraint 3FWS-PRR5S limits the motion of the feedwater line following a postulated circumferential rupture at the end of the reducer elbow (Figure 3.6-21). The restraint prevents the whipping pipe from impacting the steam generator.

The restraint is a U configuration stainless steel strap having a 7 inch strainable width and 1 inch thickness. The initial clearance between the hot pipe and restraint is 0.89 inch in the outward direction, resulting in a total acceleration gap, after slack takeup, of 2.04 inches. Figure 3.6-22 shows a typical laminated strap.

The analysis was conducted using the DINASAW computer code for the mathematical model of the pipe, restraint, and intermediate structure (Figure 3.6-21). The elastic-plastic characteristics of the pipe, strap, and intermediate structure were used in the computer solution in terms of bilinear engineering stress-strain relationships with the corresponding allowable stress not exceeding half 3.6-34 Rev. 30

MPS3 UFSAR the uniform ultimate strain. Strain rate sensitivities for materials below 400°F were considered.

The fluid forcing function depicted on Figure 3.6-21 is applied at the elbow in terms of normal pressures per unit tangential length.

The restraint and intermediate structure reaction load are shown on Figures 3.6-23 through 3.6-25 which illustrate the intermediate structure loads and Figure 3.6-26 which illustrates the restraint reaction load. The depletion of the kinetic energy of the system and the steady work are indications of the convergence of the dynamic problem. This is shown on Figure 3.6-27.

3.6.2.2.5 Energy Balance Analysis The energy balance technique for analyzing pipe impact equates the work done by the escaping fluid to the energy absorbed in deforming the ruptured pipe and the impacted target. A steady state blowdown force is used for the energy balance analysis. The magnitude of the force is described in Section 3.6.2.2.3.

The input energy of the system is determined by multiplying the pipe displacement at the break end by the component of the fluid blowdown force in the direction of the displacement.

The input energy is:

Lh E = F b ( g + d ) --------------- (3.6.2.2-18)

L h - L where:

Fb = Component of blowdown force in direction of pipe displacement Lh = Length from break to plastic hinge L = Length from break to restraint g = Pipe-target gap d = Restraint deflection The strain energy absorbed during pipe whip and impact consists of the energy absorbed by pipe bending, Eb, the energy absorbed by pipe crush during impact, Ec, and the energy absorbed by deformation of the target, Et.

To determine post-impact target deformation and the peak reaction force, the input energy is equated to the strain energy absorbed by the pipe and target. The energy absorption characteristics of the pipe crush and target deformation are calculated on the basis of the displacement integral of the appropriate force-deformation curves.

3.6-35 Rev. 30

MPS3 UFSAR Sample Energy Balance Analysis Analyze the impact of a 4-inch, schedule 80 pipe into a pipe crush bumper following a circumferential break at an elbow (Figure 3.6-28). One source of energy input is recognized: the fluid blowdown force traveling through the distance moved by the ruptured end of the pipe. The input energy is:

Lh E in = F b ( g + d ) --------------- (3.6.2.2-19)

L h - L where:

Fb = Fluid blowdown force g = Acceleration gap d = Restraint deflection Lh = Length from break to plastic hinge L = Length from break to restraint The ratio Lh/(Lh-L) represents the increased pipe displacement at the break, compared to displacement at the restraint, due to the assumed pipe rotation about a plastic hinge.

The fluid force is calculated:

Fb = Kr Kf PoA = 13.4 kips (3.6.2.2-20) where:

Kr = Rebound factor (1.2)

Kf = Thrust coefficient (0.88)

Po = Initial pressure (1,106.7 psi)

A = Pipe flow area (11.497 square inches)

The maximum thrust coefficient in the period when the energy balance occurs is 1.0. However, this drops to 0.88 as soon as the decompression wave passes the elbow (t 0.001 second) and occurs when the pipe is just starting to accelerate. Since the displacement and resulting energy input are negligible during this interval, 0.88 rather than 1.0 is used as the thrust coefficient. The duration of the entire energy balance event is evaluated after the restraint is sized. This permits a quick review of the fluid force history to assure that a higher thrust coefficient did not occur during the dynamic event.

3.6-36 Rev. 30

MPS3 UFSAR Energy may be absorbed in plastic bending of the pipe and in crush of the restraint. The energy absorbed by bending at the plastic hinge is:

Eb = Mp = Mp (g + d)/(Lh-L) (3.6.2.2-21) where:

Mp = Plastic moment

= Hinge rotation The value of Mp may be obtained from rigid, perfect-plastic limit theory, but, for this application, a strain hardening moment (Gerber 1974) is appropriate. This requires an estimate of the hinge rotation (i.e., g, d, Lh and L must be determined). The acceleration gap, g, is 1.63 inches. Let the bumper pipe have the same diameter as the process pipe. Set L at 20 inches to place the restraint on the straight pipe. Using the common expression for plastic hinge length, Lh = 3 Mp/Fb, and the method described by Gerber (1974), an iterative solution shows that M equals 244.9 in-kips and Lh equals 119.6 inches.

Input energy:

Lh E in = F b ( g + d )x ------------------ 26.3 + 16.1 d in-kips (3.6.2.2-22)

( Lh - L Energy absorbed by the process pipe bending:

E b = M p ( g + d )/ ( L h - L ) = 4 + 2.5 d in-kips (3.6.2.2-23)

The difference between Ein and Eb is the energy absorbed by the crush bumper and equals 22.3 +

13.6 d in-kips.

Size the bumper pipe thickness subject to the following constraint:

r b 0.131 t b 0.75 t p ---- = 0.237 in for schedule 40 (3.6.2.2-24) r p where:

tb = Bumper pipe wall thickness tp = Process pipe wall thickness rb = Mean radius of bumper pipe 3.6-37 Rev. 30

MPS3 UFSAR rp = Mean radius of process pipe The above identity assures that the bumper pipe will crush without crushing the process pipe upon impacting. Table 3.6-33 presents the energy absorbing capacity of a 4 inch schedule 80 pipe as functions of overall displacement and impact force (Peech et al., 1977). Interpolate to find the exact point of energy balance:

d = 2.34 in Ep = 54.4 in-k F = 32.4 kips The impact force is thus 2.4 times the fluid blowdown force. A dynamic load factor of 2.0 is considered for the intermediate steel structure. Figures 3.6-29 and 3.6-30 illustrate the pipe crush bumpers.

Finally, determine the approximate time of peak restraint load to assure that the fluid force did not exceed 0.88 P0A during the energy balance event:

2mL h 3 ( g + d ) 2 1 t = ----------------------- ------------------- ------------------------- (3.6.2.2-25) 3 ( Ln - L ) g F b L h - M p where:

m = The mass per unit length of the pipe Thus:

t = 13.9 milliseconds 3.6.2.2.6 Local Pipe Indentation The local shell indentation stiffness of the pipe is usually considered where other energy-absorbing mechanisms are not available at the point of impact. Examples include impacts into rigid displacement-limiting bumpers, concrete walls, and the omnidirectional restraint weldment (the latter interposes a significant mass between the impacting pipe and the energy absorbers).

Experimentally derived crush stiffnesses have been used. These were based on a series of pseudo-static pipe crush tests covering several crush geometries and a sufficient range of pipe thicknesses and diameters to develop parametric scaling laws (Peech et al., 1977). This was augmented by analyses to determine the sensitivity to material strength, dynamics, and variations in loading geometry. Computer Program LIDOP (ME-184) Local Indentation of Piping (Appendix Section 3A.2.14) augments the static crush tests.

3.6-38 Rev. 30

MPS3 UFSAR 3.6.2.2.7 Concrete Barrier Impact In a pipe whip impact, the force on the barrier is a complex function of time depending primarily on the sudden deceleration of the pipe wall at the impact point (slug impact), the shell indentation of the pipe as it locally crushes against the wall, and the force transmitted to the impact point by the more gradual deceleration of the adjacent run of pipe. After impact, the pipe also transmits a more enduring force resulting from the continuing fluid blowdown. The concrete is affected by this much like any other missile impact, the only significant difference being the long term fluid force. To evaluate this postulated event, the pipe is transformed into an equivalent missile and the concrete is analyzed for scabbing and structural response using the procedure described in Section 3.5.3. The analysis for structural response includes the impulse of the initial impact, as well as the subsequent fluid blowdown force and other concurrent loads.

Four basic parameters must be determined to define the equivalent missile: the kinetic energy (or impulse), the impact velocity, the pipe crush stiffness, and the bearing area. The kinetic energy and velocity can be found by either of two methods:

1. Simplified Method - Use the total input energy (fluid blowdown force x distance of pipe travel) less the energy absorbed in pipe bending prior to impact. Compute the velocity using approximate formulae.
2. Lumped Parameter Dynamic Analysis (Section 3.6.2.2.4) - This method is especially suited for evaluating the impact of piping systems with complex geometries and can even consider multiple impact points. As an alternative to the kinetic energy, the impact force history (impulse) can be computed.

Regardless of which analysis method is used, the crush resistance of the equivalent missile and the bearing area are derived from the experimental data described in Section 3.6.2.2.6. These data are modified to account for the effect of dynamics and internal pressure.

3.6.2.3 Dynamic Analysis Methods to Verify Integrity and Operability Required pipe rupture loads to determine the integrity of mechanical components are determined using the analytical methods described in Section 3.6.2.2. The load combinations for the components and for break exclusion regions are presented in Sections 3.9 and 3.6.2.1, respectively. Criteria for rupture restraints are presented in Section 3.6.2.3.1.

Jet impingement loadings are determined as follows:

1. Jet forces are represented by time-dependent forcing functions. The effects of the piping geometry, capacity of the upstream energy reservoir, source pressure, and fluid enthalpy are considered in utilizing these forcing functions.
2. The steady state jet force at the postulated break has a magnitude of:

T = KPA (3.6.2.3-1) 3.6-39 Rev. 30

MPS3 UFSAR where:

P = System pressure prior to pipe break A = Pipe break area K = Jet coefficient (theoretical maximum)

The following K values are:

a. 1.26 for saturated steam, saturated water, and steam/water mixtures blowdown (presented in 3.6.2.2.2).
b. 2.00 for nonflashing subcooled water blowdown (presented in 3.6.2.2.2).
3. In calculating the jet impingement load on an object or target, the retarding action of the surrounding air along the jet path is neglected. The jet impingement pressure on the target is calculated by taking the jet force as being constant at all distances from, and normal to, the break area and by assuming that the jet stream diverges conically at a solid angle of 20 degrees for steam or water-steam mixtures. For those cases where the 20-degree divergence assumption is shown to be unnecessarily conservative for the blowdown of steam or steam-water mixtures, Moodys asymptotic jet expansion model is utilized (Moody 1969). Jet expansion is not applicable to cases involving saturated water or subcooled water blowdown which are below the saturation temperature at the corresponding ambient pressure beyond the break.
4. The proportion of the total jet force acting on a target is determined from the fraction of the jet intercepted and by the shape factor of the target. For a target with flat surface area normal to the center axis of the jet stream, the load is the product of the impingement pressure at the target and the intercepted jet area. In those cases where the target area is such that the intercepted jet stream is deflected rather than totally stopped, a shape factor which is less than unity and is a function of the target geometry is used in calculating the total jet impingement load. For a 20-degree divergence angle and target at a distance x, the pressure intensity at the target is:

d1 2 P = P 1 d ----------------------------------- (3.6.2.3-2) d 1 + 2x tan 10 where:

P1 = Pressure intensity at the source (psi) d1 = Diameter of the assumed circular break area (inch) x = Normal distance between the break and the target (inch) 3.6-40 Rev. 30

MPS3 UFSAR Since the jet impingement force is a dynamically applied load, the target is analyzed either by static methods using an appropriate dynamic load factor, or dynamically using elastic or inelastic structural response codes (Section 3.8.3.3). The load combinations and design allowables are given in Sections 3.8.3 and 3.9.

Attenuation of fluid jet impingement for target assessment of Design Basis Accident (DBA) mitigating systems utilizes the approach presented in NUREG/CR-2913 (Ref: Weigand, Thompson, and Tomasko).

3.6.2.3.1 Pipe Rupture Restraints Two basic restraint types are used, elastic and energy-absorbing. Elastic restraints are generally used where displacements subsequent to a postulated pipe rupture must be minimized to restrict the break opening area, limit loads in the broken piping run, limit the pipe movement to protect some equipment, or to limit pressure buildup to minimize external loading on structures and equipment. Energy-absorbing restraints are used where the primary objective is to dissipate the kinetic energy of a ruptured pipe and prevent unrestricted pipe whip.

Elastic Restraints Since elastic restraints are used to minimize displacements of the broken pipe, they are close gaped. For some applications, this requires that they contact the pipe during conditions other than a postulated rupture, in which case they are also designed as a pipe support. If an elastic restraint contacts the pipe following a rupture, it is designed according to the criteria for structural steel (Section 3.8.3) which in effect limits stresses to the elastic range.

Energy-Absorbing Restraints Several approaches are used for energy absorption in pipe rupture restraints. In tension, stainless steel studs or straps are used, with a design limit of 50 percent of uniform ultimate strain. In compression, honeycomb panels or crushable pipes are used. The design limit for crushable energy absorbing components is defined in Section 3.6.2.2.1, Item 5.

Elastic intermediate structures of energy-absorbing restraints are designed to the criteria for structural steel (Section 3.8.3).

1. Pipe Crush Bumper - The pipe crush bumper absorbs impact energy in a direction toward the supporting structure. The energy absorber is a length of pipe placed normal to the axis of the process pipe. Subsequent to a rupture, the bumper pipe is crushed between its support structure and the moving process pipe. Energy is absorbed by deformation of the bumper pipe which forms a retaining recess in the bumper pipe. The bumper pipe is mechanically attached to its support by welding or bolting (Figures 3.6-29 and 3.6-30).
2. Laminated Strap Restraint - The laminated strap restraint is capable of absorbing impact loads in the outward direction from the supporting structure 3.6-41 Rev. 30

MPS3 UFSAR (Figure 3.6-22). The energy-absorbing component is a U shaped strap consisting of multiple strips (number and geometry depending on energy to be absorbed) of highly ductile material.

This laminated design exhibits great flexibility in application. The design minimizes bending strains, permitting the strap to act mainly as a membrane during the postulated rupture event.

3. Omni-Directional Restraint - The omni-directional restraint is capable of absorbing impact loads applied in any direction in the plane of the restraint (Figure 3.6-31). This restraint consists of a base weldment, an arch, ductile stainless steel holddown studs on each side of the base weldment, and a crushable pipe. The primary function of the studs is to absorb impact energy in tension. The crushable pipe absorbs energy from impact loads acting in an inward direction.

Side load impacts are absorbed by the combined action of the studs and crushable pipe.

Combinations of pipe crush bumpers and laminated straps are used to achieve energy absorption over a range of impact directions up to a full 360 degrees.

3.6.2.4 Guard Pipe Assembly Design Criteria Guard pipes were not used on Millstone 3.

3.

6.3 REFERENCES

FOR SECTION 3.6 3.6-1 De Haller, P. 1945. The Application of Graphical Method to Some Dynamic Problems in Gases. Sulzer Technical Review No. 1, p 6 24.

3.6-2 Gerber, T. L. 1974. Plastic Deformation of Piping Due to Pipe Whip Loading. ASME Paper 74 NE 1.

3.6-3 Hartree, D. R. 1952. Some Practical Methods of Using Characteristics in the Calculation of Non-Steady Compressible Flow. Los Alamos Report LA HU 1, p a 44, Los Alamos, N. Mex.

3.6-4 Jonssen, V. K.; Matthews, L.; and Spalding, D. B. 1953. Numerical Solution Procedure for Calculating the Unsteady, One-Dimensional Flow of Compressible Fluid with Allowance for the Effect of Heat Transfer and Friction. ASME Paper 73 FE 23, Report AECU 2713, USAEC, p 1 5.

3.6-5 Moody, J. F. 1961. Pipe Friction Manual (chart), 3rd Edition. Hydraulic Institute, New York, N.Y.

3.6-6 Moody, J. F. 1973. Time Dependent Pipe Forces Caused by Blowdown and Flow Stoppage. ASME Paper 73 FE 23.

3.6-42 Rev. 30

MPS3 UFSAR 3.6-7 Owzarek, J. A. 1968. Fundamentals of Gas Dynamics. International Textbook Company, Second Printing, Scranton, Pa.

3.6-8 Peech, J. M., Roemer, R. E., Pirotin, S. D., East, G. H., and Goldstein, N. A. 1977. Local Crush Rigidity of Pipes and Elbows. Paper F3/8, Transactions of the Fourth International Conference on Structural Mechanics in Reactor Technology, San Francisco, Calif.

3.6-9 Rudinger, G. 1969. Nonsteady Duct Flow-Wave Diagram Analysis. Dover Publications, Inc., New York, N.Y.

3.6-10 Shapiro, A. H. 1953. The Dynamics and Thermodynamics of Compressible Fluid Flow, Vol. I and II. Ronald Press, New York, N.Y.

3.6-11 Streeter, V. L. and Wylie, E. G. 1967. Hydraulic Transients. McGraw-Hill Book Co.,

New York, N.Y., p 32 37.

3.6-12 Letter from the NRC to Northeast Nuclear Energy Company (NNECO) approving NNECOs Request for Exemption from a Portion of General Design Criterion 4 of Appendix A to 10 CFR 50 regarding the Need to Analyze Large Primary Loop Pipe Ruptures as the Structural Design Basis for Millstone Nuclear Power Station Unit 3.

June 5, 1985.

3.6-13 Weigand, G. G., Thompson, S. L., Tomasko, D. - Two Phase Jet Loads - Sandia National Laboratories, Albuquerque, NM - January 1983 for the USNRC, NUREG/CR -

2913.

3.6-14 WCAP-14572, Revision 1-NP-A, Addendum 1, Addendum to Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Topical Report to Address Changes to Augmented Inspection Requirements.

3.6-43 Rev. 30

MPS3 UFSAR TABLE 3.6-1 HIGH-ENERGY SYSTEMS REMOTE FROM ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS MAXIMUM OPERATING MAXIMUM TEMPERATURE (1) OPERATING Approximately in Degrees PRESSURE (1)

SYSTEM CODE LOCATION Fahrenheit Approximately in psig Auxiliary Boiler Blowdown ABD Auxiliary Boiler Room, 228 7 Turbine Building Condensate Polishing Area Auxiliary Boiler Condensate ABF Auxiliary Boiler Room 228 165 Auxiliary Boiler Steam ABM Auxiliary Boiler Room 366 150 Chemical Feed Condensate (2) CNC Turbine Building 94 550 Chemical Feed Steam Generator SGF Turbine Building 446 1262 Cold Reheat CRS Turbine Building 374 183 Condensate Demineralizer CND Condensate Polishing Area 94 550 MPS3 UFSAR Mixed Bed Extraction Steam ESS Turbine Building 454 424 Feedwater Heater Relief Vents & SVH Turbine Building 442 375 Drains Feedwater Pump Recirculation and FWR Turbine Building 367 1632 Balance Drum Leakoff Feedwater Pump Turbine Steam & TFM Turbine Building 557 1092 Exhaust High Pressure Feedwater Heater HDH Turbine Building 450 407 Drains 3.6-44 Rev. 30

MPS3 UFSAR TABLE 3.6-1 HIGH-ENERGY SYSTEMS REMOTE FROM ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)

MAXIMUM OPERATING MAXIMUM TEMPERATURE (1) OPERATING Approximately in Degrees PRESSURE (1)

SYSTEM CODE LOCATION Fahrenheit Approximately in psig Hot Reheat HRS Auxiliary Boiler Room, 510 106 Turbine Bldg.

Hot Water Heating HVH Condensate Polishing Area 270 185 Turbine Building Service Building, Warehouse Condenser Air Removal ARC Turbine Building 212 125 Main Feedwater FWS Turbine Building 446 1401 Radioactive Liquid Waste (3) LWS Turbine Building 266 150 MPS3 UFSAR Waste Disposal Building, Yard Instrument Air IAS Turbine Building 350 110 Low Pressure Feedwater Heater HDL Turbine Building 374 519 Drains Condensate Demineralizer Liquid LWC Warehouse 5 250 63 Waste Turbine Plant Miscellaneous Drains DTM Turbine Building 556 1092 Turbine Plant Sample SST Turbine Building 556 1106 Main Condensate (4) CNM Turbine Building 366 551 Nitrogen Supply GSN Turbine Building, Yard Ambient 950 3.6-45 Rev. 30

MPS3 UFSAR TABLE 3.6-1 HIGH-ENERGY SYSTEMS REMOTE FROM ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)

MAXIMUM OPERATING MAXIMUM TEMPERATURE (1) OPERATING Approximately in Degrees PRESSURE (1)

SYSTEM CODE LOCATION Fahrenheit Approximately in psig Condensate Make-Up & Draw Off CNS Turbine Building, Yard 250 531 Moisture Separator Reheater Vents & DSR Turbine Building 536 936 Drains Fire Protection Low Pressure CO2 FPL Auxiliary Boiler Room, 60 300 Turbine Bldg.

Reactor Plant Aerated Vents VAS Waste Disposal Building 250 15 Moisture Separator Reheater Relief Valve Discharge and Bonnet Vent HRS Turbine Building 510 106 MPS3 UFSAR Moisture Separator Vents & Drains DSM Turbine Building 373 518 Turbine Generator Gland Seal & TME Turbine Building 557 1107 Exhaust Steam NOTES:

(1) Maximum Operating Temperature and Maximum Operating Pressure are tabulated for the Normal Plant Conditions. These are approximate values presented to classify a system as either high or moderate energy.

(2) The Chemical Feed Condensate High Energy piping extends from the Condensate Chemical Addition Feed Pump Discharge to the Condensate Lines.

(3) The High Energy portion of the Radioactive Liquid Waste piping consists of the Reboiler and the Distillate Piping from the Waste Evaporator to the Distillate Tank and back to the Waste Evaporator.

(4) The Condensate High Energy piping includes the Condensate Pump Discharge through all stages of the Low Pressure Heating System 3.6-46 Rev. 30

MPS3 UFSAR TABLE 3.6-2 HIGH-ENERGY SYSTEMS IN PROXIMITY TO ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS MAXIMUM OPERATING MAXIMUM TEMPERATURE (1) OPERATING Approximately in Degrees PRESSURE (1)

HIGH ENERGY SYSTEM CODE LOCATION Fahrenheit Approximately in psig Reactor Coolant - Normal Charging RCS Containment 557 2485 Normal Charging (2) CHS Containment, Auxiliary 190 (3) 2485 Reactor Coolant - Normal Letdown RCS Containment 557 2485 Normal Letdown (2) CHS Containment, Auxiliary 290 400 Seal Water Injection (2) CHS Containment, Auxiliary 130 2485 Reactor Coolant - Excess Letdown RCS Containment 557 2485 Reactor Coolant - Loop Fill RCS Containment 557 2485 MPS3 UFSAR Reactor Coolant - Loop Drains RCS Containment 557 2485 Reactor Coolant - High Pressure RCS Containment 557 2485 Safety Injection (4)

High Pressure Safety Injection (5) SIH Auxiliary Building Ambient 2485 Reactor Coolant - Low Pressure RCS Containment 557 2485 Safety Injection (4)

Low Pressure Safety Injection SIL Containment Ambient 695 Accumulator Discharge Reactor Coolant - Primary Coolant RCS Containment 557 2485 Reactor Coolant - Residual Heat RCS Containment 557 2485 Removal (6) 3.6-47 Rev. 30

MPS3 UFSAR TABLE 3.6-2 HIGH-ENERGY SYSTEMS IN PROXIMITY TO ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)

MAXIMUM OPERATING MAXIMUM TEMPERATURE (1) OPERATING Approximately in Degrees PRESSURE (1)

HIGH ENERGY SYSTEM CODE LOCATION Fahrenheit Approximately in psig Reactor Coolant - Accumulator RCS Containment 557 2485 Discharge Reactor Coolant - Pressurizer Surge RCS Containment 557 2485 Reactor Coolant - Pressurizer Spray RCS Containment 557 2485 Reactor Coolant - Pressurizer Safety RCS Containment 557 2485 Reactor Coolant - Pressurizer Relief RCS Containment 557 2485 Reactor Coolant - Loop Stop Valve RCS Containment 557 2485 Bypass Reactor Coolant - Loop Stop Valve RCS Containment 557 2485 MPS3 UFSAR Bypass - 1.5 inch Bypass Line Steam Generator Blowdown BDG Containment Main Steam 543 975 Valve Building (MSVB)

Auxiliary Steam ASS Engineered Safety Features 366 150 (ESF) Auxiliary Building Hot Water Heating HVH Auxiliary Building Fuel 270 185 Building Main Steam Valve Building (MSVB)

Boron Recovery (7) BRS Auxiliary Building 312 165 Containment Vacuum Pump CVS Auxiliary Building 268 1.4 Discharge Containment 3.6-48 Rev. 30

MPS3 UFSAR TABLE 3.6-2 HIGH-ENERGY SYSTEMS IN PROXIMITY TO ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)

MAXIMUM OPERATING MAXIMUM TEMPERATURE (1) OPERATING Approximately in Degrees PRESSURE (1)

HIGH ENERGY SYSTEM CODE LOCATION Fahrenheit Approximately in psig Auxiliary Condensate CNA Auxiliary Building 366 150 Reactor Plant Gaseous Drains DGS Containment Auxiliary 225 110 Building Radioactive Gaseous Waste (8) GWS Auxiliary Building 350 300 Turbine Plant Miscellaneous Drains DTM Main Steam Valve Building 575 1262 (MSVB) Engineered Safety Features (ESF)

Control Building Air Conditioning (9) HVC Control Building 115 2450 Main Steam MSS Containment, Engineered 557 1092 MPS3 UFSAR Safety Features (ESF),

Main Steam Valve Building (MSVB)

Auxiliary Feedwater FWA Containment, ESF 100 1658 Main Feedwater FWS Containment, MSVB 446 1980 Nitrogen Supply (10) GSN Auxiliary Building 90 950 Containment Emergency Diesel Generator Air Start EGA Emergency Generator Ambient 500 (11) Enclosure (EGE) 3.6-49 Rev. 30

MPS3 UFSAR TABLE 3.6-2 HIGH-ENERGY SYSTEMS IN PROXIMITY TO ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)

MAXIMUM OPERATING MAXIMUM TEMPERATURE (1) OPERATING Approximately in Degrees PRESSURE (1)

HIGH ENERGY SYSTEM CODE LOCATION Fahrenheit Approximately in psig Main Steam Safety Valve Vents & SVV Main Steam Valve Building 510 741 Drains (MSVB)

Fire Protection Low Pressure CO2 FPL Emergency Generator 60 300 Enclosure (EGE) Control Building Emergency Diesel Generator Exhaust EGD Emergency Generator 820 0 and Combustion (11) Enclosure (EGE)

Post Accident Sampling SSP Containment 617 2485 Reactor Plant Sampling SSR Auxiliary Building and 680 2485 Containment MPS3 UFSAR Reactor Plant Aerated Vents VAS Auxiliary Building 250 15 Steam Generator Chemical Feed SGF Containment, Main Steam 446 1262 Valve Building (MSVB)

NOTES:

(1) Maximum Operating Pressure and Maximum Operating Temperature is tabulated for Normal Plant Conditions. These are approximate values presented only to classify the system as high or moderate energy.

(2) The Chemical & Volume Control (CHS) System High Energy piping extends from the Reactor Coolant System isolation valves to the Charging Pump suction (via the Letdown Line through the Regenerative and Non-Regenerative Heat Exchangers and the Volume Control Tank) and from the Charging Pump Discharge to the Reactor Coolant System (via the Charging Line). The portion of piping from the Charging Pump Discharge to the Reactor Coolant Pumps (for Seal Water Injection) also is High Energy. The remainder of the Chemical & Volume Control System is Moderate Energy piping.

3.6-50 Rev. 30

MPS3 UFSAR (3) Normal Charging (CHS) Line Temperature at the outlet of the Regenerative Heat Exchanger may be as high as 500 Degrees F.

(4) The High Pressure Safety Injection (SIH) System, the Low Pressure Safety Injection (SIL) System, and the Containment Recirculation System are used only following a Loss of Coolant Accident (LOCA) thus these systems are excluded from pipe rupture analysis. Portions of piping included within the Reactor Coolant System pressure boundary are analyzed for pipe rupture effects (i.e. pipe whip and fluid jet impingement).

(5) The High Pressure Safety Injection (SIH) between the Charging Pump Discharge Header and the SIH Valves 3SIH*MV8801A, and 3SIH*MV8801B is High Energy.

(6) The Residual Heat Removal (RHS) System, although having pressure and temperature above the High Energy Threshold is classified as a Moderate Energy System as delineated in FSAR Section 3.6.2.1.2.3. Portions of piping included within the Reactor Coolant System pressure boundary are analyzed for pipe rupture effects (i.e. pipe whip and fluid jet impingement).

(7) Boron Recovery (BRS) System High Energy piping consists of the Boron Evaporator Recirculation and Distillation piping from the Boron Evaporator to the Boron Distillate Cooler and Boron Evaporator Bottom Cooler.

(8) Radioactive Gaseous Waste (GWS) System High Energy piping extends from the Degasifier Recovery Exchanger, the Degasifier, the Degasifier Recirculation Pumps and back to the Degasifier Recovery Exchanger.

(9) Control Building Air Conditioning (HVC) System High Energy components consists solely of the Control Room Air Bottle Supply.

(10) Portion of Nitrogen Gas System from Pressurized Nitrogen Gas Tubes to the Safety Injection Accumulator Tanks is High Energy piping. Pipe breaks are excluded in this line since the line is 1 inch diameter piping.

MPS3 UFSAR (11) The Emergency Diesel Generator (EGD) System and related auxiliaries is considered a single system for evaluating the effects of pipe break. A single failure is not postulated in the redundant system as discussed in FSAR Section 3.1.1.

3.6-51 Rev. 30

MPS3 UFSAR TABLE 3.6-3 MODERATE-ENERGY SYSTEMS IN REMOTE FROM ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS MAXIMUM MAXIMUM OPERATING OPERATING TEMPERATURE (1) PRESSURE (1)

Approximate Values Approximate Values MODERATE ENERGY SYSTEM CODE LOCATION Tabulated Tabulated Auxiliary Boiler Fuel Oil FOA Auxiliary Boiler Room 100°F 48 psig Condensate Makeup & Drawoff CNS Turbine Building, Yard 120°F 136 psig Condenser Air Removal ARC Turbine Building 170°F 125 psig Condenser Tube Cleaning CWA Turbine Building 95°F 29 psig Domestic Water DWS Auxiliary Boiler, Turbine 140°F 80 psig Building, Service Building Feedwater Pump & Drive Lube Oil FWL Turbine Building 145°F 20 psig MPS3 UFSAR Fire Protection - Water FPW Auxiliary Boiler Room, Turbine 110°F 122 psig Building, Service Building, Waste Disposal Building, Transformer Area Generator Hydrogen (H2) & GMH Turbine Building 55°F 70 psig Generator Carbon Dioxide (CO2)

Instrument Air IAS Auxiliary Boiler Room, Turbine 115°F 110 psig Building, Service Building, Waste Disposal Building Boron Recovery (2) BRS Waste Disposal Building, Yard 170°F 80 psig 3.6-52 Rev. 30

MPS3 UFSAR TABLE 3.6-3 MODERATE-ENERGY SYSTEMS IN REMOTE FROM ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)

MAXIMUM MAXIMUM OPERATING OPERATING TEMPERATURE (1) PRESSURE (1)

Approximate Values Approximate Values MODERATE ENERGY SYSTEM CODE LOCATION Tabulated Tabulated Main Condensate (3) CNM Turbine Building 200°F Hydrogen (H2) Gas GSH Gas Storage Area, Turbine 90°F 100 psig Building, Yard Nitrogen (N2) Gas GSN Nitrogen Storage Pad, Ambient 200 psig Auxiliary Boiler Room, Service Building Primary Grade Water PGS Primary Grade Water Pump 100°F 122 psig House, Yard MPS3 UFSAR Hot Water Pre-Heating HVG Service Building 200°F Service Air SAS Auxiliary Boiler Room, Turbine 115°F 110 psig Building, Service Building, Waste Disposal Building, Circulating Water Pump House (CWPH), Warehouse 5 Reactor Plant Aerated Vents VAS Waste Disposal Building 200°F 15 psig Reactor Plant Gaseous Vents VRS Waste Disposal Building 200°F 75 psig Reactor Plant Aerated Drains DAS Waste Disposal Building 200°F 70 psig Service Water SWP Auxiliary Boiler Room, Turbine 80°F 66 psig Building, Service Building 3.6-53 Rev. 30

MPS3 UFSAR TABLE 3.6-3 MODERATE-ENERGY SYSTEMS IN REMOTE FROM ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)

MAXIMUM MAXIMUM OPERATING OPERATING TEMPERATURE (1) PRESSURE (1)

Approximate Values Approximate Values MODERATE ENERGY SYSTEM CODE LOCATION Tabulated Tabulated Turbine Generator Lube Oil TML Turbine Building 100°F 0 psig Waste Oil Disposal WOS Turbine Building 100°F 15 psig Waste Water Treatment WTW Turbine Building Ambient Atmospheric Water Treating WTS Turbine Building, Yard 60°F 128 psig Station Vacuum Priming VPS Turbine Building 115°F 20 psig Reactor Plant Solid Waste WSS Waste Disposal Building 100°F 140 psig Chemical Feed Chlorination WTC Circulating Water Pump House 80°F 248 psig MPS3 UFSAR Turbine Plant Component Cooling CCS Turbine Building 95°F 125 psig Water Auxiliary Feedwater (4) FWA Yard 100°F 60 psig Condensate Demineralizer CCD Condensate Polishing Area 115°F 91 psig Component Cooling Quench Spray QSS Yard 75°F 158 psig Traveling Screen Wash & Disposal SWT Circulating Water Pump House 80°F 135 psig Yard Vacuum Priming VPS Circulating Water Pump House 115°F 30 psig Circulating Water CWS Turbine Building, Yard, 80°F 26 psig Circulating Water Pump House 3.6-54 Rev. 30

MPS3 UFSAR TABLE 3.6-3 MODERATE-ENERGY SYSTEMS IN REMOTE FROM ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)

MAXIMUM MAXIMUM OPERATING OPERATING TEMPERATURE (1) PRESSURE (1)

Approximate Values Approximate Values MODERATE ENERGY SYSTEM CODE LOCATION Tabulated Tabulated Low Pressure Safety Injection SIL Yard Ambient 30 psig Containment Recirculation Spray RSS Yard 115°F 225 psig Note (1) Maximum Operating Pressure and Maximum Operating Temperature for the Normal Plant Conditions are tabulated. These values are approximate and are presented to classify the system as high or moderate energy.

Note (2) Boron Recovery High Energy piping consists of the Boron Evaporator Recirculation and Distillate piping from the Boron Evaporator to the Boron Distillate Cooler. Remainder of the Boron Recovery (BRS) System is Moderate Energy piping.

Note (3) Moderate Energy piping for the Main Condensate (CNM) System extends from the Condenser to the Condensate Pump Suction, and includes the Seal Water piping to the Feed Pump Exhaust Isolation Valves, Air Ejectors Condenser Loop Seal, Extraction Nonreturn MPS3 UFSAR Valves, 6th Point Drain Loop Seal, Condenser Vacuum Breaker Valves and Loop Seals.

Note (4) Auxiliary Feedwater and Recirculation (FWA) System is Moderate Energy piping from the Demineralized Water Storage Tank (DWST) to the suction side of the Auxiliary Feedwater pumps. This Moderate Energy piping includes the Recirculation piping.

3.6-55 Rev. 30

MPS3 UFSAR TABLE 3.6-4 MODERATE-ENERGY SYSTEMS PROXIMATE TO ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS MAXIMUM MAXIMUM OPERATING OPERATING TEMPERATURE (1) PRESSURE (1)

MODERATE ENERGY SYSTEM CODE LOCATION Approximate Values Approximate Values Auxiliary Feedwater & Recirculation FWA Engineered Safety Features 100°F 48 psig (2) Building Residual Heat Removal (3) RHS Engineered Safety Features 350°F 575 psig Building, Containment Service Water SWP Auxiliary Building, Engineered 95°F 66 psig Safety Features Building, Circulating Water Pump House, Emergency Generator Enclosure, Control Building Fire Protection - Water FPW Auxiliary Building, Emergency 110°F 122 psig MPS3 UFSAR Generator Enclosure, Control Building, Fuel Building, Main Steam Valve Building, Containment Component Cooling Water CCP Auxiliary Building, Engineered 137°F 186 psig Safety Features Building, Fuel Building, Containment Control Building - Chilled Water HVK Control Building 55°F 70 psig 3.6-56 Rev. 30

MPS3 UFSAR TABLE 3.6-4 MODERATE-ENERGY SYSTEMS PROXIMATE TO ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)

MAXIMUM MAXIMUM OPERATING OPERATING TEMPERATURE (1) PRESSURE (1)

MODERATE ENERGY SYSTEM CODE LOCATION Approximate Values Approximate Values Instrument Air IAS Auxiliary Building, Engineered 115°F 110 psig Safety Features Building, Fuel Building, Emergency Generator Enclosure, Control Building, Containment, Hydrogen Recombiner Building, Main Steam Valve Building Radioactive Solid Waste WSS Auxiliary Building, Fuel 100°F 140 psig Building Charging Pump Cooling CCE Auxiliary Building 105°F 58 psig MPS3 UFSAR Fuel Pool Cooling & Purification SFC Auxiliary Building, Engineered 150°F 135 psig Safety Features Building, Containment, Fuel Building Chemical & Volume Control (4) CHS Auxiliary Building 190°F 220 psig Reactor Plant Aerated Drains DAS Auxiliary Building, Engineered 200°F 70 psig Safety Features Building, Containment, Fuel Building Domestic Water DWS Auxiliary Building, Control 140°F 80 psig Building Fire Protection - Halon FPG Control Building Ambient 36 psig 3.6-57 Rev. 30

MPS3 UFSAR TABLE 3.6-4 MODERATE-ENERGY SYSTEMS PROXIMATE TO ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)

MAXIMUM MAXIMUM OPERATING OPERATING TEMPERATURE (1) PRESSURE (1)

MODERATE ENERGY SYSTEM CODE LOCATION Approximate Values Approximate Values Containment Instrument Air IAC Containment 120°F 110 psig Containment Atmosphere Monitoring CMS Auxiliary Building, 100°F Atmospheric Containment Safety Injection Pump Cooling CCI Engineered Safety Features 110°F 35 psig Building Chilled Water, Containment Structure CDS Auxiliary Building, 90°F 140 psig Ventilation Containment, Engineered Safety Features Building Piping for Containment Purge Air HVU Auxiliary Building, 120°F 5 psig Containment MPS3 UFSAR Containment Leak Monitoring LMS Containment, Auxiliary 90°F 0 psig Building Hydrogen Recombiner HCS Hydrogen Recombiner 135°F 0 psig Building Chemical Feed Chlorination, WTC Circulating Water Pumphouse 80°F 248 psig Emergency Diesel Generator Jacket & Intercooler Water EGS Emergency Generator 180°F 50 psig Enclosure Hydrogen Gas GSH Auxiliary Building 90°F 100 psig Nitrogen Supply GSN Auxiliary Building, Ambient 200 psig Containment 3.6-58 Rev. 30

MPS3 UFSAR TABLE 3.6-4 MODERATE-ENERGY SYSTEMS PROXIMATE TO ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)

MAXIMUM MAXIMUM OPERATING OPERATING TEMPERATURE (1) PRESSURE (1)

MODERATE ENERGY SYSTEM CODE LOCATION Approximate Values Approximate Values Quench Spray QSS Engineered Safety Features 75°F 158 psig Building, Containment Neutron Shield Tank Cooling NSS Containment 135°F 20 psig Sanitary PBS Control Building 85°F Static Head Reactor Plant Gaseous Vents VRS Auxiliary Building, 200°F 75 psig Containment Service Air SAS Auxiliary Building, Engineered 115°F 110 psig Safety Features Building, Fuel Building, Emergency Generator Enclosure, Control MPS3 UFSAR Building, Hydrogen Recombiner Building Emergency Generator Fuel EGF Emergency Generator 120°F 32 psig Enclosure Low Pressure Safety Injection SIL Engineered Safety Features 115°F 235 psig Building, Containment Floor Equipment Drainage DNF Circulating Water Pumphouse, Ambient Ambient Engineered Safety Features Building, Auxiliary Building, Fuel Building 3.6-59 Rev. 30

MPS3 UFSAR TABLE 3.6-4 MODERATE-ENERGY SYSTEMS PROXIMATE TO ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)

MAXIMUM MAXIMUM OPERATING OPERATING TEMPERATURE (1) PRESSURE (1)

MODERATE ENERGY SYSTEM CODE LOCATION Approximate Values Approximate Values Roof Drainage DNR Engineered Safety Features 90°F Ambient Building, Auxiliary Building, Fuel Building, Main Steam Valve Building, Control Building Reactor Coolant Pump Oil Collection FPR Containment 160°F Atmospheric Primary Grade Water PGS Auxiliary Building, Engineered 100°F 122 psig Safety Features Building, Fuel Building, Containment Containment Recirculation Spray RSS Engineered Safety Features 115°F 225 psig MPS3 UFSAR Building, Containment Boron Recovery (5) BRS Auxiliary Building 170°F 80 psig Emergency Generator Lube Oil EGO Emergency Generator 160°F 150 psig Enclosure Condenser Air Removal ARC Auxiliary Building 170°F 125 psig Reactor Plant Gaseous Drains DGS Auxiliary Building, 165°F 104 psig Containment Reactor Plant Aerated Vents VAS Auxiliary Building 120°F 0 psig Containment Vacuum (6) CVS Auxiliary Building, 90°F 5 psig Containment 3.6-60 Rev. 30

MPS3 UFSAR TABLE 3.6-4 MODERATE-ENERGY SYSTEMS PROXIMATE TO ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)

MAXIMUM MAXIMUM OPERATING OPERATING TEMPERATURE (1) PRESSURE (1)

MODERATE ENERGY SYSTEM CODE LOCATION Approximate Values Approximate Values Feedwater Lube Oil FWL Engineered Safety Features 20°F 140 psig Building Hot Water Pre-Heating HVG Auxiliary Building 185°F 170 psig NOTES:

(1) Maximum Operating Pressure and Maximum Operating Temperature are tabulated for the Normal Plant Conditions. These are approximate values presented to classify the system as high and moderate energy.

(2) The Auxiliary Feedwater and Recirculation (FWA) System Moderate Energy piping extends from the Demineralized Water Storage Tank (DWST) to the suction side of the Auxiliary Feedwater Pumps which includes the Recirculation piping.

(3) Residual Heat Removal (RHS) System is a Moderate Energy System in accordance with the 2% Rule. The 2% Rule is defined in MPS3 UFSAR FSAR Section 3.6.2.1.2.3.

(4) The Moderate Energy piping for the Chemical Volume Control (CHS) System extends from the Letdown Heat Exchanger pressure reducing valve via the Mixed Bed Demineralizer and the Thermal Regeneration Remineralizer lines to the Volume Control Tank.

(5) Boron Recovery (BRS) System High Energy piping consists of the Boron Evaporator Recirculation and Distillate piping from the Boron Evaporator to the Boron Distillate Cooler. Remainder of the Boron Recovery (BRS) System is Moderate Energy piping.

(6) Containment Vacuum (CVS) System Moderate Energy piping extends from Inside Containment to the Containment Vacuum Pump suction. The remainder of Containment Vacuum (CVS) System piping is High Energy.

3.6-61 Rev. 30

MPS3 UFSAR TABLE 3.6-5 ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN SYSTEM BUILDING REQUIRED SAFETY FUNCTION Reactor Coolant (RCS) (1) (2) Containment Reactor Core Integrity and Heat Removal Reactor Vessel 3RCS*RV1 Containment Pressure Boundary integrity needed to maintain fuel within acceptable temperature limits Reactor Vessel 3RCS*RV1 Head Vent Containment Provide Letdown for Reactivity & Inventory Control and Reactor Coolant Pressure Control Pressurizer 3RCS*TK1 Containment Pressure Boundary - Maintain RCS Pressure Control Steam Generators 3RCS*SG1A, SG1B, SG1C, and Containment Remove Heat from Core SG1D Reactor Coolant Pumps 3RCS*P1A, P1B, P1C, and Containment Maintain Pressure Boundary Integrity P1D Reactor Coolant Loop Stop Valves 3RCS*MV8002A, Containment Maintain Pressure Boundary Integrity - Primary MPS3 UFSAR 3RCS*MV8002B, 3RCS*MV8002C, and Boundary 3RCS*MV8002D Reactor Coolant Loop Stop Valve Bypass Containment Piping and Valves on the Reactor Coolant System Up to Containment Maintain Pressure Boundary Integrity - Primary the Class 1/Class 2 Boundary as Delineated in FSAR Boundary Table 3.5-2 Pressurizer Power Operated Relief Valves Operability - Alternate Means of RCS Depressurization (PORV's) 3RCS*PCV455A, and 3RCS*PCV456 Containment Control Rod Drive Mechanism Containment Insert Control Rods to Stabilize Plant at Hot Standby Pressurizer Safety Valves 3RCS*SV8010A, B, C Containment Operability - Overpressure Protection for RCS System 3.6-62 Rev. 30

MPS3 UFSAR TABLE 3.6-5 ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)

SYSTEM BUILDING REQUIRED SAFETY FUNCTION PORV Block Valves 3RCS*MV8000A, and Containment Operability - Support PORV Function 3RCS*MV8000B Letdown/RCS Isolation Valves 3RCS*LCV459 & Containment Isolates Reactor Coolant System for Letdown Line LCV460 Break Residual Heat Removal (RHS) (1) (2) ESF Low Pressure Safety Injection Path and Reactor Coolant Heat Removal Residual Heat Removal Heat Exchanger Flow Control Operability - Supports RHS Function for Controlled Valves Cooldown 3RHS*FCV610, and 3RHS*FCV611 ESF Operability - Supports RHS Function for Controlled Cooldown 3RHS*FCV618, and 3RHS*FCV619 ESF Operability - Supports RHS Function for Controlled Cooldown MPS3 UFSAR 3RHS*HCV606 and 3RHS*HCV607 ESF Operability - Supports RHS Function for Controlled Cooldown Three RHS Series Isolation Valves 3RHS*MV8701A, Containment & ESF Operability - Initiate Second Stage of RHS by MV8701B, 8701C, 8702A, 8702B, and 8702C Opening all three Isolation Valves on either Train Residual Heat Removal (RHS) Pumps 3RHS*P1A,B ESF Operability - Takes Suction from RCS Hot Leg &

RWST RHS Heat Exchangers 3RHS*E1A, and 3RHS*E1B ESF Operability - Cools Reactor Coolant from Hot Leg and Discharges Reactor Coolant Back to Cold Leg via Safety Injection System RHS Pump Suction 3SIL*MV8812A, and ESF Operability - During Second Stage of Residual Heat 3SIL*MV8812B Removal and Long Term Sump Recirculation 3.6-63 Rev. 30

MPS3 UFSAR TABLE 3.6-5 ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)

SYSTEM BUILDING REQUIRED SAFETY FUNCTION RHS Pump Discharge Containment Isolation Valves ESF Operability - Containment Isolation and Long Term 3SIL*MV8809A, and 3SIL*MV8809B Sump Recirculation 3SIL*MV8716A and 3SIL*MV8716B ESF Operability Required to Support Long Term Sump Recirculation 3SIL*MV8804A and 3SIL*MV8804B ESF Operability Required to Support Long Term Sump Recirculation Chemical & Volume Control (CHS) (1) (2) High Pressure Safety Injection, Reactivity &

Inventory Control Normal Charging Pumps 3CHS*P3A,B,CAuxiliary Auxiliary Operability - Reactivity & Inventory Control as well as Pressure Control, and Safety Injection Boric Acid Transfer Pumps 3CHS*P2A,B Auxiliary Operability Boric Acid Tanks 3CHS*TK5A,B Auxiliary Maintain Pressure Boundary Integrity MPS3 UFSAR Boric Acid Blender 3CHS*BL1 Auxiliary Maintain Pressure Boundary Integrity Charging Pump Suction 3CHS*LCV112B, C, D, and E Auxiliary Operability - Reactivity & Inventory Control Charging Pump Discharge Containment Isolation Valves Auxiliary Operability - Containment Isolation 3CHS*MV8105, and 3CHS*MV8106 Normal Letdown Line Containment Isolation Valve Auxiliary Operability - Containment Isolation 3CHS*CV8152 and 3CHS*CV8160 Letdown Line 3CHS*AV8149A,B,C Containment Maintain Pressure Boundary Integrity 3.6-64 Rev. 30

MPS3 UFSAR TABLE 3.6-5 ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)

SYSTEM BUILDING REQUIRED SAFETY FUNCTION Reactor Coolant Pump Seal Water Injection Auxiliary Operability - Containment Isolation Containment Isolation Valves 3CHS*MV8109A, 3CHS*MV8109B, 3CHS*MV8109C, and 3CHS*MV8109D 3CHS*MV8111A, 3CHS*MV8111B, and Auxiliary Operability - Supports ECCS Function 3CHS*MV8111C 3CHS*MV8512A, and 3CHS*MV8512B Auxiliary Operability - Supports ECCS Function 3CHS*MV8511A, and 3CHS*MV8511B Auxiliary Operability - Supports ECCS Function 3CHS*MV8110 Auxiliary Operability - Supports ECCS Function High Pressure Safety Injection (SIH) (1) (2) ESF Emergency Core Cooling System (ECCS)

High Pressure Safety Injection Pumps 3SIH*P1A,B ESF Operability-Supply Borated Water to all Four RCS MPS3 UFSAR Loops Valves 3SIH*MV8801A, and 3SIH*MV8801B Auxiliary Operability - Containment Isolation 3SIH*MV8802A, 3SIH*MV8802B, and ESF Operability - Containment Isolation 3SIH*MV8835 3SIH*MV8813, 3SIH*MV8814, and 3SIH*MV8920 ESF Operability - Supports ECCS Function Low Pressure Safety Injection (SIL) (1) (2)

Safety Injection Accumulators 3SIL*TK1A, 1B, 1C, Containment Inject Borated Water to RCS Cold Leg and Large and 1D LOCA 3.6-65 Rev. 30

MPS3 UFSAR TABLE 3.6-5 ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)

SYSTEM BUILDING REQUIRED SAFETY FUNCTION Safety Injection Accumulator Isolation Valves Containment Operability - Preclude Accumulator Injection during 3SIL*MV8808A, MV8808B, MV8808C, and RCS MV8808D Depressurization - Accumulators are Isolated or Vented Safety Injection Accumulator Purge Valves Containment Operability - Accumulator Purge Valves 3SIL*SV8875A, B, C, D, E, F, G, and H 3SIL*HCV943A, and 3SIL*HCV943B Containment Operability - Purge Accumulator if Accumulator Isolation Valves Fail to Close Main Feedwater (FWS) (1) (2) Decay Heat Removal by Supplying Cooling Water to the Steam Generators Main Feedwater Valves 3FWS*CTV41A, B, C, D MSV Operability - Containment Isolation Main Feedwater Valves 3FWS*FCV510, 520, 530, and MSV Operability - Valves Close on Feedwater Isolation MPS3 UFSAR 540 Signal Main Feedwater Bypass Valves 3FWS*LV550, 560, MSV Operability - Valves Close on Feedwater Isolation 570, and 580 Signal Auxiliary Feedwater (FWA) (1) (2) Decay Heat Removal by Supplying Cooling Water to the Steam Generators Auxiliary Feedwater Motor Driven Pumps 3FWA*P1A, ESF Operability - Provide Redundant Water Supply to at and 3FWA*P1B Least Two Steam Generators Until Initiation of RHS Auxiliary Feedwater Turbine Driven Pump 3FWA*P2 ESF Operability - Alternate Path for Water Supply Demineralized Water Storage Tank 3FWA*TK1 Yard Pressure Boundary - Supply for Auxiliary Feedwater Pump Suction 3.6-66 Rev. 30

MPS3 UFSAR TABLE 3.6-5 ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)

SYSTEM BUILDING REQUIRED SAFETY FUNCTION Auxiliary Feedwater Supply Valves 3FWA*MOV35A, ESF Operability - Containment Isolation B, C, D and 3FWA*HV36A, B, C, D Auxiliary Feedwater Supply Hand Control Valves ESF Operability - Control Flow to Steam Generator Level 3FWA*HV31A, B, C, and D and 3FWA*HV32A, B, C, to Support Cooldown and D Containment Recirculation Spray (RSS) (1) (2) ESF Recirculation of Emergency Core Cooling System (ECCS) Water following a Design Basis Accident (DBA)

Containment Recirculation Spray Pumps 3RSS*P1A, ESF Operability Containment Recirculation Spray Pumps 3RSS*P1B, 3RSS*P1C, and 3RSS*P1D are started on receipt of a RWST Low-Low signal coincident with a CDA Signal Containment Recirculation Spray Coolers 3RSS*E1A, ESF Operability Containment Recirculation Water Flows 3RSS*E1B, 3RSS*E1C, and 3RSS*E1D Shell side of the Heat Exchangers MPS3 UFSAR Containment Recirculation Spray Valves ESF Operability - Containment Isolation 3RSS*MOV20A, B, C, and D Containment Recirculation Spray Suction Isolation ESF Operability - Containment Isolation Valves 3RSS*MOV23A, B, C, D 3RSS*MOV8837A, and B and 3RSS*MOV8838A, and ESF Operability - Switchover to Long Term Sump B Recirculation Quench Spray (QSS) (1) (2) ESF Maintain Integrity of the Containment by Removing Heat from Containment Atmosphere following a LOCA or a Main Steam Line or Feedwater Line Break Quench Spray Pumps 3QSS*P3A, and 3QSS*P3B ESF Operability - Provide Borated Water to QSS Headers 3.6-67 Rev. 30

MPS3 UFSAR TABLE 3.6-5 ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)

SYSTEM BUILDING REQUIRED SAFETY FUNCTION Refueling Water Storage Tank 3QSS*TK1 Yard Provide Chilled Water Supply to QSS Pump Suction and suction for the RHS, SIH, and CHS Pumps for the ECCS Function Quench Spray Valves 3QSS*MOV34A,B ESF Operability - Containment Isolation and Open subsequent to a CDA Signal to Supply QSS Headers Steam Generator Blowdown (BDG) Maintain Pressure Boundary Integrity to preserve Heat Removal via Steam Generators Steam Generator Blowdown Containment Isolation MSV Operability - Containment Isolation Valves 3BDG*CTV22A, CTV22B, CTV22C, and CTV22D Main Steam (MSS) (1) (2) Maintain Pressure Boundary of Steam Generators &

Control Steam Release for Heat Removal MPS3 UFSAR Main Steam Isolation Valves 3MSS*CTV27A,B,C,D MSV Operability - Containment Isolation Main Steam Safety Valves 3MSS*RV22A, B, C, D MSV Operability & Over Pressure Protection Main Steam Relief Valves 3MSS*RV23A, B, C, and D MSV Operability & Over Pressure Protection 3MSS*RV24A,B,C and D 3MSS*RV25A, B, C and D MSV Operability & Over Pressure Protection and 3MSS*RV26A, B, C and D Main Steam to Turbine Driven Auxiliary Feedwater ESF Operability - Containment Isolation Pump 3MSS*MOV17A,B,D and 3MSS*AOV31A, Operability - Support Turbine Driven Auxiliary 3MSS*AOV31B, and 3MSS*AOV31D Feedwater Pump Operation Main Steam Pressure Relieving Bypass Valves MSV Operability - Provides Means to Reduce Steam 3MSS*MOV74A, B, C, D Pressure during First Stage of RCS Cool Down 3.6-68 Rev. 30

MPS3 UFSAR TABLE 3.6-5 ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)

SYSTEM BUILDING REQUIRED SAFETY FUNCTION Main Steam Pressure Relieving Valves 3MSS*PV20A, MSV Operability - Primary Means to Reduce Steam 20B, 20C, and 20D Pressure during First Stage of RCS Cool Down Main Steam Pressure Relieving Block Valves MSV Operability - Isolation 3MSS*MOV 18A, B, C, D 3MSS*HV28A, 3MSS*HV28B, 3MSS*HV28C, and MSV Operability - Containment Isolation 3MSS*HV28D Reactor Plant Component Cooling Water Auxiliary Remove Heat from Safety Components and Transfer (CCP) (1) (2) Heat to the Ultimate Heat Sink Reactor Plant Component Cooling Water Pumps Auxiliary Operability - Component Cooling Water Pumps 3CCP*P1A, 3CCP*P1B, and 3CCP*P1C Circulate Water Through the Various Closed Loops Reactor Plant Component Cooling Water Heat Auxiliary Operability - Transfer Heat to Service Water System Exchanger 3CCP*E1A, 3CCP*E1B, and 3CCP*E1C MPS3 UFSAR Residual Heat Removal Heat Exchanger Discharge ESF Operability - Normally Closed Valves Prevent Flow Valves 3CCP*FV66A, and 3CCP*FV66B Through RHS Heat Exchanger During Normal Plant Operation Valves Open Upon Residual Heat Removal System Initiation to Provide Component Cooling Water to RHS Heat Exchanger Reactor Plant Component Cooling Water Surge Tank Auxiliary Pressure Boundary Integrity - Provide Sufficient 3CCP*TK1 NPSH to CCP Pumps Reactor Plant Component Cooling Water Valves Auxiliary Operability - Containment Isolation 3CCP*MOV45A, and 3CCP*MOV45B Valves 3CCP*MOV48A, and 3CCP*MOV48B Auxiliary Operability - Containment Isolation 3.6-69 Rev. 30

MPS3 UFSAR TABLE 3.6-5 ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)

SYSTEM BUILDING REQUIRED SAFETY FUNCTION Valves 3CCP*MOV49A, and 3CCP*MOV49B Auxiliary Operability - Containment Isolation Charging Pump Cooling (CCE) (1) (2) Auxiliary Operability - Supports CHS Pump Function Safety Injection Pump Cooling (CCI) (1) (2) ESF Operability - Supports SIH Pump Function Hot Water Heating (HVH) (3) Auxiliary None Isolation Valves 3HVH-AOV135A, B and 3HVH- Service Operability - A High Energy Line Break on HVH in AOV136A, B the Auxiliary Building Trips a Pressure Switch which Closes these Valves to Isolate Flow to the Auxiliary Building Auxiliary Steam (ASS) (3) Auxiliary None Isolation Valves 3ASS-AOV102A and 3ASS-AOV102B Turbine Operability - A High Energy Line Break on Auxiliary Steam (ASS) System in the Auxiliary Building Trips a MPS3 UFSAR Pressure Switch which Closes these Valves to Isolate Flow to the Auxiliary Building Service Water (SWP) (1) (2) Transfer Heat from Reactor Cooling Systems to Ultimate Heat Sink (i.e., Ocean)

Service Water Pumps 3SWP*P1A, B, C, D CWPH Operability Provide Cooling Water to Safety Related Components Containment Recirculation Coolers Service Water ESF Operability - Provide Cooling Water for Containment 3SWP*MOV54A, B, C, D and 3SWP*MOV57A, B, C, Recirculation Coolers 3RSS*E1A, E1B, E1C, and D E1D 3.6-70 Rev. 30

MPS3 UFSAR TABLE 3.6-5 ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)

SYSTEM BUILDING REQUIRED SAFETY FUNCTION Reactor Plant Component Cooling Water Heat Auxiliary Operability - Normally Open Providing Cooling Water Exchanger Service Water 3SWP*MOV50A, and for CCP and Close Subsequent to a LOCA or a HELB MOV50B Inside Containment Service Water Valves 3SWP*AOV39A and EGE Operability - Valves Open to Supply Cooling Water to 3SWP*AOV39B Diesel Generators Emergency Diesel Generator Fuel Oil (EGF) (1) (2) EGE Operation of the Diesel Generators Emergency Diesel Generator Fuel Oil (EGF) Storage EGE Operability - Supports Diesel Generator Function Motor Driven Fuel Pump 3EGF*P2A, and 3EGF*P2B Engine Driven Fuel Pump 3EGF*P3A, and 3EGF*P3B EGE Operability - Supports Diesel Generator Function Fuel Oil Storage Tanks 3EGF*TK1A, and 3EGF*TK1B Fuel Oil Vault Pressure Boundary - Supports Diesel Generator Function MPS3 UFSAR Fuel Oil Day Tank 3EGF*TK2A, and 3EGF*TK2B EGE Pressure Boundary - Supports Diesel Generator Function Fuel Oil Transfer Pumps 3EGF*P1A,1B,1C, and 1D EGE Operability - Supports Diesel Generator Function Emergency Diesel Generator Cooling Water EGE Operator of the Diesel Generators (EGS) (1) (2)

Emergency Diesel Generator Air Cooler Water Heat EGE Operability - Supports Diesel Generator Function Exchanger 3EGS*E1A, and 3EGS*E1B Emergency Diesel Generator Jacket Water Electric EGE Pressure Boundary - Supports Diesel Generator Heaters 3EGS*H1A, and 3EGS*H1B Function Emergency Diesel Generator Jacket Water Circulating EGE Pressure Boundary - Supports Diesel Generator Water Pump 3EGS*P2A, and 3EGS*P2B Function 3.6-71 Rev. 30

MPS3 UFSAR TABLE 3.6-5 ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)

SYSTEM BUILDING REQUIRED SAFETY FUNCTION Emergency Diesel Generator Fresh Water Expansion EGE Pressure Boundary - Supports Diesel Generator Tank 3EGS*TK1A, and 3EGS*TK1B Function Emergency Diesel Generator Jacket Water Cooler EGE Operability - Supports Diesel Generator Function 3EGS*E2A, and 3EGS*E2B Emergency Diesel Generator Cooling Water (EGS) Lube EGE Operability - Supports Diesel Generator Function Oil Heat Exchanger 3EGS*E3A, and 3EGS*E3B Emergency Diesel Generator Cooling Water (EGS) EGE Operability - Supports Diesel Generator Function Governor Lube Oil Cooler 3EGS*E4A, and 3EGS*E4B Emergency Diesel Generator Cooling Water (EGS) EGE Operability - Supports Diesel Generator Function Engine Driven Pump 3EGS*P1A and 3EGS*P1B Emergency Diesel Generator Cooling Water (EGS) EGE Operability - Supports Diesel Generator Function Engine Driven Intercooler Water Pump 3EGS*P3A and MPS3 UFSAR 3EGS*P3B Emergency Diesel Generator Lube Oil (EGO) (1) (2) EGE Operation of the Diesel Generators Emergency Diesel Generator Lube Oil (EGO) Lube Oil EGE Operability - Supports Diesel Generator Function Pumps 3EGO*P3A, and 3EGO*P3B Lube Oil Pumps 3EGO*P1A, and 3EGO*P1B EGE Operability - Supports Diesel Generator Function Lube Oil Pumps 3EGO*P2A, and 3EGO*P2A, and EGE Operability - Supports Diesel Generator Function 3EGO*P2B Lube Oil Pumps 3EGO*P4A, and 3EGO*P4B EGE Operability - Supports Diesel Generator Function Emergency Diesel Generator Pre Lube Oil (EGO) Filter EGE Operability - Supports Diesel Generator Function 3EGO*FLT1A, and 3EGO*FLT1B 3.6-72 Rev. 30

MPS3 UFSAR TABLE 3.6-5 ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)

SYSTEM BUILDING REQUIRED SAFETY FUNCTION Emergency Diesel Generator Lube Oil (EGO) Strainers EGE Operability - Supports Diesel Generator Function 3EGO*STR1A,5A and 3EGO*STR1B,5B Emergency Diesel Generator Lube Oil (EGO) Lube Oil EGE Operability - Supports Diesel Generator Function Suction Strainer 3EGO*STR4A, and 3EGO*STR4B Emergency Diesel Generator Lube Oil (EGO) Pre Lube EGE Operability - Supports Diesel Generator Function Oil and Filter Pump Suction Strainer 3EGO*STR2A, and 2B Emergency Diesel Generator Lube Oil (EGO) Pre Lube EGE Operability - Supports Diesel Generator Function Oil Heater 3EGO*H1A, and 3EGO*H1B Emergency Diesel Generator Air Start (EGA) (2) EGE Operation of the Diesel Generators Emergency Diesel Generator Air Start Air Receiver EGE Pressure Boundary - Air Start Function Tanks 3EGA*TK1A, 3EGA*TK1B and 3EGA*TK2A, MPS3 UFSAR 3EGA*TK2B Emergency Diesel Generator Air Start (EGA) Air Tanks EGE Pressure Boundary - Air Start Function 3EGA*TK3A and 3EGA*TK3B Diesel Generator Exhaust & Combustion EGE Operation of the Diesel Generators (EGD) (1) (2)

Emergency Diesel Generator Exhaust & Combustion EGE Pressure Boundary - Supports Diesel Generator (EGD) Exhaust Silencer (i.e. Muffler) 3EGD*SIL3A, Function 3B Emergency Diesel Generator Exhaust & Combustion EGE Pressure Boundary - Supports Diesel Generator (EGD) Intake Air Filter and Silencer 3EGD*SIL1A, and Function 1B 3.6-73 Rev. 30

MPS3 UFSAR TABLE 3.6-5 ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)

SYSTEM BUILDING REQUIRED SAFETY FUNCTION Emergency Diesel Generator Exhaust & Combustion EGE Pressure Boundary - Supports Diesel Generator (EGD) Intake Air Silencer 3EGD*SIL2A, and Function 3EGD*SIL2B Post DBA Hydrogen Recombiner Control Valves Recombiner Operability - Containment Isolation 3HCS*V2, V3, and V6, and 3HCS*V9, V10, and V13 Containment Isolation (1) (2) Containment and Maintain Containment Integrity After an Accident Connecting Structures Containment Isolation Valves (4) Containment and Operability - Powered Valves Required to Isolate Connecting Containment Structures Heating, Ventilation, and Air Conditioning and Cooling Containment and Operability (5)

Connecting MPS3 UFSAR (HVAC) (1) (2)

Structures Instrument and Controls (I&C) (1) (2) Containment and Operability (6)

Connecting Structures Electrical Equipment (1) (2) Containment and Operability (7)

Connecting Structures Building Structures (1) Support and Protect Safety Related Systems Containment Structure Containment Secondary Barrier Providing Isolation of Reactor Coolant System and the Tertiary Barrier Between the Reactor Core and the Outside Atmosphere 3.6-74 Rev. 30

MPS3 UFSAR TABLE 3.6-5 ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)

SYSTEM BUILDING REQUIRED SAFETY FUNCTION Neutron Shield Tank 3NSS*TK2 Containment Primary Barrier Providing Isolation of Reactor Core Primary Shield Wall Containment Primary Barrier Providing Isolation of Reactor Core Reinforced Concrete Internal Substructure Containment Secondary Barrier Providing Isolation of Reactor Core Reinforced Concrete Internal Floors, Walls, and Containment Secondary Barrier Providing Isolation of Reactor Enclosures Coolant System Containment Structure Liner Containment Provide a Leak Tight Membrane Containment Penetrations Containment Maintain Pressure Boundary Integrity Auxiliary Building Supports Systems required for Safe Shutdown Engineered Safety Features Building Supports Systems required for ECCS and Safe Shutdown Main Steam Valve Building Supports Systems required for Safe Shutdown MPS3 UFSAR Emergency Diesel Generator Enclosures Supports Systems required for Emergency Power Control Building Supports Systems required for Safe Shutdown Circulating Water Pumphouse Supports Systems required for Safe Shutdown Containment Enclosure Building Supports Systems required for Post Accident Mitigation NOTES:

(1) Hazards Review Program addresses Safe Shutdown Requirements which are scenario dependent. Each scenario is evaluated on a case by case basis given an initiating event.

(2) Piping and Valves are required for Essential Pressure Boundary.

(3) The Hot Water Heating (HVH) System and the Auxiliary Steam (ASS) System are Non Safety Related Systems.

3.6-75 Rev. 30

MPS3 UFSAR (4) Containment Isolation Valves are delineated in FSAR Table 6.2-65.

(5) HVAC Systems required to support Systems and Components delineated in this Table and HVAC Systems required for Accident Mitigation and Safe Shutdown are described in FSAR Chapter 9.

(6) Instruments and Control (I&C) Equipment required to support Systems and Components delineated in this Table and I&C Equipment required for Accident Mitigation and Safe Shutdown are described in FSAR Chapter 7.

(7) Electrical Equipment required to support the Systems and Components delineated in this Table and the Electrical Systems required for Accident Mitigation and Safe Shutdown are described in FSAR Chapter 8.

MPS3 UFSAR 3.6-76 Rev. 30

MPS3 UFSAR TABLE 3.6-6 POSTULATED BREAKS MAIN STEAM SYSTEM LINE BREAK TOTAL ADDITIVE DESIGNATION # BUILDING ELEVATION BREAK TYPE (1) STRESS FIGURE 3.6-3-MSS-030-92-2 1 Containment 89'-1" CB N/A - Terminal End 8 3-MSS-030-92-2 2 Containment 76'-6" CB N/A - Arbitrary Intermediate 8 3-MSS-030-92-2 3 Containment 66'-3" CB N/A - Arbitrary Intermediate 8 3-MSS-030-104-3 17 MSVB 66'-3" CB N/A - Terminal End 8 3-MSS-030-25-4 4 Turbine 56'-9" CB N/A - Arbitrary Intermediate 8 3-MSS-030-25-4 5 Turbine 56'-5" CB & LS Above Threshold 8 3-MSS-030-25-4 6 Turbine 54'-8" CB N/A - Terminal End 8 3-MSS-030-67-4 7 Turbine 54'-6" CB (2) 8 3-MSS-030-67-4 8 Turbine 54'-6" CB & LS (2) 8 3-MSS-030-67-4 9 Turbine 54'-6" CB (2) 8 MPS3 UFSAR 28" G.E. Piping 10 Turbine 40'-0" CB (2) 8 28" G.E. Piping 11 Turbine 36'-0" CB & LS (2) 8 28" G.E. Piping 12 Turbine 32'-6" CB & LS (2) 8 28" G.E. Piping 13 Turbine 31'-10" CB & LS (2) 8 28" G.E. Piping 14 Turbine 35'-4" CB & LS (2) 8 28" G.E. Piping 15 Turbine 72'-11" CB & LS (2) 8 28" G.E. Piping 16 Turbine 76'-5" CB (2) 8 3-MSS-030-93-2 1 Containment 89'-11" CB N/A - Terminal End 8 3-MSS-030-93-2 2 Containment 93'-8" CB N/A - Arbitrary Intermediate 8 3-MSS-030-93-2 3 Containment 66'-3" CB N/A - Arbitrary Intermediate 8 3.6-77 Rev. 30

MPS3 UFSAR TABLE 3.6-6 POSTULATED BREAKS MAIN STEAM SYSTEM (CONTINUED)

LINE BREAK TOTAL ADDITIVE DESIGNATION # BUILDING ELEVATION BREAK TYPE (1) STRESS FIGURE 3.6-3-MSS-030-105-3 16 MSVB 66'-3" CB N/A - Terminal End 8 3-MSS-030-26-4 4 Turbine 56'-9" CB & LS Above Threshold 8 3-MSS-030-26-4 5 Turbine 56'-6" CB & LS Above Threshold 8 3-MSS-030-26-4 6 Turbine 55'-9" CB N/A - Terminal End 8 3-MSS-030-68-4 7 Turbine 54'-6" CB N/A - Terminal End 8 3-MSS-030-68-4 8 Turbine 54'-6" CB & LS (2) 8 3-MSS-030-68-4 9 Turbine 54'-6" CB N/A - Terminal End 8 28" G.E. Piping 10 Turbine 40'-0" CB (2) 8 28" G.E. Piping 11 Turbine 36'-0" CB & LS (2) 8 28" G.E. Piping 12 Turbine 32'-6" CB & LS (2) 8 MPS3 UFSAR 28" G.E. Piping 13 Turbine 31'-10" CB & LS (2) 8 28" G.E. Piping 14 Turbine 35'-4" CB & LS (2) 8 28" G.E. Piping 15 Turbine 63'-9" CB (2) 8 3-MSS-030-94-2 1 Containment 89'-8" CB N/A - Terminal End 8 3-MSS-030-94-2 2 Containment 93'-5" CB N/A - Arbitrary Intermediate 8 3-MSS-030-94-2 3 Containment 66'-3" CB N/A - Arbitrary Intermediate 8 3-MSS-030-106-3 18 MSVB 66'-3" CB N/A - Terminal End 8 3-MSS-030-59-4 4 Turbine 66'-3" CB N/A - Arbitrary Intermediate 8 3-MSS-030-59-4 5 Turbine 57'-0" CB N/A - Arbitrary Intermediate 8 3-MSS-030-59-4 6 Turbine 55'-7" CB N/A - Terminal End 8 3.6-78 Rev. 30

MPS3 UFSAR TABLE 3.6-6 POSTULATED BREAKS MAIN STEAM SYSTEM (CONTINUED)

LINE BREAK TOTAL ADDITIVE DESIGNATION # BUILDING ELEVATION BREAK TYPE (1) STRESS FIGURE 3.6-3-MSS-030-69-4 7 Turbine 54'-6" CB N/A - Terminal End 8 3-MSS-030-69-4 8 Turbine 54'-6" CB & LS (2) 8 3-MSS-030-69-4 9 Turbine 54'-6" CB N/A - Terminal End 8 28" G.E. Piping 10 Turbine 40'-0" CB (2) 8 28" G.E. Piping 11 Turbine 36'-0" CB & LS (2) 8 28" G.E. Piping 12 Turbine 32'-6" CB & LS (2) 8 28"-G.E. Piping 13 Turbine 31'-10" CB & LS (2) 8 28" G.E. Piping 14 Turbine 35'-4" CB & LS (2) 8 28" G.E. Piping 15 Turbine 63'-9" CB (2) 8 3-MSS-030-95-2 1 Containment 89'-9" CB N/A - Terminal End 8 MPS3 UFSAR 3-MSS-030-95-2 2 Containment 74'-4" CB N/A - Arbitrary Intermediate 8 3-MSS-030-95-2 3 Containment 66'-3" CB N/A - Arbitrary Intermediate 8 3-MSS-030-107-3 19 MSVB 66'-3" CB N/A - Terminal End 8 3-MSS-030-60-4 4 Turbine 65'-8" CB N/A - Arbitrary Intermediate 8 3-MSS-030-60-4 5 Turbine 56'-8" CB & LS Above Threshold 8 3-MSS-030-60-4 6 Turbine 55'-6" CB N/A - Terminal End 8 3-MSS-030-70-4 7 Turbine 54'-6" CB N/A - Terminal End 8 3-MSS-030-70-4 8 Turbine 54'-6" CB & LS (2) 8 3-MSS-030-70-4 9 Turbine 54'-6" CB N/A - Terminal End 8 28" G.E. Piping 10 Turbine 40'-0" CB (2) 8 3.6-79 Rev. 30

MPS3 UFSAR TABLE 3.6-6 POSTULATED BREAKS MAIN STEAM SYSTEM (CONTINUED)

LINE BREAK TOTAL ADDITIVE DESIGNATION # BUILDING ELEVATION BREAK TYPE (1) STRESS FIGURE 3.6-28" G.E. Piping 11 Turbine 36'-0" CB & LS (2) 8 28" G.E. Piping 12 Turbine 32'-6" CB & LS (2) 8 28" G.E. Piping 13 Turbine 31'-10" CB & LS (2) 8 28" G.E. Piping 14 Turbine 35'-4" CB & LS (2) 8 28" G.E. Piping 15 Turbine 72'-11" CB & LS (2) 8 28" G.E. Piping 16 Turbine 76'-5" CB (2) 8 3-MSS-003-6-2 1 Containment 93'-6" CB N/A - Terminal End 8 3-MSS-003-6-2 2 Containment 86'-5" CB N/A - Arbitrary Intermediate 8 3-MSS-003-6-2 3 Containment 80'-5" CB N/A - Arbitrary Intermediate 8 3-MSS-003-6-2 4 Containment 31'-0" CB N/A - Terminal End 8 MPS3 UFSAR 3-MSS-003-4-2 1 Containment 66'-3" CB N/A - Terminal End 8 3-MSS-003-4-2 2 Containment 56'-7" CB Above Threshold 8 3-MSS-003-4-2 3 Containment 43'-4" CB N/A - Arbitrary Intermediate 8 3-MSS-003-4-2 4 Containment 31'-0" CB N/A - Terminal End 8 3-MSS-003-33-2 1 Containment 93'-6" CB N/A - Terminal End 8 3-MSS-003-33-2 2 Containment 80'-3" CB Above Threshold 8 3-MSS-003-33-2 3 Containment 71'-4" CB Above Threshold 8 3-MSS-003-33-2 4 Containment 62'-4" CB Above Threshold 8 3-MSS-003-33-2 5 Containment 42'-0" CB N/A - Arbitrary Intermediate 8 3-MSS-003-33-2 6 Containment 31'0" CB N/A - Terminal End 8 3.6-80 Rev. 30

MPS3 UFSAR NOTES:

(1) Circumferential Pipe Break (CB) and Longitudinal Pipe Split (LS) are defined in FSAR Section 3.6.2.1.3.

(2) Pipe breaks on the Main Steam System downstream of the Main Steam Manifold, 3-MSS-042, are postulated at terminal ends, fittings, valves, and any integral welded attachments. This criteria is also applicable to the Main Steam Turbine Bypass piping and the Main Steam Moisture Separator piping shown on FSAR Figure 3.6-9.

MPS3 UFSAR 3.6-81 Rev. 30

MPS3 UFSAR TABLE 3.6-7 PIPE WHIP EFFECTS MAIN STEAM SYSTEM LINE BLOWDOWN PRIMARY PROTECTION DESIGNATION BREAK # (1) SOURCE ESSENTIAL PIPE WHIP TARGET REQUIREMENT 3-MSS-030-92-2 1 3-MSS-042-63-4 Polar Crane 3MHR*CRN1 Support 3MSS-PRR7LA System 3-MSS-030-92-2 2 3RCS*SG1A None (2) 3MSS-PRR5LA 3-MSS-030-92-2 2 3-MSS-042-63-4 None (2) 3MSS-PRR2LA 3-MSS-030-92-2 3 3RCS*SG1A None (2) 3MSS-PRR3LA 3-MSS-030-92-2 3 3-MSS-042-63-4 None (2) 3MSS-PRR1LA 3-MSS-030-104-3 17 3RCS*SG1A None None 3-MSS-030-104-3 17 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR4A 3-MSS-030-25-4 4 3RCS*SG1A None (2) 3MSS-PRR2A & 3MSS-PRR3A 3-MSS-030-25-4 4 3-MSS-042-63-4 None (2) 3MSS-PRR4A & 3MSS-PRR5A MPS3 UFSAR 3-MSS-030-25-4 5 3RCS*SG1A Limit Stress to Break Exclusion Zone 3MSS-PRR2A & 3MSS-PRR3A 3-MSS-030-25-4 5 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR4A & 3MSS-PRR5A 3-MSS-030-25-4 Split (3) Limit Stress to Break Exclusion Zone 3MSS-PRR2A & 3MSS-PRR3A 3-MSS-030-25-4 6 3RCS*SG1A Control Building Boundary Wall 3MSS-PRR5A 3-MSS-030-67-4 7 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR7A & 3MSS-PRR13A 3-MSS-030-67-4 8 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR7A & 3MSS-PRR13A 3-MSS-030-67-4 9 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR13A 3-MSS-030-67-4 10 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR13A 3-MSS-030-67-4 Splits 3-MSS-042-63-4 None None 3.6-82 Rev. 30

MPS3 UFSAR TABLE 3.6-7 PIPE WHIP EFFECTS MAIN STEAM SYSTEM (CONTINUED)

LINE BLOWDOWN PRIMARY PROTECTION DESIGNATION BREAK # (1) SOURCE ESSENTIAL PIPE WHIP TARGET REQUIREMENT 28" G.E. Piping 11 3-MSS-042-63-4 None None 28" G.E. Piping 12 3-MSS-042-63-4 None None 28" G.E. Piping 13 3-MSS-042-63-4 None None 28" G.E. Piping 14 3-MSS-042-63-4 None None 28" G.E. Piping 15 3-MSS-042-63-4 None None 28" G.E. Piping 16 3-MSS-042-63-4 None None 28" G.E. Piping Splits 3-MSS-042-63-4 None None 3-MSS-030-93-2 1 3-MSS-042-63-4 Polar Crane 3MHR*CRN1 Support 3MSS-PRR7SB System 3-MSS-030-93-2 2 3RCS*SG1B None - (2) None MPS3 UFSAR 3-MSS-030-93-2 2 3-MSS-042-63-4 None - (2) 3MSS-PRR6SB 3-MSS-030-93-2 3 3RCS*SG1B None - (2) 3MSS-PRR5SB 3-MSS-030-93-2 3 3-MSS-042-63-4 None - (2) None 3-MSS-030-105-3 16 3RCS*SG1B None None 3-MSS-030-105-3 16 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR4B 3-MSS-030-26-4 4 3RCS*SG1B Limit Stress to Break Exclusion Zone 3MSS-PRR2B & 3MSS-PRR3B 3-MSS-030-26-4 4 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR4B 3-MSS-030-26-4 Split (3) Limit Stress to Break Exclusion Zone 3MSS-PRR2B & 3MSS-PRR3B 3-MSS-030-26-4 5 3RCS*SG1B Limit Stress to Break Exclusion Zone 3MSS-PRR2B & 3MSS-PRR3B 3.6-83 Rev. 30

MPS3 UFSAR TABLE 3.6-7 PIPE WHIP EFFECTS MAIN STEAM SYSTEM (CONTINUED)

LINE BLOWDOWN PRIMARY PROTECTION DESIGNATION BREAK # (1) SOURCE ESSENTIAL PIPE WHIP TARGET REQUIREMENT 3-MSS-030-26-4 5 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR5B 3-MSS-030-26-4 5 (3) Limit Stress to Break Exclusion Zone 3MSS-PRR2B & 3MSS-PRR3B 3-MSS-030-26-4 6 3RCS*SG1B Control Building Boundary Wall 3MSS-PRR5B 3-MSS-030-68-4 7 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR7B & 3MSS-PRR13B 3-MSS-030-68-4 8 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR7B & 3MSS-PRR13B 3-MSS-030-68-4 9 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR7B & 3MSS-PRR13B 3-MSS-030-68-4 10 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR13B 3-MSS-030-68-4 Splits 3-MSS-042-63-4 None None 28" G.E. Piping 11 3-MSS-042-63-4 None None 28" G.E. Piping 12 3-MSS-042-63-4 None None MPS3 UFSAR 28" G.E. Piping 13 3-MSS-042-63-4 None None 28" G.E. Piping 14 3-MSS-042-63-4 None None 28" G.E. Piping 15 3-MSS-042-63-4 None None 28" G.E. Piping Splits 3-MSS-042-63-4 None None 3-MSS-030-094-2 1 3-MSS-042-63-4 Polar Crane 3MHR*CRN1 Support 3MSS-PRR7SC System 3-MSS-030-094-2 2 3RCS*SG1C None (2) None 3-MSS-030-094-2 2 3-MSS-042-63-4 None (2) 3MSS-PRR6SC 3.6-84 Rev. 30

MPS3 UFSAR TABLE 3.6-7 PIPE WHIP EFFECTS MAIN STEAM SYSTEM (CONTINUED)

LINE BLOWDOWN PRIMARY PROTECTION DESIGNATION BREAK # (1) SOURCE ESSENTIAL PIPE WHIP TARGET REQUIREMENT 3-MSS-030-094-2 3 3RCS*SG1C None (2) 3MSS-PRR5SC 3-MSS-030-094-2 3 3-MSS-042-63-4 None (2) None 3-MSS-030-106-3 18 3RCS*SG1C None None 3-MSS-030-106-3 18 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR4C 3-MSS-030-59-4 4 3RCS*SG1C None (2) 3MSS-PRR1C & 3MSS-PRR3C 3-MSS-030-59-4 4 3-MSS-042-63-4 None (2) 3MSS-PRR4C 3-MSS-030-59-4 5 3RCS*SG1C None (2) 3MSS-PRR1C & 3MSS-PRR3C 3-MSS-030-59-4 5 3-MSS-042-63-4 None (2) 3MSS-PRR5C 3-MSS-030-59-4 6 3RCS*SG1C Control Building Boundary Wall 3MSS-PRR5C 3-MSS-030-69-4 7 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR7C & 3MSS-PRR13C MPS3 UFSAR 3-MSS-030-69-4 8 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR7C & 3MSS-PRR13C 3-MSS-030-69-4 9 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR7C & 3MSS-PRR13C 3-MSS-030-69-4 10 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR13C 3-MSS-030-69-4 Splits 3-MSS-042-63-4 None None 28" G.E. Piping 11 3-MSS-042-63-4 None None 28" G.E. Piping 12 3-MSS-042-63-4 None None 28" G.E. Piping 13 3-MSS-042-63-4 None None 28" G.E. Piping 14 3-MSS-042-63-4 None None 28" G.E. Piping 15 3-MSS-042-63-4 None None 3.6-85 Rev. 30

MPS3 UFSAR TABLE 3.6-7 PIPE WHIP EFFECTS MAIN STEAM SYSTEM (CONTINUED)

LINE BLOWDOWN PRIMARY PROTECTION DESIGNATION BREAK # (1) SOURCE ESSENTIAL PIPE WHIP TARGET REQUIREMENT 28" G.E. Piping Splits 3-MSS-042-63-4 None None 3-MSS-030-95-2 1 3-MSS-042-63-4 Polar Crane 3MHR*CRN1 Support 3MSS-PRR7LD System 3-MSS-030-95-2 2 3RCS*SG1D None (2) 3MSS-PRR5LD 3-MSS-030-95-2 2 3-MSS-042-63-4 None (2) 3MSS-PRR2LD 3-MSS-030-95-2 3 3RCS*SG1D None (2) 3MSS-PRR3LD 3-MSS-030-95-2 3 3-MSS-042-63-4 None (2) 3MSS-PRR1LD 3-MSS-030-107-3 19 3RCS*SG1D None None 3-MSS-030-107-3 19 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR4D 3-MSS-030-60-4 4 3RCS*SG1D None (2) 3MSS-PRR1D & 3MSS-PRR3D MPS3 UFSAR 3-MSS-030-60-4 4 3-MSS-042-63-4 None (2) 3MSS-PRR4D 3-MSS-030-60-4 5 3RCS*SG1D Limit Stress to Break Exclusion Zone 3MSS-PRR1D & 3MSS-PRR3D 3-MSS-030-60-4 5 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR5D 3-MSS-030-60-4 Split (3) Limit Stress to Break Exclusion Zone 3MSS-PRR1D & 3MSS-PRR3D 3-MSS-030-60-4 6 3RCS*SG1D Control Building Boundary Wall 3MSS-PRR5D 3-MSS-030-70-4 7 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR7D & 3MSS-PRR13D 3-MSS-030-70-4 8 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR7D & 3MSS-PRR13D 3-MSS-030-70-4 9 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR7D & 3MSS-PRR13D 3-MSS-030-70-4 10 3-MSS-042-63-4 Control Building Boundary Wall 3MSS-PRR13D 3.6-86 Rev. 30

MPS3 UFSAR TABLE 3.6-7 PIPE WHIP EFFECTS MAIN STEAM SYSTEM (CONTINUED)

LINE BLOWDOWN PRIMARY PROTECTION DESIGNATION BREAK # (1) SOURCE ESSENTIAL PIPE WHIP TARGET REQUIREMENT 3-MSS-030-70-4 Splits 3-MSS-042-63-4 None None 28" G.E. Piping 11 3-MSS-042-63-4 None None 28" G.E. Piping 12 3-MSS-042-63-4 None None 28" G.E. Piping 13 3-MSS-042-63-4 None None 28" G.E. Piping 14 3-MSS-042-63-4 None None 28" G.E. Piping 15 3-MSS-042-63-4 None None 28" G.E. Piping 16 3-MSS-042-63-4 None None 28" G.E. Piping Splits 3-MSS-042-63-4 None None 3-MSS-003-6-2 1 None (4) None None 3-MSS-003-6-2 2 3-MSS-030-92-2 None (2) None MPS3 UFSAR 3-MSS-003-6-2 2 None (4) None None 3-MSS-003-6-2 3 3-MSS-030-92-2 None (2) None 3-MSS-003-6-2 3 None (4) None None 3-MSS-003-6-2 4 3-MSS-030-92-2 None None 3-MSS-003-6-2 4 None (4) None None 3-MSS-003-4-2 1 None (4) None None 3-MSS-003-4-2 2 3-MSS-030-93-2 None None 3-MSS-003-4-2 2 None (4) None None 3-MSS-003-4-2 3 3-MSS-030-93-2 None (2) None 3.6-87 Rev. 30

MPS3 UFSAR TABLE 3.6-7 PIPE WHIP EFFECTS MAIN STEAM SYSTEM (CONTINUED)

LINE BLOWDOWN PRIMARY PROTECTION DESIGNATION BREAK # (1) SOURCE ESSENTIAL PIPE WHIP TARGET REQUIREMENT 3-MSS-003-4-2 3 None (4) None None 3-MSS-003-4-2 4 3-MSS-030-93-2 None None 3-MSS-003-4-2 4 None (4) None None 3-MSS-003-33-2 1 None (4) None None 3-MSS-003-33-2 2 3-MSS-030-95-2 None None 3-MSS-003-33-2 2 None (4) None None 3-MSS-003-33-2 3 3-MSS-030-95-2 None None 3-MSS-003-33-2 3 None (4) None None 3-MSS-003-33-2 4 3-MSS-030-95-2 None None 3-MSS-003-33-2 4 None (4) None None MPS3 UFSAR 3-MSS-003-33-2 5 3-MSS-030-95-2 None (2) None 3-MSS-003-33-2 5 None (4) None None 3-MSS-003-33-2 6 3-MSS-030-95-2 None None 3-MSS-003-33-2 6 None (4) None None NOTES:

(1) Repetition of Break Numbers is used to identify separate fluid reservoirs which provide a constant pressure source to maintain system pressure subsequent to the postulated pipe break. These reservoirs maintain system blowdown during the transient event as well as in the steady state.

(2) USNRC Generic Letter 87-11 eliminates the need to evaluate the pipe whip effects subsequent to Arbitrary Intermediate Breaks (AIB's). Therefore pipe whip effects from AIB's are not evaluated.

3.6-88 Rev. 30

MPS3 UFSAR (3) The Longitudinal Split 5 on 3-MSS-030-25-4 results in blowdown from both the Steam Generator 3RCS*SG1A and the Main Steam Header 3-MSS-042-63-4. The Longitudinal Split 5 on 3-MSS-030-25-4 does not result in pipe severance rather the split results in motion in and out of the plane of the pipe. This effect loads the pipe break exclusion zone but the load is much less severe than the circumferential break.

(4) The 3" Main Steam lines are used to run the Turbine Driven Auxiliary Feedwater Pump, 3FWA*P2, in the Engineered Safety Features Building. The limited capacity of the Turbine Driven Auxiliary Feedwater Pump does not sustain pipe whip and is not a contained fluid energy reservoir in accordance with FSAR Section 3.6.2.2.1 Item 3.

MPS3 UFSAR 3.6-89 Rev. 30

MPS3 UFSAR TABLE 3.6-8 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM ESSENTIAL JET LINE BREAK # IMPINGEMENT DISTANCE TO JET INTENSITY JET LOAD ON PROTECTION DESIGNATION (1) TARGET TARGET AT THE TARGET THE TARGET REQUIREMENT 3-MSS-030-92-2 1 None N/A N/A N/A None 3-MSS-030-92-2 2 Crane Wall 4 Feet 340 psi 1033 Kips None (2) 3-MSS-030-92-2 2 3RCS*SG1C 13 Feet 65 psi 564 Kips None (2) 3-MSS-030-92-2 3 None N/A N/A N/A None (3) 3-MSS-030-92-2 3 None N/A N/A N/A None (3) 3-MSS-030-104-3 17 None N/A N/A N/A None 3-MSS-030-104-3 17 None N/A N/A N/A None 3-MSS-030-25-4 4 None N/A N/A N/A None 3-MSS-030-25-4 4 None N/A N/A N/A None MPS3 UFSAR 3-MSS-030-25-4 5 None N/A N/A N/A None (4) 3-MSS-030-25-4 5 None N/A N/A N/A None (4) 3-MSS-030-25-4 Split None N/A N/A N/A None (4) 3-MSS-030-25-4 6 Control Building (5) None (6) 3-MSS-030-67-4 7 Control Building (5) None (6) 3-MSS-030-67-4 8 Control Building (5) None (6) 3-MSS-030-67-4 Split None N/A N/A N/A None (4) 3-MSS-030-67-4 9 None N/A N/A N/A None (4) 3.6-90 Rev. 30

MPS3 UFSAR TABLE 3.6-8 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM (CONTINUED)

ESSENTIAL JET LINE BREAK # IMPINGEMENT DISTANCE TO JET INTENSITY JET LOAD ON PROTECTION DESIGNATION (1) TARGET TARGET AT THE TARGET THE TARGET REQUIREMENT 28" G.E. Piping 10 None N/A N/A N/A None (4) 28" G.E. Piping 11 None N/A N/A N/A None (4) 28" G.E. Piping Split None N/A N/A N/A None (4) 28" G.E. Piping 12 None N/A N/A N/A None (4) 28" G.E. Piping Split None N/A N/A N/A None (4) 28" G.E. Piping 13 None N/A N/A N/A None (4) 28" G.E. Piping Split None N/A N/A N/A None (4) 28" G.E. Piping 14 None N/A N/A N/A None (4) 28" G.E. Piping Split None N/A N/A N/A None (4) MPS3 UFSAR 28" G.E. Piping 15 None N/A N/A N/A None (4) 28" G.E. Piping Split None N/A N/A N/A None (4) 28" G.E. Piping 16 None N/A N/A N/A None (4) 3-MSS-030-93-2 1 None N/A N/A N/A None 3-MSS-030-93-2 1 None N/A N/A N/A None 3-MSS-030-93-2 2&3 None N/A N/A N/A None (3) 3-MSS-030-105-3 16 None N/A N/A N/A None 3-MSS-030-105-3 16 None N/A N/A N/A None 3-MSS-030-26-4 4 None N/A N/A N/A None (4) 3.6-91 Rev. 30

MPS3 UFSAR TABLE 3.6-8 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM (CONTINUED)

ESSENTIAL JET LINE BREAK # IMPINGEMENT DISTANCE TO JET INTENSITY JET LOAD ON PROTECTION DESIGNATION (1) TARGET TARGET AT THE TARGET THE TARGET REQUIREMENT 3-MSS-030-26-4 4 None N/A N/A N/A None (4) 3-MSS-030-26-4 Split None N/A N/A N/A None (4) 3-MSS-030-26-4 5 None N/A N/A N/A None (4) 3-MSS-030-26-4 5 None N/A N/A N/A None (4) 3-MSS-030-26-4 Split None N/A N/A N/A None (4) 3-MSS-030-26-4 6 Control Building (5) None (6) 3-MSS-030-68-4 7 Control Building (5) None (6) 3-MSS-030-68-4 8 Control Building (5) None (6) 3-MSS-030-68-4 Split None N/A N/A N/A None (4) MPS3 UFSAR 3-MSS-030-68-4 9 None N/A N/A N/A None (4) 28" G.E. Piping 10 None N/A N/A N/A None (4) 28" G.E. Piping 11 None N/A N/A N/A None (4) 28" G.E. Piping Split None N/A N/A N/A None (4) 28" G.E. Piping 12 None N/A N/A N/A None (4) 28" G.E. Piping Split None N/A N/A N/A None (4) 28" G.E. Piping 13 None N/A N/A N/A None (4) 28" G.E. Piping Split None N/A N/A N/A None (4) 28" G.E. Piping 14 None N/A N/A N/A None (4) 3.6-92 Rev. 30

MPS3 UFSAR TABLE 3.6-8 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM (CONTINUED)

ESSENTIAL JET LINE BREAK # IMPINGEMENT DISTANCE TO JET INTENSITY JET LOAD ON PROTECTION DESIGNATION (1) TARGET TARGET AT THE TARGET THE TARGET REQUIREMENT 28" G.E. Piping Split None N/A N/A N/A None (4) 28" G.E. Piping 15 None N/A N/A N/A None (4) 3-MSS-030-94-2 1 None N/A N/A N/A None 3-MSS-030-94-2 2&3 None N/A N/A N/A None (3) 3-MSS-030-106-3 18 None N/A N/A N/A None 3-MSS-030-106-3 18 None N/A N/A N/A None 3-MSS-030-59-4 4 None N/A N/A N/A None 3-MSS-030-59-4 4 None N/A N/A N/A None 3-MSS-030-59-4 5 None N/A N/A N/A None MPS3 UFSAR 3-MSS-030-59-4 5 None N/A N/A N/A None 3-MSS-030-59-4 6 Control Building (5) None (6) 3-MSS-030-69-4 7 Control Building (5) None (6) 3-MSS-030-69-4 8 Control Building (5) None (6) 3-MSS-030-69-4 Split None N/A N/A N/A None (4) 3-MSS-030-69-4 9 Control Building (5) None (6) 28" G.E. Piping 10 None N/A N/A N/A None (4) 28" G.E. Piping 11 None N/A N/A N/A None (4) 28" G.E. Piping Split None N/A N/A N/A None (4) 3.6-93 Rev. 30

MPS3 UFSAR TABLE 3.6-8 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM (CONTINUED)

ESSENTIAL JET LINE BREAK # IMPINGEMENT DISTANCE TO JET INTENSITY JET LOAD ON PROTECTION DESIGNATION (1) TARGET TARGET AT THE TARGET THE TARGET REQUIREMENT 28" G.E. Piping 12 None N/A N/A N/A None (4) 28" G.E. Piping Split None N/A N/A N/A None (4) 28" G.E. Piping 13 None N/A N/A N/A None (4) 28" G.E. Piping Split None N/A N/A N/A None (4) 28" G.E. Piping 14 None N/A N/A N/A None (4) 28" G.E. Piping Split None N/A N/A N/A None (4) 28" G.E. Piping 15 None N/A N/A N/A None (4) 3-MSS-030-95-2 1 None N/A N/A N/A None 3-MSS-030-95-2 1 None N/A N/A N/A None MPS3 UFSAR 3-MSS-030-95-2 2 Crane Wall 4 Feet 340 psi 1033 Kips None (2) 3-MSS-030-95-2 2 3RCS*SG1C 13 Feet 65 psi 564 Kips None (2) 3-MSS-030-95-2 3 None N/A N/A N/A None (3) 3-MSS-030-107-3 19 None N/A N/A N/A None 3-MSS-030-107-3 19 None N/A N/A N/A None 3-MSS-030-60-4 4 None N/A N/A N/A None (4) 3-MSS-030-60-4 4 None N/A N/A N/A None (4) 3-MSS-030-60-4 5 None N/A N/A N/A None (4) 3-MSS-030-60-4 5 None N/A N/A N/A None (4) 3.6-94 Rev. 30

MPS3 UFSAR TABLE 3.6-8 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM (CONTINUED)

ESSENTIAL JET LINE BREAK # IMPINGEMENT DISTANCE TO JET INTENSITY JET LOAD ON PROTECTION DESIGNATION (1) TARGET TARGET AT THE TARGET THE TARGET REQUIREMENT 3-MSS-030-60-4 Split None N/A N/A N/A None (4) 3-MSS-030-60-4 6 Control Building 10 Feet 86 psi 843 Kips None (6) 3-MSS-030-70-4 7 Control Building 10 Feet 86 psi 843 Kips None (6) 3-MSS-030-70-4 8 Control Building 10 Feet 86 psi 843 Kips None (6) 3-MSS-030-70-4 Split None N/A N/A N/A None (4) 3-MSS-030-70-4 9 Control Building 10 Feet 86 psi 843 Kips None (6) 28" G.E. Piping 10 Control Building 10 Feet 86 psi 843 Kips None (6) 28" G.E. Piping 11 None N/A N/A N/A None (4) 28" G.E. Piping Split None N/A N/A N/A None (4) MPS3 UFSAR 28" G.E. Piping 12 None N/A N/A N/A None (4) 28" G.E. Piping Split None N/A N/A N/A None (4) 28" G.E. Piping 13 None N/A N/A N/A None (4) 28" G.E. Piping Split None N/A N/A N/A None (4) 28" G.E. Piping 14 None N/A N/A N/A None (4) 28" G.E. Piping Split None N/A N/A N/A None (4) 28" G.E. Piping 15 None N/A N/A N/A None (4) 28" G.E. Piping Split None N/A N/A N/A None (4) 28" G.E. Piping 16 None N/A N/A N/A None (4) 3.6-95 Rev. 30

MPS3 UFSAR TABLE 3.6-8 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM (CONTINUED)

ESSENTIAL JET LINE BREAK # IMPINGEMENT DISTANCE TO JET INTENSITY JET LOAD ON PROTECTION DESIGNATION (1) TARGET TARGET AT THE TARGET THE TARGET REQUIREMENT 3-MSS-003-6-2 1 Crane Wall (7) None 3-MSS-003-6-2 2&3 Crane Wall (7) None 3-MSS-003-6-2 4 Crane Wall (7) None 3-MSS-003-6-2 4 Crane Wall (7) None 3-MSS-003-4-2 1 Crane Wall (7) None 3-MSS-003-4-2 2&3 None N/A N/A N/A None 3-MSS-003-4-2 4 None N/A N/A N/A None 3-MSS-003-4-2 4 None N/A N/A N/A None 3-MSS-003-33-2 1 Crane Wall (7) None MPS3 UFSAR 3-MSS-003-33-2 2 None N/A N/A N/A None 3-MSS-003-33-2 3 None N/A N/A N/A None 3-MSS-003-33-2 4 None N/A N/A N/A None 3-MSS-003-33-2 5 None N/A N/A N/A None 3-MSS-003-33-2 6 None N/A N/A N/A None NOTES:

(1) Repetition of Break Numbers is used to identify fluid reservoirs which maintain system pressure subsequent of the postulated pipe break. The reservoirs sustain blowdown during the transient event and in the steady state.

3.6-96 Rev. 30

MPS3 UFSAR (2) Intermediate Break Number 2 is postulated at a location proximate to an Integral Welded Attachment (IWA) on Main Steam Lines 3-MSS-030-92-2 and 3-MSS-030-95-2. The Crane Wall remains structurally integral since this barrier is designed for larger loads over a smaller area at the pipe rupture restraint locations. The Steam Generation retain pressure boundary integrity and do not catastrophically fail as the upper support system is designed for much larger loads.

(3) USNRC Generic Letter 87-11 eliminates the need to evaluate jet impingement effects subsequent to Arbitrary Intermediate Breaks (AIB's). Therefore, jet impingement effects on the Main Steam System are not evaluated for AIB's.

(4) No essential systems are located in the Turbine Building. Jet impingement on an adjacent Main Steam pipe does not cause pressure boundary failure for the reasons provided in FSAR Section 3.6.2.1.

(5) Fluid jet impingement subsequent to this postulated break on this target is enveloped by Break Number 6 on 3-MSS-030-60-4 and Circumferential Breaks (CB) 7,8,9, and 10 on 3-MSS-030-70-4. The indicated break is partially shielded by adjacent Main Steam lines and is further from the Control Building than either 3-MSS-030-60-4 or 3-MSS-030-70-4.

(6) Fluid jet impingement on the Control Building Concrete Wall is not excessive based upon Generic Structural Evaluation of fluid jet impingement on concrete barriers described in NERM-069 Revision 1 Hazards Review Program Summary for Millstone Unit 3 Section 4.4.

(7) Fluid jet impingement subsequent to this postulated break on this target is enveloped by postulated break number 2 on either of the Main Steam System lines 3-MSS-030-92-2 or 3-MSS-030-95-2 loading the same target.

MPS3 UFSAR 3.6-97 Rev. 30

MPS3 UFSAR TABLE 3.6-9 POSTULATED BREAKS MAIN FEEDWATER SYSTEM LINE BREAK TYPE TOTAL ADDITIVE DESIGNATION BREAK # BUILDING ELEVATION (1) STRESS FIGURE 3.6-3-FWS-020-18-2 1 Containment 64'-7" CB N/A - Terminal End 10 3-FWS-020-18-2 2 Containment 64'-7" CB & LS Above Threshold 10 3-FWS-020-18-2 3 Containment 61'-0" CB N/A - Arbitrary 10 Intermediate 3-FWS-018-16-3 9 MSVB 44'-3" CB N/A - Terminal End 10 3-FWS-020-12-4 4 (2) Turbine 51'-0" CB N/A - Arbitrary 10 Intermediate 3-FWS-020-12-4 5 (2) Turbine 51'-0" CB N/A - Arbitrary 10 Intermediate 3-FWS-016-74-2 1 Containment 64'-7" CB N/A - Terminal End 10 3-FWS-016-74-2 2 Containment 64'-7" CB & LS Above Threshold 10 3-FWS-016-74-2 3 Containment 64'-7" CB & LS Above Threshold 10 MPS3 UFSAR 3-FWS-018-20-3 9 MSVB 44'-3" CB N/A - Terminal End 10 3-FWS-016-73-2 1 Containment 64'-7" CB N/A - Terminal End 10 3-FWS-016-73-2 2 Containment 64'-7" CB & LS Above Threshold 10 3-FWS-016-73-2 3 Containment 64'-7" CB & LS Above Threshold 10 3-FWS-018-24-3 9 MSVB 44'-3" CB N/A - Terminal End 10 3-FWS-020-30-2 1 Containment 64'-7" CB N/A - Terminal End 10 3-FWS-020-30-2 2 Containment 64'-7" CB & LS Above Threshold 10 3-FWS-020-30-2 3 Containment 60'-10" CB N/A - Arbitrary 10 Intermediate 3-FWS-018-28-3 9 MSVB 44'-3" CB N/A - Terminal End 10 3.6-98 Rev. 30

MPS3 UFSAR TABLE 3.6-9 POSTULATED BREAKS MAIN FEEDWATER SYSTEM (CONTINUED)

LINE BREAK TYPE TOTAL ADDITIVE DESIGNATION BREAK # BUILDING ELEVATION (1) STRESS FIGURE 3.6-3-FWS-020-15-4 6 (2) Turbine 51'-0" CB N/A - Arbitrary 10 Intermediate 3-FWS-020-15-4 7 (2) Turbine 51'-0" CB N/A - Arbitrary 10 Intermediate 3-FWS-036-11-4 8 (2) Turbine 51'-0" CB N/A Terminal End 10 NOTES:

(1) Circumferential Pipe Break (CB) and Longitudinal Pipe Split (LS) are defined in FSAR Section 3.6.2.1.3.

(2) The Main Feedwater piping in the Turbine Building is nonnuclear but is analyzed as part of the Main Feedwater piping in the Main Steam Valve Building (MSVB). The Main Feedwater piping in the Main Steam Valve Building is ASME Class 2 and 3. The criteria specified in FSAR Section 3.6.2.1.2.3.2.b is applied to the Main Feedwater piping in the Turbine Building. The approach results in fewer pipe break locations than if a fitting criteria is applied.

MPS3 UFSAR 3.6-99 Rev. 30

MPS3 UFSAR TABLE 3.6-10 PIPE WHIP EFFECTS - MAIN FEEDWATER SYSTEM LINE ESSENTIAL PIPE PROTECTION DESIGNATION BREAK# (1) BLOWDOWN SOURCE WHIP TARGET REQUIREMENT 3-FWS-020-18-2 1 3-FWS-036-11-4 Steam Generator Cubicle 3FWS-PRR12LA &

Wall 3FWS-PRR13LA 3-FWS-020-18-2 2 3RCS*SG1A None None 3-FWS-020-18-2 2 3-FWS-036-11-4 Operating Floor at 3FWS-PRR5LA (2)

Elevation 51'-4" 3-FWS-020-18-2 Split 3RCS*SG1A None None 3-FWS-020-18-2 3 3RCS*SG1A None (3) 3FWS-PRR5LA 3-FWS-020-18-2 3 3-FWS-036-11-4 None (3) 3FWS-PRR3LA 3-FWS-018-16-3 9 3RCS*SG1A None None MPS3 UFSAR 3-FWS-018-16-3 9 3-FWS-036-11-4 None None 3-FWS-020-12-4 4 3RCS*SG1A None (3) None 3-FWS-020-12-4 4 3-FWS-036-11-4 None (3) None 3-FWS-020-12-4 5 3RCS*SG1A None (3) None 3-FWS-020-12-4 5 3-FWS-036-11-4 None (3) None 3FWS-016-74-2 1 3-FWS-036-11-4 Steam Generator Cubicle 3FWS-PRR8SB Wall 3-FWS-016-74-2 2 3RCS*SG1B 3RCS*SG1B 3FWS-PRR5SB 3-FWS-016-74-2 2 3-FWS-036-11-4 None 3FWS-PRR8SB 3.6-100 Rev. 30

MPS3 UFSAR TABLE 3.6-10 PIPE WHIP EFFECTS - MAIN FEEDWATER SYSTEM (CONTINUED)

LINE ESSENTIAL PIPE PROTECTION (1)

DESIGNATION BREAK# BLOWDOWN SOURCE WHIP TARGET REQUIREMENT 3-FWS-016-74-2 Split 3RCS*SG1B None None 3-FWS-016-74-2 3 3-FWS-036-11-4 None 3FWS-PRR8SB 3-FWS-016-74-2 3 3RCS*SG1B None 3-FWS-PRR5SB 3-FWS-016-74-2 Split 3RCS*SG1B None None 3-FWS-018-20-3 9 3RCS*SG1B None None 3-FWS-018-20-3 9 3-FWS-036-11-4 None None 3-FWS-016-73-2 1 3-FWS-036-11-4 Steam Generator Cubicle 3FWS-PRR8SC Wall 3-FWS-016-73-2 2 3RCS*SG1C 3RCS*SG1C 3FWS-PRR5SC 3-FWS-016-73-2 2 3-FWS-036-11-4 None 3FWS-PRR8SC MPS3 UFSAR 3-FWS-016-73-2 Split 3RCS*SG1C None None 3-FWS-016-73-2 3 3-FWS-036-11-4 None 3FWS-PRR8SC 3-FWS-016-73-2 3 3RCS*SG1C None 3FWS-PRR5SC 3-FWS-016-73-2 Split 3RCS*SG1C None None 3-FWS-018-24-3 9 3RCS*SG1C None None 3-FWS-018-24-3 9 3-FWS-036-11-4 None None 3-FWS-020-30-2 1 3-FWS-036-11-4 Steam Generator Cubicle 3FWS-PRR12LD &

Wall 3FWS-PRR13LD 3-FWS-020-30-2 2 3RCS*SG1D None None 3.6-101 Rev. 30

MPS3 UFSAR TABLE 3.6-10 PIPE WHIP EFFECTS - MAIN FEEDWATER SYSTEM (CONTINUED)

LINE ESSENTIAL PIPE PROTECTION (1)

DESIGNATION BREAK# BLOWDOWN SOURCE WHIP TARGET REQUIREMENT 3-FWS-020-30-2 2 3-FWS-036-11-4 Operating Floor at 3FWS-PRR5LD (4)

Elevation 51'-4" 3-FWS-020-30-2 Split 3RCS*SG1D None None 3-FWS-020-30-2 3 3-FWS-036-11-4 None (3) None 3-FWS-020-30-2 3 3RCS*SG1D None (3) None 3-FWS-018-28-3 9 3RCS*SG1D None None 3-FWS-018-28-3 9 3-FWS-036-11-4 None None 3-FWS-020-15-4 6 3-FWS-036-11-4 None (3) None 3-FWS-020-15-4 6 3RCS*SG1D None (3) None MPS3 UFSAR 3-FWS-020-15-4 7 3-FWS-036-11-4 None (3) None 3-FWS-020-15-4 7 3RCS*SG1D None (3) None 3-FWS-036-11-4 8 3RCS*SG1A & 1B None None 3-FWS-036-11-4 8 3RCS*SG1C & 1D None None NOTES:

(1) Repetition of Break Numbers is used to identify separate fluid reservoirs which provide a constant pressure source to maintain system pressure subsequent to the postulated pipe break. These reservoirs maintain system blowdown during the transient event as well as in the steady state.

3.6-102 Rev. 30

MPS3 UFSAR (2) The Main Feedwater piping in the Turbine Building is nonnuclear but is analyzed as part of the Main Feedwater piping in the Main Steam Valve Building (MSVB). The Main Feedwater piping in the Main Steam Valve Building is ASME Class 2 and 3. The criteria specified in FSAR Section 3.6.2.1.2.3.2.b is applied to the Main Feedwater piping in the Turbine Building. The approach results in fewer pipe break locations than if a fitting criteria is applied.

(3) USNRC Generic Letter 87-11 eliminates the requirement to evaluate pipe whip effects subsequent to Arbitrary Intermediate Breaks (AIB's). Therefore, pipe whip effects subsequent to AIB's on the Main Feedwater System are not evaluated.

(4) The Intermediate Break Number 2 is postulated to occur at any location along the reducing elbow. A Circumferential Break (CB) and a Longitudinal Split (LS) are postulated but not concurrently. The Circumferential Break (CB) results in an impact of the Main Feedwater Line 3-FWS-020-30-2 to the Operating Floor Slab at Elevation 51'-4". Unrestrained impact would result in catastrophic failure of the Operating Floor and result in secondary missile ejection into the Reactor Coolant System cubicles. This unrestrained impact to the Operating Floor is prevented by the rupture restraint.

MPS3 UFSAR 3.6-103 Rev. 30

MPS3 UFSAR TABLE 3.6-11 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM ESSENTIAL JET BREAK IMPINGEMENT DISTANCE TO JET INTENSITY AT JET LOAD ON PROTECTION LINE DESIGNATION # (1) TARGET TARGET THE TARGET THE TARGET REQUIREMENT 3-FWS-020-18-2 1 None N/A N/A N/A None 3-FWS-020-18-2 2 None N/A N/A N/A None 3-FWS-020-18-2 2 None N/A N/A N/A None 3-FWS-020-18-2 Split None N/A N/A N/A None 3-FWS-020-18-2 3 None (2) N/A N/A N/A None 3-FWS-020-18-2 3 None (2) N/A N/A N/A None 3-FWS-018-16-3 9 None N/A N/A N/A None 3-FWS-018-16-3 9 None N/A N/A N/A None 3-FWS-020-12-4 4 None (2) N/A N/A N/A None 3-FWS-020-12-4 4 N/A N/A N/A None MPS3 UFSAR None (2) 3-FWS-020-12-4 5 None (2) N/A N/A N/A None 3-FWS-020-12-4 5 None (2) N/A N/A N/A None 3-FWS-016-74-2 1 None N/A N/A N/A None 3-FWS-016-74-2 2 None N/A N/A N/A None 3-FWS-016-74-2 2 None N/A N/A N/A None 3-FWS-016-74-2 Split None N/A N/A N/A None 3-FWS-016-74-2 3 None N/A N/A N/A None 3-FWS-016-74-2 3 None N/A N/A N/A None 3-FWS-016-74-2 Split None N/A N/A N/A None 3.6-104 Rev. 30

MPS3 UFSAR TABLE 3.6-11 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM (CONTINUED)

ESSENTIAL JET BREAK IMPINGEMENT DISTANCE TO JET INTENSITY AT JET LOAD ON PROTECTION LINE DESIGNATION # (1) TARGET TARGET THE TARGET THE TARGET REQUIREMENT 3-FWS-018-20-3 9 None N/A N/A N/A None 3-FWS-018-20-3 9 None N/A N/A N/A None 3-FWS-016-73-2 1 None N/A N/A N/A None 3-FWS-016-73-2 2 None N/A N/A N/A None 3-FWS-016-73-2 2 None N/A N/A N/A None 3-FWS-016-73-2 Split None N/A N/A N/A None 3-FWS-016-73-2 3 None N/A N/A N/A None 3-FWS-016-73-2 3 None N/A N/A N/A None 3-FWS-016-73-2 Split None N/A N/A N/A None 3-FWS-018-24-3 9 None N/A N/A N/A None MPS3 UFSAR 3-FWS-018-24-3 9 None N/A N/A N/A None 3-FWS-020-30-2 1 None N/A N/A N/A None 3-FWS-020-30-2 2 None N/A N/A N/A None 3-FWS-020-30-2 2 None N/A N/A N/A None 3-FWS-020-30-2 Split None N/A N/A N/A None 3-FWS-020-30-2 3 None (2) N/A N/A N/A None 3-FWS-020-30-2 3 None (2) N/A N/A N/A None 3-FWS-018-28-3 9 None N/A N/A N/A None 3-FWS-018-28-3 9 None N/A N/A N/A None 3-FWS-020-15-4 6 None (2) N/A N/A N/A None 3.6-105 Rev. 30

MPS3 UFSAR TABLE 3.6-11 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM (CONTINUED)

ESSENTIAL JET BREAK IMPINGEMENT DISTANCE TO JET INTENSITY AT JET LOAD ON PROTECTION LINE DESIGNATION # (1) TARGET TARGET THE TARGET THE TARGET REQUIREMENT 3-FWS-020-15-4 6 None (2) N/A N/A N/A None 3-FWS-020-15-4 7 None (2) N/A N/A N/A None 3-FWS-020-15-4 7 None (2) N/A N/A N/A None 3-FWS-036-11-4 8 None N/A N/A N/A None 3-FWS-036-11-4 8 None N/A N/A N/A None NOTES:

(1) Repetition of Break Numbers is used to identify separate fluid reservoirs which provide a constant pressure source to maintain system pressure subsequent to the postulated pipe break. These reservoirs maintain system blowdown during the transient event as well as in the steady state.

(2) USNRC Generic Letter 87-11 eliminates the requirement to evaluate the jet impingement effects subsequent to Arbitrary Intermediate Break (AIB's). Therefore jet impingement effects subsequent to AIB's are not evaluated.

MPS3 UFSAR 3.6-106 Rev. 30

MPS3 UFSAR TABLE 3.6-12 POSTULATED BREAKS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP FILL PIPING, LOOP DRAIN PIPING. LETDOWN LINE, AND NORMAL CHARGING LINE THRESHOLD DESIGNATION BREAK # BUILDING ELEVATION BREAK TYPE (1) STRESS USAGE FACTOR FIGURE EQN 12 OR EQN 10 13 3-RCS-008-24-1 13 Containment 23'-0" CB & LS Above Threshold 3.6-12 3-RCS-008-25-1 14 Containment 31'-6" CB & LS Above Threshold 3.6-12 3-RCS-150-27-1 13 Containment 23'-0" CB Terminal End 3.6-12 3-RCS-150-27-1 15 Containment 23'-0" CB N/A - Arbitrary Intermediate 3.6-12 3-RCS-150-27-1 16 Containment 22'-2" CB N/A - Arbitrary Intermediate 3.6-12 3-RCS-150-27-1 17 Containment 17'-6" CB Terminal End 3.6-12 3-RCS-008-29-1 13 Containment 23'-0" CB & LS Above Threshold 3.6-12 MPS3 UFSAR 3-RCS-008-30-1 14 Containment 31'-6" CB & LS Above Threshold 3.6-12 3-RCS-150-32-1 13 Containment 23'-0" CB Terminal End 3.6-12 3-RCS-150-32-1 15 Containment 23'-0" CB N/A - Arbitrary Intermediate 3.6-12 3-RCS-150-32-1 16 Containment 22'-2" CB N/A - Arbitrary Intermediate 3.6-12 3-RCS-150-32-1 17 Containment 17'-6" CB Terminal End 3.6-12 3-RCS-008-34-1 13 Containment 23'-0" CB & LS Above Threshold 3.6-12 3-RCS-008-35-1 14 Containment 31'-6" CB & LS Above Threshold 3.6-12 3-RCS-150-37-1 13 Containment 23'-0" CB Terminal End 3.6-12 3.6-107 Rev. 30

MPS3 UFSAR TABLE 3.6-12 POSTULATED BREAKS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP FILL PIPING, LOOP DRAIN PIPING. LETDOWN LINE, AND NORMAL CHARGING (CONTINUED)

LINE THRESHOLD DESIGNATION BREAK # BUILDING ELEVATION BREAK TYPE (1) STRESS USAGE FACTOR FIGURE EQN 12 OR EQN 10 13 3-RCS-150-37-1 15 Containment 23'-0" CB N/A - Arbitrary Intermediate 3.6-12 3-RCS-150-37-7 16 Containment 22'-2" CB N/A - Arbitrary Intermediate 3.6-12 3-RCS-150-37-1 17 Containment 17'-6" CB Terminal End 3.6-12 3-RCS-008-39-1 13 Containment 23'-0" CB & LS Above Threshold 3.6-12 3-RCS-008-40-1 14 Containment 31'-6" CB & LS Above Threshold 3.6-12 3-RCS-150-42-1 13 Containment 23'-0" CB Terminal End 3.6-12 3-RCS-150-42-1 15 Containment 23'-0" CB N/A - Arbitrary Intermediate 3.6-12 MPS3 UFSAR 3-RCS-150-42-1 16 Containment 22'-2" CB N/A - Arbitrary Intermediate 3.6-12 3-RCS-150-42-1 17 Containment 17'-6" CB Terminal End 3.6-12 3-RCS-002-127-1 TP Containment 5'-5" CB N/A (2) 3.6-33 3-RCS-002-127-1 IP Containment 4'-9" CB N/A (2) 3.6-33 3-RCS-002-127-1 TP Containment 4'-9" CB N/A (2) 3.6-33 3-RCS-002-130-1 TP Containment 5'-5" CB N/A (2) 3.6-33 3-RCS-002-130-1 IP Containment 4'-9" CB N/A (2) 3.6-33 3-RCS-002-130-1 TP Containment 4'-9" CB N/A (2) 3.6-33 3.6-108 Rev. 30

MPS3 UFSAR TABLE 3.6-12 POSTULATED BREAKS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP FILL PIPING, LOOP DRAIN PIPING. LETDOWN LINE, AND NORMAL CHARGING (CONTINUED)

LINE THRESHOLD DESIGNATION BREAK # BUILDING ELEVATION BREAK TYPE (1) STRESS USAGE FACTOR FIGURE EQN 12 OR EQN 10 13 3-RCS-002-135-1 TP Containment 5'-5" CB N/A (2) 3.6-33 3-RCS-002-135-1 IP Containment 4'-9" CB N/A (2) 3.6-33 3-RCS-002-135-1 TP Containment 4'-9" CB N/A (2) 3.6-33 3-RCS-002-143-1 TP Containment 5'-5" CB N/A (2) 3.6-33 3-RCS-002-143-1 IP Containment 4'-9" CB N/A (2) 3.6-33 3-RCS-002-143-1 TP Containment 4'-9" CB N/A (2) 3.6-33 MPS3 UFSAR 3-RCS-002-126-1 TP Containment 15'-10" CB N/A (3) 3.6-33 3-RCS-002-126-1 IP Containment 15'-2" CB N/A (3) 3.6-33 3-RCS-002-126-1 TP Containment 10'-3" CB N/A (3) 3.6-33 3-RCS-002-129-1 TP Containment 15'-10" CB N/A (3) 3.6-33 3-RCS-002-129-1 IP Containment 15'-2" CB N/A (3) 3.6-33 3-RCS-002-129-1 TP Containment 10'-3" CB N/A (3) 3.6-33 3-RCS-002-134-1 TP Containment 15'-10" CB N/A (3) 3.6-33 3.6-109 Rev. 30

MPS3 UFSAR TABLE 3.6-12 POSTULATED BREAKS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP FILL PIPING, LOOP DRAIN PIPING. LETDOWN LINE, AND NORMAL CHARGING (CONTINUED)

LINE THRESHOLD DESIGNATION BREAK # BUILDING ELEVATION BREAK TYPE (1) STRESS USAGE FACTOR FIGURE EQN 12 OR EQN 10 13 3-RCS-002-134-1 IP Containment 15'-2" CB N/A (3) 3.6-33 3-RCS-002-134-1 TP Containment 10'-3" CB N/A (3) 3.6-33 3-RCS-002-142-1 TP Containment 15'-10" CB N/A (3) 3.6-33 3-RCS-002-142-1 IP Containment 15'-2" CB N/A (3) 3.6-33 3-RCS-002-142-1 TP Containment 10'-3" CB N/A (3) 3.6-33 3-RCS-002-128-1 TP Containment 8'-10" CB N/A (4) 3.6-33 MPS3 UFSAR 3-RCS-002-128-1 IP Containment 11'-7" CB N/A (4) 3.6-33 3-RCS-002-128-1 TP Containment 11'-7" CB N/A (4) 3.6-33 3-RCS-002-131-1 TP Containment 8'-10" CB N/A (4) 3.6-33 3-RCS-002-131-1 IP Containment 11'-7" CB N/A (4) 3.6-33 3-RCS-002-131-1 TP Containment 11'-7" CB N/A (4) 3.6-33 3-RCS-002-136-1 TP Containment 8'-10" CB N/A (4) 3.6-33 3-RCS-002-136-1 IP Containment 11'-7" CB N/A (4) 3.6-33 3.6-110 Rev. 30

MPS3 UFSAR TABLE 3.6-12 POSTULATED BREAKS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP FILL PIPING, LOOP DRAIN PIPING. LETDOWN LINE, AND NORMAL CHARGING (CONTINUED)

LINE THRESHOLD DESIGNATION BREAK # BUILDING ELEVATION BREAK TYPE (1) STRESS USAGE FACTOR FIGURE EQN 12 OR EQN 10 13 3-RCS-002-136-1 TP Containment 11'-7" CB N/A (4) 3.6-33 3-RCS-002-144-1 TP Containment 8'-10" CB N/A (4) 3.6-33 3-RCS-002-144-1 IP Containment 11'-7" CB N/A (4) 3.6-33 3-RCS-002-144-1 TP Containment 11'-7" CB N/A (4) 3.6-33 3-RCS-003-137-1 1 Containment 15'-9" CB N/A Terminal End 3.6-16 3-RCS-003-137-1 2 Containment 10'-1" CB Above Threshold 3.6-16 MPS3 UFSAR 3-RCS-003-137-1 3 Containment 8'-7" CB Above Threshold 3.6-16 3-RCS-003-137-1 4 Containment 7'-0" CB Above Threshold 3.6-16 3-RCS-003-137-1 5 Containment 7'-0" CB Above Threshold 3.6-16 3-RCS-003-171-1 6 Containment 7'-0" CB Above Threshold 3.6-16 3-RCS-003-171-1 7 Containment 7'-0" CB Above Threshold 3.6-16 3-RCS-003-171-1 8 Containment 7'-0" CB Above Threshold 3.6-16 3-RCS-003-145-1 1 Containment 17'-6" CB N/A Terminal End 3.6-15 3-RCS-003-145-1 2 Containment 17'-6" CB Above Threshold 3.6-15 3-RCS-003-145-1 3 Containment 17'-6" CB Above Threshold 3.6-15 3.6-111 Rev. 30

MPS3 UFSAR TABLE 3.6-12 POSTULATED BREAKS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP FILL PIPING, LOOP DRAIN PIPING. LETDOWN LINE, AND NORMAL CHARGING (CONTINUED)

LINE THRESHOLD DESIGNATION BREAK # BUILDING ELEVATION BREAK TYPE (1) STRESS USAGE FACTOR FIGURE EQN 12 OR EQN 10 13 3-RCS-003-145-1 4 Containment 17'-6" CB Above Threshold 3.6-15 3-RCS-003-145-1 5 Containment 17'-2" CB Above Threshold 3.6-15 3-RCS-003-145-1 6 Containment 16'-7" CB Above Threshold 3.6-15 3-RCS-003-145-1 7 Containment 8'-0" CB Above Threshold 3.6-15 3-RCS-003-145-1 8 Containment 8'-0" CB Above Threshold 3.6-15 3-RCS-003-145-1 9 Containment 8'-0" CB Above Threshold 3.6-15 3-RCS-003-149-1 11 Containment 17'-6" CB N/A Terminal End 3.6-15 MPS3 UFSAR 3-RCS-003-149-1 12 Containment 17'-6" CB Above Threshold 3.6-15 3-RCS-003-149-1 13 Containment 17'-6" CB Above Threshold 3.6-15 3-RCS-003-149-1 14 Containment 17'-0" CB Above Threshold 3.6-15 3-RCS-003-149-1 15 Containment 16'-7" CB Above Threshold 3.6-15 3-RCS-003-149-1 16 Containment 16'-2" CB Above Threshold 3.6-15 3-RCS-003-149-1 17 Containment 7'-1" CB Above Threshold 3.6-15 3-RCS-003-149-1 18 Containment 7'-1" CB Above Threshold 3.6-15 3-RCS-003-149-1 19 Containment 7'-1" CB Above Threshold 3.6-15 3.6-112 Rev. 30

MPS3 UFSAR NOTES:

(1) Circumferential Pipe Break (CB) and Longitudinal Pipe Split (LS) are defined in FSAR Section 3.6.2.1.3.

(2) Breaks on the Reactor Coolant System Excess Letdown Piping off each Crossover Leg (TP) are postulated at any location (IP) up to the normally closed valve, 3RCS*AV8037A (TP). The portion of the Reactor Coolant System Excess Letdown Piping downstream of the normally closed valve, 3RCS*AV8037A is moderate energy based upon the 2% rule. The 2% rule is defined in FSAR Section 3.6.2.1.2 under Moderate Energy Piping Systems.

(3) Breaks on the Reactor Coolant System Drain Lines off each Crossover Leg (TP) are postulated at any location (IP) along the piping up to the normally closed valve, 3RCS*V203 (TP). The portion of the Reactor Coolant System Drain Lines off each Crossover Leg downstream of the normally closed valve, 3RCS*V203 is moderate energy based upon the 2% rule. The 2% rule is defined in FSAR Section 3.6.2.1.2 under Moderate Energy Piping Systems.

(4) Breaks on the Reactor Coolant System Loop Fill Piping to each Crossover Leg (TP) are postulated at any location (IP) along the Reactor Coolant System portion of this line and at the Terminal Ends (TP) at each Crossover Leg connection and at the normally closed valves, 3RCS*AV8036A, 3RCS*AV8036B, 3RCS*AV8036C, and 3RCS*AV8036D. The portion of the Loop Fill Piping upstream of the normally closed valves, 3RCS*AV8036A, 3RCS*AV8036B, 3RCS*AV8036C, and 3RCS*AV8036D is moderate energy under the 2% rule and is part of the Chemical & Volume Control System. The 2% rule is defined in FSAR Section 3.6.2.1.2 under Moderate Energy Piping Systems.

MPS3 UFSAR 3.6-113 Rev. 30

MPS3 UFSAR TABLE 3.6-13 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP FILL PIPING, LOOP DRAIN PIPING, LETDOWN AND NORMAL CHARGING LINE BLOWDOWN ESSENTIAL PIPE WHIP DESIGNATION BREAK # (1) SOURCE TARGET PRIMARY REQUIREMENT 3-RCS-008-24-1 13 3-RCS-275-4-1 None None 3-RCS-008-24-1 13 3-RCS-029-2-1 (2) 3RCS-PRRBA & 3RCS-PRRFA 3-RCS-008-24-1 Split Hot & Cold Legs (2) 3RCS-PRRCA & 3RCS-PRREA 3-RCS-008-25-1 14 3-RCS-275-4-1 (2) 3RCS-PRRA1A & 3RCS-PRRA2A 3-RCS-008-25-1 14 3-RCS-029-2-1 (2) 3RCS-PRR4A 3-RCS-008-25-1 Split Hot & Cold Legs (2) 3RCS-PRRCA & 3RCS-PRREA 3-RCS-150-27-1 13 Hot & Cold Legs None None 3-RCS-150-27-1 15 Hot & Cold Legs None None 3-RCS-150-27-1 16 Hot & Cold Legs None None MPS3 UFSAR 3-RCS-150-27-1 17 Hot & Cold Legs None None 3-RCS-008-29-1 13 3-RCS-275-9-1 None None 3-RCS-008-29-1 13 3-RCS-029-7-1 (2) 3RCS-PRRBB & 3RCS-PRRFB 3-RCS-008-29-1 Split Hot & Cold Legs (2) 3RCS-PRRCB & 3RCS-PRREB 3-RCS-008-30-1 14 3-RCS-275-9-1 (2) 3RCS-PRRA1B & 3RCS-PRRA2B 3-RCS-008-30-1 14 3-RCS-029-7-1 (2) 3RCS-PRR48 3-RCS-008-30-1 Split Hot & Cold Legs (2) 3RCS-PRRCB & 3RCS-PRREB 3-RCS-150-32-1 13 Hot & Cold Legs None None 3-RCS-150-32-1 15 Hot & Cold Legs None None 3.6-114 Rev. 30

MPS3 UFSAR TABLE 3.6-13 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP FILL PIPING, LOOP DRAIN PIPING, LETDOWN AND NORMAL CHARGING (CONTINUED)

LINE BLOWDOWN ESSENTIAL PIPE WHIP DESIGNATION BREAK # (1) SOURCE TARGET PRIMARY REQUIREMENT 3-RCS-150-32-1 16 Hot & Cold Legs None None 3-RCS-150-32-1 17 Hot & Cold Legs None None 3-RCS-008-34-1 13 3-RCS-275-14-1 None None 3-RCS-008-34-1 13 3-RCS-029-12-1 (2) 3-RCS-PRRBC & 3RCS-PRRFC 3-RCS-008-34-1 Split Hot & Cold Legs (2) 3RCS-PRRCC & 3RCS-PRREC 3-RCS-008-35-1 14 3-RCS-275-14-1 (2) 3RCS-PRRA1C & 3RCS-PRRA2C 3-RCS-008-35-1 14 3-RCS-029-12-1 (2) 3RCS-PRR4C 3-RCS-008-35-1 Split Hot & Cold Legs (2) 3-RCS-PRRCC & 3RCS-PRREC 3-RCS-150-37-1 13 Hot & Cold Legs None None MPS3 UFSAR 3-RCS-150-37-1 15 Hot & Cold Legs None None 3-RCS-150-37-1 16 Hot & Cold Legs None None 3-RCS-150-37-1 17 Hot & Cold Legs None None 3-RCS-008-39-1 13 3-RCS-275-19-1 None None 3-RCS-008-39-1 13 3-RCS-029-17-1 (2) 3RCS-PRRBD & 3RCS-PRRFD 3-RCS-008-39-1 Split Hot & Cold Legs (2) 3RCS-PRRCD & 3RCS-PRRED 3-RCS-008-40-1 14 3-RCS-275-19-1 (2) 3RCS-PRRA1D & 3RCS-PRRA2D 3-RCS-008-40-1 14 3-RCS-029-17-1 (2) 3RCS-PRR4D 3-RCS-008-40-1 Split Hot & Cold Legs (2) 3-RCS-PRRCD & 3RCS-PRRED 3-RCS-150-42-1 13 Hot & Cold Legs None None 3.6-115 Rev. 30

MPS3 UFSAR TABLE 3.6-13 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP FILL PIPING, LOOP DRAIN PIPING, LETDOWN AND NORMAL CHARGING (CONTINUED)

LINE BLOWDOWN ESSENTIAL PIPE WHIP DESIGNATION BREAK # (1) SOURCE TARGET PRIMARY REQUIREMENT 3-RCS-150-42-1 15 Hot & Cold Legs None None 3-RCS-150-42-1 16 Hot & Cold Legs None None 3-RCS-150-42-1 17 Hot & Cold Legs None None 3-RCS-002-127-1 TP 3-RCS-031-3-1 None None 3-RCS-002-127-1 IP 3-RCS-031-3-1 None None 3-RCS-002-127-1 TP 3-RCS-031-3-1 None None 3-RCS-002-130-1 TP 3-RCS-031-8-1 None None 3-RCS-002-130-1 IP 3-RCS-031-8-1 None None 3-RCS-002-130-1 TP 3-RCS-031-8-1 None None MPS3 UFSAR 3-RCS-002-135-1 TP 3-RCS-031-13-1 None None 3-RCS-002-135-1 IP 3-RCS-031-13-1 None None 3-RCS-002-135-1 TP 3-RCS-031-13-1 None None 3-RCS-002-143-1 TP 3-RCS-031-18-1 None None 3-RCS-002-143-1 IP 3-RCS-031-18-1 None None 3-RCS-002-143-1 TP 3-RCS-031-18-1 None None 3-RCS-002-126-1 TP 3-RCS-029-2-1 None None 3-RCS-002-126-1 IP 3-RCS-029-2-1 None None 3-RCS-002-126-1 TP 3-RCS-029-2-1 None None 3-RCS-002-129-1 TP 3-RCS-029-7-1 None None 3.6-116 Rev. 30

MPS3 UFSAR TABLE 3.6-13 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP FILL PIPING, LOOP DRAIN PIPING, LETDOWN AND NORMAL CHARGING (CONTINUED)

LINE BLOWDOWN ESSENTIAL PIPE WHIP DESIGNATION BREAK # (1) SOURCE TARGET PRIMARY REQUIREMENT 3-RCS-002-129-1 IP 3-RCS-029-7-1 None None 3-RCS-002-129-1 TP 3-RCS-029-7-1 None None 3-RCS-002-134-1 TP 3-RCS-029-12-1 None None 3-RCS-002-134-1 IP 3-RCS-029-12-1 None None 3-RCS-002-134-1 TP 3-RCS-029-12-1 None None 3-RCS-002-142-1 TP 3-RCS-029-17-1 None None 3-RCS-002-142-1 IP 3-RCS-029-17-1 None None 3-RCS-002-142-1 TP 3-RCS-029-17-1 None None 3-RCS-002-128-1 TP 3-RCS-031-3-1 None None MPS3 UFSAR 3-RCS-002-128-1 IP 3-RCS-031-3-1 None None 3-RCS-002-128-1 TP 3-RCS-031-3-1 None None 3-RCS-002-131-1 TP 3-RCS-031-8-1 None None 3-RCS-002-131-1 IP 3-RCS-031-8-1 None None 3-RCS-002-131-1 TP 3-RCS-031-8-1 None None 3-RCS-002-136-1 TP 3-RCS-031-13-1 None None 3-RCS-002-136-1 IP 3-RCS-031-13-1 None None 3-RCS-002-136-1 TP 3-RCS-031-13-1 None None 3-RCS-002-144-1 TP 3-RCS-031-18-1 None None 3-RCS-002-144-1 IP 3-RCS-031-18-1 None None 3.6-117 Rev. 30

MPS3 UFSAR TABLE 3.6-13 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP FILL PIPING, LOOP DRAIN PIPING, LETDOWN AND NORMAL CHARGING (CONTINUED)

LINE BLOWDOWN ESSENTIAL PIPE WHIP DESIGNATION BREAK # (1) SOURCE TARGET PRIMARY REQUIREMENT 3-RCS-002-144-1 TP 3-RCS-031-18-1 None None 3-RCS-003-137-1 1 3-RCS-275-15-1 None None 3-RCS-003-137-1 2 3-RCS-275-15-1 None None 3-RCS-003-137-1 3 3-RCS-275-15-1 None None 3-RCS-003-137-1 4 3-RCS-275-15-1 None None 3-RCS-003-137-1 5 3-RCS-275-15-1 None None 3-RCS-003-171-1 6 3-RCS-275-15-1 None None 3-RCS-003-171-1 7 3-RCS-275-15-1 None None 3-RCS-003-171-1 8 3-RCS-275-15-1 None None MPS3 UFSAR 3-RCS-003-145-1 1 3-RCS-275-20-1 None None 3-RCS-003-145-1 2 3-RCS-275-20-1 None None 3-RCS-003-145-1 3 3-RCS-275-20-1 None None 3-RCS-003-145-1 4 3-RCS-275-20-1 None None 3-RCS-003-145-1 5 3-RCS-275-20-1 None None 3-RCS-003-145-1 6 3-RCS-275-20-1 None None 3-RCS-003-145-1 7 3-RCS-275-20-1 None None 3-RCS-003-145-1 8 3-RCS-275-20-1 None None 3-RCS-003-145-1 9 3-RCS-275-20-1 None None 3-RCS-003-149-1 11 3-RCS-275-5-1 None None 3.6-118 Rev. 30

MPS3 UFSAR TABLE 3.6-13 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP FILL PIPING, LOOP DRAIN PIPING, LETDOWN AND NORMAL CHARGING (CONTINUED)

LINE BLOWDOWN ESSENTIAL PIPE WHIP DESIGNATION BREAK # (1) SOURCE TARGET PRIMARY REQUIREMENT 3-RCS-003-149-1 12 3-RCS-275-5-1 None None 3-RCS-003-149-1 13 3-RCS-275-5-1 None None 3-RCS-003-149-1 14 3-RCS-275-5-1 None None 3-RCS-003-149-1 15 3-RCS-275-5-1 None None 3-RCS-003-149-1 16 3-RCS-275-5-1 None None 3-RCS-003-149-1 17 3-RCS-275-5-1 None None 3-RCS-003-149-1 18 3-RCS-275-5-1 None None 3-RCS-003-149-1 19 3-RCS-275-5-1 None None NOTES: MPS3 UFSAR (1) Repetition of Break Numbers is used to identify separate fluid reservoirs which provide a constant pressure source to maintain system pressure subsequent to the postulated pipe break. These reservoirs maintain system blowdown during the transient event as well as in the steady state.

(2) A Large Loss of Coolant Accident (LOCA) occurs subsequent to this postulated break. The NSSS System Standard Design Criteria defines the protection requirements subsequent to this break. Compliance with the NSSS System Standard Design Criteria is deemed essential for the effects of this postulated break. The NSSS System Standard Design Criteria requires that leg to leg propagation on the affected loop shall not exceed 20% of the flow area of the broken pipe. The pipe rupture restraints are provided to ensure compliance with this requirement. This requirement although not strictly defined as essential as stated in FSAR Section 3.6.1 is met such as not to increase the severity of the postulated LOCA.

3.6-119 Rev. 30

MPS3 UFSAR TABLE 3.6-14 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP DRAIN PIPING, LOOP FILL PIPING, LETDOWN LINE, AND NORMAL CHARGING LINE ESSENTIAL JET DISTANCE TO JET INTENSITY JET LOAD ON PROTECTION DESIGNATION BREAK # IMPINGEMENT TARGET TARGET AT THE TARGET THE TARGET REQUIREMENT 3-RCS-008-24-1 13 None N/A N/A N/A None 3-RCS-008-25-1 14 None N/A N/A N/A None 3-RCS-150-27-1 13 None N/A N/A N/A None 3-RCS-150-27-1 15 None N/A N/A N/A None 3-RCS-150-27-1 16 None N/A N/A N/A None 3-RCS-150-27-1 17 None N/A N/A N/A None 3-RCS-008-29-1 13 None N/A N/A N/A None 3-RCS-008-30-1 14 None N/A N/A N/A None 3-RCS-150-32-1 13 None N/A N/A N/A None MPS3 UFSAR 3-RCS-150-32-1 15 None N/A N/A N/A None 3-RCS-150-32-1 16 None N/A N/A N/A None 3-RCS-150-32-1 17 None N/A N/A N/A None 3-RCS-008-34-1 13 None N/A N/A N/A None 3-RCS-008-35-1 14 None N/A N/A N/A None 3-RCS-150-37-1 13 None N/A N/A N/A None 3-RCS-150-37-1 15 None N/A N/A N/A None 3-RCS-150-37-1 16 None N/A N/A N/A None 3-RCS-150-37-1 17 None N/A N/A N/A None 3-RCS-008-39-1 13 None N/A N/A N/A None 3.6-120 Rev. 30

MPS3 UFSAR TABLE 3.6-14 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP DRAIN PIPING, LOOP FILL PIPING, LETDOWN LINE, AND NORMAL CHARGING (CONTINUED)

LINE ESSENTIAL JET DISTANCE TO JET INTENSITY JET LOAD ON PROTECTION DESIGNATION BREAK # IMPINGEMENT TARGET TARGET AT THE TARGET THE TARGET REQUIREMENT 3-RCS-008-40-1 14 None N/A N/A N/A None 3-RCS-150-42-1 13 None N/A N/A N/A None 3-RCS-150-42-1 15 None N/A N/A N/A None 3-RCS-150-42-1 16 None N/A N/A N/A None 3-RCS-150-42-1 17 None N/A N/A N/A None 3-RCS-002-127-1 TP None N/A N/A N/A None 3-RCS-002-127-1 IP None N/A N/A N/A None 3-RCS-002-127-1 TP None N/A N/A N/A None 3-RCS-002-130-1 TP None N/A N/A N/A None MPS3 UFSAR 3-RCS-002-130-1 IP None N/A N/A N/A None 3-RCS-002-130-1 TP None N/A N/A N/A None 3-RCS-002-135-1 TP None N/A N/A N/A None 3-RCS-002-135-1 IP None N/A N/A N/A None 3-RCS-002-135-1 TP None N/A N/A N/A None 3-RCS-002-143-1 TP None N/A N/A N/A None 3-RCS-002-143-1 IP None N/A N/A N/A None 3-RCS-002-143-1 TP None N/A N/A N/A None 3-RCS-002-126-1 TP None N/A N/A N/A None 3-RCS-002-126-1 IP None N/A N/A N/A None 3.6-121 Rev. 30

MPS3 UFSAR TABLE 3.6-14 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP DRAIN PIPING, LOOP FILL PIPING, LETDOWN LINE, AND NORMAL CHARGING (CONTINUED)

LINE ESSENTIAL JET DISTANCE TO JET INTENSITY JET LOAD ON PROTECTION DESIGNATION BREAK # IMPINGEMENT TARGET TARGET AT THE TARGET THE TARGET REQUIREMENT 3-RCS-002-126-1 TP None N/A N/A N/A None 3-RCS-002-129-1 TP None N/A N/A N/A None 3-RCS-002-129-1 IP None N/A N/A N/A None 3-RCS-002-129-1 TP None N/A N/A N/A None 3-RCS-002-134-1 TP None N/A N/A N/A None 3-RCS-002-134-1 IP None N/A N/A N/A None 3-RCS-002-134-1 TP None N/A N/A N/A None 3-RCS-002-142-1 TP None N/A N/A N/A None 3-RCS-002-142-1 IP None N/A N/A N/A None MPS3 UFSAR 3-RCS-002-142-1 TP None N/A N/A N/A None 3-RCS-002-128-1 TP None N/A N/A N/A None 3-RCS-002-128-1 IP None N/A N/A N/A None 3-RCS-002-128-1 TP None N/A N/A N/A None 3-RCS-002-131-1 TP None N/A N/A N/A None 3-RCS-002-131-1 IP None N/A N/A N/A None 3-RCS-002-131-1 TP None N/A N/A N/A None 3-RCS-002-136-1 TP None N/A N/A N/A None 3-RCS-002-136-1 IP None N/A N/A N/A None 3-RCS-002-136-1 TP None N/A N/A N/A None 3.6-122 Rev. 30

MPS3 UFSAR TABLE 3.6-14 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP DRAIN PIPING, LOOP FILL PIPING, LETDOWN LINE, AND NORMAL CHARGING (CONTINUED)

LINE ESSENTIAL JET DISTANCE TO JET INTENSITY JET LOAD ON PROTECTION DESIGNATION BREAK # IMPINGEMENT TARGET TARGET AT THE TARGET THE TARGET REQUIREMENT 3-RCS-002-144-1 TP None N/A N/A N/A None 3-RCS-002-144-1 IP None N/A N/A N/A None 3-RCS-002-144-1 TP None N/A N/A N/A None 3-RCS-003-137-1 1 None N/A N/A N/A None 3-RCS-003-137-1 2 None N/A N/A N/A None 3-RCS-003-137-1 3 None N/A N/A N/A None 3-RCS-003-137-1 4 None N/A N/A N/A None 3-RCS-003-137-1 5 None N/A N/A N/A None 3-RCS-003-137-1 6 None N/A N/A N/A None MPS3 UFSAR 3-RCS-003-137-1 7 None N/A N/A N/A None 3-RCS-003-137-1 8 None N/A N/A N/A None 3-RCS-003-145-1 1 None None None None None 3-RCS-003-145-1 2 None None None None None 3-RCS-003-145-1 3 3RCS*V132 on a None 3-RCS-150-042-1 3-RCS-003-145-1 4 3RCS*V132 on (a) None 3-RCS-150-042-1 3-RCS-003-145-1 5 None None None None None 3-RCS-003-145-1 6 None None None None None 3.6-123 Rev. 30

MPS3 UFSAR TABLE 3.6-14 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP DRAIN PIPING, LOOP FILL PIPING, LETDOWN LINE, AND NORMAL CHARGING (CONTINUED)

LINE ESSENTIAL JET DISTANCE TO JET INTENSITY JET LOAD ON PROTECTION DESIGNATION BREAK # IMPINGEMENT TARGET TARGET AT THE TARGET THE TARGET REQUIREMENT 3-RCS-003-145-1 7 3RCS*V132 on (a) None 3-RCS-150-042-1 3-RCS-003-145-1 8 3RCS*V132 on (a) None 3-RCS-150-042-1 3-RCS-003-145-1 9 3RCS*V132 on (a) None 3-RCS-150-042-1 3-RCS-003-149-1 11 3RCS*V25 on b None 3-RCS-750-110-2 3-RCS-003-149-1 12 3RCS*V25 on (b) None 3-RCS-750-110-2 MPS3 UFSAR 3-RCS-003-149-1 13 None None None None None 3-RCS-003-149-1 14 None None None None None 3-RCS-003-149-1 15 None None None None None 3-RCS-003-149-1 16 None None None None None 3-RCS-003-149-1 17 None None None None None 3-RCS-003-149-1 18 None None None None None 3-RCS-003-149-1 19 None None None None None 3.6-124 Rev. 30

MPS3 UFSAR

a. The NSSS System Standard Design Criteria states that break propagation to an unaffected leg of an affected loop be prevented. The valve 3RCS*V25 and line 3-RCS-150-042-1 are part of the Loop Stop Valve Bypass System connecting to the Hot Leg. Loss of pressure boundary is admitted and since this connection is to the Hot Leg, the break propagation is transmitted to an unaffected leg of an affected loop. This result was transmitted via NEU-6039 Dated December 23, 1985. This condition is enveloped by Case 1.1 in NES-40190 and evaluated on Page 2 of NEU-6039. The break propagation for 3-RCS-150-042-1, namely 1.4 square inches, and the given break area of the ruptured pipe, 3-RCS-003-145-1, namely 5.4 square inches are enveloped by the Case 1.1 analyzed in NEU-6039. The conclusion of the NSSS vendor is that when breaks of equal size occur on the Hot Leg and the Cold Leg the calculated Emergency Core Cooling System (ECCS) performance is less severe than is predicted for a Cold Leg break with an area equal to the total of the two leg areas combined. In Case 1.1, the Hot Leg break area propagation is 37.3 square inches exceeds the Steam Generator Cold Leg break area, 20 square inches. For a combined Hot Leg/Cold Leg break case of 20 square inches in each loop, the total break areas is 40 square inches which corresponds to a 7.1" equivalent diameter break. Millstone 3 analysis demonstrates that a 6" equivalent diameter Cold leg break in calculated Emergency Core Cooling System (ECCS) performance is less severe than a 4" break, thus a 7.1" break is expected to be less limiting than the 6" or 4" breaks.
b. The NSSS System Standard Design Criteria states that break propagation to an unaffected leg of an affected loop be prevented. The valve, 3RCS*V25 and line 3-RCS-750-110-2 are part of the Reactor Plant Sampling System connection from the Hot Leg. Loss of pressure boundary is admitted and since this connection is to the Hot Leg, the break propagation is transmitted to an unaffected leg of an affected loop. This result was transmitted to the NSSS Vendor via NES-40190 Dated November 27, 1985. A review was MPS3 UFSAR performed by the NSSS Vendor concerning this propagation and the results are transmitted via NEU-6039 Dated December 23, 1985. This condition is identified as Case 3.2 in NES-40190 and evaluated on Page 2 of NEU-6039. The NSSS Vendor conclusion is that the propagation break area, namely 0.3 square inches for 3-RCS-750-110-2 is trivial compared to the break area of the ruptured pipe, 3-RCS-003-149-1, namely 5.4 square inches. Therefore phenomena associated at the initial location are not affected by the propagation and calculated Emergency Core Cooling System (ECCS) performance is about the same as it would be without any propagation. The interaction is acceptable therefore protective hardware is not required.

3.6-125 Rev. 30

MPS3 UFSAR TABLE 3.6-15 POSTULATED BREAKS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING BREAK USAGE LINE DESIGNATION # BUILDING ELEVATION BREAK TYPE (1) THRESHOLD STRESS FACTOR FIGURE 3.6-EQN 12 OR EQN 10 13 3-RCS-014-64-1 11 Containment 19'-14" CB N/A Terminal End 14 3-RCS-014-64-1 37 Containment 20'-2" CB & LS Above Threshold 14 3-RCS-014-64-1 38 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 39 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 40 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 41 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 42 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 43 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 44 Containment 22'-4" CB & LS Above Threshold 14 MPS3 UFSAR 3-RCS-014-64-1 45 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 46 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 47 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 48 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 49 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 50 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 51 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 52 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 53 Containment 22'-4" CB & LS Above Threshold 14 3.6-126 Rev. 30

MPS3 UFSAR TABLE 3.6-15 POSTULATED BREAKS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING BREAK USAGE LINE DESIGNATION # BUILDING ELEVATION BREAK TYPE (1) THRESHOLD STRESS FACTOR FIGURE 3.6-EQN 12 OR EQN 10 13 3-RCS-014-64-1 54 Containment 22-4" CB & LS Above Threshold 14 3-RCS-014-64-1 55 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 56 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 57 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 58 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 59 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 60 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 61 Containment 22'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 62 Containment 22'-4" CB & LS Above Threshold 14 MPS3 UFSAR 3-RCS-014-64-1 63 Containment 24'-4" CB & LS Above Threshold 14 3-RCS-014-64-1 64 Containment 25'-11" CB N/A Terminal End 14 3-RCS-004-224-1 1 Containment 78'-6" CB N/A Terminal End 14 3-RCS-004-224-1 2 Containment 81'-3" CB & LS Above Threshold 14 3-RCS-004-224-1 3 Containment 82'-11" CB & LS Above Threshold 14 3-RCS-006-68-1 4 Containment 62'-3" CB & LS Above Threshold 14 3-RCS-004-61-1 5 Containment 52'-1" CB & LS Above Threshold 14 3-RCS-004-22-1 6 Containment 37'-8" CB N/A Terminal End 14 3-RCS-004-22-1 7 Containment 18'-0" CB N/A - Arbitrary Intermediate 14 3.6-127 Rev. 30

MPS3 UFSAR TABLE 3.6-15 POSTULATED BREAKS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING BREAK USAGE LINE DESIGNATION # BUILDING ELEVATION BREAK TYPE (1) THRESHOLD STRESS FACTOR FIGURE 3.6-EQN 12 OR EQN 10 13 3-RCS-004-22-1 8 Containment 17'-6" CB N/A - Arbitrary Intermediate 14 3-RCS-004-22-1 9 Containment 17'-6" CB N/A Terminal End 14 3-RCS-004-60-1 10 Containment 54'-4" CB & LS Above Threshold 14 3-RCS-004-21-1 11 Containment 52'-1" CB & LS Above Threshold 14 3-RCS-004-21-1 12 Containment 41'-6" CB N/A Terminal End 14 3-RCS-004-21-1 13 Containment 17'-6" CB & LS Above Threshold 14 3-RCS-004-21-1 14 Containment 17'-6" CB N/A - Arbitrary Intermediate 14 3-RCS-004-21-1 15 Containment 17'-6" CB N/A Terminal End 14 3-RCS-002-170-1 16 Containment 16'-6" CB N/A Terminal End 14 MPS3 UFSAR 3-RCS-002-150-1 17 Containment 62'-3" CB N/A Terminal End 14 3-RCS-006-65-1 18 Containment 77'-1" CB N/A Terminal End 14 3-RCS-006-65-1 20 Containment 77'-9" CB & LS Above Threshold 14 3-RCS-003-66-1 21 Containment 75'-0" CB Above Threshold 14 3-RCS-003-69-1 22 Containment 75'-0" CB N/A Terminal End 14 3-RCS-003-67-1 23 Containment 75'-0" CB Above Threshold 14 3-RCS-003-70-1 24 Containment 75'-0" CB N/A Terminal End 14 3-RCS-003-69-1 37 Containment 75'-0" CB Above Threshold 14 3-RCS-003-69-1 38 Containment 75'-0" CB Above Threshold 14 3.6-128 Rev. 30

MPS3 UFSAR TABLE 3.6-15 POSTULATED BREAKS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING BREAK USAGE LINE DESIGNATION # BUILDING ELEVATION BREAK TYPE (1) THRESHOLD STRESS FACTOR FIGURE 3.6-EQN 12 OR EQN 10 13 3-RCS-003-70-1 39 Containment 75'-0" CB Above Threshold 14 3-RCS-006-82-1 25 Containment 77'-1" CB N/A Terminal End 14 3-RCS-006-82-1 26 Containment 74'-11" CB & LS Above Threshold 14 3-RCS-006-82-1 27 Containment 74'-11" CB N/A - Arbitrary Intermediate 14 3-RCS-006-82-1 28 Containment 77'-5" CB N/A Terminal End 14 3-RCS-006-83-1 29 Containment 77'-1" CB N/A Terminal End 14 3-RCS-006-83-1 30 Containment 74'-11" CB & LS Above Threshold 14 3-RCS-006-83-1 31 Containment 74'-11" CB N/A - Arbitrary Intermediate 14 3-RCS-006-83-1 32 Containment 77'-5" CB N/A Terminal End 14 MPS3 UFSAR 3-RCS-006-84-1 33 Containment 77'-1" CB N/A Terminal End 14 3-RCS-006-84-1 34 Containment 74'-11" CB & LS Above Threshold 14 3-RCS-006-84-1 35 Containment 74'-11" CB N/A - Arbitrary Intermediate 14 3-RCS-006-84-1 36 Containment 77'-5" CB N/A Terminal End 14 NOTE:

(1) Circumferential Pipe Break (CB) and Longitudinal Pipe Split (LS) are defined in FSAR Section 3.6.2.1.3.

3.6-129 Rev. 30

MPS3 UFSAR TABLE 3.6-16 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Line Designation Break # (1) Blowdown Source Essential Pipe Whip Target Protection Requirement 3-RCS-014-64-1 11 3RCS*TK1 3RCS*SG1B Support System 3RCS-PRR1 3-RCS-014-64-1 37 3RCS*TK1 3RCS*SG1B Support System 3RCS-PRR1 3-RCS-014-64-1 37 3-RCS-029-6-1 None None 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 38 3RCS*TK1 None None 3-RCS-014-64-1 38 3-RCS-029-6-1 None None 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 39 3RCS*TK1 None None 3-RCS-014-64-1 39 3-RCS-029-6-1 None None 3-RCS-014-64-1 Split (2) None None MPS3 UFSAR 3-RCS-014-64-1 40 3RCS*TK1 None None 3-RCS-014-64-1 40 3-RCS-029-6-1 None None 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 41 3RCS*TK1 None None 3-RCS-014-64-1 41 3-RCS-029-6-1 None None 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 42 3RCS*TK1 None None 3-RCS-014-64-1 42 3-RCS-029-6-1 None None 3.6-130 Rev. 30

MPS3 UFSAR TABLE 3.6-16 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Line Designation Break # (1) Blowdown Source Essential Pipe Whip Target Protection Requirement 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 43 3RCS*TK1 Pressurizer Cubicle Wall (3) 3RCS-PRR3 3-RCS-014-64-1 43 3-RCS-029-6-1 None None 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 44 3RCS*TK1 Pressurizer Cubicle Wall (3) 3RCS-PRR3 3-RCS-014-64-1 44 3-RCS-029-6-1 None None 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 45 3RCS*TK1 Pressurizer Cubicle Wall (3) 3RCS-PRR3 3-RCS-014-64-1 45 3-RCS-029-6-1 None None (2)

MPS3 UFSAR 3-RCS-014-64-1 Split None None 3-RCS-014-64-1 46 3RCS*TK1 Pressurizer Cubicle Wall (3) 3RCS-PRR3 3-RCS-014-64-1 46 3-RCS-029-6-1 None None 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 47 3RCS*TK1 Pressurizer Cubicle Wall (3) 3RCS-PRR3 3-RCS-014-64-1 47 3-RCS-029-6-1 None None 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 48 3RCS*TK1 Pressurizer Cubicle Wall (3) 3RCS-PRR3 3.6-131 Rev. 30

MPS3 UFSAR TABLE 3.6-16 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Line Designation Break # (1) Blowdown Source Essential Pipe Whip Target Protection Requirement 3-RCS-014-64-1 48 3-RCS-029-6-1 None None 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 49 3RCS*TK1 Crane Wall (3) 3RCS-PRR5 3-RCS-014-64-1 49 3-RCS-029-6-1 Pressurizer Cubicle Wall (3) 3RCS-PRR3 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 50 3RCS*TK1 Crane Wall (3) 3RCS-PRR5 3-RCS-014-64-1 50 3-RCS-029-6-1 Pressurizer Cubicle Wall (3) 3RCS-PRR3 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 51 3RCS*TK1 Crane Wall (3) 3RCS-PRR5 MPS3 UFSAR 3-RCS-014-64-1 51 3-RCS-029-6-1 Pressurizer Cubicle Wall (3) 3RCS-PRR3 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 52 3RCS*TK1 Crane Wall (3) 3RCS-PRR5 3-RCS-014-64-1 52 3-RCS-029-6-1 Pressurizer Cubicle Wall (3) 3RCS-PRR3 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 53 3RCS*TK1 Crane Wall - (3) 3RCS-PRR6 3-RCS-014-64-1 53 3-RCS-029-6-1 Pressurizer Cubicle Wall (3) 3RCS-PRR4 3.6-132 Rev. 30

MPS3 UFSAR TABLE 3.6-16 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Line Designation Break # (1) Blowdown Source Essential Pipe Whip Target Protection Requirement 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 54 3RCS*TK1 Crane Wall (3) 3RCS-PRR6 3-RCS-014-64-1 54 3-RCS-029-6-1 Pressurizer Cubicle Wall (3) 3RCS-PRR4 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 55 3RCS*TK1 Crane Wall (3) 3RCS-PRR6 3-RCS-014-64-1 55 3-RCS-029-6-1 Pressurizer Cubicle Wall (3) 3RCS-PRR4 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 56 3RCS*TK1 Crane Wall (3) 3RCS-PRR6 3-RCS-014-64-1 56 3-RCS-029-6-1 Pressurizer Cubicle Wall (3) 3RCS-PRR4 MPS3 UFSAR 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 57 3RCS*TK1 Crane Wall (3) 3RCS-PRR6 3-RCS-014-64-1 57 3-RCS-029-6-1 Pressurizer Cubicle Wall (3) 3RCS-PRR4 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 58 3RCS*TK1 Crane Wall (3) 3RCS-PRR6 3-RCS-014-64-1 58 3-RCS-029-6-1 Pressurizer Cubicle Wall (3) 3RCS-PRR7 3-RCS-014-64-1 Split (2) None None 3.6-133 Rev. 30

MPS3 UFSAR TABLE 3.6-16 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Line Designation Break # (1) Blowdown Source Essential Pipe Whip Target Protection Requirement 3-RCS-014-64-1 59 3RCS*TK1 Crane Wall (3) 3RCS-PRR6 3-RCS-014-64-1 59 3-RCS-029-6-1 Pressurizer Cubicle Wall (3) 3RCS-PRR7 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 60 3RCS*TK1 Crane Wall (3) 3RCS-PRR6 3-RCS-014-64-1 60 3-RCS-029-6-1 Pressurizer Cubicle Wall (3) 3RCS-PRR7 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 61 3RCS*TK1 Pressurizer Cubicle Wall (3) 3RCS-PRR8 3-RCS-014-64-1 61 3-RCS-029-6-1 Pressurizer Cubicle Wall (3) 3RCS-PRR7 3-RCS-014-64-1 Split (2) None None MPS3 UFSAR 3-RCS-014-64-1 62 3RCS*TK1 Pressurizer Cubicle Wall (3) 3RCS-PRR8 3-RCS-014-64-1 62 3-RCS-029-6-1 Pressurizer Cubicle Wall (3) 3RCS-PRR7 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 63 3RCS*TK1 None None 3-RCS-014-64-1 63 3-RCS-029-6-1 Pressurizer Cubicle Wall (3) 3RCS-PRR8 3-RCS-014-64-1 Split (2) None None 3-RCS-014-64-1 64 3-RCS-029-6-1 Pressurizer Cubicle Wall (3) 3RCS-PRR8 3.6-134 Rev. 30

MPS3 UFSAR TABLE 3.6-16 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Line Designation Break # (1) Blowdown Source Essential Pipe Whip Target Protection Requirement 3-RCS-004-224-1 1 (4) None None 3-RCS-004-224-1 2 (4) None None 3-RCS-004-224-1 2 3RCS*TK1 None None 3-RCS-004-224-1 Split (4) None None 3-RCS-004-224-1 3 (4) None None 3-RCS-004-224-1 3 3RCS*TK1 None None 3-RCS-004-224-1 Split (4) None None 3-RCS-006-68-1 4 (4) None None 3-RCS-006-68-1 4 3RCS*TK1 None None 3-RCS-006-68-1 Split (4) None None MPS3 UFSAR 3-RCS-004-61-1 5 (4) None None 3-RCS-004-61-1 5 3RCS*TK1 None None 3-RCS-004-61-1 Split (4) None None 3-RCS-004-22-1 6 (4) None None 3-RCS-004-22-1 6 3RCS*TK1 None None 3-RCS-004-22-1 7 3RCS*TK1 None (5) None 3-RCS-004-22-1 7 3-RCS-275-10-1 None (5) None 3.6-135 Rev. 30

MPS3 UFSAR TABLE 3.6-16 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Line Designation Break # (1) Blowdown Source Essential Pipe Whip Target Protection Requirement 3-RCS-004-22-1 8 3RCS*TK1 None (5) None 3-RCS-004-22-1 8 3-RCS-275-10-1 None (5) None 3-RCS-004-22-1 9 3RCS*TK1 None None 3-RCS-004-60-1 10 3RCS*TK1 None None 3-RCS-004-60-1 10 (4) None None 3-RCS-004-60-1 Split (4) None None 3-RCS-004-21-1 11 3RCS*TK1 None None 3-RCS-004-21-1 11 3-RCS-275-10-1 None None 3-RCS-004-21-1 Split (4) None None 3-RCS-004-21-1 12 3RCS*TK1 None None MPS3 UFSAR 3-RCS-004-21-1 12 3-RCS-275-5-1 None None 3-RCS-004-21-1 13 3RCS*TK1 None None 3-RCS-004-21-1 13 3-RCS-275-5-1 None None 3-RCS-004-21-1 Split (4) None None 3-RCS-004-21-1 14 3RCS*TK1 None (5) None 3-RCS-004-21-1 14 3-RCS-275-5-1 None (5) None 3-RCS-004-21-1 15 3RCS*TK1 None None 3.6-136 Rev. 30

MPS3 UFSAR TABLE 3.6-16 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Line Designation Break # (1) Blowdown Source Essential Pipe Whip Target Protection Requirement 3-RCS-004-21-1 15 3-RCS-275-5-1 None None 3-RCS-002-170-1 16 (4) None None 3-RCS-002-150-1 17 (4) None None 3-RCS-006-65-1 18 3RCS*TK1 None None 3-RCS-006-65-1 Split 3RCS*TK1 None None 3-RCS-006-65-1 20 3RCS*TK1 3-RCS-004-224-1(6) 3-RCS-PRR906 3-RCS-003-66-1 21 3RCS*TK1 None None 3-RCS-003-69-1 22 3RCS*TK1 None None 3-RCS-003-67-1 23 3RCS*TK1 None None 3-RCS-003-70-1 24 3RCS*TK1 None None MPS3 UFSAR 3-RCS-006-82-1 25 3RCS*TK1 None None 3-RCS-006-82-1 26 3RCS*TK1 None None 3-RCS-006-82-1 Split 3RCS*TK1 None None 3-RCS-006-82-1 27 3RCS*TK1 None (5) None 3-RCS-006-82-1 28 3RCS*TK1 None None 3-RCS-006-83-1 29 3RCS*TK1 None None 3-RCS-006-83-1 30 3RCS*TK1 None None 3-RCS-006-83-1 Split 3RCS*TK1 None None 3.6-137 Rev. 30

MPS3 UFSAR TABLE 3.6-16 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Line Designation Break # (1) Blowdown Source Essential Pipe Whip Target Protection Requirement 3-RCS-006-83-1 31 3RCS*TK1 None (5) None 3-RCS-006-83-1 32 3RCS*TK1 None None 3-RCS-006-84-1 33 3RCS*TK1 None None 3-RCS-006-84-1 34 3RCS*TK1 3RCS-004-224-1 (6) 3RCS-PRR945 3-RCS-006-84-1 Split 3RCS*TK1 None None 3-RCS-006-84-1 35 3RCS*TK1 None (5) None 3-RCS-006-84-1 36 3RCS*TK1 None None 3-RCS-003-69-1 37 3RCS*TK1 None None 3-RCS-003-69-1 38 3RCS*TK1 None None MPS3 UFSAR 3-RCS-003-70-1 39 3RCS*TK1 None None NOTES:

(1) Repetition of Break Numbers is used to identify separate fluid reservoirs which provide a constant pressure source to maintain system pressure subsequent to a postulated pipe break. These reservoirs maintain system blowdown during the transient event as well as in the steady state.

(2) Longitudinal Splits (LS) on the Pressurizer Surge Line result in blowdown from both the Hot Leg 3-RCS-029-6-1 and the Pressurizer 3RCS*TK1 since the piping remains intact for a LS as stated in FSAR Section 3.6.2.1.3.

(3) The structure identified if subjected to unrestrained pipe whip of the Pressurizer Surge Line would not remain structurally integral and as a result damage to essential systems subsequent to secondary missiles etc. is credible. The pipe rupture restraints are provided to prevent the unrestrained impact to the structure.

3.6-138 Rev. 30

MPS3 UFSAR (4) Circumferential Breaks (CB) on the Pressurizer Spray Line have Cold Legs 3-RCS-275-5-1 and 3-RCS-275-10-1 as blowdown sources. Longitudinal Splits (LS) on the Pressurizer Spray Line have both Cold Legs 3-RCS-275-5-1 and 3-RCS-275-10-1 and the Pressurizer 3RCS*TK1 as blowdown sources.

(5) USNRC Generic Letter 87-11 eliminates the need to evaluate the effects of pipe whip for Arbitrary Intermediate Breaks (AIBs).

herefore pipe whip effects for AIBs on the Pressurizer Safety Lines are not evaluated.

(6) The NSSS System Standard Design Criteria does not permit break propagation from the Pressurizer Relief Line (Steam Phase) to the Pressurizer Spray Line (Liquid Phase). The pipe rupture restraint is provided to prevent the interaction.

MPS3 UFSAR 3.6-139 Rev. 30

MPS3 UFSAR TABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Break # Essential Jet Jet Intensity at Jet Load on Protection Line Designation (1) Impingement Target Distance to Target the Target the Target Requirement 3-RCS-014-64-1 11 3-RCS-008-30-1 (2) None 3-RCS-014-64-1 11 3-RCS-006-119-1 (3) None 3-RCS-014-64-1 37 3-RCS-008-30-1 (2) None 3-RCS-014-64-1 37 3-RCS-006-119-1 (3) None 3-RCS-014-64-1 Split Primary Shield Wall 2 Feet 1274 Psi 774 Kips None 3-RCS-014-64-1 Split 3-RCS-008-30-1 (4) None 3-RCS-014-64-1 38 Primary Shield Wall Less Severe than Longitudinal Split Number 37 (5) None 3-RCS-014-64-1 38 None N/A N/A N/A None 3-RCS-014-64-1 Split None N/A N/A N/A None 3-RCS-014-64-1 39 Primary Shield Wall Less Severe than Longitudinal Split Number 37 (5) None MPS3 UFSAR 3-RCS-014-64-1 39 None N/A N/A N/A None 3-RCS-014-64-1 Split None N/A N/A N/A None 3-RCS-014-64-1 40 Primary Shield Wall Less Severe than Longitudinal Split Number 37 (5) None 3-RCS-014-64-1 40 None N/A N/A N/A None 3-RCS-014-64-1 Split None N/A N/A N/A None 3-RCS-014-64-1 41 Primary Shield Wall Less Severe than Longitudinal Split Number 37 (5) None 3-RCS-014-64-1 41 None N/A N/A N/A None 3-RCS-014-64-1 Split None N/A N/A N/A None 3-RCS-014-64-1 42 Primary Shield Wall Less Severe than Longitudinal Split Number 37 (5) None 3.6-140 Rev. 30

MPS3 UFSAR TABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Break # Essential Jet Jet Intensity at Jet Load on Protection Line Designation (1) Impingement Target Distance to Target the Target the Target Requirement 3-RCS-014-64-1 42 None N/A N/A N/A None 3-RCS-014-64-1 Split None N/A N/A N/A None 3-RCS-014-64-1 43 None N/A N/A N/A None 3-RCS-014-64-1 43 None N/A N/A N/A None 3-RCS-014-64-1 Split None N/A N/A N/A None 3-RCS-014-64-1 44 North-South Cubicle Less Severe than Longitudinal Split Number 37 (5) None Wall 3-RCS-014-64-1 44 None N/A N/A N/A None 3-RCS-014-64-1 Split Removable Concrete Less Severe than Longitudinal Split Number 48 None (6)

Slab at Elevation 12'-9" MPS3 UFSAR 3-RCS-014-64-1 45 Primary Shield Wall Less Severe than Longitudinal Split Number 37 (5) None 3-RCS-014-64-1 45 Steam Generator Less Severe than Longitudinal Split Number 37 (5) None Cubicle Wall 3-RCS-014-64-1 Split Removable Concrete 9 Feet 15 psi 65 Kips None (6)

Slab at Elevation 12'-9" 3-RCS-014-64-1 46 North-South Cubicle Less Severe than Longitudinal Split Number 37 (5) None Wall 3-RCS-014-64-1 46 Steam Generator Less Severe than Longitudinal Split Number 37 (5) None Cubicle Wall 3.6-141 Rev. 30

MPS3 UFSAR TABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Break # Essential Jet Jet Intensity at Jet Load on Protection Line Designation (1) Impingement Target Distance to Target the Target the Target Requirement 3-RCS-014-64-1 Split Removable Concrete 9 Feet 15 psi 101 Kips None (6)

Slab at Elevation 12'-9" 3-RCS-014-64-1 47 North-South Cubicle Less Severe than Longitudinal Split Number 37 (5) None Wall 3-RCS-014-64-1 47 Steam Generator Less Severe than Longitudinal Split Number 37 (5) None Cubicle Wall 3-RCS-014-64-1 Split Removable Concrete 9 Feet 15 psi 125 Kips None Slab at Elevation 12'-9" 3-RCS-014-64-1 48 North-South Cubicle Less Severe than Longitudinal Split Number 37 (5) None Wall 3-RCS-014-64-1 48 Steam Generator None MPS3 UFSAR Less Severe than Longitudinal Split Number 37 (5)

Cubicle Wall 3-RCS-014-64-1 Split Removable Concrete 9 Feet 15 psi 125 Kips None Slab at Elevation 12'-9" 3-RCS-014-64-1 49 3-RCS-004-21-1 Less Severe than Longitudinal Split Number 49 None 3-RCS-014-64-1 49 North-South Cubicle Less Severe than Longitudinal Split Number 37 (5) None Wall 3-RCS-014-64-1 Split 3-RCS-004-21-1 4 Feet 586 psi 6 Kips None (7) 3-RCS-014-64-1 50 3-RCS-004-21-1 Less Severe than Longitudinal Split Number 50 None 3-RCS-014-64-1 50 North-South Cubicle Less Severe than Longitudinal Split Number 37 (5) None Wall 3.6-142 Rev. 30

MPS3 UFSAR TABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Break # Essential Jet Jet Intensity at Jet Load on Protection Line Designation (1) Impingement Target Distance to Target the Target the Target Requirement 3-RCS-014-64-1 Split 3-RCS-004-21-1 4 Feet 586 psi 6 Kips None (7) 3-RCS-014-64-1 51 3-RCS-004-21-1 Less Severe than Longitudinal Split Number 51 None (8) 3-RCS-014-64-1 51 North-South Cubicle Less Severe than Longitudinal Split Number 37 (5) None (8)

Wall 3-RCS-014-64-1 Split 3-RCS-004-21-1 4 Feet 586 psi 6 Kips None (7) 3-RCS-014-64-1 52 3-RCS-004-21-1 Less Severe than Longitudinal Split Number 52 None (8) 3-RCS-014-64-1 52 North-South Cubicle Less Severe than Longitudinal Split Number 37 (5) None (8)

Wall 3-RCS-014-64-1 Split 3-RCS-004-21-1 4 Feet 586 psi 6 Kips None (7) 3-RCS-014-64-1 53 Crane Wall Less Severe than Longitudinal Split Number 37 (5) None (8) MPS3 UFSAR 3-RCS-014-64-1 53 Removable Concrete Less Severe than Longitudinal Split Number 48 None (8)

Slab at Elevation 12'-9" 3-RCS-014-64-1 Split Removable Concrete Less Severe than Longitudinal Split Number 48 None (6)

Slab at Elevation 12'-9" 3-RCS-014-64-1 54 Crane Wall Less Severe than Longitudinal Split Number 37 (5) None (8) 3-RCS-014-64-1 54 Removable Concrete Less Severe than Longitudinal Split Number 48 None (8)

Slab at Elevation 12'-9" 3.6-143 Rev. 30

MPS3 UFSAR TABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Break # Essential Jet Jet Intensity at Jet Load on Protection Line Designation (1) Impingement Target Distance to Target the Target the Target Requirement 3-RCS-014-64-1 Split Removable Concrete Less Severe than Longitudinal Split Number 48 None (6)

Slab at Elevation 12'-9" 3-RCS-014-64-1 55 Crane Wall Less Severe than Longitudinal Split Number 37 (5) None (8) 3-RCS-014-64-1 55 Removable Concrete Less Severe than Longitudinal Split Number 48 None (8)

Slab at Elevation 12'-9" 3-RCS-014-64-1 Split Removable Concrete 9 Feet 15 psi 66 Kips None (6)

Slab at Elevation 12'-9" 3-RCS-014-64-1 56 Crane Wall Less Severe than Longitudinal Split Number 37 (5) None (8) 3-RCS-014-64-1 56 Steam Generator Less Severe than Longitudinal Split Number 37 (5) None (8)

MPS3 UFSAR Cubicle Wall 3-RCS-014-64-1 Split Removable Concrete Less Severe than Longitudinal Split Number 48 None (6)

Slab at Elevation 12'-9" 3-RCS-014-64-1 57 Steam Generator Less Severe than Longitudinal Split Number 37 (5) None (8)

Cubicle Wall 3-RCS-014-64-1 57 Removable Concrete Less Severe than Longitudinal Split Number 48 None (8)

Slab at Elevation 12'-9" 3-RCS-014-64-1 Split Removable Concrete Less Severe than Longitudinal Split Number 48 None (6)

Slab at Elevation 12'-9" 3.6-144 Rev. 30

MPS3 UFSAR TABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Break # Essential Jet Jet Intensity at Jet Load on Protection Line Designation (1) Impingement Target Distance to Target the Target the Target Requirement 3-RCS-014-64-1 58 Steam Generator Less Severe than Longitudinal Split Number 37 (5) None (8)

Cubicle Wall 3-RCS-014-64-1 58 Removable Concrete Less Severe than Longitudinal Split Number 48 None (8)

Slab at Elevation 12'-9" 3-RCS-014-64-1 Split Removable Concrete Less Severe than Longitudinal Split Number 48 None (6)

Slab at Elevation 12'-9" 3-RCS-014-64-1 59 Steam Generator Less Severe than Longitudinal Split Number 37 (5) None (8)

Cubicle Wall 3-RCS-014-64-1 59 Removable Concrete Less Severe than Longitudinal Split Number 48 None (8)

Slab at Elevation 12'-9" MPS3 UFSAR 3-RCS-014-64-1 Split Removable Concrete Less Severe than Longitudinal Split Number 48 None (8)

Slab at Elevation 12'-9" 3-RCS-014-64-1 60 Steam Generator Less Severe than Longitudinal Split Number 37 (5) None (8)

Cubicle Wall 3-RCS-014-64-1 60 Removable Concrete Less Severe than Longitudinal Split Number 48 None (8)

Slab at Elevation 12'-9" 3-RCS-014-64-1 Split Removable Concrete Less Severe than Longitudinal Split Number 48 None (6)

Slab at Elevation 12'-9" 3.6-145 Rev. 30

MPS3 UFSAR TABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Break # Essential Jet Jet Intensity at Jet Load on Protection Line Designation (1) Impingement Target Distance to Target the Target the Target Requirement 3-RCS-014-64-1 61 Crane Wall Less Severe than Longitudinal Split Number 37 (5) None (8) 3-RCS-014-64-1 61 Steam Generator Less Severe than Longitudinal Split Number 37 (5) None (8)

Cubicle Wall 3-RCS-014-64-1 Split Removable Concrete Less Severe than Longitudinal Split Number 48 None (6)

Slab at Elevation 12'-9" 3-RCS-014-64-1 62 Crane Wall Less Severe than Longitudinal Split Number 37 (5) None (8) 3-RCS-014-64-1 62 Removable Concrete Less Severe than the Longitudinal Split Number 62 None (8)

Slab at Elevation 12'-9" 3-RCS-014-64-1 Split Removable Concrete 9 Feet 15 psi 131 Kips None (6)

Slab at Elevation MPS3 UFSAR 12'-9" 3-RCS-014-64-1 63 Crane Wall Less Severe than Longitudinal Split Number 37 (5) None (8) 3-RCS-014-64-1 63 Steam Generator Less Severe than Longitudinal Split Number 37 (5) None (8)

Cubicle Wall 3-RCS-014-64-1 Split Removable Concrete Less Severe than Longitudinal Split Number 37 (5) None Slab at Elevation 12'-9" 3-RCS-014-64-1 64 None N/A N/A N/A None 3-RCS-004-224-1 1 None N/A N/A N/A None 3-RCS-004-224-1 1 None N/A N/A N/A None 3-RCS-004-224-1 2 None N/A N/A N/A None 3.6-146 Rev. 30

MPS3 UFSAR TABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Break # Essential Jet Jet Intensity at Jet Load on Protection Line Designation (1) Impingement Target Distance to Target the Target the Target Requirement 3-RCS-004-224-1 2 None N/A N/A N/A None 3-RCS-004-224-1 Split None N/A N/A N/A None 3-RCS-004-224-1 3 None N/A N/A N/A None 3-RCS-004-224-1 3 None N/A N/A N/A None 3-RCS-004-224-1 Split None N/A N/A N/A None 3-RCS-006-68-1 4 None N/A N/A N/A None 3-RCS-006-68-1 4 None N/A N/A N/A None 3-RCS-006-68-1 Split None N/A N/A N/A None 3-RCS-004-61-1 5 None N/A N/A N/A None 3-RCS-004-61-1 5 None N/A N/A N/A None MPS3 UFSAR 3-RCS-004-61-1 Split None N/A N/A N/A None 3-RCS-004-22-1 6 None N/A N/A N/A None 3-RCS-004-22-1 6 None N/A N/A N/A None 3-RCS-004-22-1 7 None N/A N/A N/A None 3-RCS-004-22-1 7 None N/A N/A N/A None 3-RCS-004-22-1 8 None N/A N/A N/A None 3-RCS-004-22-1 8 None N/A N/A N/A None 3-RCS-004-22-1 9 None N/A N/A N/A None 3-RCS-004-22-1 9 None N/A N/A N/A None 3-RCS-004-60-1 10 None N/A N/A N/A None 3.6-147 Rev. 30

MPS3 UFSAR TABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Break # Essential Jet Jet Intensity at Jet Load on Protection Line Designation (1) Impingement Target Distance to Target the Target the Target Requirement 3-RCS-004-60-1 10 None N/A N/A N/A None 3-RCS-004-60-1 Split None N/A N/A N/A None 3-RCS-004-21-1 11 None N/A N/A N/A None 3-RCS-004-21-1 11 None N/A N/A N/A None 3-RCS-004-21-1 Split None N/A N/A N/A None 3-RCS-004-21-1 12 None N/A N/A N/A None 3-RCS-004-21-1 12 None N/A N/A N/A None 3-RCS-004-21-1 13 North-South Cubicle Less Severe than Longitudinal Split Number 37 (5) None Wall 3-RCS-004-21-1 13 None N/A N/A N/A None 3-RCS-004-21-1 Split None N/A N/A N/A None MPS3 UFSAR 3-RCS-004-21-1 14 None (9) N/A N/A N/A None 3-RCS-004-21-1 14 None N/A N/A N/A None 3-RCS-004-21-1 15 None N/A N/A N/A None 3-RCS-004-21-1 15 None N/A N/A N/A None 3-RCS-002-170-1 16 None N/A N/A N/A None 3-RCS-002-150-1 17 None N/A N/A N/A None 3-RCS-006-65-1 18 3-RCS-004-224-1 Less Severe than Either Longitudinal Split Numbers 49 None (10)

Through 52 3-RCS-006-65-1 20 None N/A N/A N/A None 3-RCS-006-65-1 Split None N/A N/A N/A None 3.6-148 Rev. 30

MPS3 UFSAR TABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Break # Essential Jet Jet Intensity at Jet Load on Protection Line Designation (1) Impingement Target Distance to Target the Target the Target Requirement 3-RCS-003-66-1 21 None N/A N/A N/A None 3-RCS-003-69-1 22 None N/A N/A N/A None 3-RCS-003-67-1 23 None N/A N/A N/A None 3-RCS-003-70-1 24 None N/A N/A N/A None 3-RCS-003-69-1 37 None N/A N/A N/A None 3-RCS-003-69-1 38 None N/A N/A N/A None 3-RCS-003-70-1 39 None N/A N/A N/A None 3-RCS-006-82-1 25 None N/A N/A N/A None 3-RCS-006-82-1 26 None N/A N/A N/A None 3-RCS-006-82-1 Split None N/A N/A N/A None MPS3 UFSAR 3-RCS-006-82-1 27 None (9) N/A N/A N/A None 3-RCS-006-82-1 28 None N/A N/A N/A None 3-RCS-006-83-1 29 None N/A N/A N/A None 3-RCS-006-83-1 30 None N/A N/A N/A None 3-RCS-006-83-1 Split None N/A N/A N/A None 3-RCS-006-83-1 31 None (9) N/A N/A N/A None 3-RCS-006-83-1 32 None N/A N/A N/A None 3-RCS-006-84-1 33 None N/A N/A N/A None 3-RCS-006-84-1 34 None N/A N/A N/A None 3-RCS-006-84-1 Split None N/A N/A N/A None 3.6-149 Rev. 30

MPS3 UFSAR TABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSURIZER CUBICLE PIPING Break # Essential Jet Jet Intensity at Jet Load on Protection Line Designation (1) Impingement Target Distance to Target the Target the Target Requirement 3-RCS-006-84-1 35 None (9) N/A N/A N/A None 3-RCS-006-84-1 36 None N/A N/A N/A None NOTES:

(1) Repetition of Break Numbers is used to identify separate fluid reservoirs which provide a constant pressure source to maintain system pressure subsequent to the postulated pipe break. These reservoirs maintain system blowdown during the transient event as well as in the steady state.

(2) Terminal End break is a Large LOCA. The NSSS System Standard Design Criteria requirements are deemed essential. The requirement for a Large LOCA is to prevent leg to leg propagation within an affected loop to less than 20% of the flow area of the line that is broken. Jet impingement on the Loop Stop Valve Bypass Line 3-RCS-008-30-1 does not violate this criteria since the target pipe is loaded proximate to the Hot Leg nozzle and the remainder of the Loop Stop Valve Bypass: Line 3-RCS-008-30-1 is well supported. The jet impingement on the Loop Stop Valve Bypass, Line 3-RCS-008-30-1, is short duration and the initial break area is much smaller than a full double-ended rupture since pipe rupture restraint 3RCS-PRR1 limits separation and results in less than a full flow area exposed thus minimizing the jet loads on the Loop Stop Valve Bypass Line 3-RCS-008-30-1.

MPS3 UFSAR (3) The NSSS requirement delineated in Note 1 is applicable for this interaction. The Low Pressure Safety Injection Line 3-RCS-006-119-1 to the Hot Leg does not lose pressure boundary since the portion of the Low Pressure Safety Injection Line 3-RCS-006-119-1 to the Hot Leg is well supported by the Hot Leg. The pipe rupture restraint 3RCS-PRR1 limits separation and thus limits the amount of flow area exposed thus limiting the applied load on the Low Pressure Safety Injection Line 3-RCS-006-119-1 to the Hot Leg.

(4) Jet impingement subsequent to this postulated split targets the Loop Stop Valve Bypass Line. The criteria delineated in Note 1 must be complied with. The jet impingement loading is distributed as discussed in FSAR Section 3.6.2.1 on the Loop Stop Valve Bypass line. This line is well supported by pipe rupture restraints and by the Hot Leg to preclude pressure boundary failure.

(5) Jet impingement on structural barrier walls does not cause either barrier failure (i.e., back face scabbing or punching shear) or gross catastrophic failure of the barrier. This conclusion is based upon generic calculations which conclude that extremely high intensities are required to cause local barrier failure. The intensity, distance, and load for jet impingement effects on the Pressurizer Cubicle Piping is governed by the Longitudinal Split (LS) Number 37 which is the closest to the indicated barrier and produces the largest loads.

3.6-150 Rev. 30

MPS3 UFSAR (6) The barrier identified as essential for this break is a removable slab at Elevation 12'-9" in the Pressurizer Cubicle. This slab provides a radiation protection function and cannot fail catastrophically. The indicated jet intensity does not cause either local barrier failure (i.e., back face scabbing or punching shear) or gross catastrophic failure of the barrier. This is based upon generic calculations which conclude that much larger intensities are required to cause local barrier failure.

(7) This postulated Longitudinal Split (LS) results in fluid jet impingement on the Pressurizer Spray Line associated with another loop.

The NSSS System Standard Design Criteria does not permit propagation of a pipe break to an unaffected loop. This NSSS requirement although not strictly defined as essential per FSAR Section 3.6.1 is complied with so as not to increase the severity of the LOCA. The jet impingement load in combination with other loads in accordance with FSAR Section 3.9B-10 results in a stress less than 3Sm. Compliance with this stress allowable demonstrates pressure boundary integrity.

(8) This circumferential break (CB) results in radial jet impingement. Radial jet impingement is less severe than the fluid jet impingement associated with Longitudinal Splits Numbers 37 or 48 since pipe rupture restraints on the Pressurizer Surge Line, 3-RCS-014-64-1, limit both axial and lateral displacement. Limiting the displacement limits the amount of flow area exposed to less than the flow area of the Pressurizer Surge Line, 3-RCS-014-64-1 thus reducing jet intensity on the target.

(9) USNRC Generic Letter 87-11 eliminates the requirement to evaluate jet impingement effects subsequent to Arbitrary Intermediate Breaks (AIBs). Therefore, jet impingement effects are not evaluated by AIBs on the Pressurizer Safety Lines.

(10) This postulated break results in a Large LOCA resulting in a release of saturated steam. The NSSS System Standard Design Criteria does not permit propagation from a steam phase to liquid phase LOCA. Liquid phase LOCA would occur if the Pressurizer Spray Line ruptures subsequent to fluid jet impingement from the Pressurizer Relief Line. This particular interaction is enveloped MPS3 UFSAR by the postulated Longitudinal Splits (LS) Numbers 49 through 52 on the Pressurizer Surge Line, 3-RCS-014-64-1. Pressure boundary integrity is maintained based upon the discussion provided in Note 5 since the Pressurizer Surge Line Splits are more severe.

3.6-151 Rev. 30

MPS3 UFSAR TABLE 3.6-18 POSTULATED PIPE BREAKS - REACTOR COOLANT SYSTEM - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING Threshold Stress Line Designation Break # Building Elevation Break Type (1) Eqn 10 & Either Eqn 12 or 13 Usage Factor Figure 3.6-3-RCS-010-122-1 10 Containment 18'-5" CB N/A Terminal End 13 3-RCS-010-122-1 Split Containment 18'-9" LS (2) 13 3-SIL-010-44-2 11 Containment -22'-0" CB N/A Terminal End 13 3-SIL-010-45-1 2 Containment 11'-10" CB & LS Above Threshold 13 3-SIL-010-45-1 3 Containment 17'-3" CB & LS Above Threshold 13 3-SIL-006-139-1 TP Containment 13'-7" CB N/A Terminal End 13 3-RCS-010-132-1 10 Containment 18'-3" CB N/A Terminal End 13 3-RCS-010-132-1 Split Containment 18'-9" LS (2) 13 3-SIL-010-46-2 11 Containment -22'-0" CB N/A Terminal End 13 MPS3 UFSAR 3-SIL-010-47-1 2 Containment 11'-10" CB & LS Above Threshold 13 3-SIL-010-47-1 3 Containment 17'-3" CB & LS Above Threshold 13 3-SIL-006-140-1 TP Containment 13'-7" CB N/A Terminal End 13 3-RCS-010-138-1 10 Containment 18'-1" CB N/A Terminal End 13 3-RCS-010-138-1 Split Containment 18'-9" LS (2) 13 3-SIL-010-48-2 11 Containment -22'-0" CB N/A Terminal End 13 3-SIL-010-49-1 2 Containment 11'-10" CB & LS Above Threshold 13 3-SIL-010-49-1 3 Containment 17'-3" CB & LS Above Threshold 13 3-SIL-006-145-1 TP Containment 13'-7" CB N/A Terminal End 13 3.6-152 Rev. 30

MPS3 UFSAR TABLE 3.6-18 POSTULATED PIPE BREAKS - REACTOR COOLANT SYSTEM - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)

Threshold Stress Line Designation Break # Building Elevation Break Type (1) Eqn 10 & Either Eqn 12 or 13 Usage Factor Figure 3.6-3-RCS-010-146-1 10 Containment 18'-5" CB N/A Terminal End 13 3-RCS-010-146-1 Split Containment 18'-9" LS (2) 13 3-SIL-010-50-2 11 Containment -22'-0" CB N/A Terminal End 13 3-SIL-010-51-1 2 Containment 11'-10" CB & LS Above Threshold 13 3-SIL-010-51-1 3 Containment 17'-3" CB & LS Above Threshold 13 3-SIL-006-146-1 TP Containment 13'-7" CB N/A Terminal End 13 3-RCS-012-123-1 9 Containment 15'-4" CB N/A Terminal End 13 3-RCS-012-123-1 4 Containment 13'-7" CB & LS Above Threshold 13 3-RCS-012-123-1 5 Containment 13'-7" CB & LS Above Threshold 13 MPS3 UFSAR 3-RCS-012-123-1 6 Containment 13'-7" CB & LS Above Threshold 13 3-RCS-012-123-1 7 Containment 13'-7" CB N/A Terminal End 13 3-RCS-006-124-1 8 Containment 9'-6" CB N/A Terminal End 13 3-RCS-012-103-1 9 Containment 15'-4" CB N/A Terminal End 13 3-RCS-012-103-1 4 Containment 13'-7" CB & LS Above Threshold 13 3-RCS-012-103-1 5 Containment 13'-7" CB & LS Above Threshold 13 3-RCS-012-103-1 6 Containment 13'-7" CB & LS Above Threshold 13 3-RCS-012-103-1 7 Containment 13'-7" CB N/A Terminal End 13 3-RCS-006-140-1 8 Containment 9'-6" CB N/A Terminal End 13 3.6-153 Rev. 30

MPS3 UFSAR TABLE 3.6-18 POSTULATED PIPE BREAKS - REACTOR COOLANT SYSTEM - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)

Threshold Stress Line Designation Break # Building Elevation Break Type (1) Eqn 10 & Either Eqn 12 or 13 Usage Factor Figure 3.6-3-RCS-006-119-1 TP Containment 19'-1" CB N/A (3) N/A (3) 13 3-RCS-006-119-1 IP Containment 20'-10" CB & LS N/A (3) N/A (3) 13 3-RCS-006-119-1 TP Containment 20'-10" CB N/A (3) N/A (3) 13 3-RCS-006-120-1 TP Containment 19'-1" CB N/A (3) N/A (3) 13 3-RCS-006-120-1 IP Containment 20'-10" CB & LS N/A (3) N/A (3) 13 3-RCS-006-120-1 TP Containment 20'-10" CB N/A (3) N/A (3) 13 3-SIH-004-16-2 12 Auxiliary 12'-6" CB N/A Terminal End (4) 3-SIH-004-22-2 13 Auxiliary 12'-6" CB N/A Terminal End (4) MPS3 UFSAR 3-SIH-004-16-2 14 Auxiliary 12'-6" CB N/A - Arbitrary Intermediate (4) 3-SIH-004-16-2 15 Auxiliary 12'-6" CB N/A - Arbitrary Intermediate (4) 3-SIH-004-16-2 16 Auxiliary 18'-5" CB N/A Terminal End (4) 3-RCS-003-133-1 TP Containment 17'-6" CB N/A (5) N/A (5) Not Shown 3-RCS-003-133-1 IP Containment 17'-6" CB N/A (5) N/A (5) Not Shown 3-RCS-150-221-1 IP Containment 17'-6" CB N/A (5) N/A (5) Not Shown 3-RCS-150-221-1 TP Containment 17'-6" CB N/A (5) N/A (5) Not Shown 3-RCS-003-139-1 TP Containment 17'-6" CB N/A (5) N/A (5) Not Shown 3.6-154 Rev. 30

MPS3 UFSAR TABLE 3.6-18 POSTULATED PIPE BREAKS - REACTOR COOLANT SYSTEM - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)

Threshold Stress Line Designation Break # Building Elevation Break Type (1) Eqn 10 & Either Eqn 12 or 13 Usage Factor Figure 3.6-3-RCS-003-139-1 IP Containment 17'-6" CB N/A (5) N/A (5) Not Shown 3-RCS-150-222-1 IP Containment 17'-6" CB N/A (5) N/A (5) Not Shown 3-RCS-150-222-1 TP Containment 17'-6" CB N/A (5) N/A (5) Not Shown 3-RCS-003-147-1 TP Containment 17'-6" CB N/A (5) N/A (5) Not Shown 3-RCS-003-147-1 IP Containment 16'-6" CB N/A (5) N/A (5) Not Shown 3-RCS-150-223-1 IP Containment 9'-10" CB N/A (5) N/A (5) Not Shown 3-RCS-150-223-1 TP Containment 9'-10" CB N/A (5) N/A (5) Not Shown 3-RCS-003-121-1 TP Containment 17'-6" CB N/A (5) N/A (5) Not Shown MPS3 UFSAR 3-RCS-003-121-1 IP Containment 17'-6" CB N/A (5) N/A (5) Not Shown 3-RCS-150-220-1 IP Containment 16'-3" CB N/A (5) N/A (5) Not Shown 3-RCS-150-220-1 TP Containment 16'-3" CB N/A (5) N/A (5) Not Shown NOTES:

(1) Circumferential Pipe Break (CB) and Longitudinal Pipe Split (LS) are defined in FSAR Section 3.6.2.1.3.

(2) Only a Longitudinal Split (LS) is postulated along the intrados of the cut elbow of the Accumulator Discharge Line connection to each Cold Leg. A detailed stress analysis was performed as described in FSAR Section 3.6.2.1.3.1 and the stress analysis indicates that a LS is preferential at this location 3.6-155 Rev. 30

MPS3 UFSAR (3) Breaks are postulated at any location (IP) along the Low Pressure Safety Injection System to the Hot Leg on Loops B & C. Both Circumferential Breaks (CB) and Longitudinal Splits (LS) are postulated. The Terminal Point (TP) at Check Valves 3RCS*V8949B and C and at each connection to the Hot Leg and Intermediate Point (IP) at any location for the Low Pressure Safety Injection System to the Hot Legs 3-RCS-029-6-1 (Loop B) and 3-RCS-029-11-1 (Loop C).

(4) High Pressure Safety Injection (SIH) System from the Normal Charging Pumps to the Containment Isolation Valves 3SIH*MV8801A and 3SIH*MV8801B is a High Energy System since the lines remain pressurized during Normal Plant Operating Conditions and are reviewed for postulated pipe break.

(5) Breaks are postulated at any location (IP) along the High Pressure Safety Injection System into each Cold Leg. Only Circumferential Breaks (CB) are postulated since the High Pressure Safety Injection System has a nominal pipe size less than 4".

Terminal Points (TP) at the Cold Leg and Check Valves 3RCS*V8900A, B, C, or D and Intermediate Points (IP) at any location are postulated on the High Pressure Safety Injection System into each Cold Leg.

MPS3 UFSAR 3.6-156 Rev. 30

MPS3 UFSAR TABLE 3.6-19 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING Line Designation Break # Blowdown Source Essential Pipe Whip Target Protection Requirement 3-RCS-010-122-1 10 3-RCS-275-5-1 3-RCS-008-24-1 (1) 3RCS-PRR6A 3-RCS-010-122-1 Split 3-RCS-275-5-1 None None 3-RCS-010-44-1 11 3SIL*TK1A None None 3-SIL-010-45-1 2 3SIL*TK1A None None 3-SIL-010-45-1 Split 3SIL*TK1A (2) 3SIL-PRR4A 3-SIL-010-45-1 3 3SIL*TK1A None None 3-SIL-010-45-1 Split 3SIL*TK1A (2) 3SIL-PRR4A 3-SIL-006-139-1 TP 3SIL*TK1A (2) 3SIL-PRR4A 3-RCS-010-132-1 10 3-RCS-275-10-1 3-RCS-008-30-1 (1) 3RCS-PRR6B MPS3 UFSAR 3-RCS-010-132-1 Split 3-RCS-275-10-1 None None 3-RCS-010-46-1 11 3SIL*TK1B None None 3-SIL-010-47-1 2 3SIL*TK1B None None 3-SIL-010-47-1 Split 3SIL*TK1B (2) 3SIL-PRR4B 3-SIL-010-47-1 3 3SIL*TK1B None None 3-SIL-010-47-1 Split 3SIL*TK1B (2) 3SIL-PRR4B 3-SIL-006-140-1 TP 3SIL*TK1B (2) 3SIL-PRR4B 3-RCS-010-138-1 10 3-RCS-275-15-1 3-RCS-008-35-1 (1) 3RCS-PRR6C 3-RCS-010-138-1 Split 3-RCS-275-15-1 None None 3.6-157 Rev. 30

MPS3 UFSAR TABLE 3.6-19 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)

Line Designation Break # Blowdown Source Essential Pipe Whip Target Protection Requirement 3-RCS-010-48-1 11 3SIL*TK1C None None 3-SIL-010-49-1 2 3SIL*TK1C None None 3-SIL-010-49-1 Split 3SIL*TK1C (2) 3SIL-PRR4C 3-SIL-010-49-1 3 3SIL*TK1C None None 3-SIL-010-49-1 Split 3SIL*TK1C (2) 3SIL-PRR4C 3-SIL-006-145-1 TP 3SIL*TK1C (2) 3SIL-PRR4C 3-RCS-010-146-1 10 3-RCS-275-20-1 3-RCS-008-40-1 (1) 3RCS-PRR6D 3-RCS-010-146-1 Split 3-RCS-275-20-1 None None 3-RCS-010-50-1 11 3SIL*TK1D None None MPS3 UFSAR 3-SIL-010-51-1 2 3SIL*TK1D None None 3-SIL-010-51-1 Split 3SIL*TK1D (2) 3SIL-PRR4D 3-SIL-010-51-1 3 3SIL*TK1D None None 3-SIL-010-51-1 Split 3SIL*TK1D (2) 3SIL-PRR4D 3-SIL-006-146-1 TP 3SIL*TK1D (2) 3SIL-PRR4D 3-RCS-006-119-1 TP 3-RCS-029-6-1 None None 3-RCS-006-119-1 IP 3-RCS-029-6-1 None None 3-RCS-006-119-1 TP 3-RCS-029-6-1 None None 3-RCS-006-120-1 TP 3-RCS-029-11-1 None None 3.6-158 Rev. 30

MPS3 UFSAR TABLE 3.6-19 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)

Line Designation Break # Blowdown Source Essential Pipe Whip Target Protection Requirement 3-RCS-006-120-1 IP 3-RCS-029-11-1 None None 3-RCS-006-120-1 TP 3-RCS-029-11-1 None None 3-RCS-003-133-1 TP 3-RCS-275-10-1 None None 3-RCS-003-133-1 IP 3-RCS-275-10-1 None None 3-RCS-150-221-1 IP 3-RCS-275-10-1 None None 3-RCS-150-221-1 TP 3-RCS-275-10-1 None None 3-RCS-003-139-1 TP 3-RCS-275-15-1 None None 3-RCS-003-139-1 IP 3-RCS-275-15-1 None None 3-RCS-150-222-1 IP 3-RCS-275-15-1 None None MPS3 UFSAR 3-RCS-150-222-1 TP 3-RCS-275-15-1 None None 3-RCS-003-147-1 TP 3-RCS-275-20-1 None None 3-RCS-003-147-1 IP 3-RCS-275-20-1 None None 3-RCS-150-223-1 IP 3-RCS-275-20-1 None None 3-RCS-150-223-1 TP 3-RCS-275-20-1 None None 3-RCS-003-121-1 TP 3-RCS-275-5-1 None None 3-RCS-003-121-1 IP 3-RCS-275-5-1 None None 3-RCS-150-220-1 IP 3-RCS-275-5-1 None None 3-RCS-150-220-1 TP 3-RCS-275-5-1 None None 3.6-159 Rev. 30

MPS3 UFSAR TABLE 3.6-19 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)

Line Designation Break # Blowdown Source Essential Pipe Whip Target Protection Requirement 3-RCS-012-123-1 9 3-RCS-029-1-1 Containment Liner (3) 3RHS-PRS1A 3-RCS-012-123-1 4 3-RCS-029-1-1 Containment Liner (3) 3RHS-PRS2A 3-RCS-012-123-1 5 3-RCS-029-1-1 Containment Liner (3) 3RHS-PRS2A 3-RCS-012-123-1 6 3-RCS-029-1-1 Containment Liner (3) 3RHS-PRS2A 3-RCS-012-123-1 7 3-RCS-029-1-1 Containment Liner (3) 3RHS-PRS2A 3-RCS-012-123-1 Splits 3-RCS-029-1-1 None None 3-RCS-006-124-1 8 3-RCS-029-1-1 East-West Cubicle Wall (4) 3RCS-PRR861 3-RCS-012-103-1 9 3-RCS-029-16-1 Containment Liner (3) 3RHS-PRS1D MPS3 UFSAR 3-RCS-012-103-1 4 3-RCS-029-16-1 Containment Liner (3) 3RHS-PRS2D 3-RCS-012-103-1 5 3-RCS-029-16-1 Containment Liner (3) 3RHS-PRS2D 3-RCS-012-103-1 6 3-RCS-029-16-1 Containment Liner (3) 3RHS-PRS2D 3-RCS-012-103-1 7 3-RCS-029-16-1 Containment Liner (3) 3RHS-PRS2D 3-RCS-012-103-1 Splits 3-RCS-029-16-1 None None 3-RCS-006-124-1 8 3-RCS-029-16-1 East-West Cubicle Wall (4) 3RCS-PRR862 3-SIH-004-16-2 12 None (5) None (6) None 3-SIH-004-22-2 13 None (5) None (6) None 3.6-160 Rev. 30

MPS3 UFSAR TABLE 3.6-19 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)

Line Designation Break # Blowdown Source Essential Pipe Whip Target Protection Requirement 3-SIH-004-16-2 14 None (5) None (7) None 3-SIH-004-16-2 15 None (5) None (7) None 3-SIH-004-16-2 16 None (5) None (6) None NOTES:

(1) This terminal end break occurs at the Cold Leg Nozzle and results in a large LOCA. The NSSS System Standard Design Criteria requires that break propagation in the affected loop cannot exceed 20% of the flow area of the broken line. The Reactor Coolant System Loop Stop Valve Bypass Line identified has a nominal pipe size of 8" and therefore, if the 8" Loop Stop Valve Bypass Line is impacted pressure boundary integrity must be assured. The pipe rupture restraint indicated prevents the impact.

(2) The Terminal End (TP) break or Longitudinal Split (LS) does not results in a LOCA. The NSSS System Standard Design Criteria delineates requirements for a Large Line Break that does not result in a LOCA. The requirement is that a Non-LOCA cannot propagate into a LOCA. If unrestrained, the broken pipe would induce significant loading upon the Reactor Coolant System MPS3 UFSAR pressure boundary. The pipe rupture restraint is provided to prevent this effect.

(3) A break on the Reactor Coolant System Residual Heat Removal Piping is a Large LOCA. The NSSS System Standard design Criteria provides the acceptance criteria for Large LOCA breaks. The main criterion or this break is to ensure that Containment Leak Tight Integrity is maintained. The pipe whip of the Reactor Coolant System Residual Heat Removal Piping results in potential impact with the Containment Liner. This is an unacceptable interaction and is prevented by the pipe rupture restraints.

(4) Unrestrained pipe whip impact into the East - West Cubicle Wall would cause catastrophic failure of the wall. Therefore, pipe rupture restraints are provided to prevent this impact.

3.6-161 Rev. 30

MPS3 UFSAR (5) The High Pressure Safety Injection (SIH) System in the Auxiliary Building is a High Energy System since the piping is maintained pressurized during Normal Plant Operation between the Normal Charging Pumps and the Normally Closed Containment Isolation Valves, 3SIH*MV8801A and 3SIH*MV8801B. The High Pressure Safety Injection (SIH) System in the Auxiliary Building connects to the Normal Charging Pumps which is not considered a constant pressure source (i.e., reservoir) to sustain system pressure subsequent to a pipe break on the SIH System based upon FSAR Section 3.6.2.2.1, Item 3. The portion between the Normally Closed Containment Isolation Valves, 3SIH*MV8801A and 3SIH*MV8801B and the Reactor Coolant System Pressure Boundary is normally empty.

(6) The High Pressure Safety Injection (SIH) System in the Auxiliary Building does not whip since the High Pressure Safety Injection (SIH) System in the Auxiliary Building connects t the Normal Charging Pumps which is not considered a constant pressure source (i.e., reservoir) to sustain system pressure subsequent to a pipe break on the SIH System based upon FSAR Section 3.6.2.2.1, Item

3. The portion between the Normally Closed Containment Isolation Valves, 3SIH*MV8801A and 3SIH*MV8801B and the Reactor Coolant System Pressure Boundary is normally empty.

(7) USNRC Generic Letter 87-11 eliminates the requirement to evaluate the dynamic effects (i.e., pipe whip) of Arbitrary Intermediate Breaks (AIBs). Therefore, pipe whip effects subsequent to AIBs are not evaluated.

MPS3 UFSAR 3.6-162 Rev. 30

MPS3 UFSAR TABLE 3.6-20 JET IMPINGEMENT EFFECTS - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING Essential Jet Impingement Jet Intensity Jet Load on Protection Line Designation Break # Target Distance to Target at the Target the Target Requirement 3-RCS-010-122-1 10 None N/A N/A N/A None 3-RCS-010-122-1 Split None N/A N/A N/A None 3-RCS-010-44-1 11 None N/A N/A N/A None 3-SIL-010-45-1 2 None N/A N/A N/A None 3-SIL-010-45-1 Split None N/A N/A N/A None 3-SIL-010-45-1 3 None N/A N/A N/A None 3-SIL-010-45-1 Split None N/A N/A N/A None 3-SIL-006-139-1 TP None N/A N/A N/A None 3-RCS-010-132-1 10 None N/A N/A N/A None MPS3 UFSAR 3-RCS-010-132-1 Split None N/A N/A N/A None 3-RCS-010-46-1 11 None N/A N/A N/A None 3-SIL-010-47-1 2 None N/A N/A N/A None 3-SIL-010-47-1 Split None N/A N/A N/A None 3-SIL-010-47-1 3 None N/A N/A N/A None 3-SIL-010-47-1 Split None N/A N/A N/A None 3-SIL-006-140-1 TP None N/A N/A N/A None 3-RCS-010-138-1 10 None N/A N/A N/A None 3-RCS-010-138-1 Split None N/A N/A N/A None 3-RCS-010-48-1 11 None N/A N/A N/A None 3.6-163 Rev. 30

MPS3 UFSAR TABLE 3.6-20 JET IMPINGEMENT EFFECTS - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)

Essential Jet Impingement Jet Intensity Jet Load on Protection Line Designation Break # Target Distance to Target at the Target the Target Requirement 3-SIL-010-49-1 2 None N/A N/A N/A None 3-SIL-010-49-1 Split None N/A N/A N/A None 3-SIL-010-49-1 3 None N/A N/A N/A None 3-SIL-010-49-1 Split None N/A N/A N/A None 3-SIL-006-145-1 TP None N/A N/A N/A None 3-RCS-010-146-1 10 None N/A N/A N/A None 3-RCS-010-146-1 Split None N/A N/A N/A None 3-RCS-010-50-1 11 None N/A N/A N/A None 3-SIL-010-51-1 2 None N/A N/A N/A None MPS3 UFSAR 3-SIL-010-51-1 Split None N/A N/A N/A None 3-SIL-010-51-1 3 None N/A N/A N/A None 3-SIL-010-51-1 Split None N/A N/A N/A None 3-SIL-006-146-1 TP None N/A N/A N/A None 3-RCS-006-119-1 TP None N/A N/A N/A None 3-RCS-006-119-1 IP None N/A N/A N/A None 3-RCS-006-119-1 TP None N/A N/A N/A None 3-RCS-006-120-1 TP None N/A N/A N/A None 3-RCS-006-120-1 IP None N/A N/A N/A None 3-RCS-006-120-1 TP None N/A N/A N/A None 3.6-164 Rev. 30

MPS3 UFSAR TABLE 3.6-20 JET IMPINGEMENT EFFECTS - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)

Essential Jet Impingement Jet Intensity Jet Load on Protection Line Designation Break # Target Distance to Target at the Target the Target Requirement 3-RCS-003-133-1 TP None N/A N/A N/A None 3-RCS-003-133-1 IP (1) (1) None 3-RCS-150-221-1 IP 3RCS*V984 and 3RCS*V985 (2) None 3-RCS-150-221-1 TP 3RCS*V984 and 3RCS*V985 (2) None 3-RCS-003-139-1 TP None N/A N/A N/A None 3-RCS-003-139-1 IP None N/A N/A N/A None 3-RCS-150-222-1 IP None N/A N/A N/A None 3-RCS-150-222-1 TP 3RCS*V979 (2) None 3-RCS-003-147-1 TP None N/A N/A N/A None MPS3 UFSAR 3-RCS-003-147-1 IP None N/A N/A N/A None 3-RCS-150-223-1 IP None N/A N/A N/A None 3-RCS-150-223-1 TP None N/A N/A N/A None 3-RCS-003-121-1 TP None N/A N/A N/A None 3-RCS-003-121-1 IP None N/A N/A N/A None 3-RCS-150-220-1 IP None N/A N/A N/A None 3-RCS-150-220-1 TP None N/A N/A N/A None 3-RCS-012-103-1 9 None N/A N/A N/A None 3-RCS-012-103-1 4 3-SSR-375-026-2 (3) None 3-RCS-012-103-1 5 3-SSR-375-026-2 (3) None 3.6-165 Rev. 30

MPS3 UFSAR TABLE 3.6-20 JET IMPINGEMENT EFFECTS - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)

Essential Jet Impingement Jet Intensity Jet Load on Protection Line Designation Break # Target Distance to Target at the Target the Target Requirement 3-RCS-012-103-1 6 3-SSR-375-026-2 (3) None 3-RCS-012-103-1 7 3-SSR-375-026-2 (3) None 3-RCS-012-103-1 Splits 3-SSR-375-026-2 (3) None 3-RCS-006-124-1 8 None N/A N/A N/A None 3-RCS-012-123-1 9 None N/A N/A N/A None 3-RCS-012-123-1 4 3-SSR-375-035-2 (3) None 3-RCS-012-123-1 5 3-SSR-375-035-2 (3) None 3-RCS-012-123-1 6 3-SSR-375-035-2 (3) None 3-RCS-012-123-1 7 3-SSR-375-035-2 (3) None MPS3 UFSAR 3-RCS-012-123-1 Splits 3-SSR-375-035-2 (3) None 3-RCS-006-124-1 8 None N/A N/A N/A None 3-SIH-004-16-2 12 None (4) N/A N/A N/A None (5) 3-SIH-004-22-2 13 None (4) N/A N/A N/A None (5) 3-SIH-004-16-2 14 None (6) N/A N/A N/A None (5) 3-SIH-004-16-2 15 None (6) N/A N/A N/A None (5) 3-SIH-004-16-2 16 None (4) N/A N/A N/A None (5) 3.6-166 Rev. 30

MPS3 UFSAR NOTES:

(1) The NSSS System Standard Design Criteria states for a Small LOCA that break propagation to the unaffected leg of the affected loop should be prevented. 3-RCS-150-032-1 Loop Stop Valve Bypass Line is the essential target and is not isolated to the Hot Leg.

Loss of pressure boundary is admitted. In addition to this target, Disc Pressurization Valves 3RCS*V984 and 3RCS*V985 are targeted. Break propagation for an initiating event on the High Pressure Safety Injection Line is not permitted. These results were transmitted to the NSSS Vendor via NES-40190, dated November 27, 1985. A review was performed by the NSSS Vendor concerning these interactions and the results of the review transmitted via NEU-6039, dated December 23, 1985. This condition is identified as Case 3.3 in NES-40190 and evaluated on Page 2 of NEU-6039. The NSSS Vendor's conclusion is that Case 3.3 involves break propagation of a Small LOCA to other legs of the affected loop and that the break propagation areas are small compared to the break area of the ruptured line. Phenomena associated with the break at the original location are not affected by the propagation and calculated Emergency Core Cooling System (ECCS) performance will be basically the same as it would with no propagation. Therefore, calculated Emergency Core Cooling System (ECCS) performance conservatively bounds Case 3.3 in NES-40190 and the interactions are acceptable. Therefore, no protective hardware is required.

(2) The NSSS System Standard Design Criteria states for a Small LOCA that break propagation for an initiating break on the High Pressure Safety Injection Line is not permitted. Disc Pressurization Valves 3RCS*V984, 3RCS*V985, and 3RCS*V979 are targeted. Loss of pressure boundary is admitted. This interaction was transmitted to the NSSS Vendor via NES-40190, dated November 27, 1985. A review was performed by the NSSS Vendor concerning this interaction and the results are transmitted via NEU-6039, dated December 23, 1985. This condition is identified as Case 3.3 in NES-40190 and evaluated on Page 2 of NEU- MPS3 UFSAR 6039. The NSSS Vendor's conclusion is that Case 3.3 involves break propagation of a Small LOCA to other legs of the affected loop. The break propagation areas are small compared to the break area of the ruptured line. Phenomena associated with the break at the original location are not affected by the propagation and calculated Emergency Core Cooling System (ECCS) performance will be basically the same as it would with no propagation. Therefore, calculated Emergency Core Cooling System (ECCS) performance conservatively bounds Case 3.3 in NES-40190 and the interaction is acceptable. Therefore, no protective hardware is required.

(3) The NSSS System Standard Design Criteria states for a Large LOCA that damage to the Steam System be prevented. The 3-SSR-375-026-2 or 3-SSR-375-035-2 line is part of the Steam Generator Blowdown Sampling System. Loss of pressure boundary is admitted. This result was transmitted to the NSSS Vendor via NES-40190, dated November 27, 1985. A review was performed by the NSSS Vendor concerning this interaction and the results are transmitted via NEU-6039, dated December 23, 1985. This condition is identified as Case 2.1 in NES-40190 and evaluated on Page 1 of NEU-6039. The NSSS Vendors conclusion is that heat transfer to the secondary side is unimportant for 6: and larger LOCA breaks. For Millstone 3, the limiting case is the 4" equivalent diameter cold leg break which exhibits a 1400 second time interval between break inception and final core recovery. Over 1400 seconds only about 5% of the initial steam generator secondary side inventory would be lost through a Steam Generator Blowdown Sampling System break of this size. Moreover once Auxiliary Feedwater injection begins, following an S signal mass addition to 3.6-167 Rev. 30

MPS3 UFSAR the Steam Generator secondary will exceed the mass depletion through the broken Steam Generator Blowdown Sampling line.

Since the Steam Generator inventory remains about the same Emergency Core Cooling System (ECCS) performance will be basically unaffected by the pressure boundary loss of the Steam Generator Blowdown Sampling System. Therefore this break is less limiting than the 4" equivalent diameter cold leg break and the interaction is acceptable. Therefore, no protective hardware is required.

(4) The High Pressure Safety Injection (SIH) System in the Auxiliary Building does not have sustained fluid jet impingement since the High Pressure Safety Injection (SIH) System in the Auxiliary Building connects to the Normal Charging Pumps which is not considered a constant pressure source (i.e., reservoir) to sustain system pressure subsequent to a pipe break on the SIH System based upon FSAR Section 3.6.2.2.1, Item 3. The portion between the Normally Closed Containment Isolation Valves, 3SIH*MV8801A and 3SIH*MV8801B and the Reactor Coolant System Pressure Boundary is normally empty.

(5) The High Pressure Safety Injection (SIH) System in the Auxiliary Building is a High Energy System since the piping is maintained pressurized during Normal Plant Operation between the Normal Charging Pumps and the Normally Closed Containment Isolation Valves, 3SIH*MV8801A and 3SIH*MV8801B. The High Pressure Safety Injection (SIH) System in the Auxiliary Building connects to the Normal Charging Pumps which is not considered a constant pressure source (i.e., reservoir) to sustain system pressure subsequent to a pipe break on the SIH System based upon FSAR Section 3.6.2.2.1, Item 3. The portion between the Normally Closed Containment Isolation Valves, 3SIH*MV8801A and 3SIH*MV8801B and the Reactor Coolant System Pressure Boundary is normally empty.

(6) USNRC Generic Letter 87-11 eliminates the requirement to evaluate the dynamic effects (i.e., fluid jet impingement) of Arbitrary MPS3 UFSAR Intermediate Breaks (AIBs). Therefore, fluid jet impingement effects subsequent to AIBs are not evaluated.

3.6-168 Rev. 30

MPS3 UFSAR TABLE 3.6-21 POSTULATED BREAKS - CHEMICAL VOLUME CONTROL SYSTEM - NORMAL CHARGING Break Type Line Designation Break # (1) Building Elevation (2) Total Additive Stress Figure 3.6 3-CHS-003-662-2 10 Containment -3'-6" CB N/A Terminal End 15 3-CHS-003-661-2 20 Containment -3'-6" CB N/A Terminal End 15 3-CHS-003-662-2 21 Containment -3'-6" CB N/A Arbitrary Intermediate 15 3-CHS-003-662-2 22 Containment -3'-6" CB N/A Arbitrary Intermediate 15 3-CHS-003-662-2 23 Containment -5'-10" CB N/A Terminal End 15 3-CHS-002-73-2 24 Containment -10'-1" CB Above Threshold 15 3-CHS-002-73-2 25 Containment -10'-1" CB N/A Terminal End 15 3-CHS-003-661-2 26 Containment -2'-0" CB Above Threshold 15 3-CHS-003-661-2 27 Containment -2'-0" CB N/A Arbitrary Intermediate 15 3-CHS-003-661-2 28 Containment -2'-0" CB N/A Terminal End 15 MPS3 UFSAR 3-CHS-003-661-2 29 Containment -10'-2" CB Above Threshold 15 3-CHS-003-76-2 30 Containment -9'-9" CB N/A Terminal End 15 3-CHS-003-72-2 31 Containment -1'-5" CB N/A Terminal End 15 3-CHS-003-72-2 32 Containment 7'-6" CB N/A Arbitrary Intermediate 15 3-CHS-003-72-2 33 Containment 11'-3" CB N/A Arbitrary Intermediate 15 3-CHS-003-70-2 34 Auxiliary 12'-6" CB N/A Arbitrary Intermediate 15 3-CHS-003-70-2 35 Auxiliary 15'-4" CB N/A Arbitrary Intermediate 15 3-CHS-003-69-2 36 Auxiliary 16'-11" CB N/A Terminal End 15 3-CHS-002-237-2 37 Auxiliary 18' - 8" CB N/A Terminal End 15 3.6-169 Rev. 30

MPS3 UFSAR TABLE 3.6-21 POSTULATED BREAKS - CHEMICAL VOLUME CONTROL SYSTEM - NORMAL CHARGING Break Type Line Designation Break # (1) Building Elevation (2) Total Additive Stress Figure 3.6 3-CHS-002-237-2 38 Auxiliary 20'-1" CB N/A Arbitrary Intermediate 15 3-CHS-002-237-2 39 Auxiliary 15'-10" CB N/A Arbitrary Intermediate 15 3-CHS-002-238-2 40 Auxiliary 12'-3" CB N/A Terminal End 15 3-CHS-002-238-2 41 Auxiliary 13'-0" CB N/A Arbitrary Intermediate 15 3-CHS-002-238-2 42 Auxiliary 18'-7" CB N/A Arbitrary Intermediate 15 3-CHS-003-72-2 43 Containment 12'-6" CB N/A Terminal End 15 3-CHS-003-72-2 44 Auxiliary 12'-6" CB N/A Terminal End 15 3-CHS-004-68-2 (3) Auxiliary Varies CB and LS 15 N/A (3) 3-CHS-004-435-2 (3) Auxiliary Varies CB and LS 15 N/A (3) 3-CHS-003-434-2 (3) Auxiliary Varies CB 15 N/A (3)

MPS3 UFSAR 3-CHS-004-28-2 (3) Auxiliary Varies CB and LS 15 N/A (3) 3-CHS-004-30-2 (3) Auxiliary Varies CB and LS 15 N/A (3) 3-CHS-004-29-2 (3) Auxiliary Varies CB and LS 15 N/A (3) 3-CHS-003-65-2 (3) Auxiliary Varies CB 15 N/A (3) 3-CHS-004-433-2 (3) Auxiliary Varies CB and LS 15 N/A (3) 3-CHS-004-682-2 (3) Auxiliary Varies CB and LS N/A (3) 15 3-CHS-004-678-2 (3) Auxiliary Varies CB and LS 15 N/A (3) 3-CHS-004-677-2 (3) Auxiliary Varies CB and LS N/A (3) 15 3.6-170 Rev. 30

MPS3 UFSAR TABLE 3.6-21 POSTULATED BREAKS - CHEMICAL VOLUME CONTROL SYSTEM - NORMAL CHARGING Break Type Line Designation Break # (1) Building Elevation (2) Total Additive Stress Figure 3.6 3-CHS-004-679-2 (3) Auxiliary Varies CB and LS N/A (3) 15 3-CHS-003-67-2 (3) Auxiliary Varies CB 15 N/A (3) 3-CHS-003-31-2 (3) Auxiliary Varies CB 15 N/A (3) 3-CHS-003-32-2 (3) Auxiliary Varies CB N/A (3) 15 3-CHS-002-61-2 (3) Auxiliary Varies CB 15 N/A (3) 3-CHS-002-59-2 (3) Auxiliary Varies CB 15 N/A (3) 3-CHS-002-63-2 (3) Auxiliary Varies CB 15 N/A (3)

NOTES:

(1) A portion of the Chemical & Volume Control System Normal Charging is part of the Reactor Coolant System. Breaks 1 through 9 MPS3 UFSAR and 11 through 19 are postulated on the Reactor Coolant System and are listed in FSAR Table 3.6-12.

(2) Circumferential Pipe Break (CB) and Longitudinal Pipe Split (LS) are defined in FSAR Section 3.6.2.1.3.

(3) Pipe breaks on the Discharge Lines of the Normal Charging System in each Charging Pump Cubicle are postulated at Terminal Ends (TPs), and at each Valve, Fitting, and any Integral Welded Attachment in accordance with the criteria delineated in FSAR Section 3.6.2.1.2.3.b.

3.6-171 Rev. 30

MPS3 UFSAR TABLE 3.6-22 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - NORMAL CHARGING Line Designation Break # Blowdown Source Essential Pipe Whip Target Protection Requirement 3-CHS-003-662-2 10 None (1) None None 3-CHS-003-661-2 20 None (1) None None 3-CHS-003-662-2 21 None (1) None (2) None 3-CHS-003-662-2 22 None (1) None (2) None 3-CHS-003-662-2 23 None (1) None None 3-CHS-002-73-2 24 None (1) None None 3-CHS-002-73-2 25 None (1) None None 3-CHS-003-661-2 26 None (1) None None 3-CHS-003-661-2 27 None (1) None (2) None MPS3 UFSAR 3-CHS-003-661-2 28 None (1) None None 3-CHS-003-661-2 29 None (1) None None 3-CHS-003-76-2 30 None (1) None None 3-CHS-003-72-2 31 None (1) None None 3-CHS-003-72-2 32 None (1) None (2) None 3-CHS-003-72-2 33 None (1) None (2) None 3-CHS-003-70-2 34 None (1) None (2) None 3.6-172 Rev. 30

MPS3 UFSAR TABLE 3.6-22 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - NORMAL CHARGING Line Designation Break # Blowdown Source Essential Pipe Whip Target Protection Requirement 3-CHS-003-70-2 35 None (1) None (2) None 3-CHS-003-69-2 36 None (1) None None 3-CHS-002-237-2 37 None (1) None None 3-CHS-002-237-2 38 None (1) None (2) None 3-CHS-002-237-2 39 None (1) None (2) None 3-CHS-002-238-2 40 None (1) None None 3-CHS-002-238-2 41 None (1) None (2) None 3-CHS-002-238-2 42 None (1) None (2) None 3-CHS-003-72-2 43 None (1) None None MPS3 UFSAR 3-CHS-003-72-2 44 None (1) None None 3-CHS-004-68-2 (3) None None (1) None (4) 3-CHS-004-435-2 (3) None (1) None (4) None 3-CHS-003-434-2 (3) None None (1) None (4) 3-CHS-004-28-2 (3) None None (1) None (4) 3-CHS-004-30-2 (3) None None (1) None (4) 3-CHS-004-29-2 (3) None None (1) None (4) 3-CHS-003-65-2 (3) None (1) None (4) None 3.6-173 Rev. 30

MPS3 UFSAR TABLE 3.6-22 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - NORMAL CHARGING Line Designation Break # Blowdown Source Essential Pipe Whip Target Protection Requirement 3-CHS-004-433-2 (3) None (1) None (4) None 3-CHS-004-682-2 (3) None None (1) None (4) 3-CHS-004-678-2 (3) None (1) None (4) None 3-CHS-004-677-2 (3) None (1) None (4) None 3-CHS-004-679-2 (3) None None (1) None (4) 3-CHS-003-67-2 (3) None None (1) None (4) 3-CHS-003-31-2 (3) None None (1) None (4) 3-CHS-003-32-2 (3) None None (1) None (4) 3-CHS-002-61-2 (3) None None (1) None (4)

MPS3 UFSAR 3-CHS-002-59-2 (3) None (1) None (4) None 3-CHS-002-63-2 (3) None None (1) None (4)

NOTES:

(1) Sustained blowdown on the Chemical & Volume Control System Normal Charging Line does not occur. Dual inline check valves prevent backflow from the Reactor Coolant System and the Charging Pumps have limited capacity to sustain the system pressure.

Therefore pipe whip does not occur.

(2) USNRC Generic Letter 87-11 eliminates the requirement to evaluate the pipe whip effects from Arbitrary Intermediate Breaks (AIBs). Therefore pipe whip effects for AIBs are not evaluated.

(3) Pipe breaks on the Discharge Lines of the Normal Charging System in each Charging Pump Cubicle are postulated at Terminal Ends (TPs), and at each Valve, Fitting, and any Integral Welded Attachments in accordance with the criteria delineated in FSAR Section 3.6.2.1.2.3.b.

3.6-174 Rev. 30

MPS3 UFSAR (4) NERM-069 Revision 1, Attachment 5, Figure 10A shows the location of the Charging Pumps. NERM-069 Revision 1, Attachment 5, Figures 10A and 10B, show that Charging Pump 3CHS*P3A is located in Cubicle 092, Charging Pump 3CHS*P3B is located in Cubicle 093, and Charging Pump 3CHS*P3C is located in Cubicle 094. This layout demonstrates separation in accordance with FSAR Section 3.6.1.1.4.2. Therefore, protective hardware is not required as redundant safety-related systems, and components are separated from one another.

MPS3 UFSAR 3.6-175 Rev. 30

MPS3 UFSAR TABLE 3.6-23 JET IMPINGEMENT EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - NORMAL CHARGING Essential Jet Distance to Jet Intensity at the Jet Load on the Protection Line Designation Break # Impingement Target Target Target Target Requirement 3-CHS-003-662-2 10 None (1) N/A N/A N/A None 3-CHS-003-661-2 20 None (1) N/A N/A N/A None 3-CHS-003-662-2 21 None (1) N/A N/A N/A None (2) 3-CHS-003-662-2 22 None (1) N/A N/A N/A None (2) 3-CHS-003-662-2 23 None (1) N/A N/A N/A None 3-CHS-002-73-2 24 None (1) N/A N/A N/A None 3-CHS-002-73-2 25 None (1) N/A N/A N/A None 3-CHS-003-661-2 26 None (1) N/A N/A N/A None 3-CHS-003-661-2 27 None (1) N/A N/A N/A None (2)

MPS3 UFSAR 3-CHS-003-661-2 28 None (1) N/A N/A N/A None 3-CHS-003-661-2 29 None (1) N/A N/A N/A None 3-CHS-003-76-2 30 None (1) N/A N/A N/A None 3-CHS-003-72-2 31 None (1) N/A N/A N/A None 3-CHS-003-72-2 32 None (1) N/A N/A N/A None (2) 3-CHS-003-72-2 33 None (1) N/A N/A N/A None (2) 3-CHS-003-70-2 34 None (1) N/A N/A N/A None (2) 3-CHS-003-70-2 35 None (1) N/A N/A N/A None (2) 3.6-176 Rev. 30

MPS3 UFSAR TABLE 3.6-23 JET IMPINGEMENT EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - NORMAL CHARGING Essential Jet Distance to Jet Intensity at the Jet Load on the Protection Line Designation Break # Impingement Target Target Target Target Requirement 3-CHS-003-69-2 36 None (1) N/A N/A N/A None 3-CHS-002-237-2 37 None (1) N/A N/A N/A None 3-CHS-002-237-2 38 None (1) N/A N/A N/A None (2) 3-CHS-002-237-2 39 None (1) N/A N/A N/A None (2) 3-CHS-002-238-2 40 None (1) N/A N/A N/A None 3-CHS-002-238-2 41 None (1) N/A N/A N/A None (2) 3-CHS-002-238-2 42 None (1) N/A N/A N/A None (2) 3-CHS-003-72-2 43 None (1) N/A N/A N/A None 3-CHS-003-72-2 44 None (1) N/A N/A N/A None MPS3 UFSAR 3-CHS-004-68-2 (3) N/A N/A N/A None None (1) 3-CHS-004-435-2 (3) N/A N/A N/A None None (1) 3-CHS-003-434-2 (3) N/A N/A N/A None None (1) 3-CHS-004-28-2 (3) N/A N/A N/A None None (1) 3-CHS-004-30-2 (3) N/A N/A N/A None None (1) 3-CHS-004-29-2 (3) None (1) N/A N/A N/A None 3-CHS-003-65-2 (3) N/A N/A N/A None None (1) 3-CHS-004-433-2 (3) None (1) N/A N/A N/A None 3.6-177 Rev. 30

MPS3 UFSAR TABLE 3.6-23 JET IMPINGEMENT EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - NORMAL CHARGING Essential Jet Distance to Jet Intensity at the Jet Load on the Protection Line Designation Break # Impingement Target Target Target Target Requirement 3-CHS-004-682-2 (3) None (1) N/A N/A N/A None 3-CHS-004-678-2 (3) N/A N/A N/A None None (1) 3-CHS-004-677-2 (3) N/A N/A N/A None None (1) 3-CHS-004-679-2 (3) N/A N/A N/A None None (1) 3-CHS-003-67-2 (3) N/A N/A N/A None None (1) 3-CHS-003-31-2 (3) N/A N/A N/A None None (1) 3-CHS-003-32-2 (3) None (1) N/A N/A N/A None 3-CHS-002-61-2 (3) N/A N/A N/A None None (1) 3-CHS-002-59-2 (3) N/A N/A N/A None None (1) MPS3 UFSAR 3-CHS-002-63-2 (3) N/A N/A N/A None None (1)

NOTES:

(1) Sustained blowdown on the Chemical & Volume Control System Normal Charging Line does not occur. Dual inline check valves, 3RCS*V8378A and 3RCS*V8378B to Cold Leg 3-RCS-275-5-1 and 3RCS*V8379A and 3RCS*V8379B to Cold Leg 3-RCS-275-20-1 at the Chemical & Volume Control System/Reactor Coolant System (i.e., Class 1/Class 2) boundary prevent backflow from the Reactor Coolant System and the Charging Pumps have limited capacity and cannot sustain the system pressure. Since sustained blowdown does not occur, dynamic analyses for the effects of fluid jet impingement is not performed based upon FSAR Section 3.6.2.2.1, Item 3.

(2) USNRC Generic Letter 87-11 eliminates the need to evaluate jet impingement effects subsequent to Arbitrary Intermediate Breaks (AIBs). Therefore jet impingement effects are not evaluated for AIBs.

3.6-178 Rev. 30

MPS3 UFSAR (3) Pipe breaks on the Discharge Lines of the Normal Charging System in each Charging Pump Cubicle are postulated at Terminal Ends (TPs), and at each Valve, Fitting, and any Integral Welded Attachments in accordance with the criteria delineated in FSAR Section 3.6.2.1.2.3.b.

MPS3 UFSAR 3.6-179 Rev. 30

MPS3 UFSAR TABLE 3.6-24 POSTULATED BREAKS - CHEMICAL VOLUME CONTROL SYSTEM - LETDOWN LINE Line Designation Break #(1) Building Elevation Break Type (2) Total Additive Stress Figure 3.6-3-CHS-003-1-2 9 Containment -9'-9" CB N/A Terminal End 16 3-CHS-003-416-2 10 Containment -1'-5" CB N/A Terminal End 16 3-CHS-003-416-2 11 Containment 4'-7" CB Above Threshold 16 3-CHS-003-416-2 12 Containment 4'-7" CB Above Threshold 16 3-CHS-003-416-2 13 Containment 3'-7" CB Above Threshold 16 3-CHS-002-2-2 14 Containment 3'-3" CB Above Threshold 16 3-CHS-002-590-2 15 Containment 0'-8" CB N/A Terminal End 16 3-CHS-002-7-2 16 Containment 0'-8" CB N/A Terminal End 16 3-CHS-002-4-2 17 Containment 4'-4" CB Above Threshold 16 3-CHS-002-589-2 18 Containment 0'-8" CB N/A Terminal End 16 MPS3 UFSAR 3-CHS-002-6-2 19 Containment 0'-8" CB N/A Terminal End 16 3-CHS-002-6-2 20 Containment 0'-6" CB Above Threshold 16 3-CHS-002-3-2 21 Containment 4'-4" CB Above Threshold 16 3-CHS-002-588-2 22 Containment 0'-8" CB N/A Terminal End 16 3-CHS-002-5-2 23 Containment 0'-8" CB N/A Terminal End 16 3-CHS-002-5-2 24 Containment 0'-6" CB Above Threshold 16 3-CHS-025-304-2 25 Containment 7'-8" CB N/A Terminal End 16 3-CHS-003-8-2 26 Auxiliary Bldg. 19'-9" CB N/A Arbitrary Intermediate 16 3-CHS-003-8-2 27 Auxiliary Bldg. 20'-2" CB N/A Arbitrary Intermediate 16 3-CHS-003-8-2 28 Auxiliary Bldg. 11'-2" CB N/A Terminal End 16 3.6-180 Rev. 30

MPS3 UFSAR TABLE 3.6-24 POSTULATED BREAKS - CHEMICAL VOLUME CONTROL SYSTEM - LETDOWN LINE (CONTINUED)

Line Designation Break #(1) Building Elevation Break Type (2) Total Additive Stress Figure 3.6-3-CHS-003-8-2 29 Auxiliary Bldg. 12'-7" CB N/A Arbitrary Intermediate 16 3-CHS-003-78-2 30 Auxiliary Bldg. 32'-11" CB N/A Terminal End 16 3-CHS-003-77-2 31 Auxiliary Bldg. 32'-11" CB N/A Terminal End 16 3-CHS-003-9-2 32 Auxiliary Bldg. 22'-8" CB N/A Terminal End 16 3-CHS-002-81-2 33 Auxiliary Bldg. 17'-3" CB N/A Arbitrary Intermediate 16 3-CHS-002-81-2 34 Auxiliary Bldg. 16'-3" CB N/A Terminal End 16 3-CHS-002-80-2 35 Auxiliary Bldg. 6'-0" CB N/A Arbitrary Intermediate 16 3-CHS-002-80-2 36 Auxiliary Bldg. 6'-0" CB N/A Terminal End 16 3-CHS-002-80-2 37 Auxiliary Bldg. 17'-2" CB N/A Arbitrary Intermediate 16 3-CHS-002-80-2 38 Auxiliary Bldg. 20'-1" CB N/A Arbitrary Intermediate 16 3-CHS-002-80-2 39 Auxiliary Bldg. 20'-1" CB N/A Terminal End 16 MPS3 UFSAR 3-CHS-002-80-2 40 Auxiliary Bldg. 12'-1" CB N/A Arbitrary Intermediate 16 3-CHS-002-80-2 41 Auxiliary Bldg. 11'-10" CB N/A Arbitrary Intermediate 16 3-CHS-002-80-2 42 Auxiliary Bldg. -3'-8" CB N/A Terminal End 16 3-CHS-002-80-2 43 Auxiliary Bldg. 9'-10" CB Above Threshold 16 3-CHS-002-80-2 44 Auxiliary Bldg. 10'-1" CB Above Threshold 16 3-CHS-002-80-2 45 Auxiliary Bldg. 13'-0" CB N/A Terminal End 16 3-CHS-002-80-2 46 Auxiliary Bldg. 13'-10" CB N/A Arbitrary Intermediate 16 3-CHS-002-80-2 47 Auxiliary Bldg. 14'-0" CB Above Threshold 16 3-CHS-002-80-2 48 Auxiliary Bldg. 14'-0" CB N/A Terminal End 16 3.6-181 Rev. 30

MPS3 UFSAR TABLE 3.6-24 POSTULATED BREAKS - CHEMICAL VOLUME CONTROL SYSTEM - LETDOWN LINE (CONTINUED)

Line Designation Break #(1) Building Elevation Break Type (2) Total Additive Stress Figure 3.6-3-CHS-002-80-2 49 Auxiliary Bldg. 14'-0" CB Above Threshold 16 3-CHS-002-80-2 50 Auxiliary Bldg. 14'-0" CB Above Threshold 16 3-CHS-002-80-2 51 Auxiliary Bldg. 14'-0" CB Above Threshold 16 3-CHS-002-80-2 52 Auxiliary Bldg. 13'-9" CB Above Threshold 16 3-CHS-002-80-2 53 Auxiliary Bldg. 12'-10" CB N/A Terminal End 16 3-CHS-003-8-2 54 Auxiliary Bldg. 12'-6" CB N/A Terminal End 16 3-CHS-003-659-2 55 Containment 12'-6" CB N/A Terminal End 16 NOTES:

(1) A portion of the Chemical & Volume Control System Normal Letdown Line is part of the Reactor Coolant System. Breaks 1 through 8 are postulated on the Reactor Coolant System portion of the Normal Letdown Line and are listed in FSAR Table 3.6-12. MPS3 UFSAR (2) Circumferential Pipe Break (CB) is defined in FSAR Section 3.6.2.1.3.

3.6-182 Rev. 30

MPS3 UFSAR TABLE 3.6-25 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - LETDOWN LINE Line Designation Break # (1) Blowdown Source Essential Pipe Whip Target Protection Requirement 3-CHS-003-1-2 9 3-RCS-275-15-1 (2) 3CHS-PRR858 & 3RCS-PRR870 3-CHS-003-416-2 10 3-RCS-275-15-1 None None 3-CHS-003-416-2 11 3-RCS-275-15-1 None None 3-CHS-002-3-2 11 3-RCS-275-15-1 None None 3-CHS-003-416-2 12 3-RCS-275-15-1 None None 3-CHS-002-4-2 12 3-RCS-275-15-1 None None 3-CHS-003-416-2 13 3-RCS-275-15-1 None None 3-CHS-002-2-2 14 3-RCS-275-15-1 None None 3-CHS-002-590-2 15 3-RCS-275-15-1 None None 3-CHS-002-7-2 16 3-RCS-275-15-1 None None MPS3 UFSAR 3-CHS-002-4-2 17 3-RCS-275-15-1 None None 3-CHS-002-589-2 18 3-RCS-275-15-1 None None 3-CHS-002-6-2 19 3-RCS-275-15-1 None None 3-CHS-002-6-2 20 3-RCS-275-15-1 None None 3-CHS-002-3-2 21 3-RCS-275-15-1 None None 3-CHS-002-588-2 22 3-RCS-275-15-1 None None 3-CHS-002-5-2 23 3-RCS-275-15-1 None None 3-CHS-002-5-2 24 3-RCS-275-15-1 None None 3-CHS-025-304-2 25 3-RCS-275-15-1 None None 3-CHS-003-8-2 26 3-RCS-275-15-1 (3) (4) 3CHS-PRR967 3.6-183 Rev. 30

MPS3 UFSAR TABLE 3.6-25 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - LETDOWN LINE (CONTINUED)

Line Designation Break # (1) Blowdown Source Essential Pipe Whip Target Protection Requirement 3-CHS-003-8-2 27 3-RCS-275-15-1 (4) None 3-CHS-003-8-2 28 3-RCS-275-15-1 None None 3-CHS-003-8-2 29 3-RCS-275-15-1 (4) None 3-CHS-003-78-2 30 None None None 3-CHS-003-77-2 31 3-RCS-275-15-1 None None 3-CHS-003-9-2 32 3-RCS-275-15-1 None None 3-CHS-002-81-2 33 None (4) None 3-CHS-002-81-2 34 None None None 3-CHS-002-80-2 35 None (4) None 3-CHS-002-80-2 36 None None None 3-CHS-002-80-2 37 None (4) None MPS3 UFSAR 3-CHS-002-80-2 38 None (4) None 3-CHS-002-80-2 39 None None None 3-CHS-002-80-2 40 None (4) None 3-CHS-002-80-2 41 None (4) None 3-CHS-002-80-2 42 None None None 3-CHS-002-80-2 43 None None None 3-CHS-002-80-2 44 None None None 3-CHS-002-80-2 45 None None None 3-CHS-002-80-2 46 None (4) None 3.6-184 Rev. 30

MPS3 UFSAR TABLE 3.6-25 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - LETDOWN LINE (CONTINUED)

Line Designation Break # (1) Blowdown Source Essential Pipe Whip Target Protection Requirement 3-CHS-002-80-2 47 None None None 3-CHS-002-80-2 48 None None None 3-CHS-002-80-2 49 None None None 3-CHS-002-80-2 50 None None None 3-CHS-002-80-2 51 None None None 3-CHS-002-80-2 52 None None None 3-CHS-002-80-2 53 None None None 3-CHS-003-8-2 54 3-RCS-275-15-1 None None 3-CHS-003-659-2 55 3-RCS-275-15-1 None None NOTES:

(1) Repetition of Break Numbers is used to identify separate fluid reservoirs which provide a constant pressure source to maintain MPS3 UFSAR system pressure subsequent to the postulated pipe break. These reservoirs maintain system blowdown during the transient event as well as in the steady state.

(2) Pipe rupture restraints prevent propagation of a Non-LOCA into a LOCA. Restraints prevent excessive loads transmitted from the Class 2 line that is postulated to break to the Class 1 valve and piping.

(3) Pipe rupture restraint protects the Break Exclusion Zone for this system.

(4) USNRC Generic Letter 87-11 eliminates the need to evaluate the pipe whip effects of Arbitrary Intermediate Breaks (AIBs). Based upon this generic letter, pipe whip effects for these AIBs are not evaluated and no protection is necessary.

3.6-185 Rev. 30

MPS3 UFSAR TABLE 3.6-26 JET IMPINGEMENT EFFECTS - CHEMICAL AND VOLUME CONTROL SYSTEM - LETDOWN LINE PIPING Essential Jet Distance to Jet Intensity at Jet Load on the Protection Line Designation Break # Impingement Target Target the Target Target Requirement 3-CHS-003-1-2 9 None N/A N/A N/A None 3-CHS-003-416-2 10 None N/A N/A N/A None 3-CHS-003-416-2 11 None N/A N/A N/A None 3-CHS-003-416-2 12 None N/A N/A N/A None 3-CHS-003-416-2 13 None N/A N/A N/A None 3-CHS-002-2-2 14 None N/A N/A N/A None 3-CHS-002-590-2 15 None N/A N/A N/A None 3-CHS-002-7-2 16 None N/A N/A N/A None 3-CHS-002-4-2 17 None N/A N/A N/A None 3-CHS-002-589-2 18 None N/A N/A N/A None MPS3 UFSAR 3-CHS-002-6-2 19 None N/A N/A N/A None 3-CHS-002-6-2 20 None N/A N/A N/A None 3-CHS-002-3-2 21 None N/A N/A N/A None 3-CHS-002-588-2 22 None N/A N/A N/A None 3-CHS-002-5-2 23 None N/A N/A N/A None 3-CHS-002-5-2 24 None N/A N/A N/A None 3-CHS-025-304-2 25 None N/A N/A N/A None 3-CHS-003-8-2 26 None (1) N/A N/A N/A None 3-CHS-003-8-2 27 None (1) N/A N/A N/A None 3.6-186 Rev. 30

MPS3 UFSAR TABLE 3.6-26 JET IMPINGEMENT EFFECTS - CHEMICAL AND VOLUME CONTROL SYSTEM - LETDOWN LINE PIPING (CONTINUED)

Essential Jet Distance to Jet Intensity at Jet Load on the Protection Line Designation Break # Impingement Target Target the Target Target Requirement 3-CHS-003-8-2 28 None N/A N/A N/A None 3-CHS-003-8-2 29 None (1) N/A N/A N/A None 3-CHS-003-78-2 30 None N/A N/A N/A None 3-CHS-003-77-2 31 None N/A N/A N/A None 3-CHS-003-9-2 32 None N/A N/A N/A None 3-CHS-002-81-2 33 None (1) N/A N/A N/A None 3-CHS-002-81-2 34 None N/A N/A N/A None 3-CHS-002-80-2 35 None (1) N/A N/A N/A None 3-CHS-002-80-2 36 None N/A N/A N/A None MPS3 UFSAR 3-CHS-002-80-2 37 None (1) N/A N/A N/A None 3-CHS-002-80-2 38 None (1) N/A N/A N/A None 3-CHS-002-80-2 39 None N/A N/A N/A None 3-CHS-002-80-2 40 None (1) N/A N/A N/A None 3-CHS-002-80-2 41 None (1) N/A N/A N/A None 3-CHS-002-80-2 42 None N/A N/A N/A None 3-CHS-002-80-2 43 None N/A N/A N/A None 3-CHS-002-80-2 44 None N/A N/A N/A None 3-CHS-002-80-2 45 None N/A N/A N/A None 3.6-187 Rev. 30

MPS3 UFSAR TABLE 3.6-26 JET IMPINGEMENT EFFECTS - CHEMICAL AND VOLUME CONTROL SYSTEM - LETDOWN LINE PIPING (CONTINUED)

Essential Jet Distance to Jet Intensity at Jet Load on the Protection Line Designation Break # Impingement Target Target the Target Target Requirement 3-CHS-002-80-2 46 None (1) N/A N/A N/A None 3-CHS-002-80-2 47 None N/A N/A N/A None 3-CHS-002-80-2 48 None N/A N/A N/A None 3-CHS-002-80-2 49 None N/A N/A N/A None 3-CHS-002-80-2 50 None N/A N/A N/A None 3-CHS-002-80-2 51 None N/A N/A N/A None 3-CHS-002-80-2 52 None N/A N/A N/A None 3-CHS-002-80-2 53 None N/A N/A N/A None 3-CHS-003-8-2 54 None N/A N/A N/A None 3-CHS-003-659-2 55 None N/A N/A N/A None MPS3 UFSAR NOTE:

(1) USNRC Generic Letter 87-11 eliminates the requirement to evaluate jet impingement effects for Arbitrary Intermediate Breaks (AIBs). Therefore, jet impingement effects subsequent to AIBs are not evaluated.

3.6-188 Rev. 30

MPS3 UFSAR TABLE 3.6-27 POSTULATED BREAKS - SEAL WATER INJECTION SYSTEM Line Designation Break # Building Elevation Break Type (1) Total Additive Stress Figure 3-CHS-003-32-2 1 Auxiliary Bldg. 27'-3" CB N/A Terminal End 3.6-17 3-CHS-003-32-2 2 Auxiliary Bldg. 18'-5" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-003-32-2 3 Auxiliary Bldg. 15'-2" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-003-32-2 4 Auxiliary Bldg. 19'-6" CB N/A Terminal End 3.6-17 3-CHS-002-446-2 5 Auxiliary Bldg. 21'-8" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-446-2 6 Auxiliary Bldg. 22'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-34-2 7 Auxiliary Bldg. 34'-11" CB N/A Terminal End 3.6-17 3-CHS-002-34-2 8 Auxiliary Bldg. 36'-4" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-34-2 9 Auxiliary Bldg. 36'-4" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-34-2 10 Auxiliary Bldg. 36'-4" CB N/A Terminal End 3.6-17 MPS3 UFSAR 3-CHS-002-33-2 11 Auxiliary Bldg. 34'-3" CB N/A Terminal End 3.6-17 3-CHS-002-33-2 12 Auxiliary Bldg. 36'-4" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-33-2 13 Auxiliary Bldg. 36'-4" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-33-2 14 Auxiliary Bldg. 36'-4" CB N/A Terminal End 3.6-17 3-CHS-002-447-2 15 Auxiliary Bldg. 36'-4" CB N/A Terminal End 3.6-17 3-CHS-002-445-2 16 Auxiliary Bldg. 36'-4" CB N/A Terminal End 3.6-17 3-CHS-003-38-2 17 Auxiliary Bldg. 18'-0" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-003-38-2 18 Auxiliary Bldg. 18'-5" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-39-2 19 Auxiliary Bldg. 20'-1" CB N/A Terminal End 3.6-17 3-CHS-002-41-2 20 Auxiliary Bldg. 12'-9" CB N/A Arbitrary Intermediate 3.6-17 3.6-189 Rev. 30

MPS3 UFSAR TABLE 3.6-27 POSTULATED BREAKS - SEAL WATER INJECTION SYSTEM (CONTINUED)

Line Designation Break # Building Elevation Break Type (1) Total Additive Stress Figure 3-CHS-002-42-2 21 Auxiliary Bldg. 12'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-638-2 22 Containment 11'-8" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-638-2 23 Containment 12'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-638-2 24 Containment 12'-6" CB N/A Terminal End 3.6-17 3-CHS-002-638-2 25 Containment 9'-3" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-638-2 26 Containment 12'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-638-2 27 Containment 12'-6" CB N/A Terminal End 3.6-17 3-CHS-002-638-2 28 Containment 20'-3" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-638-2 29 Containment 20'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-003-38-2 33 Auxiliary Bldg. 18'-5" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-44-2 34 Auxiliary Bldg. 20'-0" CB N/A Arbitrary Intermediate 3.6-17 MPS3 UFSAR 3-CHS-002-44-2 35 Auxiliary Bldg. 18'-11" CB N/A Terminal End 3.6-17 3-CHS-002-46-2 36 Auxiliary Bldg. 12'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-637-2 37 Containment 11'-6" CB Above Threshold 3.6-17 3-CHS-002-637-2 38 Containment 11'-9" CB Above Threshold 3.6-17 3-CHS-003-38-2 41 Auxiliary Bldg. 18'-5" CB N/A Terminal End 3.6-17 3-CHS-002-51-2 42 Auxiliary Bldg. 13'-4" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-51-2 43 Auxiliary Bldg. 12'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-52-2 44 Containment 11'-8" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-651-2 45 Containment 12'-6" CB N/A Arbitrary Intermediate 3.6-17 3.6-190 Rev. 30

MPS3 UFSAR TABLE 3.6-27 POSTULATED BREAKS - SEAL WATER INJECTION SYSTEM (CONTINUED)

Line Designation Break # Building Elevation Break Type (1) Total Additive Stress Figure 3-CHS-002-651-2 46 Containment 12'-6" CB N/A Terminal End 3.6-17 3-CHS-002-651-2 47 Containment 12'-9" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-651-2 48 Containment 20'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-54-2 52 Auxiliary 19'-6" CB N/A Terminal End 3.6-17 3-CHS-002-56-2 53 Auxiliary 12'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-57-2 54 Containment 12'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-650-2 55 Containment 11'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-650-2 56 Containment 11'-6" CB N/A Terminal End 3.6-17 3-CHS-002-650-2 57 Containment 11'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-650-2 58 Containment 11'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-650-2 59 Containment 11'-6" CB N/A Terminal End 3.6-17 MPS3 UFSAR 3-CHS-002-650-2 60 Containment 13'-7" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-650-2 61 Containment 19'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-43-1 30 Containment 20'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-43-1 31 Containment 20'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-150-689-1 32 Containment 20'-6" CB N/A Terminal End 3.6-17 3-CHS-002-48-1 39 Containment 20'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-150-688-1 40 Containment 20'-6" CB N/A Terminal End 3.6-17 3-CHS-002-53-1 49 Containment 20'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-53-1 50 Containment 20'-6" CB N/A Arbitrary Intermediate 3.6-17 3.6-191 Rev. 30

MPS3 UFSAR TABLE 3.6-27 POSTULATED BREAKS - SEAL WATER INJECTION SYSTEM (CONTINUED)

Line Designation Break # Building Elevation Break Type (1) Total Additive Stress Figure 3-CHS-150-687-1 51 Containment 20'-6" CB N/A Terminal End 3.6-17 3-CHS-002-58-1 62 Containment 20'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-002-58-1 63 Containment 20'-6" CB N/A Arbitrary Intermediate 3.6-17 3-CHS-150-686-1 64 Containment 20'-6" CB N/A Terminal End 3.6-17 3-CHS-002-42-2 65 Auxiliary 12'-6" CB N/A Terminal End 3.6-17 3-CHS-002-42-2 66 Containment 12'-6" CB N/A Terminal End 3.6-17 3-CHS-002-47-2 67 Auxiliary 12'-6" CB N/A Terminal End 3.6-17 3-CHS-002-47-2 68 Containment 12'-6" CB N/A Terminal End 3.6-17 3-CHS-002-57-2 69 Auxiliary 12'-6" CB N/A Terminal End 3.6-17 3-CHS-002-57-2 70 Containment 12'-6" CB N/A Terminal End 3.6-17 3-CHS-002-52-2 71 Auxiliary 12'-6" CB N/A Terminal End 3.6-17 MPS3 UFSAR 3-CHS-002-52-2 72 Containment 12'-6" CB N/A Terminal End 3.6-17 NOTE:

(1) Circumferential Pipe Break (CB) is defined in FSAR Section 3.6.2.1.3.

3.6-192 Rev. 30

MPS3 UFSAR TABLE 3.6-28 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - SEAL WATER INJECTION LINE Line Designation Break # Blowdown Source Essential Pipe Whip Target Protection Requirement 3-CHS-003-32-2 1 None (1) None None 3-CHS-003-32-2 2 None (1) None None (2) 3-CHS-003-32-2 3 None (1) None None (2) 3-CHS-003-32-2 4 None (1) None None 3-CHS-002-446-2 5 None (1) None None (2) 3-CHS-002-446-2 6 None (1) None None (2) 3-CHS-002-34-2 7 None (1) None None 3-CHS-002-34-2 8 None (1) None None (2) 3-CHS-002-34-2 9 None (1) None None (2) MPS3 UFSAR 3-CHS-002-34-2 10 None (1) None None 3-CHS-002-33-2 11 None (1) None None 3-CHS-002-33-2 12 None (1) None None (2) 3-CHS-002-33-2 13 None (1) None None (2) 3-CHS-002-33-2 14 None (1) None None 3-CHS-002-447-2 15 None (1) None None 3-CHS-002-445-2 16 None (1) None None 3-CHS-003-38-2 17 None (1) None None (2) 3.6-193 Rev. 30

MPS3 UFSAR TABLE 3.6-28 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - SEAL WATER INJECTION LINE Line Designation Break # Blowdown Source Essential Pipe Whip Target Protection Requirement 3-CHS-003-38-2 18 None (1) None None (2) 3-CHS-002-39-2 19 None (1) None None 3-CHS-002-41-2 20 None (1) None None (2) 3-CHS-002-42-2 21 None (1) None None (2) 3-CHS-002-638-2 22 None (1) None None (2) 3-CHS-002-638-2 23 None (1) None None (2) 3-CHS-002-638-2 24 None (1) None None 3-CHS-002-638-2 25 None (1) None None (2) 3-CHS-002-638-2 26 None (1) None None (2) MPS3 UFSAR 3-CHS-002-638-2 27 None (1) None None 3-CHS-002-638-2 28 None (1) None None (2) 3-CHS-002-638-2 29 None (1) None None (2) 3-CHS-003-38-2 33 None (1) None None (2) 3-CHS-002-44-2 34 None (1) None None (2) 3-CHS-002-44-2 35 None (1) None None 3-CHS-002-46-2 36 None (1) None None (2) 3-CHS-002-637-2 37 None (1) None None 3.6-194 Rev. 30

MPS3 UFSAR TABLE 3.6-28 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - SEAL WATER INJECTION LINE Line Designation Break # Blowdown Source Essential Pipe Whip Target Protection Requirement 3-CHS-002-637-2 38 None (1) None None 3-CHS-003-38-2 41 None (1) None None 3-CHS-002-51-2 42 None (1) None None (2) 3-CHS-002-51-2 43 None (1) None None (2) 3-CHS-002-52-2 44 None (1) None None (2) 3-CHS-002-651-2 45 None (1) None None (2) 3-CHS-002-651-2 46 None (1) None None 3-CHS-002-651-2 47 None (1) None None (2) 3-CHS-002-651-2 48 None (1) None None (2) MPS3 UFSAR 3-CHS-002-54-2 52 None (1) None None 3-CHS-002-56-2 53 None (1) None None (2) 3-CHS-002-57-2 54 None (1) None None (2) 3-CHS-002-650-2 55 None (1) None None (2) 3-CHS-002-650-2 56 None (1) None None 3-CHS-002-650-2 57 None (1) None None (2) 3-CHS-002-650-2 58 None (1) None None (2) 3-CHS-002-650-2 59 None (1) None None 3.6-195 Rev. 30

MPS3 UFSAR TABLE 3.6-28 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - SEAL WATER INJECTION LINE Line Designation Break # Blowdown Source Essential Pipe Whip Target Protection Requirement 3-CHS-002-650-2 60 None (1) None None (2) 3-CHS-002-650-2 61 None (1) None None (2) 3-CHS-002-43-1 30 None (1) None None (2) 3-CHS-002-43-1 31 None (1) None None (2) 3-CHS-150-689-1 32 None (1) None None 3-CHS-002-48-1 39 None (1) None None (2) 3-CHS-150-688-1 40 None (1) None None 3-CHS-002-53-1 49 None (1) None None (2) 3-CHS-002-53-1 50 None (1) None None (2) MPS3 UFSAR 3-CHS-150-687-1 51 None (1) None None 3-CHS-002-58-1 62 None (1) None None (2) 3-CHS-002-58-1 63 None (1) None None (2) 3-CHS-150-686-1 64 None (1) None None 3-CHS-002-42-2 65 None (1) None None 3-CHS-002-42-2 66 None (1) None None 3-CHS-002-47-2 67 None (1) None None 3-CHS-002-47-2 68 None (1) None None 3.6-196 Rev. 30

MPS3 UFSAR TABLE 3.6-28 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - SEAL WATER INJECTION LINE Line Designation Break # Blowdown Source Essential Pipe Whip Target Protection Requirement 3-CHS-002-57-2 69 None (1) None None 3-CHS-002-57-2 70 None (1) None None 3-CHS-002-52-2 71 None (1) None None 3-CHS-002-52-2 72 None (1) None None NOTES:

(1) Sustained blowdown of the Chemical & Volume Control System - Seal Water Injection Line does not occur. The Seal Water Injection Line connects the Charging Pump to the Reactor Coolant pump. The limited capacity of the pumps does not sustain the system pressure; therefore, pipe whip does not occur for the Seal Water Injection Line and is not evaluated in accordance with FSAR Section 3.6.2.2.1, Item 3.

(2) USNRC Generic Letter 87-11 eliminates the requirement to evaluate pipe whip for Arbitrary Intermediate Breaks (AIBs) therefore pipe whip is not evaluated for AIBs. MPS3 UFSAR 3.6-197 Rev. 30

MPS3 UFSAR TABLE 3.6-29 JET IMPINGEMENT EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - SEAL WATER INJECTION LINE Essential Jet Impingement Distance to Jet Intensity at Jet Load on the Protection Line Designation Break # Target Target the Target Target Requirement 3-CHS-003-32-2 1 None - See Note 1 N/A N/A N/A None 3-CHS-003-32-2 2 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-003-32-2 3 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-003-32-2 4 None - See Note 1 N/A N/A N/A None 3-CHS-002-446-2 5 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-446-2 6 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-34-2 7 None - See Note 1 N/A N/A N/A None 3-CHS-002-34-2 8 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-34-2 9 None - See Note 1 N/A N/A N/A None - See Note 2 MPS3 UFSAR 3-CHS-002-34-2 10 None - See Note 1 N/A N/A N/A None 3-CHS-002-33-2 11 None - See Note 1 N/A N/A N/A None 3-CHS-002-33-2 12 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-33-2 13 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-33-2 14 None - See Note 1 N/A N/A N/A None 3-CHS-002-447-2 15 None - See Note 1 N/A N/A N/A None 3-CHS-002-445-2 16 None - See Note 1 N/A N/A N/A None 3-CHS-003-38-2 17 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-003-38-2 18 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-39-2 19 None - See Note 1 N/A N/A N/A None 3.6-198 Rev. 30

MPS3 UFSAR TABLE 3.6-29 JET IMPINGEMENT EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - SEAL WATER INJECTION LINE (CONTINUED)

Essential Jet Impingement Distance to Jet Intensity at Jet Load on the Protection Line Designation Break # Target Target the Target Target Requirement 3-CHS-002-41-2 20 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-42-2 21 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-638-2 22 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-638-2 23 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-638-2 24 None - See Note 1 N/A N/A N/A None 3-CHS-002-638-2 25 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-638-2 26 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-638-2 27 None - See Note 1 N/A N/A N/A None 3-CHS-002-638-2 28 None - See Note 1 N/A N/A N/A None - See Note 2 MPS3 UFSAR 3-CHS-002-638-2 29 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-003-38-2 33 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-44-2 34 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-44-2 35 None - See Note 1 N/A N/A N/A None 3-CHS-002-46-2 36 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-637-2 37 None - See Note 1 N/A N/A N/A None 3-CHS-002-637-2 38 None - See Note 1 N/A N/A N/A None 3-CHS-003-38-2 41 None - See Note 1 N/A N/A N/A None 3-CHS-002-51-2 42 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-51-2 43 None - See Note 1 N/A N/A N/A None - See Note 2 3.6-199 Rev. 30

MPS3 UFSAR TABLE 3.6-29 JET IMPINGEMENT EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - SEAL WATER INJECTION LINE (CONTINUED)

Essential Jet Impingement Distance to Jet Intensity at Jet Load on the Protection Line Designation Break # Target Target the Target Target Requirement 3-CHS-002-52-2 44 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-651-2 45 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-651-2 46 None - See Note 1 N/A N/A N/A None 3-CHS-002-651-2 47 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-651-2 48 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-54-2 52 None - See Note 1 N/A N/A N/A None 3-CHS-002-56-2 53 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-57-2 54 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-650-2 55 None - See Note 1 N/A N/A N/A None - See Note 2 MPS3 UFSAR 3-CHS-002-650-2 56 None - See Note 1 N/A N/A N/A None 3-CHS-002-650-2 57 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-650-2 58 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-650-2 59 None - See Note 1 N/A N/A N/A None 3-CHS-002-650-2 60 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-650-2 61 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-43-1 30 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-43-1 31 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-150-689-1 32 None - See Note 1 N/A N/A N/A None 3-CHS-002-48-1 39 None - See Note 1 N/A N/A N/A None - See Note 2 3.6-200 Rev. 30

MPS3 UFSAR TABLE 3.6-29 JET IMPINGEMENT EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - SEAL WATER INJECTION LINE (CONTINUED)

Essential Jet Impingement Distance to Jet Intensity at Jet Load on the Protection Line Designation Break # Target Target the Target Target Requirement 3-CHS-150-688-1 40 None - See Note 1 N/A N/A N/A None 3-CHS-002-53-1 49 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-53-1 50 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-150-687-1 51 None - See Note 1 N/A N/A N/A None 3-CHS-002-58-1 62 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-002-58-1 63 None - See Note 1 N/A N/A N/A None - See Note 2 3-CHS-150-686-1 64 None - See Note 1 N/A N/A N/A None 3-CHS-002-42-2 65 None - See Note 1 N/A N/A N/A None 3-CHS-002-42-2 66 None - See Note 1 N/A N/A N/A None MPS3 UFSAR 3-CHS-002-47-2 67 None - See Note 1 N/A N/A N/A None 3-CHS-002-47-2 68 None - See Note 1 N/A N/A N/A None 3-CHS-002-57-2 69 None - See Note 1 N/A N/A N/A None 3-CHS-002-57-2 70 None - See Note 1 N/A N/A N/A None 3-CHS-002-52-2 71 None - See Note 1 N/A N/A N/A None 3-CHS-002-52-2 72 None - See Note 1 N/A N/A N/A None NOTES:

(1) Sustained blowdown does not occur on the Chemical & Volume Control System Seal Water Injection Line. The Seal Water Injection Line connects the Charging Pump to each Reactor Coolant Pump. In each case, the pump has limited capacity to sustain the system pressure. Therefore, jet impingement effects are not sustained and are not evaluated as stated in FSAR Section 3.6.2.2.1 Item 3.

3.6-201 Rev. 30

MPS3 UFSAR TABLE 3.6-30 POSTULATED BREAKS - STEAM GENERATOR BLOWDOWN SYSTEM Line Designation Break # Building Elevation Break Type (1) Total Additive Stress Figure 3.6-3-BDG-002-45-2 1 Containment 24'-7" CB N/A Terminal End 18 3-BDG-004-9-2 2 Containment 21'-1" CB N/A Arbitrary Intermediate 18 3-BDG-004-79-2 3 Containment 20'-4" CB N/A Terminal End 18 3-BDG-004-9-2 4 Containment 22'-7" CB & LS Above Threshold 18 3-BDG-004-17-4 5 MSVB 20'-0" CB N/A Arbitrary Intermediate 18 3-BDG-004-17-4 6 MSVB 19'-6" CB N/A Arbitrary Intermediate 18 3-BDG-004-17-4 7 MSVB 19'-6" CB N/A Terminal End 18 3-BDG-002-44-2 8 Containment 24'-7" CB N/A Terminal End 18 3-BDG-002-44-2 9 Containment 24'-3" CB Above Threshold 18 3-BDG-004-10-2 10 Containment 20'-6" CB N/A Arbitrary Intermediate 18 3-BDG-004-81-2 11 Containment 20'-5" CB N/A Terminal End 18 MPS3 UFSAR 3-BDG-004-18-4 12 MSVB 20'-0" CB N/A Arbitrary Intermediate 18 3-BDG-004-18-4 13 MSVB 19'-6" CB N/A Arbitrary Intermediate 18 3-BDG-004-18-4 14 MSVB 19'-6" CB N/A Terminal End 18 3-BDG-002-46-2 15 Containment 24'-7" CB N/A Terminal End 18 3-BDG-002-46-2 16 Containment 24'-2" CB Above Threshold 18 3-BDG-004-11-2 17 Containment 21'-1" CB N/A Arbitrary Intermediate 18 3-BDG-004-83-2 18 Containment 20'-3" CB N/A Terminal End 18 3-BDG-004-19-4 19 MSVB 20'-0" CB N/A Arbitrary Intermediate 18 3-BDG-004-19-4 20 MSVB 19'-6" CB N/A Arbitrary Intermediate 18 3-BDG-004-19-4 21 MSVB 19'-6" CB N/A Terminal End 18 3-BDG-002-47-2 22 Containment 24'-7" CB N/A Terminal End 18 3.6-202 Rev. 30

MPS3 UFSAR TABLE 3.6-30 POSTULATED BREAKS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)

Line Designation Break # Building Elevation Break Type (1) Total Additive Stress Figure 3.6-3-BDG-004-85-2 23 Containment 19'-9" CB N/A Terminal End 18 3-BDG-004-12-2 24 Containment 16'-6" CB & LS Above Threshold 18 3-BDG-004-12-2 25 Containment 20'-0" CB N/A Arbitrary Intermediate 18 3-BDG-004-20-4 26 MSVB 20'-0" CB N/A Arbitrary Intermediate 18 3-BDG-004-20-4 27 MSVB 19'-6" CB N/A Arbitrary Intermediate 18 3-BDG-004-20-4 28 MSVB 19'-5" CB N/A Terminal End 18 3-BDG-004-17-4 29 MSVB 20'-6" CB N/A Terminal End 18 3-BDG-004-18-4 30 MSVB 20'-6" CB N/A Terminal End 18 3-BDG-004-19-4 31 MSVB 20'-6" CB N/A Terminal End 18 3-BDG-004-20-4 32 MSVB 20'-6" CB N/A Terminal End 18 3-BDG-004-78-2 33 Containment 72'-4" CB N/A Terminal End 18 MPS3 UFSAR 3-BDG-004-78-2 34 Containment 72'-4" CB N/A - Arbitrary Intermediate 18 2" LINE TO SG1A 35 Containment 69'-1" CB N/A - Arbitrary Intermediate 18 2" LINE TO SG1A 36 Containment 68'-10" CB N/A Terminal End 18 3-BDG-004-80-2 37 Containment 74'-7" CB N/A Terminal End 18 2" LINE TO SG1B 38 Containment 69'-1" CB Above Threshold 18 2" LINE TO SG1B 39 Containment 68'-11" CB Above Threshold 18 2" LINE TO SG1B 40 Containment 68'-8" CB N/A Terminal End 18 3-BDG-004-82-2 41 Containment 74'-8" CB N/A Terminal End 18 3-BDG-004-82-2 42 Containment 74'-8" CB N/A - Arbitrary Intermediate 18 2" LINE TO SG1C 43 Containment 69'-7" CB N/A - Arbitrary Intermediate 18 2" LINE TO SG1C 44 Containment 69'-4" CB N/A Terminal End 18 3.6-203 Rev. 30

MPS3 UFSAR TABLE 3.6-30 POSTULATED BREAKS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)

Line Designation Break # Building Elevation Break Type (1) Total Additive Stress Figure 3.6-3-BDG-004-84-2 45 Containment 72'-4" CB N/A Terminal End 18 2" LINE TO SG1D 46 Containment 69'-2" CB Above Threshold 18 2" LINE TO SG1D 47 Containment 68'-11" CB Above Threshold 18 2" LINE TO SG1D 48 Containment 68'-8" CB N/A Terminal End 18 3-BDG-004-55-2 49 Containment 20'-6" CB N/A Terminal End 18 3-BDG-004-54-2 50 Containment 20'-6" CB N/A Terminal End 18 3-BDG-004-56-2 51 Containment 20'-6" CB N/A Terminal End 18 3-BDG-004-57-2 52 Containment 20'-6" CB N/A Terminal End 18 3-BDG-004-17-4 IP MSVB 38' to 56' CB & LS N/A (2) 18 3-BDG-004-18-4 IP MSVB 38' to 56' CB & LS N/A (2) 18 3-BDG-004-19-4 IP MSVB 38' to 56' CB & LS N/A (2) 18 MPS3 UFSAR 3-BDG-004-20-4 IP MSVB 38' to 56' CB & LS N/A (2) 18 3-BDG-004-21-4 IP MSVB & Turbine 56'-0" CB & LS N/A (2) 18 3-BDG-004-22-4 IP MSVB & Turbine 56'-0" CB & LS N/A (2) 18 3-BDG-004-23-4 IP MSVB & Turbine 56'-0" CB & LS N/A (2) 18 3-BDG-004-24-4 IP MSVB & Turbine 56'-0" CB & LS N/A (2) 18 3-BDG-003-49-4 IP MSVB 19'-1" CB N/A (2) 18 3-BDG-003-48-4 IP MSVB 19'-2" CB N/A (2) 18 3-BDG-003-50-4 IP MSVB 18'-10" CB N/A (2) 18 3-BDG-003-51-4 IP MSVB 18'-9" CB N/A (2) 18 3.6-204 Rev. 30

MPS3 UFSAR TABLE 3.6-30 POSTULATED BREAKS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)

Line Designation Break # Building Elevation Break Type (1) Total Additive Stress Figure 3.6-3-BDG-003-49-4 TP MSVB 17'-3" CB N/A (3) 18 3-BDG-003-48-4 TP MSVB 17'-3" CB N/A (3) 18 3-BDG-003-50-4 TP MSVB 16'-11" CB N/A (3) 18 3-BDG-003-51-4 TP MSVB 16'-10" CB N/A (3) 18 3-BDG-008-29-4 None Turbine 56'-0" None N/A (4) 18 NOTES:

(1) Circumferential Pipe Break (CB) and Longitudinal Pipe Split (LS) are defined in FSAR Section 3.6.2.1.3.

(2) Pipe breaks are postulated on the Nonnuclear (i.e., Nonsafety and Nonseismic) portion of the Steam Generator Blowdown System in the Main Steam Valve Building at all valves, fittings, and integral welded attachments. Use of a fitting criteria is in accordance with FSAR Section 3.6.2.1.2.3.2.a. Break Number Designation IP is for an Intermediate Break Location and the Break Number Designation TP is a Terminal Point. MPS3 UFSAR (3) Terminal End (TP) pipe breaks are postulated at the normally closed valves 3BDG-V978, 3BDG-V979, 3BDG-V977, and 3BDG-V976 in accordance with FSAR Section 3.6.2.1.2.3.1.

(4) The Steam Generator Blowdown Tank Manifold, 3-BDG-008-29-4, is located in the Turbine Building. No safety-related structures, systems, or components are located in the Turbine Building. Steam Generator Blowdown Tank Manifold, 3-BDG-008-29-4 is therefore remote from essential structures, system, or component. Postulated pipe break on the Steam Generator Blowdown Tank Manifold, 3-BDG-008-29-4 is not assessed.

3.6-205 Rev. 30

MPS3 UFSAR TABLE 3.6-31 PIPE WHIP EFFECTS - STEAM GENERATOR BLOWDOWN SYSTEM Primary Protection Line Designation Break # (1) Blowdown Source Essential Pipe Whip Target Requirement Figure 3-BDG-002-45-2 1 3-BDG-008-29-4 Steam Generator 3RCS*SG1A Shell 3BDG-PRR975A 3.6-18 3-BDG-004-9-2 2 3-BDG-008-29-4 East-West Concrete Wall None 3.6-18 3-BDG-004-9-2 2 3RCS*SG1A Steam Generator 3RCS*SG1A Shell 3BDG-PRR972A 3.6-18 3-BDG-004-79-2 3 3RCS*SG1A None None 3.6-18 3-BDG-004-9-2 4 3RCS*SG1A None None 3.6-18 3-BDG-004-9-2 4 3-BDG-008-29-4 None None 3.6-18 3-BDG-004-17-4 5 3-BDG-008-29-4 None None 3.6-18 3-BDG-004-17-4 5 3RCS*SG1A Limit Stress to Break Exclusion Zone 3BDG-PRR886 3.6-18 3-BDG-004-17-4 6 3-BDG-008-29-4 None None 3.6-18 MPS3 UFSAR 3-BDG-004-17-4 6 3RCS*SG1A Limit Stress to Break Exclusion Zone 3BDG-PRR941 3.6-18 3-BDG-004-17-4 7 3-BDG-008-29-4 None None 3.6-18 3-BDG-004-17-4 7 3RCS*SG1A Limit Stress to Break Exclusion Zone 3BDG-PRR948 3.6-18 3-BDG-004-17-4 29 3RCS*SG1A Limit Stress to Break Exclusion Zone 3BDG-PRR885 3.6-18 3-BDG-004-17-4 29 3-BDG-008-29-4 None None 3.6-18 3-BDG-002-44-2 8 3-BDG-008-29-4 Steam Generator 3RCS*SG1B Shell 3BDG-PRR975B 3.6-18 3-BDG-002-44-2 9 3-BDG-008-29-4 Steam Generator 3RCS*SG1B Shell 3BDG-PRR975B 3.6-18 3-BDG-002-44-2 9 3RCS*SG1B None None 3.6-18 3-BDG-004-10-2 10 3RCS*SG1B Crane Wall Thimble 18-1 None 3.6-18 3-BDG-004-10-2 10 3-BDG-008-29-4 Prevent Pipe Whip into Containment Liner 3BDG-PRR973 3.6-18 3.6-206 Rev. 30

MPS3 UFSAR TABLE 3.6-31 PIPE WHIP EFFECTS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)

Primary Protection Line Designation Break # (1) Blowdown Source Essential Pipe Whip Target Requirement Figure 3-BDG-004-81-2 11 3RCS*SG1B None None 3.6-18 3-BDG-004-18-4 12 3-BDG-008-29-4 None None 3.6-18 3-BDG-004-18-4 12 3RCS*SG1B Limit Stress to Break Exclusion Zone 3BDG-PRR888 3.6-18 3-BDG-004-18-4 13 3-BDG-008-29-4 None None 3.6-18 3-BDG-004-18-4 13 3RCS*SG1B Limit Stress to Break Exclusion Zone 3BDG-PRR942 3.6-18 3-BDG-004-18-4 14 3RCS*SG1B Limit Stress to Break Exclusion Zone 3BDG-PRR950 3.6-18 3-BDG-004-18-4 14 3-BDG-008-29-4 None None 3.6-18 3-BDG-004-18-4 30 3RCS*SG1B Limit Stress to Break Exclusion Zone 3BDG-PRR887 3.6-18 3-BDG-004-18-4 30 3-BDG-008-29-4 None None 3.6-18 3-BDG-002-46-2 15 3-BDG-008-29-4 Steam Generator 3RCS*SG1C Shell 3BDG-PRR975C 3.6-18 MPS3 UFSAR 3-BDG-002-46-2 16 3RCS*SG1C None None 3.6-18 3-BDG-002-46-2 16 3-BDG-008-29-4 Steam Generator 3RCS*SG1C Shell 3BDG-PRR975C 3.6-18 3-BDG-004-11-2 17 3RCS*SG1C Steam Generator 3RCS*SG1C Shell 3BDG-PRR972C 3.6-18 3-BDG-004-11-2 17 3-BDG-008-29-4 East-West Concrete Wall None 3.6-18 3-BDG-004-83-2 18 3RCS*SG1C None None 3.6-18 3-BDG-004-19-4 19 3RCS*SG1C Limit Stress to Break Exclusion Zone 3BDG-PRR883 3.6-18 3-BDG-004-19-4 19 3-BDG-008-29-4 None None 3.6-18 3-BDG-004-19-4 20 3-BDG-008-29-4 None None 3.6-18 3-BDG-004-19-4 20 3RCS*SG1C Limit Stress to Break Exclusion Zone 3BDG-PRR943 3.6-18 3.6-207 Rev. 30

MPS3 UFSAR TABLE 3.6-31 PIPE WHIP EFFECTS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)

Primary Protection Line Designation Break # (1) Blowdown Source Essential Pipe Whip Target Requirement Figure 3-BDG-004-19-4 21 3RCS*SG1C Limit Stress to Break Exclusion Zone 3BDG-PRR947 3.6-18 3-BDG-004-19-4 21 3-BDG-008-29-4 None None 3.6-18 3-BDG-004-19-4 31 3RCS*SG1C Limit Stress to Break Exclusion Zone 3BDG-PRR884 3.6-18 3-BDG-004-19-4 31 3-BDG-008-29-4 None None 3.6-18 3-BDG-002-47-2 22 3-BDG-008-29-4 Steam Generator 3RCS*SG1D Shell 3BDG-PRR975D 3.6-18 3-BDG-004-85-2 23 3RCS*SG1D None None 3.6-18 3-BDG-004-12-2 24 3-BDG-008-29-4 None None 3.6-18 3-BDG-004-12-2 24 3RCS*SG1D None None 3.6-18 3-BDG-004-12-2 25 3-BDG-008-29-4 Prevent Pipe Whip into Containment Liner 3BDG-PRR974 3.6-18 3-BDG-004-12-2 25 3RCS*SG1D None None 3.6-18 MPS3 UFSAR 3-BDG-004-20-4 26 3-BDG-008-29-4 None None 3.6-18 3-BDG-004-20-4 26 3RCS*SG1D Limit Stress to Break Exclusion Zone 3BDG-PRR881 3.6-18 3-BDG-004-20-4 27 3-BDG-008-29-4 None None 3.6-18 3-BDG-004-20-4 27 3RCS*SG1D Limit Stress to Break Exclusion Zone 3BDG-PRR944 3.6-18 3-BDG-004-20-4 28 3RCS*SG1D Limit Stress to Break Exclusion Zone 3BDG-PRR949 3.6-18 3-BDG-004-20-4 28 3-BDG-008-29-4 None None 3.6-18 3-BDG-004-20-4 32 3RCS*SG1D Limit Stress to Break Exclusion Zone 3BDG-PRR882 3.6-18 3-BDG-004-20-4 32 3-BDG-008-29-4 None None 3.6-18 3-BDG-004-55-2 49 3RCS*SG1A None None 3.6-18 3.6-208 Rev. 30

MPS3 UFSAR TABLE 3.6-31 PIPE WHIP EFFECTS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)

Primary Protection Line Designation Break # (1) Blowdown Source Essential Pipe Whip Target Requirement Figure 3-BDG-004-54-2 50 3RCS*SG1B None None 3.6-18 3-BDG-004-56-2 51 3RCS*SG1C Prevent Pipe Whip into Steam Generator 3BDG-PRR970 3.6-18 Support 3-BDG-004-57-2 52 3RCS*SG1D None None 3.6-18 NOTES:

(1) Repetition of Break Numbers is used to identify separate fluid reservoirs which provide a constant pressure source to maintain system pressure subsequent to the postulated pipe break. These reservoirs maintain system blowdown during the transient event as well as in the steady state. The Steam Generator Blowdown Line runs from each Steam Generator Inside Containment to a common header 3-BDG-008-29-4 in the Turbine Building and Eventually to the Steam Generator Blowdown Tank 3BDG*TK1 in the Turbine Building. The common header is assumed to be a constant pressure source. The Steam Generator Blowdown Tank is low pressure (i.e., Relief Valve 3BDG-RV30 set at 75 psig) and cannot sustain the system pressure. MPS3 UFSAR (2) Steam Generator Blowdown (BDG) System - Wet Lay Up Piping is a High Energy System from each connection to the Steam Generators to the Normally Closed Valves 3BDG*V887, 3BDG*V889, 3BDG*V891, and 3BDG*V893. The West Lay Up Piping is separated from adjacent Steam Generators so any pipe whip in one Steam Generator Cubicle does not damage essential structures, systems, or components associated with an adjacent Steam Generator based upon the Millstone Unit 3 Hazards Review Program Summary, NERM-069 Revision 1. The Wet Lay Up Piping is above the Floor Slab at Elevation 51'-4" which provides separation between the Reactor Coolant Pump Cubicles and the Upper Steam Generator Cubicles. Consequently pipe whip effects on the Wet Lay Up Piping cannot propagate into a Loss of Coolant Accident (LOCA) based upon separation documented in the Millstone Unit 3 Hazards Review Program Summary, NERM-069, Revision 1.

3.6-209 Rev. 30

MPS3 UFSAR TABLE 3.6-32 JET IMPINGEMENT EFFECTS - STEAM GENERATOR BLOWDOWN SYSTEM Essential Jet Jet Intensity at Jet Load on Protection Line Designation Break # (1) Impingement Target Distance to Target the Target the Target Requirement 3-BDG-002-45-2 1 None N/A N/A N/A None 3-BDG-004-9-2 2 None (2) N/A N/A N/A None 3-BDG-004-9-2 2 None (2) N/A N/A N/A None 3-BDG-004-79-2 3 None N/A N/A N/A None 3-BDG-004-9-2 4 None N/A N/A N/A None 3-BDG-004-9-2 4 None N/A N/A N/A None 3-BDG-004-9-2 Split None N/A N/A N/A None 3-BDG-004-17-4 5 None (2) N/A N/A N/A None 3-BDG-004-17-4 5 None (2) N/A N/A N/A None 3-BDG-004-17-4 6 N/A N/A N/A None MPS3 UFSAR None (2) 3-BDG-004-17-4 6 None (2) N/A N/A N/A None 3-BDG-004-17-4 7 None N/A N/A N/A None 3-BDG-004-17-4 7 None N/A N/A N/A None 3-BDG-004-17-4 29 None N/A N/A N/A None 3-BDG-004-17-4 29 None N/A N/A N/A None 3-BDG-002-44-2 8 None N/A N/A N/A None 3-BDG-002-44-2 9 None N/A N/A N/A None 3-BDG-002-44-2 9 None N/A N/A N/A None 3-BDG-004-10-2 10 None (2) N/A N/A N/A None 3.6-210 Rev. 30

MPS3 UFSAR TABLE 3.6-32 JET IMPINGEMENT EFFECTS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)

Essential Jet Jet Intensity at Jet Load on Protection Line Designation Break # (1) Impingement Target Distance to Target the Target the Target Requirement 3-BDG-004-10-2 10 None (2) N/A N/A N/A None 3-BDG-004-81-2 11 None N/A N/A N/A None 3-BDG-004-18-4 12 None (2) N/A N/A N/A None 3-BDG-004-18-4 12 None (2) N/A N/A N/A None 3-BDG-004-18-4 13 None (2) N/A N/A N/A None 3-BDG-004-18-4 13 None (2) N/A N/A N/A None 3-BDG-004-18-4 14 None N/A N/A N/A None 3-BDG-004-18-4 14 None N/A N/A N/A None 3-BDG-004-18-4 30 None N/A N/A N/A None 3-BDG-004-18-4 30 None N/A N/A N/A None MPS3 UFSAR 3-BDG-002-46-2 15 None N/A N/A N/A None 3-BDG-002-46-2 16 None N/A N/A N/A None 3-BDG-002-46-2 16 None N/A N/A N/A None 3-BDG-004-11-2 17 None (2) N/A N/A N/A None 3-BDG-004-11-2 17 None (2) N/A N/A N/A None 3-BDG-004-83-2 18 None N/A N/A N/A None 3-BDG-004-19-4 19 None (2) N/A N/A N/A None 3-BDG-004-19-4 19 None (2) N/A N/A N/A None 3.6-211 Rev. 30

MPS3 UFSAR TABLE 3.6-32 JET IMPINGEMENT EFFECTS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)

Essential Jet Jet Intensity at Jet Load on Protection Line Designation Break # (1) Impingement Target Distance to Target the Target the Target Requirement 3-BDG-004-19-4 20 None (2) N/A N/A N/A None 3-BDG-004-19-4 20 None (2) N/A N/A N/A None 3-BDG-004-19-4 21 None N/A N/A N/A None 3-BDG-004-19-4 21 None N/A N/A N/A None 3-BDG-004-19-4 31 None N/A N/A N/A None 3-BDG-004-19-4 31 None N/A N/A N/A None 3-BDG-002-47-2 22 None N/A N/A N/A None 3-BDG-004-85-2 23 None N/A N/A N/A None 3-BDG-004-12-2 24 None N/A N/A N/A None 3-BDG-004-12-2 24 None N/A N/A N/A None MPS3 UFSAR 3-BDG-004-12-2 Split None N/A N/A N/A None 3-BDG-004-12-2 25 None (2) N/A N/A N/A None 3-BDG-004-12-2 25 None (2) N/A N/A N/A None 3-BDG-004-20-4 26 None (2) N/A N/A N/A None 3-BDG-004-20-4 26 None (2) N/A N/A N/A None 3-BDG-004-20-4 27 None (2) N/A N/A N/A None 3-BDG-004-20-4 27 None (2) N/A N/A N/A None 3-BDG-004-20-4 28 None N/A N/A N/A None 3-BDG-004-20-4 28 None N/A N/A N/A None 3.6-212 Rev. 30

MPS3 UFSAR TABLE 3.6-32 JET IMPINGEMENT EFFECTS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)

Essential Jet Jet Intensity at Jet Load on Protection Line Designation Break # (1) Impingement Target Distance to Target the Target the Target Requirement 3-BDG-004-20-4 32 None N/A N/A N/A None 3-BDG-004-20-4 32 None N/A N/A N/A None 3-BDG-004-55-2 49 None N/A N/A N/A None 3-BDG-004-54-2 50 None N/A N/A N/A None 3-BDG-004-56-2 51 None N/A N/A N/A None 3-BDG-004-57-2 52 None N/A N/A N/A None 3-BDG-004-78-2 33 None (3) N/A N/A N/A None 3-BDG-004-78-2 34 None (3) N/A N/A N/A None 2" LINE TO SG1A 35 None (3) N/A N/A N/A None 2" LINE TO SG1A 36 None (3) N/A N/A N/A None MPS3 UFSAR 3-BDG-004-80-2 37 None (3) N/A N/A N/A None 2" LINE TO SG1B 38 None (3) N/A N/A N/A None 2" LINE TO SG1B 39 None (3) N/A N/A N/A None 2" LINE TO SG1B 40 None (3) N/A N/A N/A None 3-BDG-004-82-2 41 None (3) N/A N/A N/A None 3-BDG-004-82-2 42 None (3) N/A N/A N/A None 2" LINE TO SG1C 43 None (3) N/A N/A N/A None 2" LINE TO SG1C 44 None (3) N/A N/A N/A None 3.6-213 Rev. 30

MPS3 UFSAR TABLE 3.6-32 JET IMPINGEMENT EFFECTS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)

Essential Jet Jet Intensity at Jet Load on Protection Line Designation Break # (1) Impingement Target Distance to Target the Target the Target Requirement 3-BDG-004-84-2 45 None (3) N/A N/A N/A None 2" LINE TO SG1D 46 None (3) N/A N/A N/A None 2" LINE TO SG1D 47 None (3) N/A N/A N/A None 2" LINE TO SG1D 48 None (3) N/A N/A N/A None 3-BDG-004-17-4 IP None N/A N/A N/A None 3-BDG-004-17-4 IP None N/A N/A N/A None 3-BDG-004-17-4 Split None N/A N/A N/A None 3-BDG-004-18-4 IP None N/A N/A N/A None 3-BDG-004-18-4 IP None N/A N/A N/A None 3-BDG-004-18-4 Split None N/A N/A N/A None MPS3 UFSAR 3-BDG-004-19-4 IP None N/A N/A N/A None 3-BDG-004-19-4 IP None N/A N/A N/A None 3-BDG-004-19-4 Split None N/A N/A N/A None 3-BDG-004-20-4 IP None N/A N/A N/A None 3-BDG-004-20-4 IP None N/A N/A N/A None 3-BDG-004-20-4 Split None N/A N/A N/A None 3-BDG-004-21-4 IP None N/A N/A N/A None 3-BDG-004-21-4 IP None N/A N/A N/A None 3-BDG-004-21-4 Split None N/A N/A N/A None 3.6-214 Rev. 30

MPS3 UFSAR TABLE 3.6-32 JET IMPINGEMENT EFFECTS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)

Essential Jet Jet Intensity at Jet Load on Protection Line Designation Break # (1) Impingement Target Distance to Target the Target the Target Requirement 3-BDG-004-22-4 IP None N/A N/A N/A None 3-BDG-004-22-4 IP None N/A N/A N/A None 3-BDG-004-22-4 Split None N/A N/A N/A None 3-BDG-004-23-4 IP None N/A N/A N/A None 3-BDG-004-23-4 IP None N/A N/A N/A None 3-BDG-004-23-4 Split None N/A N/A N/A None 3-BDG-004-24-4 IP None N/A N/A N/A None 3-BDG-004-24-4 IP None N/A N/A N/A None 3-BDG-004-24-4 Split None N/A N/A N/A None 3-BDG-003-49-4 IP None N/A N/A N/A None MPS3 UFSAR 3-BDG-003-48-4 IP None N/A N/A N/A None 3-BDG-003-50-4 IP None N/A N/A N/A None 3-BDG-003-51-4 IP None N/A N/A N/A None 3-BDG-003-49-4 TP None N/A N/A N/A None 3-BDG-003-48-4 TP None N/A N/A N/A None 3-BDG-003-50-4 TP None N/A N/A N/A None 3-BDG-003-51-4 TP None N/A N/A N/A None NOTES:

(1) Repetition of Break Numbers is used to identify separate fluid reservoirs which provide a constant pressure source to maintain system pressure subsequent to the postulated pipe break. These reservoirs maintain system blowdown during the transient event as well as in the steady state.

3.6-215 Rev. 30

MPS3 UFSAR (2) USNRC Generic Letter 87-11 eliminates the need to study jet impingement effects subsequent to Arbitrary Intermediate Breaks (AIBs). Therefore, no targets are evaluated.

(3) Steam Generator Blowdown (BDG) System - Wet Lay Up Piping is a High Energy System from each connection to the Steam Generators to the Normally Closed Valves 3BDG*V887, 3BDG*V889, 3BDG*V891, and 3BDG*V893. The Wet Lay Up Piping is separated from adjacent Steam Generators so any fluid jet impingement in one Steam Generator Cubicle does not damage essential structures, systems, or components associated with an adjacent Steam Generator based upon the Millstone Unit 3 Hazards Review Program Summary, NERM-069 Revision 1. The Wet Lay Up Piping is above the Floor Slab at Elevation 51'-4" which provides separation between the Reactor Coolant Pump Cubicles and the Upper Steam Generator Cubicles. Consequently fluid jet impingement effects on the Wet Lay Up Piping cannot propagate into a Loss of Coolant Accident (LOCA) based upon separation documented in the Millstone Unit 3 Hazards Review Program Summary, NERM-069, Revision 1.

MPS3 UFSAR 3.6-216 Rev. 30

MPS3 UFSAR TABLE 3.6-33 ENERGY ABSORBING CAPACITY OF A 4 INCH SCHEDULE 80 PIPE Overall Displacement d (in) Energy Absorbed Ep (in-k) Impact Force F (kips) 0.0 0.0 0.0 0.52 6.60 18.15 1.05 17.68 23.69 1.57 31.18 27.74 2.09 46.45 30.75 2.62 63.43 34.32 MPS3 UFSAR 3.6-217 Rev. 30

MPS3 UFSAR TABLE 3.6-34 POSTULATED BREAKS AUXILIARY FEEDWATER SYSTEM Line Designation Break # Building Elevation Break Type (1) Total Additive Stress Figure 3FWA*P1A Discharge Flange 1 ESF 26'-5" CB N/A Terminal End 3.6-32 3-FWA-006-2-3 2 ESF 26'-5" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-006-2-3 3 ESF 23'-2" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-006-2-3 4 ESF 22'-5" CB N/A Terminal End 3.6-32 3-FWA-006-8-3 5 ESF 22'-6" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-003-32-3 6 ESF 25'-11" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-003-39-3 7 ESF 25'-11" CB N/A Terminal End 3.6-32 3-FWA-003-39-3 8 ESF 25'-1" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-004-121-3 9 ESF 22'-6" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-004-121-3 10 ESF 22'-6" CB N/A Terminal End 3.6-32 3-FWA-004-42-2 13 Containment 43'-0" CB N/A Arbitrary Intermediate 3.6-32 MPS3 UFSAR 3-FWA-004-42-2 14 Containment 43'-0" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-006-135-2 15 Containment 58'-4" CB N/A Terminal End 3.6-32 3-FWA-003-33-3 16 ESF 25'-11" CB N/A Terminal End 3.6-32 3-FWA-003-33-3 17 ESF 25'-1" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-004-123-3 18 ESF 22'-6" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-004-123-3 19 ESF 22'-6" CB N/A Terminal End 3.6-32 3-FWA-004-36-2 22 Containment 43'-0" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-004-36-2 23 Containment 43'-0" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-006-100-2 24 Containment 58'-5" CB N/A Terminal End 3.6-32 3-FWA-004-125-3 25 ESF 22'-6" CB N/A Terminal End 3.6-32 3.6-218 Rev. 30

MPS3 UFSAR TABLE 3.6-34 POSTULATED BREAKS AUXILIARY FEEDWATER SYSTEM (CONTINUED)

Line Designation Break # Building Elevation Break Type (1) Total Additive Stress Figure 3-FWA-004-120-2 28 Containment 20'-0" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-004-120-2 29 Containment 20'-0" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-004-120-2 30 Containment 20'-0" CB N/A Terminal End 3.6-32 3-FWA-004-120-2 31 Containment 20'-0" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-004-120-2 32 Containment 21'-6" CB & LS Above Threshold 3.6-32 3-FWA-004-120-2 33 Containment 21'-6" CB N/A Terminal End 3.6-32 3-FWA-004-54-2 34 Containment 21'-6" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-006-102-2 35 Containment 52'-11" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-006-138-2 36 Containment 58'-4" CB N/A Terminal End 3.6-32 3-FWA-004-127-3 37 ESF 22'-6" CB N/A Terminal End 3.6-32 3-FWA-004-119-2 40 Containment 20'-5" CB N/A Arbitrary Intermediate 3.6-32 MPS3 UFSAR 3-FWA-004-119-2 41 Containment 21'-0" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-004-119-2 42 Containment 21'-0" CB N/A Terminal End 3.6-32 3-FWA-004-48-2 43 Containment 21'-0" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-006-101-2 44 Containment 21'-9" CB N/A Arbitrary Intermediate 3.6-32 3-FWA-006-137-2 45 Containment 58'-4" CB N/A Terminal End 3.6-32 3-FWA-006-79-3 46 ESF 22'-6" CB N/A Terminal End 3.6-32 3-FWA-004-117-2 47 Containment 43'-0" CB N/A Terminal End 3.6-32 3-FWA-004-118-2 48 Containment 43'-0" CB N/A Terminal End 3.6-32 3-FWA-004-119-2 49 Containment 20'-0" CB N/A Terminal End 3.6-32 3-FWA-004-120-2 50 Containment 20'-0" CB N/A Terminal End 3.6-32 3-FWA-004-41-2 51 ESF 43'-0" CB N/A Terminal End 3.6-32 3.6-219 Rev. 30

MPS3 UFSAR TABLE 3.6-34 POSTULATED BREAKS AUXILIARY FEEDWATER SYSTEM (CONTINUED)

Line Designation Break # Building Elevation Break Type (1) Total Additive Stress Figure 3-FWA-004-35-2 52 ESF 43'-0" CB N/A Terminal End 3.6-32 3-FWA-004-47-2 53 ESF 20'-0" CB N/A Terminal End 3.6-32 3-FWA-004-53-2 54 ESF 20'-0" CB N/A Terminal End 3.6-32 3-FWA-004-122-3 55 ESF 12'-0" CB N/A Terminal End 3.6-32 3-FWA-003-66-3 56 ESF 26'-0" CB N/A Terminal End 3.6-32 3-FWA-004-124-3 57 ESF 12'-0" CB N/A Terminal End 3.6-32 3-FWA-003-71-3 58 ESF 26'-0" CB N/A Terminal End 3.6-32 3-FWA-006-8-3 59 ESF 26'-6" CB N/A Terminal End 3.6-32 3-FWA-004-126-3 60 ESF 22'-0" CB N/A Terminal End 3.6-32 3-FWA-003-56-3 61 ESF 26'-0" CB N/A Terminal End 3.6-32 3-FWA-004-128-3 62 ESF 22'-0" CB N/A Terminal End 3.6-32 MPS3 UFSAR 3-FWA-003-61-3 63 ESF 26'-0" CB N/A Terminal End 3.6-32 3-FWA-008-14-3 64 ESF 22'-6" CB N/A Terminal End 3.6-32 3-FWA-008-14-3 65 ESF 26'-5" CB N/A Terminal End 3.6-32 3-FWA-004-160-3 66 ESF 25'-5" CB N/A Terminal End 3.6-32 NOTE:

(1) Circumferential Pipe Break (CB) and Longitudinal Pipe Split (LS) are defined in FSAR Section 3.6.2.1.3.

3.6-220 Rev. 30

MPS3 UFSAR TABLE 3.6-35 PIPE WHIP EFFECTS AUXILIARY FEEDWATER SYSTEM Line Designation Break # Blowdown Source Essential Pipe Whip Target Protection Requirement 3FWA*P1A Discharge Flange 1 None (1) None None 3-FWA-006-2-3 2 None (1) None None 3-FWA-006-2-3 3 None (1) None None 3-FWA-006-2-3 4 None (1) None None 3-FWA-006-8-3 5 None (1) None None 3-FWA-003-32-3 6 None (1) None None 3-FWA-003-39-3 7 None (1) None None 3-FWA-003-39-3 8 None (1) None None 3-FWA-004-121-3 9 None (1) None None MPS3 UFSAR 3-FWA-004-121-3 10 None (1) None None 3-FWA-004-42-2 13 None (1) None None 3-FWA-004-42-2 14 None (1) None None 3-FWA-006-135-2 15 None (1) None None 3-FWA-003-33-3 16 None (1) None None 3-FWA-003-33-3 17 None (1) None None 3-FWA-004-123-3 18 None (1) None None 3-FWA-004-123-3 19 None (1) None None 3.6-221 Rev. 30

MPS3 UFSAR TABLE 3.6-35 PIPE WHIP EFFECTS AUXILIARY FEEDWATER SYSTEM (CONTINUED)

Line Designation Break # Blowdown Source Essential Pipe Whip Target Protection Requirement 3-FWA-004-36-2 22 None (1) None None 3-FWA-004-36-2 23 None (1) None None 3-FWA-004-36-2 23 3-FWS-020-18-2 None None 3-FWA-006-100-2 24 None (1) None None 3-FWA-004-125-3 25 None (1) None None 3-FWA-004-120-2 28 None (1) None None 3-FWA-004-120-2 29 None (1) None None 3-FWA-004-120-2 30 None (1) None None 3-FWA-004-120-2 31 None (1) None None MPS3 UFSAR 3-FWA-004-120-2 32 Split None (1) None None 3-FWA-004-120-2 32 None (1) None None 3-FWA-004-120-2 33 None (1) None None 3-FWA-004-54-2 34 None (1) None None 3-FWA-004-54-2 34 3-FWS-020-26-2 None None 3-FWA-006-102-2 35 None (1) None None 3-FWA-006-102-2 35 3-FWS-020-26-2 None None 3-FWA-006-138-2 36 None (1) None None 3-FWA-004-127-3 37 None (1) None None 3.6-222 Rev. 30

MPS3 UFSAR TABLE 3.6-35 PIPE WHIP EFFECTS AUXILIARY FEEDWATER SYSTEM (CONTINUED)

Line Designation Break # Blowdown Source Essential Pipe Whip Target Protection Requirement 3-FWA-004-119-2 40 None (1) None None 3-FWA-004-119-2 41 None (1) None None 3-FWA-004-119-2 42 None (1) None None 3-FWA-004-48-2 43 None (1) None None 3-FWA-004-48-2 43 3-FWS-020-22-2 None None 3-FWA-006-101-2 44 None (1) None None 3-FWA-006-101-2 44 3-FWS-020-22-2 None None 3-FWA-006-137-2 45 None (1) None None 3-FWA-006-79-3 46 None (1) None None 3-FWA-004-117-2 47 None None MPS3 UFSAR None (1) 3-FWA-004-118-2 48 None (1) None None 3-FWA-004-119-2 49 None (1) None None 3-FWA-004-120-2 50 None (1) None None 3-FWA-004-41-2 51 None (1) None None 3-FWA-004-35-2 52 None (1) None None 3-FWA-004-47-2 53 None (1) None None 3-FWA-004-53-2 54 None (1) None None 3-FWA-004-122-3 55 None (1) None None 3.6-223 Rev. 30

MPS3 UFSAR TABLE 3.6-35 PIPE WHIP EFFECTS AUXILIARY FEEDWATER SYSTEM (CONTINUED)

Line Designation Break # Blowdown Source Essential Pipe Whip Target Protection Requirement 3-FWA-003-66-3 56 None (1) None None 3-FWA-004-124-3 57 None (1) None None 3-FWA-003-71-3 58 None (1) None None 3-FWA-006-8-3 59 None (1) None None 3-FWA-004-126-3 60 None (1) None None 3-FWA-003-56-3 61 None (1) None None 3-FWA-004-128-3 62 None (1) None None 3-FWA-003-61-3 63 None (1) None None 3-FWA-008-14-3 64 None (1) None None MPS3 UFSAR 3-FWA-008-14-3 65 None (1) None None 3-FWA-004-160-3 66 None (1) None None NOTE:

(1) Sustained blowdown of the Auxiliary Feedwater System does not occur. Dual inline check valves prevent backflow from the Main Feedwater System and the Auxiliary Feedwater Pumps have limited capacity to sustain the system pressure. Therefore, pipe whip does not occur.

3.6-224 Rev. 30

MPS3 UFSAR TABLE 3.6-36 JET IMPINGEMENT EFFECTS AUXILIARY FEEDWATER SYSTEMS Essential Jet Distance to Jet Intensity Jet Load on Protection Line Designation Break # Impingement Target Target at the Target the Target Requirement 3FWA*P1A Discharge Flange 1 None (1) N/A N/A N/A None 3-FWA-006-2-3 2 None (1) N/A N/A N/A None (2) 3-FWA-006-2-3 3 None (1) N/A N/A N/A None (2) 3-FWA-006-2-3 4 None (1) N/A N/A N/A None 3-FWA-006-8-3 5 None (1) N/A N/A N/A None (2) 3-FWA-003-32-3 6 None (1) N/A N/A N/A None (2) 3-FWA-003-39-3 7 None (1) N/A N/A N/A None 3-FWA-003-39-3 8 None (1) N/A N/A N/A None (2) 3-FWA-004-121-3 9 None (1) N/A N/A N/A None (2)

MPS3 UFSAR 3-FWA-004-121-3 10 None (1) N/A N/A N/A None 3-FWA-004-42-2 13 None (1) N/A N/A N/A None (2) 3-FWA-004-42-2 14 None (1) N/A N/A N/A None (2) 3-FWA-006-135-2 15 None (1) N/A N/A N/A None 3-FWA-003-33-3 16 None (1) N/A N/A N/A None 3-FWA-003-33-3 17 None (1) N/A N/A N/A None (2) 3-FWA-004-123-3 18 None (1) N/A N/A N/A None (2) 3-FWA-004-123-3 19 None (1) N/A N/A N/A None 3.6-225 Rev. 30

MPS3 UFSAR TABLE 3.6-36 JET IMPINGEMENT EFFECTS AUXILIARY FEEDWATER SYSTEMS (CONTINUED)

Essential Jet Distance to Jet Intensity Jet Load on Protection Line Designation Break # Impingement Target Target at the Target the Target Requirement 3-FWA-004-36-2 22 None (1) N/A N/A N/A None (2) 3-FWA-004-36-2 23 None N/A N/A N/A None (2) 3-FWA-006-100-2 24 None (1) N/A N/A N/A None 3-FWA-004-125-3 25 None (1) N/A N/A N/A None 3-FWA-004-120-2 28 None (1) N/A N/A N/A None (2) 3-FWA-004-120-2 29 None (1) N/A N/A N/A None (2) 3-FWA-004-120-2 30 None (1) N/A N/A N/A None 3-FWA-004-120-2 31 None (1) N/A N/A N/A None (2) 3-FWA-004-120-2 32 Split None (1) N/A N/A N/A None MPS3 UFSAR 3-FWA-004-120-2 32 None (1) N/A N/A N/A None 3-FWA-004-120-2 33 None (1) N/A N/A N/A None 3-FWA-004-54-2 34 None N/A N/A N/A None (2) 3-FWA-006-102-2 35 None N/A N/A N/A None (2) 3-FWA-006-138-2 36 None (1) N/A N/A N/A None 3-FWA-004-127-3 37 None (1) N/A N/A N/A None 3-FWA-004-119-2 40 None (1) N/A N/A N/A None (2) 3-FWA-004-119-2 41 None (1) N/A N/A N/A None (2) 3.6-226 Rev. 30

MPS3 UFSAR TABLE 3.6-36 JET IMPINGEMENT EFFECTS AUXILIARY FEEDWATER SYSTEMS (CONTINUED)

Essential Jet Distance to Jet Intensity Jet Load on Protection Line Designation Break # Impingement Target Target at the Target the Target Requirement 3-FWA-004-119-2 42 None (1) N/A N/A N/A None 3-FWA-004-48-2 43 None N/A N/A N/A None (2) 3-FWA-006-101-2 44 None N/A N/A N/A None (2) 3-FWA-006-137-2 45 None (1) N/A N/A N/A None 3-FWA-006-79-3 46 None (1) N/A N/A N/A None 3-FWA-004-117-2 47 None (1) N/A N/A N/A None 3-FWA-004-118-2 48 None (1) N/A N/A N/A None 3-FWA-004-119-2 49 None (1) N/A N/A N/A None 3-FWA-004-120-2 50 None (1) N/A N/A N/A None MPS3 UFSAR 3-FWA-004-41-2 51 None (1) N/A N/A N/A None 3-FWA-004-35-2 52 None (1) N/A N/A N/A None 3-FWA-004-47-2 53 None (1) N/A N/A N/A None 3-FWA-004-53-2 54 None (1) N/A N/A N/A None 3-FWA-004-122-3 55 None (1) N/A N/A N/A None 3-FWA-003-66-3 56 None (1) N/A N/A N/A None 3-FWA-004-124-3 57 None (1) N/A N/A N/A None 3-FWA-003-71-3 58 None (1) N/A N/A N/A None 3.6-227 Rev. 30

MPS3 UFSAR TABLE 3.6-36 JET IMPINGEMENT EFFECTS AUXILIARY FEEDWATER SYSTEMS (CONTINUED)

Essential Jet Distance to Jet Intensity Jet Load on Protection Line Designation Break # Impingement Target Target at the Target the Target Requirement 3-FWA-006-8-3 59 None (1) N/A N/A N/A None 3-FWA-004-126-3 60 None (1) N/A N/A N/A None 3-FWA-003-56-3 61 None (1) N/A N/A N/A None 3-FWA-004-128-3 62 None (1) N/A N/A N/A None 3-FWA-003-61-3 63 None (1) N/A N/A N/A None 3-FWA-008-14-3 64 None (1) N/A N/A N/A None 3-FWA-008-14-3 65 None (1) N/A N/A N/A None 3-FWA-004-160-3 66 None (1) N/A N/A N/A None NOTES: MPS3 UFSAR (1) Sustained blowdown of the Auxiliary Feedwater System does not occur. Dual inline check valves inside containment prevent backflow from the Main Feedwater System and the Auxiliary Feedwater Pumps have limited capacity to sustain the system pressure.

(2) USNRC Generic Letter 87-11 eliminates the requirement to evaluate the jet impingement effects of Arbitrary Intermediate Breaks (AIBs). Therefore, jet impingement effects subsequent to AIBs are not evaluated.

3.6-228 Rev. 30

tions whose identification number includes the letter B contain material within nce-of-plant (BOP) scope, while sections whose identification number includes the letter N tain material within the nuclear steam supply system (NSSS) scope (including material aining to the reactor coolant loop piping out to the first weld of any connecting piping).

3.7-1 Rev. 30

B.1.1 Design Response Spectra horizontal design response spectra used for seismic analysis are shown on Figures 3.7B-1 and B-2. The spectra for the safe shutdown earthquake (SSE) correspond to a maximum rock eleration of 0.17g, and the spectra for the operating basis earthquake (OBE) correspond to a imum rock acceleration of 0.09g. These spectra are based on seismic analysis data rmined during the construction permit stage. The vertical design response spectra are taken to wo-thirds of the horizontal design response spectra. Regulatory Guide 1.60 does not apply as cribed in Section 1.8.

B.1.2 Design Time History horizontal design time histories produce spectra which envelop the horizontal design onse spectra. These spectra are shown on Figures 3.7B-3 through 3.7B-8. The vertical design e histories are taken to be two-thirds of the horizontal design time histories for both the SSE the OBE.

ere soil structure interaction analysis is performed, the time history is applied at bedrock in the field. Where analysis is performed using subgrade springs to represent rock stiffness, the time ory is applied at the foundation level.

horizontal design time history is checked by comparing its spectral values with those of the design response spectra at 250 oscillator periods between 0.0167 second and 5.0 seconds. The llator periods are distributed logarithmically according to the following expression:

Ti = Ti-1 (3.7B-1)

= (T250/T1)1/249 = (5/0.0167)1/249

= 1.023 re:

Ti-1 and Ti = Any two consecutive oscillator periods Ti/Ti-1 = The period ratio (ratio between the ith period, Ti, and the (i-1) period, Ti-1).

B.1.3 Critical Damping Values values of the percentage of critical damping used in the analysis of Seismic Category I ctures, systems, and components depends on the stress levels resulting from the seismic input ion (SSE or OBE) used in the analysis.

values of damping used for the input motion are listed in Table 3.7B-1. Structural damping is gned to be 2.0 percent for the OBE and 5.0 percent for the SSE.

higher damping values (Table 3.7B-1) are used only where justified by detailed study (either ing or calculated stress levels). Damping values utilized in the analysis of Seismic Category I 3.7B-2 Rev. 30

B.1.4 Supporting Media for Seismic Category I Structures founding materials for major plant structures are listed in Table 2.5.4-14. Most of the major ty related structures are founded on bedrock, with the exception of the control building, rgency diesel generator building, and the hydrogen recombiner building. The control building unded on 1 to 4 feet of compacted structural backfill overlying basal till of thickness varying ween 1 foot on the east side and 15 feet on the west. The emergency diesel generator building unded on basal till varying in thickness from less than 10 feet to 30 feet overlying bedrock.

hydrogen recombiner is founded on concrete fill overlying bedrock.

rge portion of the circulating water discharge tunnel and the service water intake lines are nded on bedrock. However, some sections are founded on soil, particularly near the intake and harge points in the vicinity of Niantic Bay. When soil was encountered as a founding material, nsuitable overburden was removed to sound ablation or basal till. In the event that the invert ation was higher than the excavated grade, compacted structural backfill was placed in thin to the subgrade elevation in accordance with procedures described in Section 2.5.4.5.2.

escription of each of the founding materials at the site is presented in Section 2.5.4.2. The mic velocities and moduli values of these materials are presented in Section 2.5.4.4.3. The perties of structural backfill materials are described in Section 2.5.4.5.2. The use of these perties in the analysis of soil-structure interaction is discussed in Section 3.7B.2.4.

B.2 SEISMIC SYSTEM ANALYSIS mic systems discussed in this section are those major Category I structures that are considered onjunction with foundation media in forming a soil-structure interaction model. Other systems components are considered seismic subsystems (Sections 3.7B.3 and 5.4.14).

B.2.1 Seismic Analysis Methods erally, analysis of seismic systems were performed using either the modal analysis response ctra method or the modal analysis time history method. Response spectrum analysis uses the ral frequencies, mode shapes, and weighted modal dampings to determine the maximum mic response of multi-degree-of-freedom systems with lumped masses and elastic connecting mbers. Time history modal analysis uses the same free vibration characteristics and damping ors as the spectrum analysis, primarily to determine acceleration response spectra at selected tions within the structure. Those buildings that required finite element soil-structure raction were analyzed using the frequency domain time history method (Section 3.7B.2.4).

methods used for seismic analysis of particular Seismic Category I structures are summarized able 3.7B-2.

hematical models of three buildings representative of Seismic Category I structures are ussed in detail. Models selected are the containment structure (Figure 3.7B-9), the main steam e building (Figure 3.7B-10), and the emergency generator enclosure (Figures 3.7B-11 and B-12). These buildings are described in Sections 3.8.1 and 3.8.3.

3.7B-3 Rev. 30

-structure interaction analysis was performed. In this case, a planar model was employed in h of the horizontal direction. All other Seismic Category I structures were analyzed sidering six degrees-of-freedom; three translation, two rocking, and one torsional.

se buildings that were modeled using finite element soil structure interaction techniques uded the control building and the emergency generator enclosure. The computer program for e analyses was PLAXLY. Structures founded on rock, which are discussed in Section B.2.4, were analyzed using discrete springs suggested by Richart et al., 1970, Whitman et al.,

7, Wass 1972, and Kausel et al., 1975. Springs associated with the degrees-of-freedom in a bal orthogonal coordinate system were used.

eveloping the lumped-mass models for Seismic Category I structures, an adequate number of rees-of-freedom was ensured by providing a lumped mass at the mat, floor, and roof ations. Additional masses were used when dictated by changes in geometry or stiffness. In the s of the containment shell, eight lumped masses were selected at intervals uniformly ributed along the height of the shell. Studies have been conducted which show the difference ase shear and overturning moment between a 3-mass and a 15-mass model to be less than 3 ent. Comparisons between 3-, 5-, 10-, and 15-mass models have shown a difference in natural uency for the first two modes of less than 1 percent.

analysis of the models, twice the number of modes with frequencies less than 33 Hz were sidered.

h Seismic Category I building has a foundation mat which serves as a single structural port, except the emergency generator enclosure which has spread footings that serve the same pose. Two inch (above grade) and 1 inch (below grade) shake spaces, larger than any predicted cture-to-structure displacements, are provided between all Seismic Category I structures. In ition, 4 inch shake spaces are provided between all Seismic Category I structures and the tainment structure. Relative displacements were considered for all Category I systems and ponents which run between structures. Seismic effects such as hydrodynamic loads and linear responses were considered where appropriate (Section 2.5.4).

B.2.2 Natural Frequencies and Response Loads ponse spectrum analyses for the containment structure and the main steam valve building, are marized below, as well as the finite element soil structure interaction analysis of the rgency generator enclosure. They are representative samples of Category I structures.

B.2.2.1 Containment Summary response spectrum analysis of the containment and internal structures included the eloping of two discrete analyses. The dynamic model of the containment (Figure 3.7B-9) was lyzed with cracked and uncracked concrete properties.

significant natural frequencies and modal participation factors resulting from the uncracked cracked analyses are presented in Tables 3.7B-3 and 3.7B-4. Significant mode shapes lting from these analyses are presented in Tables 3.7B-5 and 3.7B-6 and the response loads ined by the square root of the sum of the squares method (SRSS) are summarized in les 3.7B-7 and 3.7B-8. Table 3.7B-9 presents the degrees-of-freedom corresponding to x, y, 3.7B-4 Rev. 30

obtained by taking the absolute sum of the SRSS modal responses due to each direction of itation and enveloping the cracked and uncracked analysis (Tables 3.7B-10 and 3.7B-11).

amplified response spectra (ARS) were produced using the modal analysis time history hod. Typical ARS resulting from the enveloping of the cracked and uncracked analyses are vided at the mat, steam generators support slab, top of the primary shield wall, operating floor crane wall of internal structure, as well as the springline and dome apex of containment cture for horizontal and vertical safe shutdown earthquake (SSE) excitations (Figures 3.7B-13 ugh 3.7B-33).

B.2.2.2 Main Steam Valve Building Summary dynamic model of the main steam valve building (Figure 3.7B-10) was excited separately by vertical and two orthogonal horizontal excitations. The resulting natural frequencies, icipation factors, and mode shapes are listed in Tables 3.7B-12 and 3.7B-13. The closely ced modes (CSM) modal responses are given in Tables 3.7B-14 and 3.7B-15. The CSM modal bination method is explained in Section 3.7B.2.7. Final lumped mass accelerations and lacements are obtained by taking the absolute sum of the CSM modal responses due to each ction of excitation (Table 3.7B-16). For the main steam valve building analysis, le 3.7B-17 presents the relationship between the degrees-of-freedom and the motion of the ped masses given in Tables 3.7B-14, 3.7B-15, and 3.7B-16.

mples of the ARS resulting from each direction of excitation produced using the modal lysis time history method are provided at the mat (Figures 3.7B-34 thru 3.7B-36), the floor at el 41 feet 0 inch (Figures 3.7B-37 thru 3.7B-39), and the roof slab at el 85 feet 4 inches ures 3.7B-40 through 3.7B-42).

B.2.2.3 Emergency Generator Enclosure Summary seismic analysis of the emergency generator enclosure used a lumped mass finite element del (Section 3.7B.2.4). The dynamic model shown on Figures 3.7B-11 and 3.7B-12 was ited separately by one vertical and two orthogonal horizontal excitations. Since the method d for the analysis was a frequency domain solution rather than a modal analysis, no natural uencies, participation factors, or mode shapes are tabulated. The solution gives transfer ctions which represent the ratio of the amplitude of any particular mass point of the structure he amplitude of the input motion at bedrock. Typical transfer functions (including imaginary

) for each mass point of the dynamic model are presented on Figures 3.7B-43 thru 3.7B-51 for h direction of excitation. Final absolute sum response loads are presented in Table 3.7B-18 and B-19. Also, since diesel generators are located on independent foundations, a separate analysis performed for them.

ical ARS at elevation 24 feet 6 inches, 51 feet 0 inches, and 66 feet 0 inches, produced using frequency domain time history method, are shown on Figures 3.7B-52 thru 3.7B-60. ARS lting from the independent analysis of the diesel generator isolation mats are presented on ures 3.7B-61 and 3.7B-62. Examples of the input time history at bedrock of the plant site and resulting time history at the base of the structure are shown on Figures 3.7B-63 and 3.7B-64.

3.7B-5 Rev. 30

dynamic model of a Seismic Category I structure is constructed to obtain a satisfactory esentation of the dynamic behavior of the structure. Major Seismic Category I structures that considered in conjunction with foundation media in forming a soil-structure interaction model defined as seismic systems. Other Seismic Category I systems and components that are not gnated as seismic systems are considered as seismic subsystems (R). In most cases, ipment and components come under the definition of seismic subsystems and are analyzed as coupled system from the primary structure. To define criteria for decoupling subsystems, Rm, mass ratio and Rf, the frequency ratio are significant. Rm and Rf are defined as:

R m = Total mass of the supported subsystem-

(3.7B-2)

Mass that supports the subsystem R f = Fundamental frequency of the support subsystem-

(3.7B-3)

Frequency of the dominant support motion following criteria are used.

1. If Rm < 0.01, decoupling is acceptable for any Rf,
2. If 0.01 Rm 0.1, decoupling is acceptable if Rf 0.8 or if Rf 1.25,
3. If Rm > 0.1, an approximate model of the subsystem is included in the primary system model.

e subsystem is comparatively rigid and also rigidly connected to the primary system, only the s of the subsystem is included at the support point in the primary system model. In the case of bsystem supported by very flexible connections (e.g., a pipe supported by hangers), the system is not included in the primary model; however additional mass is considered.

ost cases, the equipment and components which come under the definition of subsystems e analyzed separately from the primary structure, and the seismic input for the subsystem was ained from the analysis of the structure. One important exception to this procedure is the lysis of the reactor coolant system (RCS), which was considered to be a subsystem but was lyzed using a coupled model of the RCS and primary structure with ground motion as the mic input (Section 5).

dynamic model of a seismic system consists of a set of lumped masses, generally having six rees-of-freedom per mass connected by weightless elastic members. Masses are usually ped at floor levels and include the masses of the floors, walls, columns, equipment, and ng. The floors are treated as rigid diaphragms that transfer the earthquake inertia forces to es and shearwalls, which in turn transfer the loads to the foundation mat and the subgrade.

m theory, combining the effects of shear, flexure, torsion, and axial deformation, is used to blish the stiffness characteristics of the frame-wall systems. Eccentricities between the centers ass (CM) and centers of rigidity (CR) are considered.

an example, the lumped mass model of the containment is shown on Figure 3.7B-9. The model constructed so that it properly represents the free vibration of a cantilevered structure in shear 3.7B-6 Rev. 30

esented by springs as discussed in Section 3.7B.2.4. Masses M10 through M17 represent the e and the cylindrical portion of the containment shell and the dome. M1 consists of the mat portions of the walls and columns which are attached to the mat. The internal structure, sisting of equipment, primary shield wall, cubicle walls, crane wall, etc., were modeled by ses M2 through M9.

B.2.4 Soil-Structure Interaction supporting media for Seismic Category I structures has been discussed in Section 3.7B.1.4.

ndicated in Table 2.5.4-14, Seismic Category I structures are founded directly on bedrock h a few structures founded on shallow soil overburden over rock.

te element soil-structure interaction analysis is required for shallowly embedded structures on low soil overburden over rock (NRC 1975). The control building and the emergency erator enclosure were analyzed using the finite element method.

the finite element analysis, the emergency generator enclosure lumped mass model ction 3.7B.2.3) was attached to a finite element representation of the soil subgrade ures 3.7B-11 and 3.7B-12). These models were then subjected to horizontal and vertical itations by computer program, PLAXLY. The bottom boundary was taken at bedrock and the boundaries used PLAXLYs energy transmitting boundary corresponding to layers of soil nding laterally to infinity. The material properties of the soil were assumed to be linearly oelastic, and the dynamic equations were solved in the frequency domain using Fourier sformation techniques. Nonlinear behavior in the subgrade was accounted for by the use of computer program SHAKE, which determines the stain corrected soil properties.

tion 3A.1.7 of Appendix 3A give a detailed discussion of PLAXLY.

structures founded directly on bedrock analyses using discrete subgrade springs based on ations suggested by Richart et al. (1970), Whitman et al. (1967), Wass (1972), and Kausel

l. (1975) were performed. A value of 10 percent (translational) and 5 percent (rotational) imum subgrade damping assured the conservatism of the discrete spring method.

B.2.5 Development of Floor Response Spectra or response spectra were developed separately for one vertical and two mutually perpendicular zontal earthquake motions using the modal analysis time history method discussed in tion 3.7B.2.1 for all buildings except the emergency generator enclosure. Floor response ctra for the emergency generator enclosure were developed using the modal analysis method n PLAXLY (Section 3.7B.2.4).

mbination of the three components of floor response spectra for use in subsystem seismic lysis is discussed in Section 3.7B.3.

3.7B-7 Rev. 30

cedures to combine three components of earthquake motion to determine the response of mic systems meet the recommendations of Regulatory Guide 1.92, Rev. 1, to the extent cribed in Section 1.8.1.92.

B.2.7 Combination of Modal Responses eneral, for those Seismic Category I systems analyzed by the modal analysis response ctrum method, the influence of closely spaced modes was considered by use of the double method, described in Regulatory Guide 1.92, Rev. 1, to the extent described in tion 1.8.1.92 and in Singh et al. (1973).

B.2.8 Interaction of Non-Category I Structures with Seismic Category I Structures acent Seismic Category I structures are separated from Non-Category I structures by a 1-inch ke space below grade and a 2-inch Shake space above grade.

-Category I structures are sufficiently remote from Seismic Category I structures to preclude raction or are designed (Section 3.8.3) to avoid collapse during an SSE, thus preventing ificant damage to adjacent Category I structures.

B.2.9 Effects of Parameter Variations on Floor Response Spectra rder to consider the effects of variations of structural properties, dampings, and soil perties, peak resonant period values of the floor response spectra were spread +15 and -15 ent. For the containment structure which experiences substantial cracking during the ctural acceptance test, floor response spectra for the cracked and uncracked cases were eloped.

use of floor time histories in subsystem analysis is discussed in Section 5.4.14.

B.2.10 Use of Constant Vertical Static Factors amic vertical responses were calculated in all seismic system analyses, precluding the need to constant vertical static factors.

B.2.11 Method Used to Account for Torsional Effects emergency generator building was analyzed using planar models for two horizontal ctions using three degrees-of-freedom per mass. Since this building is basically symmetrical, approach is justified. All other Seismic Category I structures were analyzed dynamically g six degrees-of-freedom per mass. Effects of eccentricity between centers of mass and ters of stiffness or soil contact were included in determining appropriate member stiffnesses in dynamic models.

3.7B-8 Rev. 30

discussed in Section 3.7B.2.1, the results of the response spectrum analysis were used for the gn of most Seismic Category I structures and the results of the modal time history analyses vided the floor time histories used to create floor response spectra. The maximum responses ined from the time history analysis are compared with those resulting from the response ctrum technique in Tables 3.7B-20 through 3.7B-22 for the containment structure and main m valve building.

higher time history results are attributable to the fact that the artificial ground time history ted to envelop the Millstone 3 ground response spectra is quite conservative. The extent of conservatism is illustrated on Figures 3.7B-3 through 3.7B-8 which compares the Millstone 3 und response spectra with the response spectra generated by using the site artificial time ory as the forcing function. However, because of the method used in combining structural onses from the response spectrum analysis due to each of the three components of earthquake ion (absolute sum), a majority of the response accelerations from the response spectrum hod are larger than those from the time history analysis.

B.2.13 Methods for Seismic Analysis of Category I Dams re are no Category I dams at this site.

B.2.14 Determination of Seismic Category I Structure Overturning Moments rturning moments for Seismic Category I structures were calculated using inertia forces based he accelerations summarized in Section 3.7B.2.2 as follows:

n n (3.7B-4)

OM = Mi ahi Hi + Mi avi Di + Ma i=1 i=1 re OM = Overturning moment i = Mass point on structure Mi = Mass at i ahi = Horizontal acceleration at i in the direction being investigated avi = Vertical acceleration at i Hi = Height of point i above bottom of foundation mat Di = Distance from point i to edge of foundation mat in the direction of acceleration 3.7B-9 Rev. 30

r directions of horizontal acceleration were considered: north, south, east, and west. The rturning moment was calculated for each direction of horizontal acceleration assuming the ical acceleration to act either up or down. The maximum moment calculated from the erent combinations of the horizontal and vertical accelerations was then divided into the sting moment based on the deadweight of the structure to determine the factor of safety inst overturning. The minimum factors of safety allowed are summarized in Section 3.8.5.5.

B.2.15 Analysis Procedure for Damping equivalent viscous modal damping, reflecting the different damping rates in various portions he structure, was computed for use in structural dynamic analysis. The modal damping ratio is eighted average of member and support spring damping, based on the contribution of each to total strain energy of the mode shape. The method is based on work by Roesset et al. (1973).

following discussion is meant to serve as a general description. The more theoretical vations (Roesset et al., 1973, Whitman 1970) are found in the literature.

Damping and Strain Energy Methods important factor in determining structural response is the damping phenomenon. Two types of ping are generally recognized: viscous (in which the energy dissipated per cycle is portional to frequency) and hysteretic (in which no frequency dependence is seen). Most ctural elements display hysteretic behavior, while supporting soils appear to combine both teretic and viscous damping mechanisms.

certain applications (e.g., pipe breaks) in which foundation motion can be neglected, only teretic damping need be considered. Whitmans (1973) analysis of Biggs formula gives a ul approximation for the damping of each mode when material damping varies from element lement. His expression for the equivalent viscous modal damping is obtained by a in-energy weighting of element damping:

N j

Di E i j i=1 B eqv = ---------------------

N

- (3.7B-5) j Ei i=1 re:

v = Equivalent viscous damping ratio (fraction of critical) for structure vibrating in mode j, Number of elements, Hysteretic damping ratio for element i, 3.7B-10 Rev. 30

articular, when damping is uniform (i.e., Di = D) then Beqv = D for all modes. When damping ot uniform, modal damping is weighted toward those elements which make the largest tribution to the energy of each mode. In other applications (e.g., earthquakes) in which ndation motion is significant, viscous damping also must be considered. Current practice treats soil damping as viscous for translational motion, and hysteretic for rotational motion. Roesset

l. (1973) extended Biggs formula to include the viscous damping contributions of the soil:

N Nv j

j W- j Di E i + -------Wk B k Ek j i 1 k 1 B eqv = -------------------------------------------------------------

N

- (3.7B-6)

N v j j Ei + Ek i=1 i=1 re:

NH = Number of hysteretically damped elements, NV = Number of viscously damped elements, Wj = Frequency of structure mode j (radians per second),

WK = Frequency of element k (radians per second),

BK = Critical damping ratio of element k at frequency k.

square of the natural frequency w of the soil element, is equal to the ratio of the soil spring ness to total mass of the structure plus foundation.

s formula reflects the fact that the energy dissipation per cycle by the viscous mechanism is portional to frequency of motion. Any element which displays both hysteretic and viscous ping appears in both summations of the numerator, but is not repeated in the denominator.

h element strain energy appearing in Equation 3.7B-4 is evaluated from the element stiffness rix and the displacement of the elements boundary joints. After comparing the Biggs and sset modal damping ratios calculated from Equations 3.7B-5 and 3.7B-6, the lower value for h mode is selected for use in the dynamic analysis, thus assuring that the composite modal ping value never exceeds the hysteretic damping value. In no case do modal damping values eed 10 percent.

3.7B-11 Rev. 30

B.3.1 Seismic Analysis Methods B.3.1.1 Equipment and Components mic Category I equipment and components are documented for seismic adequacy. The basic rce of seismic design data is either the ground response spectra, the amplified response ctra, floor, or mat time history, derived through a dynamic analysis of the relevant structure ctions 3.7B.2.5 and 3.7B.2.9).

ee principal methods, which include combinations of these methods, are used for documenting quacy for Seismic Category I equipment and components:

1. Static analysis
2. Dynamic analysis
3. Testing effects of supports are reflected in the input to seismically qualified equipment and ponents. General stress limits are given in Section 3.9B.3.1. Specific component stress wables are given in Tables 3.9B-5, 3.9B-6, and 3.9B-7. Such limits either conform to, or are e conservative than, those of ASME Section III, Subsection NF.

oratory tests are performed on Seismic Category I mechanical and electrical equipment, and plex instrumentation that cannot be modeled to predict response correctly. Conversely, lytical methods described in this section are employed when mathematical modeling niques are used or when equipment characteristics (e.g., size, rating) preclude laboratory ing. Combinations of analysis and testing are also employed to assure adequacy of Seismic egory I equipment.

safe shutdown earthquake (SSE) in combination with the faulted loads are the load basis ch assure the structural integrity of Seismic Category I equipment. The operating basis hquake (1/2 SSE) in conjunction with the operating loadings are used to assure continued ration of the equipment and components.

ipment vendors and suppliers are required to formulate programs for qualification of Seismic egory I equipment in accordance with specification requirements. Documentation of the er's qualification program is reviewed and approved by the Applicant.

B.3.1.1.1 Static Analysis ic analysis is used for equipment and components that can be characterized as relatively ple structures. This type of analysis involves the multiplication of the equipment, or ponent, total weight by the specified seismic acceleration (direction dependent loading), to duce forces that are applied at the center of gravity in the horizontal and vertical directions. A ss analysis of equipment components, such as supports, holddown bolts, and other structural mbers, is performed to determine their adequacy.

3.7B-12 Rev. 30

uency. The relevant response curves are reviewed to determine a cutoff frequency which nds the rigid range from the resonance range of the response curves. Equipment and ponents having fundamental natural frequencies above the cutoff frequency of the relevant onse curve are analyzed to rigid range response accelerations.

components or equipment having a fundamental natural frequency below the cutoff uency, the accelerations used in static analysis are 1.3 times the peak acceleration value, as cated in the amplified response curve for simply supported conditions. For equipment having ti-support points in a single plane, the resonant response acceleration is multiplied by 1.5.

h of the three defined directions of earthquake input (two horizontal and one vertical taken ogonally) is evaluated separately. Horizontal and vertical seismic loads are combined using square root of the sum of the squares. Equipment is designed to withstand the combined cts of normal operating loads, acting simultaneously with the earthquake loadings, without of safety function or structural integrity.

B.3.1.1.2 Dynamic Analysis etailed dynamic analysis is performed when equipment complexity, or dynamic interaction, ludes static analysis, or when static analysis is too conservative. Dynamic analysis methods ude:

1. Response spectrum modal analysis
2. Time-history by modal superposition
3. Time-history by numerical integration response spectrum modal analysis technique is used most commonly.

Modeling lumped mass, or the consistent mass, approach is employed in the dynamic analysis. In the ped and the consistent mass idealizations, the main structure is divided into substructures and masses of these substructures are concentrated at a number of discrete points. The nature of e substructures, and the stiffness properties of the corresponding modeling elements, rmine the minimum spacing of the mass points and the degrees of freedom to be associated h each point. In accordance with minimum spacing requirements, the analyst can then choose, the model, particular mass points which reflect predominant masses of subcomponents which tribute significantly to the total response.

deling of equipment for dynamic analysis starts with the calculation of lumped masses at rete stations and the evaluation of the elastic properties (or stiffness) of connecting members.

number of discrete mass points is selected to adequately describe the dynamic characteristics natural frequencies of the equipment. General modeling guidelines indicate that good natural uency characteristics are obtained when the number of discrete mass or node points selected wice the number of the highest mode of interest. Papers prepared by Lin and Hadjian (1976),

3.7B-13 Rev. 30

Response Spectrum Modal Analysis normal mode approach is employed for seismic analysis of equipment and components.

ural frequencies, eigenvectors, participation factors, and modal member-end forces and ments of the undamped structure are calculated. The system of equations which describes the vibrations of an n-degree of freedom, undamped structure is:

[ M ] { X** } + [ K ] { X } = 0 (3.7B.3-1) re:

[M] = Mass matrix for assembled system

[K] = Stiffness matrix for assembled system

{ X** } = Nodal acceleration vector

{X} =Nodal displacement vector mode shapes and frequencies are solved in accordance with:

2

( [ K ] - n [ M ] ) [ ]n = 0 n = 1, 2, 3, N (3.7B.3-2) re:

n = Natural frequency of the nth mode

[]n = Mode shape vector for the nth mode n = Number of significant modes considered envector-eigenvalue extraction techniques, such as Householder-QR, Jacobi Reduction, and erse Iteration, are used, depending upon the total number of dynamic degrees of freedom and number of modes desired.

modal participation factor for the nth mode specific direction, i, is defined by T

ni = [--------------------------------

] [ M ] [ -] (3.7B.3-3)

T

[] [M][]

re:

[]T = Transpose of mode shape vector for the nth mode

[] = Earthquake direction vector referring to direction i modal member-end forces and moments are determined by 3.7B-14 Rev. 30

re:

[Km] = Member stiffness matrix each modal frequency, the corresponding response acceleration is determined for a given l of equipment damping from the applicable response curve.

maximum response for each mode is found by computing:

{ X** } n = ni R ni [ ] n 1

{ X* } = ------ { X** } n n

(3.7B.3-5) 1 - **

{ X } n = --------

2

{ X }n n

n R ni

{ F } n = -------------

2

- { Fn }n n

re:

{ X** } = The modal acceleration

{ X* } = The velocity

{X} = The displacement

{F} = The member-end force and moment vectors Rni = The spectral acceleration for the nth mode basis for the combination of maximum modal response is discussed in Section 3.7B.3.7.

Time-History Methods re are two separate approaches to the solution of the equations of dynamic equilibrium for e-history motions. The modal superposition method involves the modal solution of the free ation response of the system, and transformation to normal coordinates using the mode shapes he system. This procedure uncouples the equations of motion so that the response of the em in each individual mode may be evaluated independently. Total system response is rmined by combining responses of individual modes oscillating simultaneously. The second hod of time-history analysis is the direct integration solution that includes numerical gration of the simultaneous differential equations of dynamic equilibrium without sformation to normal coordinates. System response at each time point is evaluated by this nique.

3.7B-15 Rev. 30

puter program. Equations of motion for time-history solutions are contained in tion 3.7B.2.1. Computer program capabilities for time-history solutions are outlined in endix 3A, Section 3.2.

B.3.1.1.3 Testing ipment and components that use seismic testing as their qualification basis conform to the owing general instructions for earthquake testing. These requirements conform with other licable industry standards such as IEEE 344-1975, Section 3.10, or provide guidance for ing where no such standards are available. Equipment packages or components are shown to eismically adequate by being tested individually, as part of a simulated structural section, or of an assembled module or unit. In any case, the minimum acceptance criteria include:

1. No loss of safety function or ability to function before, during, or after the test.
2. No structural/electrical failure (i.e., connections and anchorages) that would compromise safety related component integrity.
3. No adverse or maloperation before, during, or after the proposed test that could result in an improper safety action.

ipment vendors and suppliers are required to formulate programs for qualifying the ipment in accordance with the conditions specified in the seismic design requirements tained in the equipment specifications. The vendor must submit a summary of the proposed rt for review and approval.

characteristics of the testing input at the equipment mounting locations are defined by lified response spectrum curves, zero period response curve levels, time-history motions, or binations of these as applicable.

use of single and multifrequency input testing is accepted as a method of seismic lification, based upon the particular plant site, structure, and floor response characteristics.

ctures, particularly at lower elevations, exhibit a broad frequency range response similar to ground motion during an earthquake. This broad range frequency motion is filtered at higher ctural elevations and response becomes more sinusoidal in nature. Knowledge of the floor onse characteristics of the structure and response characteristics of the equipment generally ate the requirements for testing. Periodic testing is applicable where periodic floor motion is cated and, conversely, random input testing is most applicable for broad frequency range input omponents. Periodic testing can be used to envelop multiple peak floor responses, as well as le peak, providing sufficiently high forcing is used. For equipment exhibiting multiple onse modes, single frequency input may be used, providing the input has sufficient intensity nvelop the floor response spectra of the individual modes of the equipment.

testing machine (fixture) setup is arranged so that the equipment tested is mounted to ulate, to the extent possible, the actual service mounting, with no dynamic coupling to the test

. Equipment is tested in the operating condition wherever possible, and functions are 3.7B-16 Rev. 30

in situ application of vibratory devices, to superimpose the seismic vibratory loadings on the plex active device for operability testing, is acceptable when application is justifiable.

test program may be based upon selectively testing a representative number of mechanical ponents according to type, load level, size, etc., on a prototype basis.

ddition to the single and multifrequency testing programs outlined, laboratory shock results, hipment shock data, or adequate historical dynamic adequacy data (i.e., previous relevant test nvironmental data) are also given consideration. The test method selected must demonstrate adequacy of principal structural and functional capability of the equipment.

eral testing guidance criteria specified for equipment include the following:

1. Single Frequency Testing Testing is performed for as much of the range between 1 and 33 Hz as practicable or justified. Input for qualification should, as a minimum, equal the zero period response curve level.
2. Sinusoidal Input
a. A frequency scan (two octaves per minute, maximum), at a constant acceleration level, is performed over the frequency range of interest. The objective of this test is to determine the natural frequencies and amplification factors of the tested equipment, and its critical components or appurtenances, and to ensure general seismic adequacy over the full frequency range of interest. The acceleration inputs used are the maximum rigid range accelerations indicated by the relevant response spectrum curves (damping independent).
b. A dwell test of the equipment at its fundamental natural frequency is included at the acceleration values specified in item (a) above.

Additionally, other frequencies are selected if amplification factors of 2.0 or more are indicated. A minimum 20 second duration is considered acceptable for each dwell.

c. Other methods of sinusoidal testing may be employed as justified. Included are exploratory tests, per item 2(a), which employ a low acceleration level input to identify equipment response characteristics, and to aid in selecting requirements for further testing. A geometrically spaced constant frequency input may be employed for further testing. Intervals of one-half octave or less are employed for this spacing.
3. Sine Beat Input A sine beat test may be performed in conjunction with a sine scan, as an alternative to the dwell portion of the program outlined in item 2(b). The sine beat test is 3.7B-17 Rev. 30

particular test frequency are chosen to produce a magnitude of equipment response most nearly equivalent to that produced by the particular floor response spectrum at justifiable damping levels.

Current practice indicates that a minimum of 10 cycles per beat should be used, unless a lower number of cycles is shown sufficient to duplicate or exceed the response spectra for the equipment at the appropriate location.

An alternative qualification program consists of applying a series of sine beats at geometrically-spaced frequency intervals of one-half octave or less over the frequency range of interest. The peak amplitude of the beat employed is, as a minimum, the maximum rigid range acceleration indicated by the relevant response spectrum curve.

4. Multi-frequency Testing Multi-frequency testing is applicable as a general qualification method. Input excitation in this category includes time history, random, power spectral density, complex wave shapes, and others as justified. For the type of input applied, the testing machine input must equal, as a minimum, the zero period acceleration of the applicable response curve. A frequency range of 1 to 33 Hz is normally considered.
5. Time-History Input An acceleration time-history of the equipment support location, or one based on a synthesized response curve, may be used as testing machine input. A 15- to 30-second time history input is normally employed. The test table input must develop a response curve which envelops the relevant response spectrum curve when a synthesized record is used.
6. Random Motion Input Random input testing is performed so that the applicable response spectrum curve is enveloped by that produced by the table motion. The input is controlled by one-third octave (or less) bandwidth filters over the frequency range of interest with a minimum of 15 seconds or greater test duration. Normally, random tests are performed to produce a response curve based on test machine input which envelops the relevant response spectrum curve. A special case random test may be performed when a power spectral density equivalent of the applicable response curve is specified.

Tests combining random input in conjunction with other waveforms may be employed as justified.

3.7B-18 Rev. 30

A complex wave test may be performed by subjecting the equipment to an input motion, generated by summing a group of decaying sinusoids spaced at one-third octave, or narrower frequency intervals, over the frequency range of interest.

Individual decay rate controls of from 0.5 to 10 percent are used. Response curves based on test table input must be shown to envelop the relevant response curve.

B.3.1.2 Piping Systems lyses of Seismic Category I (including all ASME Code Classes 1, 2, and 3 piping systems) ng are performed by the modal analysis response spectra method. Nonseismic piping is mically analyzed when its failure could result in unacceptable damage to a Seismic Category I em. Either a modal analysis response spectra method or equivalent static load method of lysis is utilized for the latter.

criteria and procedures used for modeling are described in Section 3.7B.3.3.2. A typical hematical model of a piping system is shown on Figure 3.7-65. Defined boundaries, such as ipment and pipe anchors, may be considered as isolating the piping system when it is ergoing earthquake excitation.

modal analysis response spectra method is used for dynamic analysis of Seismic Category I ng. Input for these piping systems is described as follows:

1. Amplified Response Spectra (ARS)

Obtained for discrete locations in the structure where the piping systems are supported. Enveloping and peak broadening procedures are applied on these ARS curves before input, as described below. Damping values used for piping are 0.5 percent for OBE and 1 percent for SSE (Section 3.7B.1.3), except that increased damping values may be applied on an as-needed basis for final stress reconciliation (or piping system backfits) in accordance with ASME Code Case N-411 (Figure 3.7B-71).

2. Seismic Piping Anchor Movements Seismic piping anchor movements are obtained from seismic displacements of structures at piping anchor and support locations. These movements are used as static input to calculate the resulting internal forces and moments throughout the piping system. The methods used to consider differential piping support movements at different support points are discussed in Section 3.7B.3.8.

ere a piping system is subjected to more than one response spectrum, as when support points located in different parts of the structure or in separate structures, an enveloping procedure as l as peak broadening is applied to generate a composite, or worst-case, spectrum for analysis.

k broadening of minus 15 percent and plus 15 percent of peak frequencies is provided to ount for uncertainties in the calculated values of structural frequencies. Accordingly, piping ems designed using those amplified response spectra having natural frequencies within +/-15 3.7B-19 Rev. 30

lysis provides peak response quantities for each mode which are then combined according to tion 3.7B.3.7. All significant dynamic modes of responses under seismic excitation with uencies less than 50 cps or modes less than 50, whichever is reached first, are included in the amic analysis described in Section 3.7B.3.8. The combined seismic responses, together with rnal forces and moments due to seismic anchor movements, are then combined with other ings according to ASME Section III Code, Articles NB 3600 (Class 1 piping), NC 3600 ss 2 piping), or ND 3600 (Class 3 piping).

e-history modal superposition analysis is employed for fluid-induced transient dynamic blems (e.g., water hammer and steam hammer), but is not normally used for piping seismic lysis.

all size seismic Category I piping systems (Section 3.7B.3.5.2) are seismically qualified, in

, by the application of standard span procedures. The standard span procedures are restricted mall bore piping systems (one inch and below ASME Class 1 and two inch and below ASME ss 2, 3, and ANSI B31.1) which meet the criteria of the prequalified analysis.

tests or empirical methods are used in lieu of analytical methods for all Seismic Category I ng.

ailed descriptions of analytical procedures and design criteria for Seismic Category I piping be found in Section 3.7B.3.8. The type of seismic analysis used and the criteria used are marized in Table 3.7B-26.

B.3.2 Determination of Number of Earthquake Cycles B.3.2.1 Equipment and Components ME III (NB 3112,3b) requires that the number of earthquake cycles to be used in the analysis SME Code Class 1 components be specified as part of the design mechanical loads. The owing criteria are used for all components within the jurisdication of this code:

1. A total of five OBE and one SSE is assumed.
2. A minimum of 10 maximum stress cycles per earthquake is assumed.

Alternatively, the number of cycles per earthquake may be obtained from the structural time-history analysis.

B.3.2.2 Piping Systems ASME Class 1 piping systems are designed for a minimum of 10 maximum stress cycles per mic event in the analysis. A total of five OBE and one SSE is assumed.

3.7B-20 Rev. 30

B.3.3.1 Equipment and Components procedures used for modeling of equipment and components are contained in tion 3.7B.3.1.

B.3.3.2 Piping Systems basic method of analysis used is the finite element stiffness method. In accordance with this hod, the continuous piping is mathematically idealized as an assembly of elastic structural mbers connecting discrete nodal points. Nodal points are placed in such a manner as to isolate icular types of piping elements, such as straight runs of pipe, elbows, valves, etc., for which e-deformation characteristics can be categorized. Nodal points are also placed at all ontinuities, such as piping supports, concentrated weights, branch lines, and changes in cross ion. Inertial characteristics of the piping system are simulated by discrete masses of pipe and components (including all concentrated and eccentric masses such as valves and valve rators) lumped at selected node points. System loads other than weights, such as thermal es and earthquake inertial forces, are also applied at the nodal points. The stiffness matrix of piping system is calculated based upon the elastic properties of the pipe and pipe components, nclude the effects of bending, torsional, axial, and shear deformations. The stiffness of piping ws, and certain branch connections, is modified to account for local deformation effects by flexibility factors suggested in the ASME Section III Code, Articles NB 3600 (Class 1 ng), NC 3600 (Class 2 piping), and ND 3600 (Class 3 piping).

B.3.4 Basis for Selection of Frequencies B.3.4.1 Equipment and Components plified response spectra (floor) developed for two orthogonal horizontal and vertical direction hquakes are the basic sources of seismic design accelerations. As noted in Section 3.7B.3.1.1, mic accelerations are selected from the amplified response spectra based on natural frequency ulations for the equipment or component.

B.3.4.2 Piping Systems he seismic design and multi-mass modal analysis of Seismic Category I piping systems ction 3.7B.3.8), the practice of selecting piping fundamental natural frequencies to preclude nance is not used.

B.3.5 Use of Equivalent Static Load Method of Analysis B.3.5.1 Equipment and Components se components which are considered relatively simple or rigid are designed, by virtue of ral frequency calculations, to withstand the effects of amplified seismic acceleration values 3.7B-21 Rev. 30

nant response is considered conservative, since fundamental natural frequencies do not erally coincide with the frequency at resonance of the relevant response curve. Components ing fundamental natural frequencies within the broadened response peak are designed to peak eleration values, increased by a factor of 1.3, or as justified, to account for the contribution of ignificant dynamic modes under a resonant condition. Generally, the vibratory characteristics he components, qualified by resonant static analysis, are such that no possibility exists for cent or multiple modes to exist within the relatively narrow peak of a typical response ctrum.

discussion which follows justifies the use of a factor of 1.3 as a conservative multiple to be lied to single or multiple degree-of-freedom systems having fundamental frequencies within broadened resonant response peak. Multiply supported, or continuous type, span components not part of this proof. When such cases arise, a factor of 1.5 times the peak resonant response sed.

B.3.5.1.1 Single Degree-of-Freedom Systems k broadening is intended to reflect a range of uncertainty in the precise location of the resonant k of the response curve, and not to indicate that the multiple peak resonant response is possible hin this broadened range. What is concluded is that there is a fairly equal chance that the peak he curve (singular) would fall in the specified range and, thus, what exists, in fact, is a family esonant response curves, each having only one point of peak resonant response ure 3.7B-66). If more than one system or component mode of vibration falls within the adened peak, one and only one mode (a presumed worst case) can be presumed at an actual onse peak value (Figure 3.7B-67). All other possible modes would realistically respond to er values. Using the simple vibration theory and some simplifying assumptions, it is shown a factor of 1.3 is conservative.

mple damped oscillator responds with a transmissibility:

2 1 + 2 ------

n TR = ---------------------------------------------------------------- (3.7B.3-6) 2 2 ----- 2 1 - ------ + 2 -

n n re:

n = The undamped natural circular frequency w = The frequency of the exciting force value of TR is dependent on the damping value, and the ratio of exciting frequency to llator natural frequency. When the exciting frequency equals the oscillator natural frequency, steady state input is amplified by the value of TR and the response amplitude is maximum. In ismic environment, maximum response is equal to the peak of the amplified response ctrum curve.

3.7B-22 Rev. 30

e values computed. It is shown that TR increases as the number of modes increase, and that most conservative placement of assumed modes is with one mode at the peak and other modes tered around this peak.

a shown on Figure 3.7B-68 justify the use of a factor of 1.3 as conservative for all potential ipment applications. The curves are developed for two planes representing five modes and modes assumed acting within the broadened resonant peak. These numbers are intended to w an upper bound for general equipment application. Equipment damping values of 2.0 and percent are used for static analysis. Higher damping is shown to indicate the trend and the servatism of this method.

further conservatism, all modes are considered participating equally. This is never the case in amic analysis. The higher frequencies of the component are given equal weight to the damental resonant frequency and the modes are centered on the nominal response curve. If the damental frequency were placed on the peak of the nominal curve, the results would show n lower transmissibilities.

factor 1.3 is applicable only for those components whose fundamental natural frequency falls hin the broadened response peak.

as been shown that, for the range of values associated with component and system static lysis, use of the 1.3 factor is conservative. In fact, for a predominant number of likely cases, a e far less than this could be justified on the basis of the data.

example, a value of 1.1 could be justified for most components which present only a few ificant modes of vibration within the broadened response peak. It is further emphasized that, eaching these conclusions, the most conservative assumptions regarding location of the inal response curve and the placement of response modes for the arbitrary component have n made.

B.3.5.1.2 Multi-Degree-of-Freedom Systems a conclusive supplement to the previous discussion, a study was performed utilizing rigorous amic analysis of models closely representative of typical components.

s investigation consists of computing the ratio of maximum dynamic stress to maximum static ss (i.e., the factor denoted by K) for several model beams subject to a flat response and typical lified response spectra. Since bending stress is dominant for frame/equipment construction, actual ratio employed equals:

K = maximum dynamic moment-

(3.7B.3-7) maximum static moment h SRSS and absolute (ABS) moments are computed for comparison purposes, but conclusions based solely on SRSS moments because they most closely represent actual dynamic stress.

ximum static moment corresponds, in the case of the 1 g flat response, to 1 g static load. In the of a typical amplified response, the maximum static load is based upon the following uency relationships (Figure 3.7B-69):

fo fp, g = gmax (peak acceleration) 3.7B-23 Rev. 30

re:

fo = The fundamental frequency of the model beam fp = The frequency at which the peak acceleration occurs effect of peak spreading is investigated by using a flat response, thus giving all modes the e acceleration. This is equivalent to infinite peak spreading. The importance of the uncertainty he location of the peak acceleration with respect to the fundamental mode of the model beams xamined by adjusting the fundamental frequency from well below to well above the peak nant frequency of a typical response spectrum.

model beams selected for this study are shown on Figure 3.7B-70. These beams are typical of frames and equipment combinations used in nuclear power plants. All dynamic analyses were ducted using the STRUDL-SW computer program described in Appendix 3A. Static analyses e carried out by hand, except for the simple/fixed beam with overhang. Consistent with design tice, all mountings in this study are assumed rigid.

B.3.5.1.3 Results for Flat Response le 3.7B-23 summarizes the results for a 1 g flat response applied to the model beams of ure 3.7B-70. Three K factors were computed for comparison purposes:

Maximum SRSS dynamic moment K s/c = ------------------------------------------------------------

Maximum static moment from concentrated load Maximum SRSS dynamic moment - (3.7B.3-9)

K s/u = -----------------------------------------------------------

Maximum static moment from uniform load Maximum ABS K a/u dynamic moment -

= -----------------------------------------------------------

Maximum static moment from uniform load conclusions in this study are based on Ks/u because it most closely represents the actual ratio ynamic moment to static moment. Ka/u was not chosen because, as Table 3.7B-24 illustrates, des are so widely spaced that no more than one modal frequency lies within a +/-10 percent uency band. Ks/c is shown since this is the K factor which represents a typical simplification d in component analysis (concentrated static loads at component center of gravity).

3.7B-24 Rev. 30

ter than unity.

B.3.5.1.4 Results for Amplified Response le 3.7B-25 presents the results for the simply supported/fixed model beam with 33 percent rhang subjected to the response spectra of Figure 3.7B-69. The first mode column on this table s the fundamental frequency, fo, and response acceleration, g, at fo. Note that f was adjusted density variation) from well below to well above the peak frequency, fp, of the response ctra to determine the effect on K of the uncertainty in the location of the peak frequency with ect to the fundamental frequency of the model beam. Since all values of Ks/u were less than y, it is concluded that this uncertainty has no important effects on the K factor.

B.3.5.1.5 Conclusions

1. Peak acceleration times 1.3 applied as a static load to equipment whose fundamental natural frequency is within the broadened peak of the amplified response spectra curve is conservative for simply supported systems.
2. No amount of peak spreading can itself result in a Ks/u factor significantly greater than unity.
3. Uncertainty in the frequency at which the peak response acceleration occurs itself has no important effects on the K factor.
4. Multiple supported continuous spans are not included in the scope of this study.

Components or equipment which make up a system of continuous multiple span supports utilize a factor no less than 1.5 times peak acceleration as in item 1 above, if applicable.

B.3.5.2 Piping Systems ivalent static load method is not generally used in the seismic analysis of Seismic Category I ng. However, seismic and certain nonseismic small bore piping and instrument tubing ction 3.7B.3.1.2) are seismically supported at standard intervals based on maximum spans blished from modal response spectra analysis of representative small bore piping figurations.

ome cases, when deemed appropriate, the analysis of small bore piping and instrument tubing erformed using methods based on simplified dynamic engineering formulations to ascertain r code adequacy.

simplified dynamic formulation essentially involves the application of a factor of 1.3 to the k value of ARS when the fundamental frequency of the piping or tubing configuration is less 33 Hz while accounting for the seismic effects. The factor of 1.0 is applied when the damental frequency of the piping or tubing configuration is greater than or equal to 33 Hz.

3.7B-25 Rev. 30

B.3.6.1 Equipment and Components he seismic analysis of equipment and components, each of the three defined directions of hquake input (two horizontal and one vertical taken orthogonally) is evaluated separately. The sses resulting from the orthogonal earthquake inputs are combined by the square root of the of squares (SRSS).

B.3.6.2 Piping Systems he seismic analysis of piping systems, the effects of simultaneous action of three spatial ponents of earthquake motion are considered. When the response spectrum modal analysis hod is used for seismic analysis, the maximum piping modal responses (e.g., moments and lacements) due to the three spatial components of earthquake motion are obtained by the dified SRSS method. In this method, when considering a particular vibration mode, responses particular direction due to the two horizontal direction excitations are combined first by the SS method and then combined with response (in this same direction) due to the vertical ction excitation by absolute sum method. It has been demonstrated that this modified SRSS hod always results in conservative answers as compared with the SRSS method described in ition C.2.1 of Regulatory Guide 1.92 (Chang 1973).

athematical terms, the modified SRSS method of response combination due to three spatial ponents of an earthquake at a particular node point can be expressed as:

2 2

( Rx )N = ( R xx ) N + ( R xz ) N + R xy N (3.7B.3-10) re:

(Rxx)

(Rxy) = Response in the x, y, or z direction, respectively for mode N (Rxz)

(Rx)N = Combined response in the x direction due to seismic excitation for mode N e this is accomplished, (Rx)N, (Ry)N, and (Rz)N are then combined independently for ificant modes N from 1 to K by methods described in Section 3.7B.3.7, where K is the ber of significant modes considered in the modal response combination.

B.3.7 Combination of Modal Responses following methods are applicable to piping system analysis employing the modal analysis onse spectra method.

1. If the modes are not closely spaced, the maximum responses of piping are obtained by taking the SRSS of corresponding maximum modal responses of the piping 3.7B-26 Rev. 30
2. If closely spaced modes exist, then the grouping method is employed to combine various modal responses from a dynamic modal analysis. This is in conformance with Position C.1.2.1 of Regulatory Guide 1.92.

equipment and components, the following methods of combination of modal responses are loyed for each direction of earthquake motion:

1. When performing response spectrum modal analysis, the representative maximum value of a particular response earthquake is obtained by taking the SRSS of corresponding maximum values of the response of the element attributed to individual significant modes of the structure, system, or component.

Mathematically, this can be expressed as:

N 2

R = Rk (3.7B.3-11)

K=1 where:

R = The representative maximum value of a particular response of a given element to a given component of an earthquake, Rk = The peak value of the response of the element due to the kth mode, N = The number of significant modes considered in the modal response combination.

2. When performing response spectrum modal analysis, if closely spaced modes exist, the grouping method as described in Regulatory Guide 1.92 is employed to combine modal responses. Mathematically, this can be expressed as follows:

N P j j 12 2

R = Rk + R l q R mq (3.7B.3-12) k=1 q=1 l=i m=i where:

Rlq and Rmq = Modal responses, R and Rm within the qth group, respectively; i = The number of the mode where a group starts; j = The number of the mode where a group ends; 3.7B-27 Rev. 30

P = The number of groups of closely spaced modes, excluding individual separated modes.

3. The method used for combination of three components of earthquake motion is described in Section 3.7B.3.6.

B.3.8 Analytical Procedures for Piping Systems general analytical procedure of the modal analysis response spectra method for piping ems is described in Section 3.7B.3.1.2. Basic steps and equations used in the analytical cedure are described below.

the dynamic analysis, the piping is represented by a lumped mass, multi-degree-of-freedom hematical model. The distributed piping mass is lumped at the system nodal points. The ation of motion for the system is:

[ M ] { X** } + [ C ] { X* } + [ K ] { X } = { F } (3.7B.3-13) re:

[M] = Mass matrix for assembled system

[C] = Damping matrix for assembled system

[K] = Stiffness matrix for assembled system

{X} = Nodal displacement vector = [X(t)]

{ X* } = Nodal velocity vector = { X* ( t ) }

{ X** } = Nodal acceleration vector = { X** ( t ) }

{F} = Applied dynamic force vector = {F(t)}, or = [M] { U** g } for seismic analysis

{ U** g } = Seismic acceleration vector for points of pipe support ation 3.7B.3-13 is solved for the system dynamic response as follows: First, the frequency ation, obtained by removing the forcing and damping terms from the above equation, is solved the system natural frequencies and mode shapes. Next, the natural mode shapes are used to ct an orthogonal transformation of the equation yielding a series of independent equations of ion uncoupled in the system modes. Then, the uncoupled equations are solved by the response ctrum method to obtain system response in each mode, and the individual modal results are bined to determine the total system dynamic response. The mathematical formulation of these s is described in the following subsections.

3.7B-28 Rev. 30

t, the eigenvalues (natural frequencies) and the eigenvectors (mode shapes) for each of the ral modes are calculated by solving the frequency equation:

2

( [ K ] - n [ M ] ) [ ]n = { 0 }

(3.7B.3-14) n = 1, 2, 3, , N re:

Wn = Natural frequency of the nth mode,

{}n = Mode shape vector of the nth mode,

{0} = Null vector, N = Number of significant modes considered eigenvalues and eigenvectors are obtained using the Householder-QR algorithm, Jacobi uction or Inverse iteration.

B.3.8.2 Dynamic Response t, let {n(t)] be the generalized coordinate vector, substitute {X} = [] {n} into the equation of ion and pre-multiply by []T; an orthogonal transformation results, from which the uncoupled ations of motion shown below are obtained

  • 2

+ 2 n n + n n = P n (3.7B.3-15) n = 1, 2, 3, , N re:

[] = The square matrix of mode shape vectors n = Generalized coordinate for the nth mode = nt h = Damping ratio of the nth mode expressed as percent of critical damping Pn = Generalized force of the nth mode Pn = {}nT {F}/Mn for applied dynamic force {F}

Pn = {}nT [M] { U** g } /Mn for seismic analysis Mn = Generalized mass of the nth mode = {}nT [M] {}n utions to these differential equations are obtained by the method of amplified floor response ctrum (ARS) superposition, as described below.

3.7B-29 Rev. 30

response of a piping system to seismic excitations is obtained using the method of response ctrum superposition. At any pipe support point, seismic input is produced by a set of three S, one in each global coordinate direction. These ARS are generated from the application of e-history acceleration responses obtained from the structure or equipment time-history lysis. These ARS are peak broadened (Section 3.7B.2.9) to reflect variations in structure perties. Where a piping system is subjected to more than one set of (three) ARS, such as port points located in different structures or different parts of the same structure, the eloping and peak broadening are applied to all sets of ARS at support points ction 3.7B.3.1.2). Thus, after the enveloping and peak broadening process, a set of three ARS, in each global coordinate direction, results. The maximum acceleration for the nth mode of piping system in jth global coordinate direction is then given by

{ X** } nj = { } n nj max (3.7B.3-16)

{ X** } nj = { } n nj /M n n = 1, 2, 3, ....N j = 1,2,3 corresponds to response in X,Y, or Z global coordinate direction, respectively.

re:

{ X** } nj = Maximum acceleration vector of mode n nj max = Maximum generalized coordinate acceleration of mode n nj = Modal participation factor for the nth mode in jth global coordinate direction nj = {}nT [M] {e}j

{e}j = A vector with components of unity in all directions parallel to jth global coordinate direction and zero otherwise (Sa)nj = Spectral acceleration for the nth mode, in jth global coordinate direction (from enveloped and peak broadened ARS) maximum inertia force vector for the nth mode and in jth global coordinate direction is given

{ F } nj max = M N { } n nj max (3.7B.3-17)

{ F } nj max = { } n nj ( S a ) nj se inertia forces are calculated for each of the system natural modes in all three global rdinate directions and applied as static forces in the same manner as the deadweight or ivalent thermal forces, to find internal moments and forces in each mode. The total maximum onses due to seismic excitation are then obtained by combining the modal responses cribed in Section 3.7B.3.7. The calculated primary stress range due to seismic inertial 3.7B-30 Rev. 30

ximum relative displacements in two horizontal and the vertical direction between piping ports and anchor points (i.e., between floor penetrations and equipment supports at different ations within a building, and also between buildings) are used as equivalent static lacement boundary conditions in order to calculate the secondary stresses of the piping em. Relative seismic displacements used are obtained from a dynamic analysis of the ctures, and are always considered to be out-of-phase between different buildings to obtain the t conservative piping responses.

se seismic member moments and forces are then combined with loads from deadweight, sure, thermal, and other mechanical loads to complete the stress analysis of all Seismic egory I, and some Non- Seismic Category piping. For ASME Code Class 1 piping, stress nsities, and cumulative usage factors of the piping system are computed based on the mulation specified in Subarticle NB 3600, ASME Section III for ASME Code Class 2 and 3 ng, the formulations in Subarticle NC 3600 and ND 3600, respectively, are used.

design criteria, loading combination, and stress limits for BOP Seismic Category I piping ems are described in Section 3.9B.3.

B.3.9 Multiply Supported Equipment and Components with Distinct Inputs calculate the maximum inertial response of multiply supported subsystems, an upper bound elope of all the individual response spectra for the support locations is used. In addition, the tive displacements at the support points are considered. Support displacements are imposed ically on the subsystem in the most conservative combinations. For support locations within a egory I structure, relative displacements are determined algebraically and imposed.

placements between Category I structures are considered to be out-of-phase, and the imum relative displacements between Category I structures are thus determined from olute sums of the support displacements. The stresses due to seismic inertia and relative lacements are added to those due to other appropriate loadings such as deadweight, pressure,

, and the resulting stresses are limited by allowable stresses defined in applicable codes.

B.3.10 Use of Constant Vertical Static Factors B.3.10.1 Equipment and Components stant load factors are not used for vertical floor response in the seismic design of Seismic egory I equipment and components.

B.3.10.2 Piping Systems method of applying constant static factors as vertical response loads, based on the assumption ertically rigid structures, for the seismic design of Seismic Category I piping is not used.

wever, a simplified analysis (equivalent static load method) using constant load factors for the ical and horizontal directions based on the peaks of applicable amplified response spectra is d for seismic analysis of certain nonseismic piping systems (Section 3.7B.3.5.2).

3.7B-31 Rev. 30

effect of eccentric masses such as valve operators is considered in the seismic piping analysis cribed in Section 3.7B.3.1.2. These eccentric masses are included in the mathematical model the piping analysis and the torsional effects caused by them are evaluated and included in the l piping response.

B.3.12 Buried Seismic Category I Piping Systems erforming stress analysis of buried Seismic Category I piping systems, the loadings sidered are:

1. Internal pressure
2. Soil pressure (includes dead load, and live loads due to traffic when applicable)
3. Thermal expansion
4. Differential movements between structures and adjacent soil due to settlement and seismic motion
5. Seismic wave effects cts of loadings (1) and (2) are assessed by well known methods (e.g., Terzaghi 1955; King,

. and Crocker, C. 1967, Section 21-29 to 21-33); effects of loadings (3) and (4) are accounted by a static analysis considering piping modeled together with soil springs. This is basically a m on elastic foundation approach. Loadings (4) and (5) are discussed in detail in the sections w.

basic assumptions concerning buried piping stress analysis are:

1. piping satisfies elementary theory of beams;
2. soil is linear elastic, homogeneous and isotropic; and
3. soil strain is fully transferred to the pipe; i.e., there is no slippage between the pipe and the soil.

B.3.12.1 Seismic Wave Effect in the Free Field ious seismic waves develop during an earthquake. There are compression waves (P-waves),

ar waves (S-waves), and different kinds of surface waves such as Rayleigh waves (R-waves).

en seismic waves propagate through the soil, responses of buried Seismic Category I piping calculated by making use of the analytical approach proposed by Goodling (1978, 1979, 0).

ight portions of buried piping far from the effect of external supports, bends and tees are med to move with the soil when seismic waves propagate through it. Since Rayleigh waves 3.7B-32 Rev. 30

Vm Axial strain m = ------- - (3.7B.3-18)

CR Bending curvature = am/C2R re:

Vm = peak ground velocity, in/sec am = peak ground acceleration, in/sec2 CR = Rayleigh wave velocity, in/sec refore, the stresses on the straight portions of buried piping mentioned above are given by:

EV m F max Axial Stress a = ----------- - = ----------- = m E (3.7B.3-20)

CR Am ED o a m ED o X Bending Stress b = ---------------- - = --------------

- (3.7B.3-21) 2C R 2 2 re:

E = Youngs modulus of pipe, psi Do = outside diameter of pipe, in Fmax = maximum axial force, lb Am = cross-sectional area of pipe, in2 mic wave effect on bends and tees in the free field are considered separately. For a bend, the imum stresses are determined by assuming that its longitudinal leg is in the direction of imum soil strain and its transverse leg is in the perpendicular direction.

longitudinal leg may terminate into another bend, an anchor, or a free end. The bend is sified accordingly and the actual slippage length (L') along which slippage between pipe and occurs is determined as outlined by Goodling (1978).

net relative displacement , between soil and pipe at the bend is given by:

2 S 1 L' fL - - -----------

1 = m L' - -------------- - (3.7B.3-22) 2A m E A m E 3.7B-33 Rev. 30

mL' = theoretical unrestrained relative movement at the elbow over length L'.

S1L'/AmE = the pipe elongation due to friction along the soil/pipe interface.

fL'2/2AmE = the pipe elongation due to friction along the soil/pipe interface.

f = frictional force per unit length of pipe, lb/in bend legs are considered as beams on elastic foundation for which its parameter is given k-

= 4 -------- (3.7B.3-23) 4EI re:

k = soil spring constant, per unit length, lb/in2 I = moment of inertia of pipe cross section, in4 ase of long-transverse leg (its length is greater than 3/4), the following equations are ved by incorporating the interdependence of forces, moments, soil deformation, and rotation he pipe in the immediate vicinity of the bend (Goodling 1980).

1 M = --------------------

- (3.7B.3-24)

R/K'EI k

S 1 = ------- + M (3.7B.3-25) 2 re:

M = bending moment, in-lb/in f = elbow angle, radians R = radius of elbow, in 9

K = 1 - --------------------------------

tR 2 10 + 12 ----

a t = actual pipe wall thickness, in a = outside radius of pipe, in ase of short transverse leg, the following conservative equations derived by Goodling (1978) used:

3.7B-34 Rev. 30

2A m E 1 = -----------------------------------------------------------------

- (3.7B.3-26)

KL' C3 1 + --------------- ---------------------------------

A m E 2C C - C 2 1 3 2 k 1 C3 S 1 = --------- ---------------------------------

- (3.7B.3-27) 2C 1 C 3 - C 2 2

k 1 C2 M = ------- ---------------------------------

2

- (3.7B.3-28) 2 2C 1 C 3 - C 2 re:

C1, C2, and C3 are coefficients given by Goodling (1978) maximum axial force in the longitudinal leg, Fmax, induced by the seismic motion is given Fmax = S1 + fL' (3.7B.3-29) ing determined the values of S, M and F, the stresses due to local deformation at the bend can valuated. These stresses are superimposed on stresses caused by the curvature of the pipe ng seismic wave propagation. The combined stresses at the bend are multiplied by an nsification factor (0.75 i) to account for the higher intensity of stresses at the elbow.

following expressions for stress result:

Stress at an Elbow

= 0.75i [EDo /2+M/Z] + S1/Am(3.7B.3-30)

So1 (elbow)

Stress in the Longitudinal Run So1 (long) = Fmax/Am + EDo /2 (3.7B.3-31) re:

Z = section modulus of pipe, in3 asionally, when the conservative equations used in case of short transverse leg result in cceptable stresses, the technique incorporating the passive resistance of soil, as presented by dling (1978), may be used.

milar approach is used for analyzing tees (Goodling 1978).

3.7B-35 Rev. 30

ing an earthquake, differential seismic motions occur between structures and adjacent soil.

se differential motions are obtained by the seismic analysis of structures with soil-structure raction taken into account. The effect on the buried piping systems due to these differential ions can be evaluated by considering separately the effects of different components, namely, erential motion components transverse to the direction of the piping axis, and differential ion components parallel to the direction of the piping axis.

1. Differential motion components transverse to the direction of piping axis.

The subgrade reaction approach is used here to simulate the effect of soil on the deformation and stress of the buried pipe due to differential motion components transverse to the direction of piping axis. The approach is based on Terzaghis theory (Terzaghi 1955) that soil subjected to pressure behaves like a system of uniformly spaced elastic springs with predetermined stiffness. The soil is thus represented by a series of orthogonal pairs of elastic springs in directions transverse to the piping axis and attached to the piping in the mathematical model.

The elastic springs in the vertical direction are calculated in accordance with Terzaghi (1955). The elastic springs in the horizontal direction are calculated based on the method described in Audibert and Nyman (1975). The maximum expected transverse seismic displacements at the structural penetration are used as input in the calculation. The computation is done by the computer program NUPIPE-SW, which is listed in Appendix 3A. In principle, this approach is basically a beam analysis on an elastic foundation (Hetenyi 1946).

2. Differential motion components parallel to the direction of piping axis.

The effect on buried long straight piping due to differential seismic motions along the direction of piping axis at penetration is assessed by considering the frictional force between pipe and soil, and the maximum axial stress due to this effect is (Yeh 1974):

2EF a = ---------------a-t re:

a = Stress in piping at penetration point due to axial displacement, a, of the structure at the penetration point (psi)

E = Youngs modulus of pipe (psi)

F = Frictional force between pipe and soil (psi) 3.7B-36 Rev. 30

t = Pipe thickness (inch) en bends or tees exist close to the penetration, the effect of differential motion parallel to ction of piping axis is analyzed again by the method described in Shah and Chu (1974).

B.3.12.3 Effects of Differential Movements Due to Structural Settlement lement of a structure can happen either due to its own weight over a period of time or due to arthquake. The settlement of the structure where the piping is connected, as well as the soil cent to the structure where the piping is buried, are imposed on the buried piping, and the roach outlined in differential motion components transverse to the direction of piping axis n above is used to evaluate stresses in the buried piping.

B.3.12.4 Accommodations for Buried Piping Structural Penetrations resultant loadings imposed by thermal, structural, and seismic distortions may cause severe l stresses in buried piping at structure penetration points. The piping, if anchored at the ctural wall, may be too stiff to accommodate these distortions for such locations. In such s, the buried piping design includes a structural penetration, consisting of a concrete box or a duit which is not attached to the structure, and is free to move with the soil rather than with the cture. Within the box or conduit, the piping may (if necessary) be provided with expansion ts or piping loops to accommodate relative displacements in both axial and transverse ctions.

B.3.13 Interaction of Other Systems (Piping and Equipment) with Seismic Category I Systems (Piping and Equipment) seismic category systems (piping and equipment) are designed to be isolated from Seismic egory I systems (piping and equipment). Isolation may be accomplished by physical restraints, iers, or separation. If it is not practical to isolate the Seismic Category I system from the seismic, then the potential for damage due to seismic interactions is evaluated through the owing program.

Seismic Interaction Program consists of three distinct tasks:

demonstrating the adequacy of equipment anchorages for nonseismic equipment in Seismic Category I buildings, demonstrating the structural integrity of piping and supports for selected subsystems, and performing walkdowns to identify swing/sway interactions between non-Seismic Category I piping and equipment and Seismic Category I piping and equipment.

-Seismic Category I equipment in Seismic Category I buildings is reviewed to ensure the mic adequacy of its anchorage by one of the following methods:

3.7B-37 Rev. 30

2. compare the anchorage detail to explicitly qualified anchorages; and
3. for anchorages which are not seismically designed and are not similar to seismic anchorages, calculations are performed to demonstrate adequacy.

equipment anchorages evaluated by Method 3, an effort is made to group typical anchorage ils and perform bounding calculations. The manner in which structural integrity is onstrated for each piece of interacting equipment is documented by the calculations.

eptance criteria for equipment anchorage evaluations is given in J.F. Opeka to B.J.

ngblood Letter B11844, dated October 31, 1985, Docket No. 50-423.

structural integrity of piping and supports is addressed through a program developed and lemented by Sargent and Lundy. A set of piping subsystems was selected to be representative ipe sizes, hanger configurations, and operating conditions. As discussed in more detail in the ve-referenced letter, these selected subsystems are bounding.

traint loads and pipe stresses were calculated using dynamic analysis results. They have been pared to the failure capacities associated with each subsystem to assess the inherent margin of ty in the design. Maximum dynamic lateral displacements are used to confirm interaction eria utilized during plant walkdowns.

mic interaction walkdowns are conducted to identify swing/sway interactions between

-Seismic Category I piping and equipment and Seismic Category I piping and equipment. All ractions are evaluated considering the local flexibility of the interacting equipment/piping.

se reviews address restrictions such as penetrations and interferences with structures which ld limit displacements. In all cases, interactions with active Seismic Category I components is vented.

rmation gained from the experience database is used in conjunction with the above efforts.

database shows that properly supported equipment and piping maintains its structural grity during strong motion earthquakes. Further, this program benefits from the database rmation regarding the severity of seismic interactions and the knowledge of configurations ch have not performed well in past earthquakes.

nonseismic category piping systems attached to Seismic Category I piping systems, the amic effects of the nonseismic category piping are simulated in the analysis modeling of the mic Category I piping. The nonseismic category piping is modeled in a manner consistent h the accuracy of the Category I piping analysis. The attached nonseismic category piping is designed to ensure that, during an earthquake of SSE intensity, it does not cause a failure of Seismic Category I piping system.

B.3.14 Seismic Analysis for Reactor Internals Section 3.7N.3.

B.3.15 Analysis Procedure for Damping mping values of equipment, components and piping systems are given in Section 3.7B.1.3.

3.7B-38 Rev. 30

ddition to the steady state loads imposed on the system under normal operating conditions, the gn of the equipment requires that consideration also be given to abnormal loading conditions, h as earthquakes. Seismic loadings are considered for earthquakes of two magnitudes: safe tdown earthquake (SSE) and operating basis earthquake (OBE). The SSE is defined as the imum vibratory ground motion at the plant site that can reasonably be predicted from logic and seismic evidence. The OBE is that earthquake which, considering the local geology seismology, can be reasonably expected to occur during the plant life.

the OBE loading condition, the reactor coolant system is designed to be capable of continued operation. The design for the SSE is intended to assure:

1. That the integrity of the reactor coolant pressure boundary is not compromised
2. That the capability to shut down the reactor and maintain it in a safe condition is not compromised
3. That the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures of 10 CFR 100 is not compromised seismic requirements for safety related instrumentation and electrical equipment are covered ection 3.10. The safety class definitions, classification lists, operating condition categories, the methods used for seismic qualification of mechanical equipment are given in Section 3.2.

N.1 SEISMIC INPUT N.1.1 Design Response Spectra er to Section 3.7B.1.1.

N.1.2 Design Time History er to Section 3.7B.1.2.

N.1.3 Critical Damping Values damping values given in Table 3.7N-1 are used in the analysis of Westinghouse equipment.

values for RCS components are based on testing programs, as reported in WCAP-7921-AR, ch have been accepted by the staff. This WCAP defines the Westinghouse NSSS position on ulatory Guide 1.61.

ts on fuel assembly bundles justified conservative component damping, values of 7 percent for E and 10 percent for SSE to be used in the fuel assembly component qualification.

umentation of the fuel assembly tests is found in WCAP-8236 (1973) and WCAP-8288 74).

3.7N-110 Rev. 30

size CRDM complete with rod position indicator coils, attachment to a simulated vessel head, variable gap between the top of the pressure housing support plate and a rigid bumper esenting the support.

program consisted of transient vibration tests in which the CRDM was deflected a specified al amount and suddenly released. A logarithmic decrement analysis of the decaying transient vides the effective damping of the assembly. The effect on damping of variations in the drive t axial position, upper seismic support clearance, and initial deflection amplitude was stigated.

upper support clearance had the largest affect on the CRDM damping, with the damping easing with increasing clearance. With an upper clearance of 0.06 inches, the measured ping was approximately 8 percent. The clearances in a typical upper seismic CRDM support a minimum of 0.10 inches. The increasing damping with increasing clearances trend from the results, indicated that the damping would be greater than 8 percent for both the OBE and the

, based on a comparison between typical deflections during these seismic events and the al deflections of mechanisms in the test. Component damping values of 5 percent are, efore, conservative for both the OBE and SSE.

N.1.4 Supporting Media for Seismic Category I Structures er to Section 3.7B.1.4.

N.2 SEISMIC SYSTEM ANALYSIS er to Section 3.7B.2 N.3 SEISMIC SUBSYSTEM ANALYSIS s section describes the seismic analysis performed on subsystems within Westinghouses pe of responsibility.

N.3.1 Seismic Analysis Methods se components that must remain functional in the event of the SSE (Seismic Category I) are tified by applying the criteria of Section 3.2.1.

eneral, the dynamic analyses are performed using a modal analysis in combination with a onse spectrum analysis.

N.3.1.1 Dynamic Analysis - Mathematical Model first step in any dynamic analysis is to model the structure or component, i.e., convert the real cture or component into a system of masses, springs, and dashpots suitable for mathematical lysis. The essence of this step is to select a model so that the displacements obtained are a d representation of the motion of the structure or component. Stated differently, the true inertia 3.7N-111 Rev. 30

Equations of Motion sider the multi-degree of freedom system shown in Figure 3.7N-1. Making a force balance on h mass point r, the equations of motion can be written in the form:

i i m r y** r + c ri u* i + k ri u i = 0 (3.7N-1) re:

mr = The value of the mass or mass moment of rotational inertia at mass point r y** r = Absolute translational or angular acceleration of mass point r cri = Damping coefficient - external force or moment required at mass point r to produce a unit translational or angular velocity at mass point i, maintaining zero translational or angular velocity at all other mass points. Force or moment is positive in the direction of positive translational or angular velocity.

u* i = Translational or angular velocity of mass point i relative to the base Kri = Stiffness coefficient - the external force (moment) required at mass point r to produce a unit deflection (rotation) at mass point i, maintaining zero displacement (rotation) at all other mass points Force (moment) is positive in the direction of positive displacement (rotation).

ui = Displacement (rotation) of mass point i relative to the base e:

y** = u** r + y** s (3.7N-2) re:

y** s = Absolute translational (angular) acceleration of the base u** r = Translational (angular) acceleration of mass point r relative to the base ation 3.7N-1 can be written as:

i i m r u** r + c ri u* i + k ri u i = -m r y** s (3.7N-3) 3.7N-112 Rev. 30

mu** + cu* + ku = - m y** s (3.7N-4)

N.3.1.2 Modal Analysis Natural Frequencies and Mode Shapes first step in the modal analysis method is to establish the normal modes, which were rmined by eigen solution of Equation 3.7N-3. The right hand side and the damping term are equal to zero for this purpose, as illustrated in Biggs (1964) (Pages 83 thru 111). Thus, ation 3.7N-3 becomes:

i m r u** r + k ri u i = 0 (3.7N-5) equation given for each mass point r in Equation 3.7N-5 can be written as a system of ations in matrix form as M{} + K{} (3.7N-6) re:

M = Mass and rotational inertia matrix

{} = Column matrix of the general displacment and rotation at each mass point relative to the base K = Square stiffness matrix

{}= Column matrix of general translational and angular accelerations at each mass point relative to the base, d2 {}/dt2 monic motion is assumed and the {} is expressed as:

{}={} sin wt (3.7N-7) re:

{} = Column matrix of the spatial displacement and rotation at each mass point relative to the base

= Natural frequency of harmonic motion in radians per second displacement function and its second derivative are substituted into Equation 3.7N-6 and d:

3.7N-113 Rev. 30

determinant [K] - 2 [M] is set equal to zero and is then solved for the natural frequencies.

associated mode shapes are then obtained from Equation 3.7N-8. This yields n natural uencies and mode shapes where n equals the number of dynamic degrees of freedom of the em. The mode shapes are all orthogonal to each other and are sometimes referred to as normal de vibrations. For a single degree of freedom system, the stiffness matrix and mass matrix are le terms and the determinant [K] - 2 [M] when set equal to zero yields simply:

k - 2m = 0 k-

= --- (3.7N-9) m re is the natural angular frequency in radians per second.

natural frequency in cycles per second is therefore:

1 k f = ------ ---- (3.7N-10) 2 m ind the mode shapes, the natural frequency corresponding to a particular mode, n, can be stituted in Equation 3.7N-8.

1. Modal Equations response of a structure or component is always some combination of its normal modes. Good uracy can usually be obtained by using only the first few modes of vibration. In the normal de method, the mode shapes are used as principal coordinates to reduce the equations of ion to a set of uncoupled differential equations that describe the motion of each mode n. These ations may be written as (Biggs 1964, pages 116 thru 125):

A** + 2 n p n A* n + n A n = - n y** s 2

(3.7N-11) re the modal displacement or rotation, An, is related to the displacement or rotation of mass nt r in mode n, urn, by the equation:

urn = Anrn (3.7N-12) re:

n = Natural frequency of mode n in radians per seconds pn = Critical damping ratio of mode n n = Modal participation factor of mode n given by:

3.7N-114 Rev. 30

n = ----------------------

r

- (3.7N-13) 2 mr rn re:

'rn = Value of rn in the direction of the earthquake essence of the modal analysis lies in the fact that Equation 3.7N-11 is analogous to the ation of motion for a single degree of freedom system that is developed from Equation 3.7N-4.

iding Equation 3.7N-4 by m gives:

c k u** + ---- u* + ---- u = - y** s (3.7N-14) m m critical damping ratio of the single degree of freedom system, p, is defined by the equation:

c p = ---- (3.7N-15) cc re the critical damping coefficient is given by the expression:

cc = 2m (3.7N-16) stituting Equation 3.7N-16 into Equation 3.7N-15 and solving for c/m gives:

---c- = 2p (3.7N-17) m stituting this expression and the expression for k/m given by Equation 3.7N-9 into Equation N-14 gives:

2 u** + 2pu* + u = - y** s (3.7N-18) e the similarity of Equations 3.7N-11 and 3.7N-18. Thus, each mode may be analyzed as ugh it were a single degree of freedom system and all modes are independent of each other. By method, a fraction of critical damping, i.e., c/cc, may be assigned to each mode and it is not essary to identify or evaluate individual damping coefficients, i.e., c. However, assigning only ngle damping ratio to each mode is not appropriate for a slightly damped structure supported massive moderately damped structure. There are several methods which can be used to rporate damping in the model.

method is to develop and analyze separate mathematical models for both structures, using r respective damping values. The massive moderately damped support structure is analyzed

. The calculated response at the support points for the slightly damped structures is used as a ing function for the subsequent detailed analysis. Another method is to inspect the mode pes to determine which modes correspond to the slightly damped structure and then use the 3.7N-115 Rev. 30

N.3.1.3 Response Spectrum Analysis response spectrum is a plot showing the variation in the maximum response (Thomas et al.,

3, pages 24 thru 51) (displacement, velocity, and acceleration) of a single degree of freedom em versus its natural frequency of vibration when subjected to a time history motion at its e.

response spectrum concept can be best explained by outlining the steps involved in eloping a spectrum curve. Determination of a single point on the curve requires that the onse (displacement, velocity, and acceleration) of a single degree of freedom system with a n damping and natural frequency is calculated for a given base motion.

variations in response are established and the maximum absolute value of each is plotted as rdinate with the natural frequency used as the abscissa. The process is repeated for other med values of frequency in sufficient detail to establish the complete curve. Other curves esponding to different fractions of critical damping are obtained in a similar fashion. Thus, the rmination of each point of the curve requires a complete dynamic response analysis; the rmination of a complete spectrum may involve hundreds of such analyses. However, once a onse spectrum plot is generated for the particular base motion, it may be used to analyze each cture and component with the base motion. The spectral acceleration, velocity, and lacement are related by the equation:

2 Sa = n Sv = n sd n n n ponse spectra developed for Millstone 3 are discussed in Section 3.7B.3.

N.3.2 Determination of Number of Earthquake Cycles ere fatigue analyses of mechanical systems and components are required, Westinghouse cifies in the equipment specification the number of cycles of the operating basis earthquake E) to be considered. The number of cycles for NSSS components is given in Table 3.9-1. The gue analyses are performed and presented as part of the components stress report.

N.3.3 Procedure Used for Modeling er to Section 3.7N.3.1 for modeling procedures for subsystems in Westinghouses scope of onsibility.

N.3.4 Basis for Selection of Frequencies analysis of equipment subjected to seismic loading involves several basic steps, the first of ch is the establishment of the intensity of the seismic loading. Considering that the seismic ut originates at the point of support, the response of the equipment and its associated supports 3.7N-116 Rev. 30

ee ranges of equipment/support behavior which affect the magnitude of the seismic eleration are possible:

1. If the equipment is rigid, relative to the structure, the maximum acceleration of the equipment mass approaches that of the structure at the point of equipment support.

The equipment acceleration value in this case corresponds to the low period region of the floor response spectra.

2. If the equipment is very flexible, relative to the structure, the internal distortion of the structure is unimportant and the equipment behaves as though supported on the ground.
3. If the periods of the equipment and supporting structure are nearly equal, resonance occurs and must be taken into account.

N.3.5 Use of Equivalent Static Load Method of Analysis equivalent static load or static analysis method involves the multiplication of the total weight he equipment or component member by the specified seismic acceleration coefficient. The nitude of the seismic acceleration coefficient is established on the basis of the expected amic response characteristics of the component. Components which can be adequately racterized as single degree of freedom systems are considered to have a modal participation or of one. Seismic acceleration coefficients for multi-degree of freedom systems, which may n the resonance region of the amplified response spectra curves, are increased by 50 percent to ount conservatively for the increased modal participation where the equivalent static load hod is used.

N.3.6 Three Components of Earthquake Motion seismic design of the RCS equipment includes the effect of the seismic response of the ports, equipment, structures, and components. Floor response spectra are generated for two pendicular horizontal directions (i.e., N-S, E-W) and the vertical direction. The equipment onse is determined using horizontal and vertical spectra which envelope the appropriate floor onse spectra. The total seismic response is obtained by combining the unidirectional onses in the two horizontal directions and the vertical direction using the square root of the of the squares method.

e history analysis was not used on Millstone 3.

N.3.7 Combination of Modal Responses total unidirectional seismic response is obtained by combining the individual modal onses utilizing the square root of the sum of the squares method. For systems having modes h closely spaced frequencies, this method is modified to include the possible effect of these des. The groups of closely spaced modes are chosen such that the difference between the 3.7N-117 Rev. 30

cessively higher frequencies. No one frequency is in more than one group. Combined total onse for systems which have such closely spaced modal frequencies is obtained by adding to square root of the sum of the squares of all modes the product of the responses of the modes in h group of closely spaced modes and a coupling factor. This can be represented hematically as:

N S Nj - 1 Nj 2 2 RT = Ri + 2 RK Rl EK l

(3.7N-20) i=1 j=1 K = Mj l = K+1 re:

RT = Total unidirectional response Ri = Absolute value of response of mode i N = Total number of modes considered S = Number of groups of closely spaced modes Mj = Lowest modal number associated with group j of closely spaced modes Nj = Highest modal number associated with group j of closely spaced modes Kl = Coupling factor with

' K - ' l 2 - 1 Kl = 1 + ------------------------------------- (3.7N-21)

( ' K K + l l )

2 12

' K = K [ 1 - ( ' K ) ] (3.7N-22 2 -

' K = K + ----------- (3.7N-23)

K td ere:

K = Frequency of closely spaced mode K K = Fraction of critical damping in closely spaced mode K td = Duration of the earthquake 3.7N-118 Rev. 30

er to Section 3.7B.3.9 N.3.9 Multiply Supported Equipment Components with Distinct Inputs en response spectrum methods are used to evaluate reactor coolant system primary ponents interconnected between floors, the procedures of the following paragraphs are used.

primary components of the reactor coolant system are supported at no more than two floor ations.

ynamic response spectrum analysis is first made assuming no relative displacement between port points. The response spectra used in this analysis are the most severe floor response ctra.

ondly, the effect of differential seismic movement of components interconnected between rs is considered statically in the detailed component analysis. The differential motion is luated as a free end displacement in accordance with ASME III, NB-3213.19.

results of these two steps, the dynamic inertia analysis and the static differential motion lysis, are combined absolutely with due consideration for the ASME classification of the sses.

N.3.10 Use of Constant Vertical Static Factors stant vertical load factors are not used as the vertical floor response load for the seismic gn of safety related components and equipment within Westinghouses scope of onsibility.

N.3.11 Torsional Effects of Eccentric Masses er to Section 3.7B.3.11.

N.3.12 Buried Seismic Category I Piping Systems and Tunnels er to Section 3.7B.3.12.

N.3.13 Interaction of Other Piping with Seismic Category I Piping er to Section 3.7B.3.13.

N.3.14 Seismic Analyses for Reactor Internals l assembly component stresses induced by horizontal seismic disturbances are analyzed ugh the use of finite element computer modeling.

time history floor response, based on a standard seismic time history normalized to SSE ls, is used as the seismic input. The reactor internals and the fuel assemblies are modeled as ng and lumped mass systems or beam elements. The component seismic response of the fuel mblies is analyzed to determine design adequacy. A detailed discussion of the analyses 3.7N-119 Rev. 30

l assembly lateral structural damping obtained experimentally is presented in WCAP-8236 endum 1 (1974) and WCAP-8288 Addendum 1 (1974) (Figure 3-4). The data indicate that no ping values less than 10 percent were obtained for fuel assembly displacements greater than inches.

distribution of fuel assembly amplitudes decreases as one approaches the center of the core.

average amplitude for the minimum displacement fuel assembly is well above 0.11 inches for SSE.

l assembly displacement time history for the SSE seismic input is illustrated in WCAP-8236 endum 1 (1974) and WCAP-8288 Addendum 1 (1974) (Figure 2-3).

CRDM's are seismically analyzed to confirm that system stresses under the combined loading ditions, as described in Section 3.9N.1, do not exceed allowable levels as defined by the ME Code,Section III. The CRDM is mathematically modeled as a system of lumped and ributed masses. The model is analyzed under appropriate seismic excitation and the resultant mic bending moments along the length of the CRDM are calculated. The corresponding sses are then combined with the stresses from the other loadings required and the combination hown to meet ASME Code,Section III requirements.

N.3.15 Analysis Procedure for Damping damping values and procedures used for Westinghouse scope of supply and analysis are ussed in Sections 3.7N.1.3 and 3.7N.3.1.

N.4 SEISMIC INSTRUMENTATION er to Section 3.7.4.

3.7N-120 Rev. 30

4.1 Comparison with Regulatory Guide 1.12 lstone 3 complies with Regulatory Guide 1.12 (Section 1.8) with the following exceptions.

SI/ANS 2.2 is used instead of ANSI N18.5 referenced by the Regulatory Guide, since ANSI/

S 2.2 is the final issued version of proposed standard ANSI N18.5.

y solid state digital instrumentation that will enable the processing of data at the plant site hin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the seismic event is used. No peak recording instrumentation is used.

sensor required by section C.1.c.(3) is mounted on the wall of the emergency generator losure approximately 3 feet above the mat near a corner to reduce the potential for equipment age due to flooding. This location will provide representative indication based on the stiffness he supporting structure.

ulatory Position C.3 of the guide is not applicable because the safe shutdown earthquake for lstone 3 is less than 0.3g (Section 2.5.2.6).

4.2 Location and Description of Instrumentation le 3.7-1 lists the location of the seismic instrumentation. Seismic instrumentation is located in s where it can be serviced during periods of unit shutdown. All instruments are oriented to the e azimuths used in the mathematical model to permit direct use of the data within the model.

accelerometers, recorders, and central controllers have instrument characteristics that meet or eed the requirements in Section 5 of ANSI/ANS 2.2-1978.

Triaxial Accelerograph (five provided) e triaxial time-history accelerographs capable of measuring and permanently recording olute acceleration versus time are provided. The instrumentation consists of local triaxial force nce accelerometers, solid state recorders, and a central controller located in the control room.

r accelerometers input to the two centrally located solid state dual panel recorders. The fifth elerometer inputs to a local solid state recorder located in the Auxiliary Building.

elerometers are installed at two locations on the outside of the containment structure in the ineering safety features building. One accelerometer is mounted on the base mat of the tainment at elevation (-) 24 feet - 3 inches and the other directly above the first on the tainment shell at elevation 40 feet - 6 inches.

third accelerometer is installed on the wall of the emergency generator enclosure at elevation et - 6 inches. The vault is part of the emergency generator enclosure building. This location is sen because the emergency generator enclosure building has dynamic response characteristics erent from that of the containment structure. Table 2.5.4-14 describes the differing founding ditions.

fourth accelerometer is located at elevation 46 feet - 6 inches near the charging pump surge located in the Auxiliary Building. This accelerometer replaces the response spectrum rder required at this location by Regulatory Guide 1.12 revision 1 since alternate triaxial rumentation may be provided at this location in accordance with Part 1b of Section 4.1.6 of SI/ANS 2.2-1978.

3.7-122 Rev. 30

d state recorder located in the Auxiliary Building.

Solid State Recorders o solid state dual panel recorders capable of recording motion in all three axes are provided.

se recorders are centrally located in the control room and continuously monitor the input from first four accelerometers described above. Each axis has an independently adjustable trigger.

ny one axis is triggered all three axes of all four sensors are recorded. Event data is stored on h storage cards for later analysis. Each solid state recorder has its own battery to ensure the rder can record events after a loss of AC power. The internal battery provides approximately ours of power autonomy. A main control board alarm annunciates via the alarm relay panel if containment mat accelerometer is triggered. The alarm relay panel is not powered from the S such that the main control board alarm will not annunciate following a loss of external er. A main control board alarm annunciates on a loss of external power.

local solid state recorder located in the Auxiliary Building records the data from the elerometer located on the elevation 51 foot - 4 inch slab of the containment structure internal cent to a steam generator support. The local solid state recorder is similar to the solid state l panel recorders described above except it is stand alone so it will only record data when one s monitored axes exceeds the trigger and it will not trigger the other recorders. Operations will fy power available on a daily basis and the internal battery will ensure the recorder can record nts after a loss of AC power. The internal battery provides approximately 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> of power nomy. Data collected by the local solid state recorder may be evaluated using the central troller to determine the response spectrum.

Central Controller central controller automatically retrieves data from the two dual panel solid state recorders rtly after the recorders are triggered and determines the response spectrum of the event. If the onse spectrum recorded at the containment mat exceeds the OBE criteria an LED is lit on the m relay panel and the central controller display will indicate OBE. The central controller is ered from a UPS such that data can be analyzed after a loss of power. The UPS is sized to er the central controller for more than 25 minutes. The alarm relay panel is not powered from UPS and will not indicate OBE exceedance following a loss of power. A main control board m annunciates on a loss of external power.

central controller may be used to determine the response spectrum of data recorded off the l solid state recorder.

4.3 Control Room Operator Notification recorders located in the control room and the local solid state recorder located in the iliary Building record the signals generated by the accelerometers on flash memory cards. If signals from any axis of any of the first four sensors exceed the trigger value all axes of all sensor record. If the signals from any axis of the sensor mounted on the containment steam 3.7-123 Rev. 30

ny of the first four sensors trigger or if power is lost to the seismic monitoring system (except local solid state recorder) a main control board seismic monitoring warning/trouble alarm unciates.

e containment mat accelerometer triggers, a main control board alarm will annunciate via the m relay panel. The alarm relay panel is not powered from the UPS such that the main control rd will not annunciate following a loss of external power.

central controller will automatically retrieve the data from the two solid state dual panel rders and determine the response spectrum. If the data from the containment mat exceeds E criteria a LED is lit on the alarm relay panel and the central controller display will indicate E. Data from the local solid state recorder must be collected and brought to the central troller or another personal computer with correct software installed to determine the response ctrum.

central controller will continue to operate on UPS supply for approximately 25 minutes owing a loss of external power. The operator may determine OBE exceedance from the central troller during this time. The alarm relay panel is not powered from the UPS and will not cate OBE exceedance following a loss of power.

4.4 Comparison of Measured and Predicted Responses criteria and procedures used to compare recorded data obtained from seismic instrumentation lant design parameters are based on ANSI/ANS 2.10-1979 (Section 3.7.5). Instrumentation uirements are in accordance with ANSI/ANS 2.2-1978 and the supplemental provisions of ulatory Guide 1.12, Revision 1 (Section 3.7.4.1).

5 REFERENCES FOR SECTION 3.7 1 American Nuclear Society 2.2-1978. American National Standard for Earthquake Instrumentation Criteria for Nuclear Power Plants. ANSI/ANI 2.2-1978, La Grange Park, Illinois.

2 American Nuclear Society 1979. American National Standard for Guidelines for Retrieval, Review, Processing, and Evaluation of Records Obtained from Seismic Instrumentation. ANSI/ANS 2.10-1979, La Grange Park, Illinois.

3 Audibert, J.M.E. and Nyman, K.J. 1975. Coefficients for Subgrade Reactions for the Design of Buried Piping. Second ASCE Specialty Conference on Structural Design of Nuclear Power Facilities, Vol 1-A, New Orleans, La., p 109-141, December 8-10, 1975.

4 Biggs, J.M. 1964. Introduction to Structural Dynamics. McGraw-Hill, New York.

5 Chang, T.Y. 1973. Comparison of Seismic Response Combination Procedures for Piping Systems. SWEC, October 1973.

6 Goodling, E.C., Jr. Flexibility Analysis of Buried Pipe. ASME Publication 78-PVP-82.

3.7-124 Rev. 30

8 Goodling, E.C., Jr. More on Flexibility Analysis of Buried Pipe. ASME Publication 80-C2/PVP-67.

9 Hetenyi, M. 1946. Beams on Elastic Foundation. The University of Michigan Press.

10 Johnson, J.J. and Kennedy, R.P. 1977. Earthquake Response of Nuclear Power Facilities.

Presented at the ASCE Fall Convention and Exhibit, San Francisco, Calif., October 17-21, 1977.

11 Kausel, E. and Roesset, J. 1975. Dynamic Stiffness of Circular Foundations. ASCE, Eng.

Mechanics Division.

12 King, R.C. and Crocker, C. 1967. Piping Handbook. McGraw-Hill Book Co.

13 Lin, C.W. 1974. How to Lump the Masses - A Guide to Piping Seismic Analysis. ASME Paper 74-NE-7, Presented at the Pressure Vessels and Piping Conference, Miami, Fla.

14 Lin, Y.Y. and Hanjian, A.H. 1976. Discrete Modeling of Containment Structures.

Presented at the International Symposium on Earthquake Structural Engineering, St.

Louis, Mo.

15 Newmark, N.M. 1972. Earthquake Response Analysis of Reactor Structures. Nuclear Engineering and Design, Vol 20, p 303-332.

16 Newmark, N.M. and Rosenblueth, E. 1971. Fundamentals of Earthquake Engineering.

Prentice-Hall, Inc.

17 Richart, F.E., Jr.; Hall, J.R., Jr.; and Woods, R.D. 1970. Vibrations of Soils and Foundations. Prentice-Hall, Inc., Englewood Cliffs, N.J.

18 Roesset, J.M.; Whitman, R.V.; and Doby, R. 1973. Modal Analysis for Structures with Foundation Interaction. Journal of the Structural Division, Proceedings ASCE, p 399-416.

19 Shah, H.H. and Chu, S.L. 1974. Seismic Analysis of Underground Structural Elements.

Journal of the Power Division, Proceedings of ASCE, Vol 100, No. P01, p 53-62, July 1974.

20 Singh, A.K.; Chu, S.L.; and Singh, S.L. 1973. Influence of Closely Spaced Modes in Response Spectrum Method of Analysis. Presented at the Specialty Conference on Structural Design of Nuclear Plant Facilities, Chicago, Ill.

21 Terzaghi, K. 1955. Evaluation of Coefficients of Subgrade Reaction. Geotechnique, p 297-325.

3.7-125 Rev. 30

23 U.S. Nuclear Regulatory Commission (NRC) 1975. Seismic System Analysis. Standard Review Plan, Section 3.7.2, NUREG-75/087.

24 Wass, G. 1972. Linear Two-Dimensional Analysis of Soil Dynamics in Semi-Infinite Layered Media. PhD Thesis, Univ. of Calif., Berkeley, Presented to J. Lysmer.

25 WCAP-7921-AR, 1974. Damping Values of Nuclear Plant Components.

26 WCAP-7950, 1972. Gesinski, T.L. Fuel Assembly Safety Analysis for Combined Seismic and Loss-of-Coolant Accident.

27 WCAP-8236 (Proprietary) 1973, and WCAP-8288 (Non proprietary) 1974. Gesinki, T.L.

and Chiang, D. Safety Analysis of the 17 x 17 Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident.

28 WCAP-8236, Addendum 1 (Proprietary), 1974 and WCAP-8288, Addendum 1, (Non proprietary), 1974. Safety Analysis of the 8-Grid 17 x 17 Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident.

29 Whitman, R.V. and Richart, F.E., Jr. 1967. Design Procedures for Dynamically Loaded Foundations. Journal of the Soil Mechanics and Foundation Division, Proceedings of the ASCE.

30 Whitman, R.V. 1970. Soil Structure Interaction. In: Seismic Design for Nuclear Power Plants, MIT Press, Cambridge, Mass., p 245-269.

31 Yeh, G.C.K. 1974. Seismic Analysis of Slender Buried Beams. Bulletin of the Seismological Society of America, Vol 64, No. 5, p 1, 555-1, 562.

3.7-126 Rev. 30

Triaxial Force Balance Main Control Board Instrument Location Accelerometer Annunciation Containment Structure - Outside Mat, elevation (-) 24 feet 3 inches X (1) X(2)

Containment shell, elevation 40 feet 6 inches X (2)

Containment Structure - Inside Steam Generator Support, elevation 51 feet 4 inches X (3)

Emergency Generator Enclosure Located on Mat in Diesel Fuel Oil X (1) (4)

Vault, elevation 4 feet 6 inches Auxiliary Building Surge Tank Support, elevation 46 feet 6 inches X (1)

NOTES:

(1) All four sensors record event sensed by any axis of any of the four sensors.

(2) Main control board annunciation of OBE exceedance on containment mat sensor only.

(3) Data must be collected from Etna recorder in Auxiliary Building and may be evaluated on central controller.

(4) This sensor is mounted to the wall approximately 3 feet above the mat near a corner to reduce the potential for equipment dam due to flooding. This location will provide representative indication based on the stiffness of the supporting structure.

3.7-127 Rev

TABLE 3.7B-1 DAMPING FACTORS Type of Condition of Structure, System or Percent of Critical Stress Level Component Damping Low stress, well below Steel, reinforced concrete; no 0.5 to (1) proportional limit. Stresses cracking and no slipping at below 0.25 yield point stress joints, piping or components Working stress limited to 0.5 a. Welded steel, well reinforced 2 yield point stress concrete (with only slight cracking)

b. Bolted steel 5 At or just below yield point a. Welded steel 5
b. Reinforced concrete 5
c. Bolted steel 7 At all stress levels a. Rock (translation) 10
b. Rock (rotation) 5 TE:

For final reconciliation of pipe stress analysis or piping system backfits, damping values as defined in ASME Code Case N-411 (Figure 3.7B-71) may be utilized for both OBE and SSE.

3.7-128 Rev. 30

TABLE 3.7B-2 METHODS OF SEISMIC ANALYSIS USED FOR SEISMIC CATEGORY I STRUCTURES Response Modal Time Frequency Spectrum History Domain Time Structure Analysis Analysis History Analysis actor containment and internals X X in steam valve building X X drogen recombiner building X X eguards area X X xiliary building X X el building X X ntrol building X vice building X X ergency generator enclosure X rvice water pumphouse X X fueling water storage and chemical X X dition tanks mineralized water storage tank and X X closure el building canopy X X rbine building X X ste disposal building X 3.7-129 Rev. 30

TABLE 3.7B-3 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT ODAL FREQUENCIES AND PARTICIPATION FACTORS UNCRACKED MODEL Participation Factors Mode Frequency (Hz) Horizontal Vertical N-S E-W 1 4.664 33.88 424.20 4.67 2 4.751 353.06 -30.44 -3.28 3 4.854 3.98 37.96 -4.09 4 5.565 252.24 -42.17 -32.00 5 6.135 53.40 86.74 -1.22 6 8.244 0.11 0.89 0.10 7 10.917 3.42 -177.61 3.11 8 10.921 170.86 3.10 8.60 9 11.809 13.83 -99.23 17.97 10 12.630 78.31 -58.48 399.46 11 12.881 10.41 -117.44 -228.28 12 13.295 157.36 61.05 -62.87 13 13.607 8.36 8.53 0.17 14 15.592 42.13 4.40 -241.13 15 18.751 44.54 -110.37 -0.04 16 18.755 108.26 43.75 1.45 17 22.152 2.60 -0.14 -1.10 18 22.157 0.48 0.95 -0.06 3.7-130 Rev. 30

TABLE 3.7B-4 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODAL FREQUENCIES AND PARTICIPATION FACTORS CRACKED MODEL Participation Factors Vertical Mode Frequency (Hz) Horizontal Earthquake Earthquake N-S E-W 1 3.4236 14.97 317.58 0.22 2 3.4239 315.07 -14.92 -0.25 3 4.750 22.93 277.82 6.24 4 5.544 295.89 -48.38 -32.19 5 5.863 9.33 10.31 -0.51 6 6.134 57.44 95.54 -1.06 7 7.761 15.69 155.83 -0.35 8 7.762 153.44 -15.40 2.02 9 9.371 4.21 -1.29 376.49 10 9.695 0.25 2.02 -0.03 11 11.801 15.39 -118.13 10.09 12 12.758 87.27 -136.84 79.36 13 13.135 190.02 98.59 94.36 14 13.375 37.78 -72.22 -17.73 15 13.428 4.91 3.73 -63.01 16 15.388 44.63 4.92 -324.04 17 15.763 0.88 1.23 -0.20 18 15.766 2.09 0.49 -29.84 3.7-131 Rev. 30

TABLE 3.7B-5 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) UNCRACKED MODEL Mode Joint Displacement X Y Z 1 1 0.001 0.0 0.016 2 0.012 0.006 0.153 3 0.023 0.009 0.336 4 0.016 0.005 0.338 5 0.021 0.003 0.418 6 0.043 0.018 0.442 7 0.054 0.013 0.646 8 0.075 0.011 0.916 9 0.078 0.007 1.0 10 0.008 0.0 0.098 11 0.016 0.0 0.20 12 0.025 0.0 0.310 13 0.035 0.0 0.425 14 0.044 0.0 0.536 15 0.051 0.0 0.621 16 0.063 0.0 0.776 17 0.079 0.0 0.972 2 1 0.009 0.0 0.0 2 0.031 0.005 -0.005 3 0.051 -0.006 -0.005 4 0.060 0.0 -0.007 5 0.067 -0.005 -0.006 6 0.074 -0.003 -0.007 7 0.102 -0.007 -0.009 8 0.143 -0.013 -0.011 9 0.166 -0.006 -0.014 10 0.093 0.0 -0.008 3.7-132 Rev. 30

Mode Joint Displacement X Y Z 11 0.198 0.0 -0.017 12 0.312 0.0 -0.027 13 0.430 0.0 -0.037 14 0.545 0.0 -0.047 15 0.633 0.0 -0.054 16 0.795 0.0 -0.068 17 1.0 0.0 -0.086 3 1 0.0 0.0 0.001 2 -0.010 -0.006 -0.104 3 -0.021 -0.007 -0.264 4 -0.014 -0.004 -0.257 5 -0.017 -0.002 -0.326 6 -0.038 -0.014 -0.342 7 -0.048 -0.010 -0.508 8 -0.066 -0.009 -0.722 9 -0.066 -0.006 -0.776 10 0.008 0.0 0.083 11 0.017 0.0 0.188 12 0.028 0.0 0.303 13 0.038 0.0 0.421 14 0.049 0.0 0.538 15 0.057 0.0 0.626 16 0.072 0.0 0.790 17 0.091 0.0 1.0 4 1 0.009 0.0 -0.002 2 0.132 0.007 -0.040 3 0.270 -0.053 -0.039 4 0.327 -0.016 -0.063 3.7-133 Rev. 30

Mode Joint Displacement X Y Z 5 0.370 -0.048 -0.057 6 0.417 -0.038 -0.070 7 0.590 -0.062 -0.093 8 0.858 -0.094 -0.142 9 1.0 -0.053 -0.191 10 -0.002 -0.001 0.0 11 -0.020 -0.001 0.003 12 -0.040 -0.001 0.007 13 -0.062 -0.001 0.010 14 -0.082 -0.001 0.014 15 -0.098 -0.001 0.016 16 -0.129 -0.001 0.022 17 -0.171 -0.001 0.028 5 1 0.004 0.0 0.006 2 0.010 0.012 0.107 3 0.052 -0.006 -0.010 4 0.104 0.002 0.147 5 0.121 -0.012 0.10 6 0.094 0.010 0.173 7 0.182 -0.004 0.290 8 0.326 -0.018 0.640 9 0.437 -0.008 1.0 10 0.001 0.0 0.002 11 -0.003 0.0 -0.005 12 -0.007 0.0 -0.013 13 -0.013 0.0 -0.022 14 -0.018 0.0 -0.031 15 -0.022 0.0 -0.038 3.7-134 Rev. 30

Mode Joint Displacement X Y Z 16 -0.030 0.0 -0.052 17 -0.041 0.0 -0.071 6 1 0.0 0.0 -0.004 2 -0.081 0.007 -0.285 3 -0.036 0.006 -0.300 4 -0.024 0.005 -0.240 5 -0.020 0.001 -0.235 6 -0.030 0.017 -0.193 7 -0.005 0.011 -0.061 8 0.089 0.006 0.582 9 0.166 0.005 1.0 10 0.0 0.0 -0.004 11 -0.001 0.0 -0.005 12 -0.001 0.001 -0.005 13 -0.001 0.001 -0.005 14 -0.001 0.001 -0.005 15 -0.001 0.001 -0.005 16 -0.001 0.001 -0.004 17 -0.001 0.001 -0.004 7 1 0.001 0.0 -0.028 2 0.005 0.001 -0.058 3 0.005 0.004 -0.055 4 0.003 0.002 -0.059 5 0.002 0.003 -0.051 6 0.003 0.005 -0.049 7 -0.001 0.006 -0.019 8 -0.007 0.007 0.076 9 -0.011 0.004 0.116 3.7-135 Rev. 30

Mode Joint Displacement X Y Z 10 0.004 0.001 -0.235 11 0.008 0.001 -0.429 12 0.010 0.001 -0.552 13 0.011 0.001 -0.583 14 0.009 0.001 -0.516 15 0.007 0.001 -0.391 16 -0.001 0.001 0.084 17 -0.018 0.001 1.0 8 1 -0.027 -0.001 0.0 2 -0.036 -0.002 0.0 3 -0.038 -0.012 0.0 4 -0.034 -0.007 0.0 5 -0.030 -0.012 0.0 6 -0.025 -0.010 0.0 7 -0.002 -0.015 0.0 8 0.049 -0.021 -0.001 9 0.076 -0.014 -0.001 10 -0.234 -0.001 -0.004 11 -0.429 -0.002 -0.008 12 -0.552 -0.002 -0.010 13 -0.584 -0.003 -0.011 14 -0.517 -0.003 -0.010 15 -0.393 -0.003 -0.007 16 0.083 -0.003 0.002 17 1.0 -0.004 0.019 9 1 0.003 0.003 -0.024 2 0.090 -0.002 -0.389 3 0.110 0.025 -0.337 3.7-136 Rev. 30

Mode Joint Displacement X Y Z 4 0.065 0.010 -0.496 5 0.051 0.027 -0.423 6 0.083 0.022 -0.481 7 -0.012 0.038 -0.412 8 -0.107 0.057 0.658 9 -0.165 0.029 1.0 10 0.0 0.006 0.002 11 -0.003 0.008 0.033 12 -0.005 0.011 0.058 13 -0.007 0.013 0.073 14 -0.007 0.014 0.075 15 -0.005 0.015 0.065 16 0.001 0.016 0.011 17 0.016 0.017 -0.104 10 1 0.026 0.102 -0.020 2 0.253 0.198 -0.273 3 0.408 0.350 -0.411 4 0.418 0.314 -0.331 5 0.396 0.365 -0.327 6 0.355 0.388 -0.267 7 0.193 0.459 -0.001 8 -0.315 0.551 0.227 9 -0.603 0.482 0.441 10 0.013 0.252 -0.009 11 -0.006 0.412 0.003 12 -0.023 0.556 0.015 13 -0.036 0.681 0.023 3.7-137 Rev. 30

Mode Joint Displacement X Y Z 14 -0.041 0.780 0.027 15 -0.039 0.837 0.025 16 -0.012 0.913 0.005 17 0.054 1.0 -0.045 11 1 0.004 -0.060 -0.041 2 0.058 -0.077 -0.542 3 0.077 -0.070 -0.784 4 0.091 -0.10 -0.670 5 0.086 -0.095 -0.653 6 0.061 -0.054 -0.557 7 0.035 -0.065 -0.056 8 -0.157 -0.069 0.555 9 -0.259 -0.116 1.0 10 0.002 -0.183 -0.023 11 0.0 -0.314 0.001 12 -0.002 -0.433 0.024 13 -0.003 -0.536 0.042 14 -0.004 -0.618 0.051 15 -0.004 -0.666 0.049 16 0.0 -0.729 0.013 17 0.008 -0.801 -0.080 12 1 -0.043 0.013 -0.016 2 -0.368 -0.010 -0.191 3 -0.615 -0.136 -0.250 4 -0.609 -0.087 -0.238 5 -0.575 -0.147 -0.229 6 -0.536 -0.118 -0.210 7 -0.255 -0.189 -0.065 3.7-138 Rev. 30

Mode Joint Displacement X Y Z 8 0.553 -0.272 0.211 9 1.0 -0.181 0.361 10 -0.028 0.079 -0.011 11 -0.006 0.150 -0.002 12 0.018 0.214 0.007 13 0.037 0.270 0.015 14 0.049 0.316 0.019 15 0.049 0.342 0.019 16 0.019 0.377 0.007 17 -0.064 0.417 -0.025 13 1 -0.041 0.0 -0.042 2 -0.255 -0.047 -0.364 3 -0.529 -0.168 -0.489 4 -0.549 -0.133 -0.566 5 -0.525 -0.184 -0.531 6 -0.472 -0.159 -0.545 7 -0.257 -0.225 -0.324 8 0.448 -0.305 0.636 9 0.831 -0.240 1.0 10 -0.030 0.046 -0.031 11 -0.010 0.096 -0.010 12 0.012 0.141 0.014 13 0.032 0.181 0.034 14 0.044 0.213 0.047 15 0.046 0.232 0.049 16 0.021 0.257 0.023 17 -0.052 0.286 -0.051 14 1 -0.017 0.073 -0.002 3.7-139 Rev. 30

Mode Joint Displacement X Y Z 2 -0.142 0.348 0.005 3 -0.246 0.543 0.035 4 -0.251 0.599 0.022 5 -0.243 0.627 0.022 6 -0.228 0.659 0.013 7 -0.10 0.767 -0.032 8 0.543 0.909 -0.091 9 0.854 1.0 -0.128 10 -0.017 0.010 -0.002 11 -0.015 -0.061 -0.001 12 -0.008 -0.129 0.0 13 0.001 -0.190 0.001 14 0.010 -0.240 0.002 15 0.015 -0.269 0.003 16 0.007 -0.309 0.003 17 -0.027 -0.357 0.002 15 1 -0.034 0.0 0.084 2 -0.012 0.002 0.019 3 0.009 -0.001 -0.022 4 0.014 0.0 -0.029 5 0.017 0.0 -0.036 6 0.018 0.0 -0.039 7 0.019 0.001 -0.036 8 -0.026 0.0 0.045 9 -0.048 -0.005 0.086 10 -0.263 0.0 0.650 11 -0.404 0.0 1.0 12 -0.367 0.0 0.909 3.7-140 Rev. 30

Mode Joint Displacement X Y Z 13 -0.169 0.0 0.419 14 0.105 0.0 -0.259 15 0.297 0.0 -0.736 16 0.329 0.0 -0.815 17 -0.194 0.0 0.481 16 1 0.082 0.001 0.033 2 0.029 -0.002 0.005 3 -0.030 0.003 -0.011 4 -0.043 0.003 -0.014 5 -0.049 0.004 -0.018 6 -0.053 0.004 -0.019 7 -0.051 0.004 -0.017 8 0.070 0.007 0.020 9 0.129 0.016 0.037 10 0.649 0.001 0.262 11 1.0 -0.0 0.404 12 0.910 -0.0 0.368 13 0.420 -0.001 0.170 14 -0.259 -0.001 -0.104 15 -0.736 -0.001 -0.297 16 -0.815 -0.001 -0.329 17 0.481 -0.002 -0.195 17 1 0.002 -0.001 0.0 2 -0.002 -0.017 -0.001 3 0.004 -0.001 -0.001 4 0.002 -0.009 0.0 5 -0.002 -0.002 0.001 6 -0.005 -0.004 0.002 3.7-141 Rev. 30

Mode Joint Displacement X Y Z 7 -0.012 0.001 0.002 8 0.060 0.010 -0.009 9 0.075 0.007 -0.011 10 -0.059 0.0 0.010 11 -0.155 0.0 0.027 12 -0.230 0.0 0.039 13 -0.230 0.0 0.039 14 -0.141 0.0 0.024 15 -0.020 0.0 0.004 16 0.234 0.001 -0.040 17 1.0 0.001 -0.172 18 1 0.0 0.0 0.001 2 -0.002 0.001 -0.004 3 0.0 0.0 -0.008 4 0.001 0.0 -0.006 5 0.001 0.001 -0.006 6 0.0 -0.001 -0.004 7 0.002 0.0 0.014 8 0.005 0.0 0.024 9 0.004 0.0 0.020 10 -0.010 0.0 -0.059 11 -0.027 0.0 -0.154 12 -0.039 0.0 -0.228 13 -0.040 0.0 -0.229 14 -0.024 0.0 -0.141 15 -0.004 0.0 -0.021 16 0.040 0.0 0.233 17 0.172 0.0 1.0 3.7-142 Rev. 30

TABLE 3.7B-6 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) CRACKED MODEL Mode Joint Displacement X Y Z 1 1 0.0 0.0 0.004 2 0.0 0.0 0.010 3 0.001 0.0 0.016 4 0.001 0.0 0.017 5 0.001 0.0 0.019 6 0.001 0.001 0.020 7 0.001 0.0 0.028 8 0.002 0.0 0.038 9 0.002 0.0 0.043 10 0.004 0.0 0.086 11 0.009 0.0 0.191 12 0.014 0.0 0.306 13 0.020 0.0 0.424 14 0.026 0.0 0.540 15 0.030 0.0 0.628 16 0.037 0.0 0.791 17 0.047 0.0 1.0 2 1 0.004 0.0 0.0 2 0.009 0.002 -0.001 3 0.013 -0.001 -0.001 4 0.014 0.001 -0.001 5 0.016 0.0 -0.001 6 0.017 0.0 -0.001 7 0.023 -0.001 -0.001 8 0.030 -0.002 -0.001 9 0.035 -0.001 -0.002 3.7-143 Rev. 30

Mode Joint Displacement X Y Z 10 0.086 0.0 -0.004 11 0.190 0.0 -0.009 12 0.305 0.0 -0.014 13 0.424 0.0 -0.020 14 0.540 0.0 -0.026 15 0.628 0.0 -0.030 16 0.791 0.0 -0.037 17 1.0 0.0 -0.047 3 1 0.001 0.0 0.008 2 0.012 0.007 0.145 3 0.025 0.010 0.338 4 0.017 0.005 0.335 5 0.021 0.003 0.419 6 0.045 0.018 0.441 7 0.056 0.013 0.650 8 0.079 0.011 0.922 9 0.081 0.008 1.0 10 0.001 0.0 0.006 11 0.0 0.0 0.001 12 -0.001 0.0 -0.006 13 -0.001 0.0 -0.014 14 -0.002 0.0 -0.022 15 -0.002 0.0 -0.028 16 -0.004 0.0 -0.041 17 -0.005 0.0 -0.060 4 1 0.010 -0.001 -0.002 3.7-144 Rev. 30

Mode Joint Displacement X Y Z 2 0.134 0.008 -0.040 3 0.271 -0.053 -0.038 4 0.328 -0.015 -0.062 5 0.371 -0.047 -0.057 6 0.416 -0.038 -0.069 7 0.590 -0.061 -0.091 8 0.858 -0.094 -0.138 9 1.0 -0.052 -0.186 10 0.011 -0.001 -0.002 11 0.007 -0.001 -0.001 12 0.002 -0.001 0.0 13 -0.005 -0.001 0.001 14 -0.013 -0.001 0.002 15 -0.019 -0.001 0.003 16 -0.033 -0.001 0.005 17 -0.054 -0.001 0.009 5 1 0.006 0.0 0.007 2 0.052 0.014 0.151 3 0.119 -0.017 0.043 4 0.184 -0.001 0.195 5 0.211 -0.022 0.148 6 0.195 0.002 0.224 7 0.322 -0.018 0.345 8 0.521 -0.039 0.666 9 0.659 -0.019 1.0 10 0.007 0.0 0.008 11 0.006 0.0 0.006 3.7-145 Rev. 30

Mode Joint Displacement X Y Z 12 0.003 0.0 0.003 13 -0.001 0.0 -0.002 14 -0.005 0.0 -0.007 15 -0.009 0.0 -0.011 16 -0.018 0.0 -0.021 17 -0.032 0.0 -0.037 6 1 0.004 0.0 0.006 2 0.010 0.012 0.107 3 0.050 -0.006 -0.011 4 0.102 0.002 0.146 5 0.119 -0.011 0.094 6 0.092 0.010 0.172 7 0.178 -0.004 0.288 8 0.320 -0.018 0.638 9 0.430 -0.007 1.0 10 0.005 0.0 0.008 11 0.004 0.0 0.007 12 0.003 0.0 0.004 13 0.0 0.0 0.0 14 -0.002 0.0 -0.004 15 -0.004 0.0 -0.008 16 -0.010 0.0 -0.018 17 -0.020 0.0 -0.034 7 1 -0.001 0.0 -0.013 2 -0.001 0.0 -0.011 3 -0.001 0.0 -0.007 3.7-146 Rev. 30

Mode Joint Displacement X Y Z 4 -0.001 0.0 -0.005 5 0.0 0.001 -0.003 6 0.0 0.001 7 0.001 0.001 0.008 8 0.003 0.0 0.026 9 0.004 0.0 0.036 10 -0.023 0.0 -0.222 11 -0.042 0.0 -0.422 12 -0.055 0.0 -0.552 13 -0.059 0.0 -0.589 14 -0.053 0.0 -0.526 15 -0.041 0.0 -0.403 16 0.007 0.0 0.074 17 0.101 0.0 1.0 8 1 -0.012 0.0 0.001 2 -0.009 0.0 0.0 3 -0.004 -0.003 0.0 4 -0.001 -0.002 0.0 5 0.001 -0.003 -0.001 6 0.004 -0.003 -0.001 7 0.013 -0.004 -0.001 8 0.030 -0.006 -0.002 9 0.039 -0.004 -0.002 10 -0.222 0.0 0.022 11 -0.422 0.0 0.042 12 -0.552 0.0 0.056 13 -0.589 0.0 0.059 3.7-147 Rev. 30

Mode Joint Displacement X Y Z 14 -0.526 0.0 0.053 15 -0.403 0.0 0.041 16 0.074 0.0 -0.007 17 1.0 0.0 -0.101 9 1 0.001 0.035 0.0 2 0.005 0.043 -0.002 3 0.009 0.050 -0.003 4 0.009 0.050 -0.003 5 0.009 0.052 -0.003 6 0.009 0.052 -0.003 7 0.008 0.056 -0.002 8 0.007 0.060 -0.002 9 0.006 0.060 -0.002 10 0.0 0.193 0.0 11 0.0 0.363 0.0 12 0.0 0.517 0.0 13 -0.001 0.651 0.0 14 -0.001 0.759 0.0 15 -0.001 0.821 0.0 16 0.0 0.904 0.0 17 0.0 1.0 0.0 10 1 -0.002 0.0 -0.014 2 -0.063 0.002 -0.426 3 -0.027 0.006 -0.378 4 -0.040 0.002 -0.410 5 -0.041 0.002 -0.361 3.7-148 Rev. 30

Mode Joint Displacement X Y Z 6 -0.028 0.014 -0.357 7 -0.033 0.012 -0.220 8 0.035 0.012 0.605 9 0.085 0.005 1.0 10 -0.001 -0.001 -0.010 11 0.0 -0.003 -0.004 12 0.001 -0.005 0.003 13 0.001 -0.006 0.010 14 0.002 -0.007 0.014 15 0.002 -0.008 0.015 16 0.001 -0.009 0.006 17 -0.002 -0.010 -0.018 11 1 0.004 0.002 -0.029 2 0.090 -0.004 -0.392 3 0.108 0.023 -0.341 4 0.064 0.008 -0.497 5 0.050 0.024 -0.424 6 0.082 0.019 -0.480 7 -0.012 0.035 -0.406 8 -0.106 0.054 0.657 9 -0.164 0.025 1.0 10 0.005 0.001 -0.039 11 0.005 -0.001 -0.036 12 0.004 -0.002 -0.021 13 0.001 -0.003 0.001 14 -0.002 -0.004 0.025 15 -0.004 -0.004 0.037 3.7-149 Rev. 30

Mode Joint Displacement X Y Z 16 -0.003 -0.005 0.032 17 0.007 -0.006 -0.018 12 1 0.031 0.022 -0.050 2 0.270 0.081 -0.571 3 0.424 0.211 -0.840 4 0.440 0.157 -0.696 5 0.417 0.204 -0.683 6 0.363 0.247 -0.569 7 0.193 0.299 -0.027 8 -0.410 0.372 0.541 9 -0.746 0.277 1.0 10 0.077 0.012 -0.123 11 0.097 0.001 -0.158 12 0.078 -0.010 -0.128 13 0.027 -0.021 -0.046 14 -0.036 -0.030 0.056 15 -0.077 -0.035 0.123 16 -0.076 -0.043 0.121 17 0.051 -0.052 -0.095 13 1 -0.061 -0.023 -0.032 2 -0.395 -0.069 -0.281 3 -0.648 -0.219 -0.376 4 -0.638 -0.174 -0.346 5 -0.601 -0.241 -0.334 6 -0.562 -0.206 -0.299 7 -0.266 -0.291 -0.071 3.7-150 Rev. 30

Mode Joint Displacement X Y Z 8 0.547 -0.392 0.308 9 1.0 -0.304 0.530 10 -0.287 -0.014 -0.151 11 -0.418 -0.003 -0.220 12 -0.368 0.008 -0.194 13 -0.159 0.019 -0.084 14 0.120 0.028 0.063 15 0.312 0.033 0.164 16 0.337 0.041 0.177 17 -0.199 0.051 -0.107 14 1 -0.015 0.005 0.030 2 0.040 0.022 -0.104 3 0.077 0.047 -0.162 4 0.080 0.039 -0.148 5 0.076 0.048 -0.145 6 0.070 0.055 -0.130 7 0.040 0.067 -0.036 8 -0.070 0.083 0.126 9 -0.132 0.065 0.228 10 -0.328 0.004 0.619 11 -0.530 0.001 1.0 12 -0.495 -0.002 0.933 13 -0.239 -0.004 0.450 14 0.126 -0.006 -0.238 15 0.386 -0.007 -0.728 16 0.440 -0.010 -0.830 17 -0.240 -0.012 0.451 3.7-151 Rev. 30

Mode Joint Displacement X Y Z 15 1 0.002 -0.022 0.002 2 -0.283 -0.075 -0.214 3 -0.506 -0.199 -0.297 4 -0.505 -0.167 -0.279 5 -0.480 -0.220 -0.270 6 -0.451 -0.195 -0.247 7 -0.228 -0.267 -0.071 8 0.418 -0.351 0.257 9 0.777 -0.284 0.441 10 0.603 -0.015 0.323 11 1.0 -0.004 0.534 12 0.945 0.006 0.505 13 0.466 0.016 0.248 14 -0.227 0.024 -0.121 15 -0.724 0.030 -0.387 16 -0.835 0.037 -0.446 17 0.449 0.047 0.237 16 1 -0.017 0.094 -0.002 2 -0.159 0.367 0.005 3 -0.275 0.553 0.035 4 -0.279 0.611 0.021 5 -0.270 0.635 0.021 6 -0.252 0.668 0.012 7 -0.109 0.770 -0.034 8 0.566 0.905 -0.090 9 0.897 1.0 -0.127 3.7-152 Rev. 30

Mode Joint Displacement X Y Z 10 0.007 0.074 -0.004 11 0.034 0.043 -0.006 12 0.048 0.008 -0.007 13 0.041 -0.028 -0.006 14 0.015 -0.061 -0.003 15 -0.012 -0.083 0.0 16 -0.048 -0.114 0.008 17 -0.105 -0.154 0.035 17 1 0.0 0.0 0.0 2 -0.001 0.0 -0.001 3 0.001 0.0 -0.002 4 0.002 0.0 0.0 5 0.002 0.0 0.0 6 0.002 0.0 0.001 7 0.002 0.0 0.008 8 0.002 0.0 0.008 9 0.002 0.0 0.010 10 -0.013 0.0 -0.044 11 -0.040 0.0 -0.134 12 -0.064 0.0 -0.215 13 -0.069 0.0 -0.229 14 -0.046 0.0 -0.154 15 -0.012 0.0 -0.039 16 0.064 0.0 0.214 17 0.298 0.0 1.0 18 1 -0.001 0.007 0.0 3.7-153 Rev. 30

Mode Joint Displacement X Y Z 2 -0.014 0.022 0.001 3 -0.019 0.041 0.004 4 -0.018 0.044 0.003 5 -0.018 0.047 0.003 6 -0.016 0.049 0.002 7 -0.004 0.058 -0.003 8 0.058 0.070 -0.011 9 0.085 0.077 -0.016 10 -0.044 0.006 0.013 11 -0.133 0.004 0.040 12 -0.214 0.001 0.064 13 -0.228 -0.002 0.068 14 -0.154 -0.004 0.046 15 -0.040 -0.006 0.012 16 0.212 -0.009 -0.064 17 1.0 -0.012 -0.298 3.7-154 Rev. 30

DISPLACEMENTS FOR SAFE SHUTDOWN EARTHQUAKE UNCRACKED MODEL, RESPONSE SPECTRUM ANALYSIS Degree of

  • Freedom Acceleration ** Displacement ***

Horizontal Vertical Horizontal Vertical N-S E-W N-S E-W 1 0.058 0.026 0.012 0.031xE-3 0.05xE-3 0.031xE-3 2 0.013 0.007 0.048 0.035xE-3 0.034xE-3 0.112xE-3 3 0.026 0.063 0.006 0.050xE-3 0.396xE-3 0.030xE-3 4 0.627xE-4 0.254xE-3 0.646xE-4 0.010xE-4 0.084xE-4 0.001xE-4 5 0.874xE-4 0.559xE-4 0.335xE-4 0.003xE-4 0.007xE-4 0.0 6 0.232xE-3 0.531xE-4 0.973xE-4 0.073xE-4 0.010xE-4 0.002xE-4 7 0.163 0.046 0.063 0.224xE-2 0.047xE-2 0.032xE-2 8 0.028 0.013 0.068 0.201xE-3 0.172xE-3 0.251xE-3 9 0.054 0.163 0.080 0.740xE-3 0.382xE-2 0.403xE-3 10 0.862xE-3 0.211xE-2 0.102xE-2 0.165xE-4 0.702xE-4 0.045xE-4 11 0.242xE-2 0.421xE-2 0.111xE-2 0.280xE-4 0.911xE-4 0.047xE-4 12 0.176xE-2 0.416xE-3 0.646xE-3 0.446xE-4 0.077xE-4 0.043xE-4 13 0.192 0.053 0.091 0.443xE-2 0.094xE-2 0.056xE-2 14 0.052 0.021 0.101 0.850xE-3 0.287xE-3 0.425xE-3 15 0.059 0.238 0.118 0.929xE-3 0.834xE-2 0.594xE-3 16 0.926xE-3 0.287xE-2 0.150xE-2 0.140xE-4 0.911xE-4 0.076xE-4 17 0.148xE-2 0.403xE-2 0.134xE-2 0.347xE-4 0.134xE-3 0.071xE-4 3.7-155 Rev

ANALYSIS Degree of

  • Freedom Acceleration ** Displacement ***

Horizontal Vertical Horizontal Vertical N-S E-W N-S E-W 18 0.262xE-2 0.720xE-3 0.107xE-2 0.635xE-4 0.148xE-4 0.068xE-4 19 0.222 0.056 0.092 0.532xE-2 0.101xE-2 0.061xE-2 20 0.041 0.019 0.101 0.277xE-3 0.150xE-3 0.408xE-3 21 0.061 0.239 0.098 0.123xE-2 0.839xE-2 0.50xE-3 22 0.971xE-3 0.305xE-2 0.170xE-2 0.152xE-4 0.976xE-4 0.086xE-4 23 0.153xE-2 0.415xE-2 0.129xE-2 0.364xE-4 0.137xE-3 0.069xE-4 24 0.285xE-2 0.789xE-3 0.115xE-2 0.686xE-4 0.160xE-4 0.075xE-4 25 0.244 0.059 0.089 0.601xE-2 0.116xE-2 0.064xE-2 26 0.052 0.021 0.110 0.767xE-3 0.173xE-3 0.459xE-3 27 0.062 0.289 0.096 0.125xE-2 0.104xE-1 0.049xE-2 28 0.101xE-2 0.321xE-2 0.189xE-2 0.162xE-4 0.103xE-3 0.095xE-4 29 0.159xE-2 0.427xE-2 0.124xE-2 0.380xE-4 0.141xE-3 0.068xE-4 30 0.298xE-2 0.831xE-3 0.122xE-2 0.713xE-4 0.165xE-4 0.079xE-4 31 0.269 0.064 0.082 0.675xE-2 0.299xE-2 0.066xE-2 32 0.048 0.023 0.115 0.620xE-3 0.467xE-3 0.481xE-3 33 0.071 0.305 0.081 0.144xE-2 0.110xE-1 0.042xE-2 34 0.106xE-2 0.333xE-2 0.207xE-2 0.170xE-4 0.107xE-3 0.104xE-4 3.7-156 Rev

ANALYSIS Degree of

  • Freedom Acceleration ** Displacement ***

Horizontal Vertical Horizontal Vertical N-S E-W N-S E-W 35 0.165xE-2 0.436xE-2 0.120xE-2 0.391xE-4 0.144xE-3 0.066xE-4 36 0.309xE-2 0.870xE-3 0.129xE-2 0.735xE-4 0.170xE-4 0.083xE-4 37 0.358 0.054 0.075 0.951xE-2 0.210xE-3 0.079xE-2 38 0.060 0.026 0.138 0.991xE-3 0.380xE-3 0.569xE-3 39 0.071 0.433 0.013 0.20xE-2 0.160xE-1 0.019xE-2 40 0.122xE-2 0.364xE-2 0.261xE-2 0.192xE-4 0.118xE-3 0.131xE-4 41 0.193xE-2 0.481xE-2 0.658xE-3 0.456xE-4 0.161xE-3 0.046xE-4 42 0.349xE-2 0.101xE-2 0.158xE-2 0.804xE-4 0.185xE-4 0.097xE-4 43 0.523 0.115 0.111 0.138xE-1 0.305xE-2 0.119xE-2 44 0.101 0.036 0.166 0.152xE-2 0.402xE-3 0.685xE-3 45 0.120 0.623 0.078 0.307xE-2 0.228xE-1 0.047xE-2 46 0.154xE-2 0.416xE-2 0.311xE-2 0.210xE-4 0.125xE-3 0.156xE-4 47 0.307xE-2 0.609xE-2 0.314xE-2 0.697xE-4 0.171xE-3 0.164xE-4 48 0.414xE-2 0.127xE-2 0.212xE-2 0.886xE-4 0.272xE-4 0.119xE-4 49 0.622 0.144 0.180 0.161xE-1 0.348xE-2 0.149xE-2 50 0.101 0.035 0.175 0.857xE-3 0.262xE-3 0.652xE-3 51 0.170 0.698 0.142 0.398xE-2 0.249xE-1 0.077xE-2 3.7-157 Rev

ANALYSIS Degree of

  • Freedom Acceleration ** Displacement ***

Horizontal Vertical Horizontal Vertical N-S E-W N-S E-W 52 0.162xE-2 0.448xE-2 0.319xE-2 0.213xE-4 0.126xE-3 0.160xE-4 53 0.312xE-2 0.619xE-2 0.328xE-2 0.703xE-4 0.172xE-3 0.171xE-4 54 0.425xE-2 0.133xE-2 0.224xE-2 0.895xE-4 0.204xE-4 0.122xE-4 55 0.129 0.045 0.015 0.268xE-2 0.031xE-2 0.027xE-3 56 0.024 0.020 0.078 0.093xE-3 0.096xE-3 0.270xE-3 57 0.045 0.121 0.005 0.030xE-2 0.246xE-2 0.028xE-3 58 0.233xE-3 0.776xE-3 0.829xE-4 0.032xE-4 0.258xE-4 0.003xE-4 59 0.179xE-3 0.807xE-4 0.421xE-4 0.004xE-4 0.008xE-4 0.001xE-4 60 0.864xE-3 0.196xE-3 0.127xE-3 0.282xE-4 0.032xE-4 0.002xE-4 61 0.205 0.044 0.012 0.571xE-2 0.064xE-2 0.050xE-3 62 0.036 0.033 0.110 0.160xE-3 0.163xE-3 0.443xE-3 63 0.044 0.188 0.003 0.063xE-2 0.501xE-2 0.048xE-3 64 0.355xE-3 0.127xE-2 0.390xE-4 0.053xE-4 0.419xE-4 0.004xE-4 65 0.108xE-3 0.938xE-4 0.286xE-4 0.006xE-4 0.010xE-4 0.001xE-4 66 0.147xE-2 0.314xE-3 0.789xE-4 0.480xE-4 0.054xE-4 0.004xE-4 67 0.291 0.055 0.013 0.90xE-2 0.100xE-2 0.085xE-3 68 0.047 0.045 0.133 0.222xE-3 0.224xE-3 0.60xE-3 3.7-158 Rev

ANALYSIS Degree of

  • Freedom Acceleration ** Displacement ***

Horizontal Vertical Horizontal Vertical N-S E-W N-S E-W 69 0.055 0.255 0.006 0.099xE-2 0.775xE-2 0.078xE-3 70 0.391xE-3 0.166xE-2 0.798xE-4 0.070xE-4 0.545xE-4 0.005xE-4 71 0.179xE-3 0.107xE-3 0.537xE-4 0.007xE-4 0.011xE-4 0.001xE-4 72 0.194xE-2 0.377xE-3 0.115xE-3 0.636xE-4 0.071xE-4 0.006xE-4 73 0.373 0.049 0.017 0.124xE-1 0.138xE-2 0.012xE-2 74 0.059 0.056 0.154 0.276xE-3 0.276xE-3 0.736xE-3 75 0.049 0.320 0.008 0.136xE-2 0.109xE-3 0.106xE-1 76 0.454xE-3 0.199xE-2 0.640xE-4 0.083xE-4 0.639xE-4 0.006xE-4 77 0.192xE-3 0.126xE-3 0.481xE-4 0.008xE-4 0.011xE-4 0.001xE-4 78 0.233xE-2 0.455xE-3 0.102xE-3 0.751xE-4 0.081xE-4 0.007xE-4 79 0.458 0.059 0.013 0.157xE-1 0.174xE-2 0.015xE-2 80 0.068 0.065 0.176 0.319xE-3 0.319xE-3 0.845xE-3 81 0.059 0.381 0.009 0.173xE-2 0.133xE-1 0.014xE-2 82 0.508xE-3 0.225xE-2 0.425xE-4 0.092xE-4 0.704xE-4 0.007xE-4 83 0.142xE-3 0.113xE-3 0.354xE-4 0.008xE-4 0.012xE-4 0.002xE-4 84 0.262xE-2 0.495xE-3 0.101xE-3 0.831xE-4 0.093xE-4 0.008xE-4 85 0.519 0.065 0.017 0.182xE-1 0.202xE-2 0.018xE-2 3.7-159 Rev

ANALYSIS Degree of

  • Freedom Acceleration ** Displacement ***

Horizontal Vertical Horizontal Vertical N-S E-W N-S E-W 86 0.072 0.069 0.188 0.344xE-3 0.343xE-3 0.909xE-3 87 0.064 0.428 0.009 0.20xE-2 0.154xE-1 0.016xE-2 88 0.513xE-3 0.238xE-2 0.797xE-4 0.096xE-4 0.738xE-4 0.007xE-4 89 0.198xE-3 0.125xE-3 0.491xE-4 0.008xE-4 0.012xE-4 0.001xE-4 90 0.277xE-2 0.489xE-3 0.140xE-3 0.872xE-4 0.097xE-4 0.009xE-4 91 0.646 0.089 0.010 0.229xE-1 0.254xE-2 0.022xE-2 92 0.079 0.076 0.206 0.377xE-3 0.375xE-3 0.993xE-3 93 0.089 0.526 0.006 0.251xE-2 0.193xE-1 0.019xE-2 94 0.119xE-2 0.558xE-2 0.276xE-3 0.150xE-4 0.116xE-3 0.017xE-4 95 0.168xE-3 0.155xE-3 0.462xE-4 0.011xE-4 0.017xE-4 0.002xE-4 96 0.590xE-2 0.120xE-2 0.471xE-3 0.138xE-3 0.151xE-4 0.021xE-4 97 0.832 0.093 0.017 0.288xE-1 0.319xE-2 0.029xE-2 98 0.093 0.084 0.255 0.415xE-3 0.413xE-3 0.109xE-2 99 0.092 0.686 0.014 0.316xE-2 0.242xE-1 0.025xE-2 100 0.267xE-2 0.829xE-2 0.450xE-3 0.168xE-4 0.130xE-3 0.022xE-4 101 0.290xE-3 0.253xE-3 0.780xE-4 0.016xE-4 0.020xE-4 0.004xE-4 102 0.832xE-2 0.268xE-2 0.117xE-2 0.155xE-3 0.170xE-4 0.027xE-4 3.7-160 Rev

Degrees of freedom corresponding to x-, y-, and z-direction translation and rotation about the x, y, and z axis are identified in Table 3.7B-9.

Tabulated accelerations are in units of g for translational degrees of freedom, and g/feet for rotational degrees of freedom.

Tabulated displacements are in units of feet for translational degrees of freedom, and rads for rotational degrees of freedom.

3.7-161 Rev. 30

FOR SAFE SHUTDOWN EARTHQUAKE CRACKED MODEL, RESPONSE SPECTRUM ANALYSIS Acceleration ** Displacement ***

Horizontal Vertical Horizontal Vertical Degree of

  • Freedom N-S E-W N-S E-W 1 0.068 0.022 0.015 0.296xE-3 0.049xE-3 0.024xE-3 2 0.017 0.004 0.046 0.027xE-3 0.014xE-3 0.126xE-3 3 0.022 0.068 0.004 0.049xE-3 0.284xE-3 0.012xE-3 4 ****0.418xE-4 0.136xE-3 0.284xE-4 0.007xE-4 0.065xE-4 0.001xE-4 5 0.882xE-4 0.579xE-4 0.869xE-4 0.004xE-4 0.007xE-4 0.0 6 0.168xE-3 0.358xE-4 0.761xE-4 0.066xE-4 0.007xE-4 0.003xE-4 7 0.180 0.054 0.060 0.252xE-2 0.516xE-3 0.232xE-3 8 0.023 0.011 0.075 0.20xE-3 0.157xE-3 0.263xE-3 9 0.072 0.160 0.026 0.838xE-3 0.307xE-2 0.145xE-3 10 0.904xE-3 0.206xE-2 0.745xE-3 0.184xE-4 0.568xE-4 0.026xE-4 11 0.245xE-2 0.392xE-2 0.122xE-2 0.316xE-4 0.841xE-4 0.028xE-4 12 0.201xE-2 0.543xE-2 0.502xE-3 0.506xE-4 0.089xE-4 0.038xE-4 13 0.217 0.075 0.065 0.505xE-2 0.102xE-2 0.438xE-3 14 0.057 0.028 0.103 0.986xE-3 0.284xE-3 0.383xE-3 15 0.085 0.216 0.039 0.983xE-3 0.699xE-2 0.216xE-3 16 0.119xE-2 0.297xE-2 0.512xE-3 0.151xE-4 0.750xE-4 0.026xE-4 17 0.171xE-2 0.405xE-2 0.417xE-3 0.386xE-4 0.124xE-3 0.034xE-4 3.7-162 Rev

Acceleration ** Displacement ***

Horizontal Vertical Horizontal Vertical Degree of

  • Freedom N-S E-W N-S E-W 18 0.300xE-2 0.948xE-3 0.686xE-3 0.726xE-4 0.156xE-4 0.056xE-4 19 0.252 0.078 0.065 0.608xE-2 0.114xE-2 0.502xE-3 20 0.039 0.021 0.112 0.325xE-3 0.151xE-3 0.408xE-3 21 0.083 0.214 0.032 0.134xE-2 0.695xE-2 0.204xE-3 22 0.130xE-2 0.311xE-2 0.546xE-3 0.161xE-4 0.807xE-4 0.029xE-4 23 0.176xE-2 0.418xE-2 0.405xE-3 0.405xE-4 0.127xE-3 0.034xE-4 24 0.327xE-2 0.105xE-2 0.715xE-3 0.785xE-4 0.168xE-4 0.061xE-4 25 0.278 0.078 0.065 0.687xE-2 0.129xE-2 0.548xE-3 26 0.057 0.028 0.116 0.892xE-3 0.206xE-3 0.43xE-3 27 0.082 0.257 0.032 0.131xE-2 0.867xE-2 0.213xE-3 28 0.139xE-2 0.321xE-2 0.587xE-3 0.178xE-4 0.854xE-2 0.032xE-4 29 0.182xE-2 0.429xE-2 0.414xE-3 0.423xE-4 0.131xE-3 0.035xE-4 30 0.341xE-2 0.111xE-2 0.750xE-3 0.817xE-4 0.175xE-4 0.064xE-4 31 0.307 0.080 0.065 0.773xE-2 0.161xE-2 0.598xE-3 32 0.051 0.031 0.122 0.721xE-3 0.415xE-3 0.447xE-3 33 0.084 0.271 0.030 0.153xE-2 0.914xE-2 0.209xE-3 34 0.148xE-2 0.328xE-2 0.636xE-3 0.187xE-4 0.890xE-4 0.035xE-4 35 0.186xE-2 0.438xE-2 0.425xE-3 0.435xE-4 0.134xE-3 0.036xE-4 3.7-163 Rev

Acceleration ** Displacement ***

Horizontal Vertical Horizontal Vertical Degree of

  • Freedom N-S E-W N-S E-W 36 0.353xE-2 0.116xE-2 0.790xE-3 0.842xE-4 0.181xE-4 0.066xE-4 37 0.413 0.084 0.043 0.109xE-1 0.220xE-2 0.794xE-3 38 0.070 0.038 0.142 0.115xE-2 0.374xE-3 0.515xE-3 39 0.077 0.379 0.011 0.208xE-2 0.134xE-1 0.236xE-3 40 0.181xE-2 0.346xE-2 0.853xE-3 0.214xE-4 0.982xE-4 0.043xE-4 41 0.212xE-2 0.481xE-2 0.302xE-3 0.507xE-4 0.150xE-3 0.037xE-4 42 0.396xE-2 0.137xE-2 0.106xE-2 0.922xE-4 0.199xE-4 0.076xE-4 43 0.605 0.132 0.118 0.159xE-1 0.322xE-2 0.121xE-2 44 0.101 0.50 0.173 0.175xE-2 0.435xE-3 0.608xE-3 45 0.137 0.551 0.037 0.323xE-2 0.191xE-1 0.366xE-3 46 0.220xE-2 0.429xE-2 0.989xE-3 0.237xE-4 0.105xE-3 0.051xE-4 47 0.366xE-2 0.656xE-2 0.944xE-3 0.777xE-4 0.167xE-3 0.065xE-4 48 0.474xE-2 0.176xE-2 0.175xE-2 0.102xE-3 0.220xE-4 0.096xE-4 49 0.718 0.175 0.181 0.185xE-1 0.372xE-2 0.147xE-2 50 0.083 0.038 0.196 0.995xE-3 0.282xE-3 0.649xE-3 51 0.199 0.620 0.060 0.427xE-2 0.209xE-1 0.464xE-3 52 0.228xE-2 0.481xE-2 0.105xE-2 0.240xE-4 0.105xE-3 0.052xE-4 53 0.372xE-2 0.667xE-2 0.987xE-3 0.784xE-4 0.168xE-3 0.067xE-4 3.7-164 Rev

Acceleration ** Displacement ***

Horizontal Vertical Horizontal Vertical Degree of

  • Freedom N-S E-W N-S E-W 54 0.491xE-2 0.185xE-2 0.187xE-2 0.103xE-3 0.222xE-4 0.098xE-3 55 0.145 0.052 0.026 0.439xE-2 0.341xE-3 0.097xE-3 56 0.029 0.006 0.089 0.029xE-3 0.010xE-3 0.608xE-3 57 0.052 0.127 0.015 0.340xE-3 0.443xE-2 0.059xE-3 58 0.162xE-3 0.733xE-3 0.702xE-4 0.030xE-4 0.447xE-4 0.003xE-4 59 0.366xE-3 0.863xE-4 0.335xE-3 0.010xE-4 0.012xE-4 0.001xE-4 60 0.740xE-3 0.157xE-3 0.169xE-3 0.444xE-4 0.030xE-4 0.005xE-4 61 0.221 0.065 0.034 0.972xE-2 0.707xE-3 0.15xE-3 62 0.022 0.007 0.141 0.033xE-3 0.009xE-3 0.114xE-2 63 0.065 0.207 0.020 0.706xE-3 0.980xE-2 0.090xE-3 64 0.329xE-3 0.132xE-2 0.130xE-3 0.055xE-4 0.808xE-4 0.005xE-4 65 0.220xE-3 0.106xE-3 0.197xE-3 0.018xE-4 0.021xE-4 0.001xE-4 66 0.132xE-2 0.321xE-3 0.277xE-3 0.802xE-4 0.055xE-4 0.010xE-4 67 0.289 0.065 0.032 0.155xE-1 0.108xE-2 0.140xE-3 68 0.022 0.005 0.183 0.04xE-3 0.012xE-3 0.162xE-2 69 0.064 0.286 0.018 0.108xE-2 0.156xE-1 0.082xE-3 70 0.458xE-3 0.175xE-2 0.193xE-3 0.075xE-4 0.109xE-3 0.008xE-4 71 0.283xE-3 0.136xE-3 0.235xE-3 0.026xE-4 0.030xE-4 0.001xE-4 3.7-165 Rev

Acceleration ** Displacement ***

Horizontal Vertical Horizontal Vertical Degree of

  • Freedom N-S E-W N-S E-W 72 0.176xE-2 0.450xE-3 0.378xE-3 0.108xE-3 0.075xE-4 0.015xE-4 73 0.351 0.058 0.021 0.215xE-1 0.147xE-2 0.073xE-3 74 0.028 0.005 0.227 0.047xE-3 0.017xE-3 0.204xE-2 75 0.058 0.355 0.011 0.146xE-2 0.216xE-1 0.040xE-3 76 0.538xE-3 0.211xE-2 0.240xE-3 0.090xE-4 0.130xE-3 0.011xE-4 77 0.366xE-3 0.172xE-3 0.318xE-3 0.033xE-4 0.038xE-4 0.002xE-4 78 0.213xE-2 0.530xE-3 0.450xE-3 0.129xE-3 0.090xE-4 0.019xE-4 79 0.416 0.039 0.010 0.273xE-1 0.185xE-2 0.049xE-3 80 0.017 0.007 0.263 0.053xE-3 0.220xE-3 0.238xE-2 81 0.039 0.423 0.006 0.185xE-2 0.275xE-1 0.028xE-3 82 0.519xE-3 0.243xE-2 0.255xE-3 0.101xE-4 0.145xE-3 0.011xE-4 83 0.180xE-3 0.198xE-3 0.621xE-4 0.039xE-4 0.045xE-4 0.002xE-4 84 0.244xE-2 0.583xE-3 0.481xE-3 0.144xE-3 0.101xE-4 0.020xE-4 85 0.476 0.061 0.026 0.317xE-1 0.214xE-2 0.114xE-3 86 0.020 0.008 0.282 0.058xE-3 0.026xE-3 0.257xE-2 87 0.061 0.484 0.015 0.214xE-2 0.320xE-1 0.068xE-3 88 0.584xE-3 0.262xE-2 0.248xE-3 0.106xE-4 0.153xE-3 0.011xE-4 89 0.347xE-3 0.222xE-3 0.276xE-3 0.043xE-4 0.049xE-4 0.002xE-4 3.7-166 Rev

Acceleration ** Displacement ***

Horizontal Vertical Horizontal Vertical Degree of

  • Freedom N-S E-W N-S E-W 90 0.262xE-2 0.576xE-3 0.478xE-3 0.151xE-3 0.106xE-4 0.019xE-4 91 0.580 0.058 0.029 0.399xE-1 0.269xE-2 0.135xE-3 92 0.042 0.009 0.318 0.067xE-3 0.03xE-3 0.283xE-2 93 0.058 0.584 0.016 0.268xE-2 0.403xE-1 0.077xE-3 94 0.119xE-2 0.597xE-2 0.427xE-3 0.188xE-4 0.249xE-3 0.019xE-4 95 0.444xE-3 0.533xE-3 0.200xE-4 0.105xE-4 0.123xE-4 0.004xE-4 96 0.597xE-2 0.119xE-2 0.842xE-3 0.247xE-3 0.188xE-4 0.036xE-4 97 0.770 0.067 0.031 0.506xE-1 0.343xE-2 0.141xE-3 98 0.044 0.012 0.356 0.084xE-3 0.037xE-3 0.314xE-2 99 0.067 0.776 0.013 0.342xE-2 0.510xE-1 0.061xE-3 100 0.191xE-2 0.761xE-2 0.732xE-3 0.225xE-4 0.280xE-3 0.034xE-4 101 0.531xE-3 0.645xE-3 0.221xE-4 0.125xE-4 0.148xE-4 0.005xE-4 102 0.773xE-2 0.190xE-2 0.132xE-2 0.278xE-3 0.225xE-4 0.059xE-4 NOTES:
  • Degrees of freedom corresponding to x- and y- direction translation and rotation about the z axis are identified in Table 3.7B-9
    • Tabulated accelerations are in units of g for translational degrees of freedom, and g/feet for rotational degrees of freedom.
      • Tabulated displacements are in units of feet for translational degrees of freedom, and rads for rotational degrees of freedom.
        • 10E-4 = 10-4 = 0.0010 3.7-167 Rev

TABLE 3.7B-9 CONTAINMENT AND INTERNAL STRUCTURES DEGREES OF FREEDOM Direction of Motion Mass Point Degree of Freedom (Global Coordinates)

  • 1 1 Translation X 1 2 Translation Y 1 3 Translation Z 1 4 Rotation X 1 5 Rotation Y 1 6 Rotation Z 2 7 Translation X 2 8 Translation Y 2 9 Translation Z 2 10 Rotation X 2 11 Rotation Y 2 12 Rotation Z 3 13 Translation X 3 14 Translation Y 3 15 Translation Z 3 16 Rotation X 3 17 Rotation Y 3 18 Rotation Z 4 19 Translation X 4 20 Translation Y 4 21 Translation Z 4 22 Rotation X 4 23 Rotation Y 4 24 Rotation Z 5 25 Translation X 5 26 Translation Y 5 27 Translation Z 3.7-168 Rev. 30

Direction of Motion Mass Point Degree of Freedom (Global Coordinates)

  • 5 28 Rotation X 5 29 Rotation Y 5 30 Rotation Z 6 31 Translation X 6 32 Translation Y 6 33 Translation Z 6 34 Rotation X 6 35 Rotation Y 6 36 Rotation Z 7 37 Translation X 7 38 Translation Y 7 39 Translation Z 7 40 Rotation X 7 41 Rotation Y 7 42 Rotation Z 8 43 Translation X 8 44 Translation Y 8 45 Translation Z 8 46 Rotation X 8 47 Rotation Y 8 48 Rotation Z 9 49 Translation X 9 50 Translation Y 9 51 Translation Z 9 52 Rotation X 9 53 Rotation Y 9 54 Rotation Z 10 55 Translation X 3.7-169 Rev. 30

Direction of Motion Mass Point Degree of Freedom (Global Coordinates)

  • 10 56 Translation Y 10 57 Translation Z 10 58 Rotation X 10 59 Rotation Y 10 60 Rotation Z 11 61 Translation X 11 62 Translation Y 11 63 Translation Z 11 64 Rotation X 11 65 Rotation Y 11 66 Rotation Z 12 67 Translation X 12 68 Translation Y 12 69 Translation Z 12 70 Rotation X 12 71 Rotation Y 12 72 Rotation Z 13 73 Translation X 13 74 Translation Y 13 75 Translation Z 13 76 Rotation X 13 77 Rotation Y 13 78 Rotation Z 14 79 Translation X 14 80 Translation Y 14 81 Translation Z 14 82 Rotation X 14 83 Rotation Y 3.7-170 Rev. 30

Direction of Motion Mass Point Degree of Freedom (Global Coordinates)

  • 14 84 Rotation Z 15 85 Translation X 15 86 Translation Y 15 87 Translation Z 15 88 Rotation X 15 89 Rotation Y 15 90 Rotation Z 16 91 Translation X 16 92 Translation Y 16 93 Translation Z 16 94 Rotation X 16 95 Rotation Y 16 96 Rotation Z 17 97 Translation X 17 98 Translation Y 17 99 Translation Z 17 100 Rotation X 17 101 Rotation Y 17 102 Rotation Z 3.7-171 Rev. 30

DISPLACEMENTS HORIZONTAL-MOTION

  • MassNo. Elevation (ft) Accelerations ** Displacement ***

SSE **** OBE SSE OBE N-S E-W N-S E-W N-S E-W N-S E-W Internal Structure 1 (-) 37.25 0.170 0.170 0.090 0.090 0.004 0.006 0.003 0.004 2 (-) 11.25 0.294 0.298 0.168 0.192 0.039 0.060 0.025 0.041 3 3.0 0.357 0.414 0.230 0.283 0.078 0.118 0.051 0.081 4 11.27 0.395 0.397 0.267 0.272 0.092 0.121 0.060 0.083 5 18.17 0.421 0.447 0.270 0.304 0.105 0.145 0.068 0.100 6 24.5 0.451 0.456 0.290 0.309 0.125 0.154 0.077 0.105 7 50.84 0.541 0.517 0.349 0.352 0.167 0.219 0.108 0.149 8 86.5 0.855 0.821 0.552 0.558 0.243 0.316 0.158 0.215 9 109.1 1.070 1.010 0.694 0.685 0.284 0.356 0.184 0.243 External Structure 10 (-) 6.75 0.223 0.194 0.128 0.114 0.058 0.058 0.038 0.038 11 16.25 0.321 0.292 0.196 0.180 0.127 0.127 0.083 0.083 12 39.25 0.386 0.368 0.240 0.230 0.201 0.202 0.130 0.131 13 62.25 0.439 0.424 0.294 0.268 0.276 0.278 0.180 0.181 14 85.25 0.530 0.468 0.357 0.303 0.350 0.353 0.228 0.229 15 104.0 0.601 0.560 0.407 0.358 0.408 0.410 0.265 0.267 16 133.5 0.745 0.658 0.502 0.427 0.513 0.516 0.334 0.336 17 163.3 0.942 0.855 0.642 0.556 0.650 0.654 0.423 0.425 NOTES:

  • Maximum of cracked, uncracked
    • Tabulated accelerations are in units of g 3.7-172 Rev

3.7-173 Rev. 30 TABLE 3.7B-11 CONTAINMENT AND INTERNAL STRUCTURE ENVELOPED ACCELERATIONS AND DISPLACEMENTS VERTICAL-MOTION

  • Elevation Accelerations ** Displacements ***

ass No. (ft) SSE **** OBE SSE OBE ernal Structure

(-) 37.25 0.113 0.060 0.002 0.001

(-) 11.25 0.108 0.068 0.007 0.005 3.0 0.188 0.121 0.020 0.013 11.27 0.172 0.111 0.011 0.007 18.17 0.202 0.130 0.018 0.012 24.5 0.204 0.132 0.019 0.012 50.84 0.251 0.162 0.024 0.016 86.5 0.324 0.207 0.034 0.022 109.1 0.317 0.199 0.023 0.015 ternal Structure

(-) 6.75 0.124 0.071 0.008 0.005 16.25 0.179 0.103 0.014 0.009 39.25 0.225 0.132 0.020 0.013 62.25 0.268 0.163 0.025 0.016 85.25 0.308 0.182 0.029 0.019 104.0 0.330 0.196 0.032 0.021 133.5 0.369 0.230 0.035 0.023 163.3 0.431 0.259 0.039 0.025 TES:

Maximum of cracked, uncracked Tabulated accelerations are in units of g Displacements are listed in inches 3.7-174 Rev. 30

TABLE 3.7B-12 MAIN STEAM VALVE BUILDING SIGNIFICANT MODAL FREQUENCIES AND PARTICIPATION FACTORS Mode Frequency (Hz) Participation Factors X Y Z 1 8.380 126.101 0.963 0.630 2 9.456 0.813 -2.172 -133.882 3 13.093 17.672 0.510 0.919 4 23.991 7.296 -46.766 91.569 5 24.478 108.417 5.424 -6.261 6 25.663 2.749 -159.477 -30.236 7 29.689 23.791 5.297 3.240 8 33.833 36.719 -1.254 -0.800 9 34.306 2.605 -18.092 35.379 10 50.549 2.403 -28.776 -121.336 11 51.756 125.939 0.508 1.685 12 52.469 1.052 -110.173 48.891 13 61.221 20.459 26.390 10.175 14 61.438 7.329 -51.449 -28.932 15 64.407 3.223 -2.901 0.345 16 84.853 1.025 59.045 2.653 3.7-175 Rev. 30

ABLE 3.7B-13 MAIN STEAM VALVE BUILDING SIGNIFICANT MODE SHAPES (EIGENVECTORS)

Mode Joint Displacement X Y Z 1 1 0.013 0.000 0.000 2 0.043 0.000 0.000 3 0.232 0.005 0.002 4 1.0 0.005 0.005 2 1 0.000 0.000 0.017 2 0.000 0.001 0.053 3 -0.001 0.011 0.318 4 -0.007 0.011 1.0 3 1 0.033 0.001 0.002 2 0.106 0.001 0.010 3 0.205 0.017 -0.034 4 1.0 0.020 0.091 4 1 0.007 -0.036 0.093 2 0.014 -0.050 0.167 3 0.074 -0.134 1.0 4 0.008 -0.357 -0.155 5 1 0.101 0.004 -0.006 2 0.209 0.005 -0.011 3 1.0 0.014 -0.056 4 -0.166 0.036 0.007 6 1 -0.003 0.125 0.032 2 -0.001 -0.173 0.064 3 -0.013 -0.458 0.321 4 -0.013 1.0 -0.094 7 1 0.210 0.036 0.029 2 0.167 0.047 0.066 3 0.256 0.066 0.124 3.7-176 Rev. 30

Mode Joint Displacement X Y Z 4 1.0 0.217 -0.007 8 1 0.160 -0.004 -0.003 2 0.403 -0.006 -0.009 3 1.0 0.010 -0.075 4 -0.696 -0.033 0.069 9 1 0.009 -0.048 0.124 2 0.021 -0.064 0.305 3 0.053 -0.105 1.0 4 -0.035 -0.146 -0.776 10 1 -0.012 0.107 0.599 2 -0.018 0.135 1.0 3 0.004 0.256 -0.350 4 0.0 -0.159 0.026 11 1 0.629 0.002 0.008 2 1.0 0.001 0.014 3 -0.393 0.009 -0.005 4 0.033 -0.006 0.0 12 1 -0.007 0.533 -0.315 2 -0.004 0.678 -0.369 3 -0.003 1.0 0.090 4 -0.001 -0.662 -0.053 13 1 0.857 0.829 0.426 2 -0.193 1.0 0.092 3 0.619 0.287 0.070 4 0.186 -0.467 0.094 14 1 -0.157 0.829 0.621 2 0.040 1.0 0.148 3 -0.117 0.508 0.079 3.7-177 Rev. 30

Mode Joint Displacement X Y Z 4 -0.034 -0.704 0.133 15 1 -0.769 0.519 -0.082 2 -0.055 0.617 -0.858 3 -0.026 0.155 1.0 4 -0.264 -0.387 0.042 16 1 -0.015 -0.634 -0.038 2 0.002 -0.646 0.003 3 0.0 1.0 -0.002 4 0.002 -0.224 0.008 3.7-178 Rev. 30

ABLE 3.7B-14 MAIN STEAM VALVE BUILDING CSM* ACCELERATIONS FOR SAFE SHUTDOWN EARTHQUAKE FROM RESPONSE SPECTRUM ANALYSIS Degree of Freedom ** X *** Y*** Z ***

1 0.103 0.006 0.015 2 0.007 0.059 0.045 3 0.015 0.040 0.107 4 0.0 0.001 0.002 5 0.001 0.0 0.001 6 0.002 0.0 0.0 7 0.156 0.003 0.012 8 0.008 0.072 0.057 9 0.012 0.050 0.160 10 0.001 0.002 0.003 11 0.001 0.0 0.0 12 0.003 0.001 0.001 13 0.193 0.010 0.018 14 0.008 0.115 0.089 15 0.019 0.083 0.187 16 0.001 0.003 0.004 17 0.003 0.0 0.0 18 0.003 0.002 0.001 19 0.414 0.004 0.005 20 0.014 0.168 0.095 21 0.006 0.022 0.395 22 0.001 0.002 0.006 23 0.003 0.0 0.0 24 0.005 0.001 0.001 TES:

Closely spaced modes.

The relationship between the degrees of freedom and the motion of the lumped masses is given in Table 3.7B-17.

3.7-179 Rev. 30

3.7-180 Rev. 30 ABLE 3.7B-15 MAIN STEAM VALVE BUILDING CSM* DISPLACEMENTS FOR SAFE SHUTDOWN EARTHQUAKE FROM RESPONSE SPECTRUM ANALYSIS Degree of Freedom ** X *** Y *** Z ***

0.703x10-4 0.210x10-5 0.310x10-5 0.230x10-5 0.279x10-4 0.173x10-4 0.310x10-5 0.154x10-4 0.716x10-4 0.10x10-6 0.20x10-6 0.330x10-5 0.40x10-6 0.0 0.10x10-6 0.42x10-5 0.10x10-6 0.10x10-6 0.217x10-3 0.270x10-5 0.560x10-5 0.310x10-5 0.379x10-4 0.230x10-4 0.540x10-5 0.246x10-4 0.198x10-3 0.10x10-6 0.40x10-6 0.660x10-5 0.150x10-5 0.10x10-6 0.30x10-6 0.840x10-5 0.20x10-6 0.20x10-6 0.113x10-2 0.143x10-4 0.259x10-4 0.237x10-4 0.927x10-4 0.630x10-4 0.287x10-4 0.107x10-3 0.116x10-2 0.30x10-6 0.110x10-5 0.160x10-4 0.55x10-5 0.30x10-6 0.10x10-5 0.242x10-4 0.50x10-6 0.40x10-6 3.7-181 Rev. 30

Degree of Freedom ** X *** Y *** Z ***

0.480x10-2 0.249x10-4 0.406x10-4 0.289x10-4 0.199x10-3 0.111x10-3 0.379x10-4 0.463x10-3 0.340x10-2 0.90x10-6 0.240x10-5 0.336x10-4 0.274x10-4 0.20x10-6 0.140x10-5 0.417x10-4 0.60x10-6 0.60x10-6 TES:

Closely spaced modes.

The relationship between the degrees of freedom and the motion of the lumped masses is given in Table 3.7B-17.

Tabulated displacements are in units of feet for translational degrees of freedom, and g/radians for rotational degrees.

3.7-182 Rev. 30

TABLE 3.7B-16 MAIN STEAM VALVE BUILDING ABS

1 0.170 0.750x10-4 2 0.113 0.50x10-4 3 0.170 0.917x10-4 4 0.003 0.360x10-5 5 0.002 0.50x10-6 6 0.002 0.440x10-5 7 0.170 0.225x10-3 8 0.137 0.667x10-4 9 0.222 0.225x10-3 10 0.006 0.710x10-5 11 0.001 0.190x10-5 12 0.005 0.880x10-5 13 0.221 0.117x10-2 14 0.212 0.183x10-3 15 0.290 0.129x10-2 16 0.008 0.174x10-4 17 0.003 0.680x10-5 18 0.006 0.251x10-4 19 0.424 0.487x10-2 20 0.277 0.342x10-3 21 0.423 0.368x10-2 22 0.009 0.369x10-4 3.7-183 Rev. 30

Degree of Freedom ** Acceleration *** Displacement ****

23 0.003 0.290x10-4 24 0.007 0.429x10-4 TES:

Absolute sum of each direction of excitation.

Table 3.7B-18 defines degrees of freedom.

Tabulated accelerations are in units of g for translational degrees of freedom and g/feet for rotational degrees of freedom.

  • Tabulated displacements are in units of feet for translational degrees of freedom and rads for rotational degrees.

3.7-184 Rev. 30

TABLE 3.7B-17 MAIN STEAM VALVE BUILDING DEGREES OF FREEDOM Mass Point* Degree of Freedom Direction of Motion 1 1 Translation X 1 2 Translation Y 1 3 Translation Z 1 4 Rotation X 1 5 Rotation Y 1 6 Rotation Z 2 7 Translation X 2 8 Translation Y 2 9 Translation Z 2 10 Rotation X 2 11 Rotation Y 2 12 Rotation Z 3 13 Translation X 3 14 Translation Y 3 15 Translation Z 3 16 Rotation X 3 17 Rotation Y 3 18 Rotation Z 4 19 Translation X 4 20 Translation Y 4 21 Translation Z 4 22 Rotation X 4 23 Rotation Y 4 24 Rotation Z TE:

or locations of lumped masses, refer to Figure 3.7b-10.

3.7-185 Rev. 30

BLE 3.7B-18 EMERGENCY GENERATOR ENCLOSURE ACCELERATIONS AND DISPLACEMENTS HORIZONTAL-MOTION*

Accelerations **

Mass No. Elevation (ft-in) SSE OBE N-S E-W N-S E-W 3 66-0 0.660 0.547 0.426 0.357 2 51-0 0.514 0.438 0.341 0.288 1 24-6 0.317 0.285 0.173 0.163

-S applies to X-direction, E-W applies to Z-direction.

Tabulated accelerations are in units of g.

3.7-186 Rev. 30

Displacements ***

Mass No. Elevation (ft-in) SSE OBE N-S E-W N-S E-W 3 66-0 0.1070 0.0425 0.0692 0.0278 2 51-0 0.0830 0.0340 0.0551 0.0224 1 24-6 0.0379 0.0170 0.0234 0.0102 Displacements are listed in inches.

3.7-187 Rev. 30

BLE 3.7B-19 EMERGENCY GENERATOR ENCLOSURE ACCELERATIONS AND DISPLACEMENTS VERTICAL - MOTION Mass No. Elevation ft-in Accelerations*

SSE OBE 3 66-0 0.250 0.148 2 41-0 0.220 0.128 1 24-6 0.176 0.100 Mass No. Elevation ft-in Displacements **

SSE OBE 3 66-0 0.0055 0.0033 2 51-0 0.0048 0.0028 1 24-6 0.0040 0.0024 TES:

Tabulated accelerations are in units of g.

Displacements are listed in inches 3.7-188 Rev. 30

ABLE 3.7B-20 COMPARISON OF RESPONSE SPECTRA AND TIME HISTORY ANALYSIS RESULTS CONTAINMENT AND INTERNAL STRUCTURES (UNCRACKED PROPERTIES)

Mass No.* SSE Accelerations **

Response Spectrum Time History N-S E-W Vertical N-S E-W Vertical 1 0.170 0.170 0.113 0.180 0.179 0.140 2 0.271 0.298 0.109 0.222 0.246 0.170 3 0.336 0.414 0.174 0.276 0.342 0.205 4 0.370 0.397 0.161 0.302 0.335 0.208 5 0.392 0.447 0.183 0.320 0.370 0.218 6 0.415 0.456 0.186 0.341 0.380 0.223 7 0.486 0.517 0.224 0.429 0.483 0.249 8 0.748 0.821 0.303 0.598 0.675 0.282 9 0.946 1.010 0.312 0.701 0.760 0.290 10 0.188 0.171 0.122 0.226 0.213 0.181 11 0.260 0.235 0.179 0.302 0.290 0.224 12 0.359 0.316 0.225 0.383 0.354 0.263 13 0.439 0.376 0.268 0.451 0.406 0.302 14 0.530 0.449 0.308 0.502 0.462 0.336 15 0.601 0.501 0.330 0.541 0.507 0.355 16 0.745 0.621 0.369 0.671 0.620 0.380 17 0.942 0.792 0.431 0.839 0.809 0.411 TES:

Figure 3.7B-9 shows the location of the lumped masses.

Safe shutdown earthquake accelerations are in units of g.

3.7-189 Rev. 30

ABLE 3.7B-21 COMPARISON OF RESPONSE SPECTRA AND TIME HISTORY ALYSIS RESULTS CONTAINMENT AND INTERNAL STRUCTURES (CRACKED PROPERTIES) ass No.

  • SSE Accelerations **

Response Spectrum Time History N-S E-W Vertical N-S E-W Vertical 1 0.170 0.170 0.113 0.178 0.178 0.127 2 0.294 0.258 0.108 0.225 0.242 0.161 3 0.357 0.340 0.188 0.302 0.338 0.197 4 0.395 0.329 0.172 0.332 0.339 0.204 5 0.421 0.372 0.202 0.355 0.373 0.211 6 0.451 0.385 0.204 0.380 0.382 0.215 7 0.541 0.467 0.251 0.486 0.464 0.239 8 0.855 0.724 0.324 0.679 0.638 0.270 9 1.070 0.878 0.317 0.782 0.714 0.281 10 0.223 0.194 0.125 0.215 0.215 0.159 11 0.321 0.292 0.169 0.272 0.276 0.236 12 0.386 0.368 0.210 0.358 0.361 0.304 13 0.430 0.424 0.260 0.431 0.435 0.382 14 0.465 0.468 0.287 0.490 0.498 0.437 15 0.562 0.560 0.310 0.535 0.538 0.464 16 0.667 0.658 0.369 0.717 0.711 0.497 17 0.867 0.855 0.412 0.962 0.991 0.538 TES:

Figure 3.7B-9 shows the location of the lumped masses.

Safe shutdown earthquake accelerations are in units of g.

3.7-190 Rev. 30

ABLE 3.7B-22 COMPARISON OF RESPONSE SPECTRA AND TIME HISTORY NALYSIS RESULTS MAIN STEAM VALVE BUILDING SSE ACCELERATIONS*

Response Spectrum Time History ass No.

    • N-S E-W Vertical N-S E-W Vertical 1 0.170 0.170 0.113 0.174 0.174 0.118 2 0.170 0.222 0.137 0.180 0.181 0.120 3 0.221 0.290 0.212 0.250 0.254 0.135 4 0.424 0.423 0.277 0.488 0.441 0.159 TES:

Safe shutdown earthquake accelerations are in units of g.

Figure 3.7B-10 shows the location of the lumped masses.

3.7-191 Rev. 30

Max Dynamic Max Static Fundamental Moment Moment S/C K Model Beam Frequency Sum (in-lb Load Type (in-lb) Location A/U S/U Cantilever 1 SRSS 620,000 Conc

  • 700,000 Fixed End 0.89 0.89 0.99 ABS 649,000 Unif ** 700,000 Simple-Simple 1 SRSS 179,000 Conc 348,000 Midspan 0.51 1.03 1.07 ABS 186,000 Unif 174,000 Fixed-Fixed 1 SRSS 103,000 Conc 174,000 Fixed End 0.59 0.89 0.97 ABS 112,000 Unif 116,000 Simple-Fixed - No 1 SRSS 152,000 Conc 261,000 Fixed End 0.58 0.87 0.97 Overhang ABS 169,000 Unif 174,000 Simple-Fixed - 16% 1.34 SRSS 83,200 Conc 162,000 Fixed End 0.51 0.75 1.03 Overhang ABS 114,000 Unif 111,000 Simple-Fixed - 33% 1.04 SRSS 57,000 Conc 77,400 Fixed End 0.74 0.74 1.15 Overhang ABS 89,000 Unif 77,200 Simple-Fixed - 50% 0.62 SRSS 152,000 Conc 174,000 Simple 0.87 0.87 1.01 Overhang Support ABS 176,000 Unif 174,000 NOTES:

3.7-192 Rev

Unif = Uniform load 3.7-193 Rev. 30

TABLE 3.7B-24 MODAL DENSITY, n*

Simple-Fixed 33%

Cantilever Fixed-Fixed Simple-Fixed Simple-Simple Overhang Frequency Frequency Frequency Frequency Frequency ode No (Hz) (Hz) (Hz) (Hz) (Hz) 1 1.0 1.0 1.0 1.0 1.0 2 5.8 2.7 3.2 3.8 2.9 3 15.3 4.9 6.3 8.2 6.5 4 22.8 7.5 10.2 13.6 8.4 5 28.0 10.2 14.0 19.5 13.3 6 43.2 12.9 17.9 25.4 17.6 TE:

Modal density is based on a +/-10 percent criterion.

3.7-194 Rev. 30

(For Simple-Fixed Beam with 33% Overhang)

Maximum Dynamic Moment Maximum Static First Sum Moment Moment Uniform K S/U Model Beam Mode Dynamic Load (S) (in-lb) Location Load (U) (in-lb)* A/U High Low Simple-Fixed go gmax gc 33%

Overhang fo f fc Model 6A 0.10 2.87 0.33 SRSS 20,000 Fixed 222,000 s 0.09 0.70 3-4 20 ABS 30,000 Fixed 0.13 Model 6B 0.10 2.87 0.33 SRSS 148,00 Fixed 222,000 s 0.67 1.0 3-4 20 ABS 157,00 Fixed 0.71 Model 6C 2.87 2.87 0.33 SRSS 102,00 Simple 222,000 s 0.45 Support 3.3 3-4 20 ABS 118,00 Simple 0.53 Support Model 6D 0.40 2.87 0.33 SRSS 22,000 Fixed 31,000 s 0.71 10.0 3-4 20 ABS 32,000 Fixed 1.03 Model 6E 0.33 2.87 0.33 SRSS 20,000 Fixed 25,700 s 0.78 20.0 3-4 20 ABS 27,000 Fixed 1.05 Model 6F 0.30 2.87 0.33 SRSS 18,000 Fixed 23,400 s 0.77 33.0 3-4 20 ABS 25,000 Fixed 1.07 3.7-195 Rev

g = gmax, if fo < fp; g at fo fp (where fp is the frequency at peak) 3.7-196 Rev. 30

Type of Combined Stress ASME Section III Code Class Earthquake Type of Seismic Analyses Calculations and Stress Criteria Class 1 SSE Dynamic response spectra ASME Code,Section III, Subarticle (Sizes 1.25 inch NPS and larger) NB-3600 OBE Dynamic response spectra ASME Code,Section III, Subarticle NB-3600 Class 1 SSE Simplified analyses ASME Code,Section III, Subarticle (Sizes 1 inch NPS and below) NB-3600 OBE Simplified analyses ASME Code,Section III, Subarticle NB-3600 Classes 2 and 3 SSE Dynamic response spectra ASME Code,Section III, Subarticles (Sizes 2.5 inch NPS and larger) NC-3600 and ND-3600 OBE Dynamic response spectra ASME Code,Section III, Subarticles NC-3600 and ND-3600 Classes 2 and 3 SSE Dynamic response spectra or ASME Code,Section III, Subarticles (Sizes 2 inch NPS and below) simplified analyses NC-3600 and ND-3600 OBE Dynamic response spectra or ASME Code,Section III, Subarticles simplified analyses NC-3600 and ND-3600 3.7-197 Rev

TABLE 3.7N-1 DAMPING VALUES USED FOR SEISMIC ANALYSIS FOR WESTINGHOUSE SUPPLIED EQUIPMENT Damping (Percent of Critical)

Upset Conditions Faulted Conditions Item (OBE) (SSE, DBA) mary coolant loop system components 2 4 lded steel structures 2 4 lted and/or riveted steel structures 4 7 el assemblies 7 10 DMs/CRDM supports 5 5 3.7-198 Rev. 30

1 CONCRETE CONTAINMENT s section provides the following information on concrete containment and on concrete ions of steel/concrete containments:

1. The physical description
2. The applicable design codes, standards, and specifications
3. The loading criteria, including loads and load combinations
4. The design and analysis procedures
5. The structural acceptance criteria
6. The materials, quality control programs, and special construction techniques
7. The testing and inservice inspection programs 1.1 Description of the Containment containment structure is a steel-lined conventionally reinforced concrete pressurized water tor containment structure designed to operate under sub-atmospheric conditions. The cture has a vertical cylindrical wall and hemispherical dome supported on a flat base mat ch is founded on bedrock (Figures 3.8-1 and 3.8-2).

design, analyses, and construction of the containment structure is similar to the designs for following plants:

Connecticut Yankee Atomic Power Company, Nuclear Power Plant (Docket No. 50-213)

Virginia Electric and Power Company, Surry Power Station, Units 1 and 2 (Docket Nos.

50-280 and 50-281)

Maine Yankee Atomic Power Company, Maine Yankee Atomic Power Station (Docket No. 50-309)

Duquesne Light Company, Beaver Valley Power Station, Unit No. 1 (Docket No. 50-334)

Virginia Electric and Power Company, North Anna Power Station, Units 1 and 2 (Docket Nos. 50-338 and 50-339) 3.8-1 Rev. 30

base foundation slab is 10 feet thick with a 158 foot diameter. The bottom reinforcement is a angular grid pattern and the top reinforcement consists of concentric circular bars combined h radial bars arranged to permit uniform spacing of the vertical wall reinforcing bars (rebars) ch extend into the mat. Splices in adjacent parallel bars are staggered.

floor liner plate of 0.25 inch thickness is anchored to the mat by means of rebars welded to 7 by 0.5 inch continuous vertical plates which are in turn welded to the liner seams ure 3.8-3). Where interior wall and column rebars are anchored to the mat, a 3 inch x 6 inch angular bar (bridging bar) or 1.25 inch thick plate was welded through the liner with test nnels all around to ensure the continuity of the steel membrane. The vertical rebar was welded to the top of the bar and butt-welded to the bottom of these members providing tinuity of rebar without creating multiple penetrations of the liner.

einforced concrete slab approximately 2 feet thick, was placed over and anchored through the liner to stiffen it against negative pressures and to protect it from heat associated with a DBA ilar to other pressurized water reactor containment structures (e.g., North Anna 1 and 2, ket No. 50-338 and 50-339). his slab also serves as anchorage and support for equipment ted on the lowest level of the containment.

1.1.2 Cylindrical Wall cylindrical wall is 4 feet 6 inches thick with an inside diameter of 140 feet and a height from to spring line of 131 feet 3 inches. Hoop tension in the cylindrical wall of the containment cture is resisted by horizontal bars located near both the outer and inner faces of the wall.

ces in these bars are staggered where possible. Meridional tension is resisted by rows of ical bars placed near each face of the wall. The vertical bars are placed in groups of not more 20 bars of equal length. These groups are arranged so that no adjacent group in the same face he wall has splices closer than 3 feet apart vertically.

ial shear in the containment structure wall resulting from the design basis accident (DBA) es from a maximum at the base of the wall and approaches zero at some level above the top of mat. To resist the large radial shear near the base of the wall, flat steel bars inclined at roximately 45 degrees with the horizontal are welded to the vertical reinforcement as shown in ure 3.8-4. The welded flat bars terminate at approximately elevation (-) 18 feet where the al shear load is reduced. Above this level, radial shear is resisted by Z type reinforcing steel.

s Z reinforcing continues to elevation (-) 4 feet (-) 8 inches where the shear becomes gnificant. In the lower portion of the containment structure wall, supplementary reinforcing l, normal to the face of the wall is provided to resist potential splitting of the concrete in the e of the vertical bars. Tangential shear (Vu) resulting from the earthquake loading is resisted he concrete and diagonal reinforcing bars. A maximum allowable tangential shear stress is gned to the concrete. Stresses in excess of Vc are resisted by inclined reinforcing bars hored in the mat. No diagonal rebars are required above the spring line.

3.8-2 Rev. 30

rances of less than 2 inches were reviewed on a case-by-case basis and were determined to vide adequate clearance to preclude interaction with the containment liner for the worst case gn bases movements.

etrations in the liner are provided for piping, electrical conductors, fuel transfer tube, and ge air ducts as indicated in Section 3.8.1.1.4. Access openings in the structure are the 7 foot de diameter personnel access lock and the 15 foot inside diameter equipment hatch ure 3.8-7). The reinforcement around these penetrations is shown in Figures 3.8-5 through 9.

ckets for support of internal miscellaneous small piping, and electrical conductors are attached nsert plates in the liner and/or to overlay plates on the liner. Major equipment and pipe loads, ept at the penetrations, are not carried by the exterior wall. Steel anchors were used to attach insert plates and the liner to the concrete structure. External supports on the containment wall the enclosure structure are shown in Figures 3.8-10 and 3.8-61.

1.1.3 Dome inside radius of the 2 foot 6 inch thick dome is 70 feet. The internal height from base mat to center of the dome is 201 feet 3 inches. The dome reinforcement consists of a layer of forcement placed meridionally extending from the vertical bars of the cylindrical wall near h face and horizontal hoop bars in similar layers. Mechanical anchorages are provided on the set of meridional bars extending up toward the crown of the dome. All other meridional bars terminated at a point where they are adequately developed beyond the point at which they e required for design purposes. Near the crown meridional bars are welded to a ring cast in the crete concentric with the centerline of the containment. Figure 3.8-11 shows the rebar at the e cylinder junction.

dome liner plate is 0.5 inch thick. Attachments to the dome are basically for the quench and rculation spray nozzle lines. One major connection to the dome structure is the top pivot port for the spray header maintenance truss (Figure 3.8-12). This structure is supported at the of the dome and on the polar crane rail.

1.1.4 Steel Liner and Penetrations Steel Liner The steel liner consists of a vertical cylindrical portion, closed at the top by a hemispherical dome and attached at the bottom by a mat liner portion. As shown in Figure 3.8-13, the liner is welded to a skirt ring (knuckle plate assembly) that is welded to a plate which in turn is embedded and anchored into the concrete mat. The knuckle plate-to-liner junction and the knuckle plate-to-mat anchorage is proportioned to develop the full strength of the liner.

3.8-3 Rev. 30

bars so that the overall deformation of the liner under the parameters derived from the design basis accident (DBA) and normal operation is essentially the same as that of the concrete containment structure.

The function of the liner is to act as a gas-tight membrane under conditions that can be encountered throughout the operating life of the plant. The liner is designed to resist all direct loads and accommodate deformation of the concrete containment structure without jeopardizing leak-tight integrity. Under DBA conditions, the liner is under a state of biaxial compressive strain due to thermal effects and during the test condition, the liner plate is under a state of biaxial tensile strain. The anchor studs prevent buckling of the liner and act as nodal points. Tests conducted at Northeastern University, Boston, Massachusetts, using 5/8-inch diameter studs and 3/8-inch thick plate, show that shear failure occurs in the stud adjacent to the weld connecting the stud to the plate; in no instance was the plate damaged. Tests conducted for the stud manufacturer under the direction of Dr. I.M. Viest (TRW, Inc. 1975) indicate that, with the manufacturers recommended depth of embedment of the stud in concrete, the ultimate strength of the stud material can be developed in direct tension. The reinforcement ring and liner adjacent to the hatches are anchored to the concrete containment with a denser stud pattern.

The liner pressure boundary includes embedments, insert plates, and penetrations. Liner dimensions are given in Sections 3.8.1.1.1 to 3.8.1.1.3 and shown in Figure 3.8-14. Leak chase channels are installed over penetration to liner seams and over knuckle plate to liner seams.

Embedments Three types of embedments are used to maintain the leaktightness of the steel membrane while transferring loads across the mat liner plate to the concrete mat. One is a 3 x 6 rectangular forged bar also called a bridging bar, another is a 1.25 inch thick plate, and the other is a 5 inch thick forged plate to which the neutron shield tank is mounted. Leak test channels are welded all around the embedments to ensure the leaktightness of the steel membrane. Vertical reinforcing steel is Cadwelded to the top and bottom of the embedments providing reinforcing bar continuity without creating multiple penetrations.

Insert Plates Loads from supports for piping such as the spray headers and other miscellaneous equipment are transferred to the containment concrete wall through insert plates and their anchors. Each insert plate is designed to provide a rigid base for the attached supports and anchor studs. Sufficient insert plate anchorage is provided so that the adjacent liner plate sees negligible stress from the applied support loads and the leak tight integrity is well maintained.

Penetrations 3.8-4 Rev. 30

Figures 3.8-15 through 3.8-22). Containment penetrations are anchored and transfer loads to the reinforced concrete containment wall. Leak test channels for repaired penetrations are partially cut and not replaced. These channels are welded over all seams between the penetration sleeve, and reinforcement plate and reinforcement plate to liner. These penetrations are classified as follows:

1. Sleeved Piping Penetration These penetrations have a sleeve around the outside of forged piping with integral flued head. Sleeved penetrations are used for multiple small pipes passing through one penetration and for thermally hot piping systems. Thermally hot piping is insulated to prevent the operating temperature of the concrete adjacent to the sleeve, during normal operation or any other long-term period, from exceeding 150°F except at local areas around the penetrations which are allowed to have increased temperatures not exceeding 200°F; for accident or other short-term periods, the temperatures are not to exceed 350°F for the interior surface.

However, local areas are allowed to reach 650°F from steam or water jets in the event of pipe failure. Penetrations in which the insulation would be insufficient to maintain the concrete within the allowable temperature limit are equipped with a cooling jacket located inside the sleeve. The cooling water for the cooling jacket is supplied by the component cooling water subsystem. Each penetration sleeve carrying thermally hot piping is designed with adequate space between the sleeve and the piping to allow for the required pipe insulation and for the cooling jacket.

Piping located outside the containment carrying the cooling water does not require any secondary penetration of the containment structure. The thermally hot piping is connected to the sleeve by a forged flued headed pipe providing a transition from the pipe to the sleeve and designed so that stress concentrations are minimized. The penetration sleeve is welded to a liner reinforcement plate (Figure 3.8-16). Loads are transferred from piping to the containment structure through the forged flued head to the sleeve and liner reinforcement plate. The forging is designed so that the heat flow from the pipe to the sleeve is at a minimum and structurally adequate to take the pipe load. Shear bars are provided on the outside of the sleeve to carry torsional pipe loads. Multiple small piping penetrations pass through a forged attached plate which is welded to a sleeve (Figure 3.8-17).

2. Unsleeved Piping Penetrations These penetrations consist of piping installed through the containment wall that are thermally cold piping systems and only one pipe is passing through the penetration. The process pipe is welded directly to the reinforcement plate (Figure 3.8-18).
3. Electrical Penetrations 3.8-5 Rev. 30

thermocouple leads to solid rods for power circuits. Each penetration required either a 12 inch or an 18 inch diameter steel sleeve. The sleeves are welded to liner reinforcement plates. The electrical leads are installed in the penetration assemblies which are mounted to the pipe sleeve by a welded flange. Individual conductors can be replaced without cutting the containment liner or sleeve. The penetrations are constructed and tested in accordance with IEEE Standard 317 dated 9-20-72, Electrical Penetration Assemblies in Containment Structures for Nuclear Generating Stations. Each installed penetration is periodically tested for leak tightness.

4. Fuel Transfer Tube (Mechanical Transfer System)

This penetration is provided for fuel transfer between the containment structure and the fuel building. The penetration consists of a stainless steel pipe installed inside a stainless steel enclosure with bellows expansion joints to compensate for differential movements of the buildings. The inner pipe acts as the transfer tube and connects the containment refueling canal with the spent fuel pool in the fuel building via the fuel transfer canal. The enclosure is welded to the containment liner. The bellows were selected to withstand thermal expansion differentials, seismic motions, and radial and axial differential movements of the fuel building and the containment structure. The enclosure has a tap which provides means for leak testing the system. A blind flange is provided on the inner tube of the containment side and a valve on the fuel building side (Figure 3.8-20).

5. Personnel Air Lock The personnel air lock is a double closure penetration (Figure 3.8-21). Each closure head is hinged and double gasketed with a leakage test tap between the O rings. The enclosed space between the O rings is pressurized to containment design pressure to test for leakage through the access door when it is locked in place. The personnel access lock can be independently pressurized up to containment design pressure for testing. Both doors are hydraulically latched and hydraulically swung. Both doors are interlocked so that in the event one door is opening the other cannot be actuated. Both doors are furnished with a pressure equalizing connection. The equalizing valves are manually and automatically operated by the person entering or leaving the personnel access lock. The personnel access lock is externally protected from tornado missiles by concrete shield walls and a roof.
6. Equipment Hatch The equipment hatch is a single closure penetration (Figure 3.8-22). The equipment hatch cover is mounted inside the containment structure and is double gasketed with a leakage test tap between the O rings. The enclosed space 3.8-6 Rev. 30

cover is provided with a hoist with two point suspension and a sliding rail for storage. A positive locking device is furnished to prevent circular swing. A removable concrete tornado missile shield protects the equipment hatch. These concrete shield blocks are removed during Modes 5 and 6. Based on a probabilistic analysis, the mean value of a tornado missile impacting the equipment hatch during Modes 5 and 6 is on the order of 10-7 per year.

1.1.5 Ring Girder einforced concrete ring girder is provided which encircles the containment structure and is ated from the containment wall by a compressible material (Figure 3.8-23). The ring girder is vided to prevent postulated sliding of rock wedges toward the containment wall during a mic event. Sliding would occur when seismic forces exceeded the frictional forces on the rock t and/or foliation planes. The ring is analyzed as a laterally loaded and supported ring, in ch only inward unrestrained deflection is possible. This is because the rock itself resists lateral ward movement. The ring girder does not interact with the containment structure except that containment mat gives vertical support for the ring gravitational forces. In the area of the ESF ding, radial walls in the lower chambers act as struts between the ring and the ESF building wall (Figure 3.8-24). Through these walls, relatively continuous support for the ring is vided around this area of the containment. Lateral displacements of the ring are calculated to a maximum of 0.2 inch, which is negligible with respect to the 4 inches of compressible erial which is provided to isolate the containment shell from any lateral loads above the mat.

ring girder is a seismic Category I structure and is designed for all applicable loadings uding: dead, seismic, and rock pressure loads. Loads and load combinations are in accordance h Section 3.8.1.3.

allowable shear stress for the design of the ring girder is in accordance with ACI 318-71 ause no circumferential tension exists in the ring. Materials and quality control are in ordance with Section 3.8.1.6. There are no testing or inservice inspection requirements.

1.2 Applicable Codes, Standards, and Specifications ctural design, materials, the tests of material and the methods of testing, where applicable, form to the following codes, standards, and specifications unless otherwise stated.

1.2.1 General letters ASTM, AISC, AWS, ACI, and BOCA refer to the American Society for Testing and erials, American Institute of Steel Construction, American Welding Society, American crete Institute, and Building Officials and Code Administrators, respectively.

ere multiple dates are shown for structural specifications in Section 3.8.1.2.2, these reflect uments whose applicable issue has changed during construction. The engineering cifications reflect the specific issue date in effect at any given time.

3.8-7 Rev. 30

1. ACI 211.1-70 Recommended Practice for Selecting Proportions for Normal Weight Concrete
2. ACI 214-65 Recommended Practice for Evaluation of Compression Test Results of Field Concrete
3. ACI 301-72 Specification for Structural Concrete for Buildings 4a. ACI 614-59 Recommended Practice for Measuring, Mixing, and Placing Concrete 4b. ACI 304-73 Recommended Practice for Measuring, Mixing, Transporting, and Placing Concrete
5. ACI 305-72 Recommended Practice for Hot Weather Concreting
6. ACI 306-66 Recommended Practice for Cold Weather Concreting
7. ACI 318-71 Building Code Requirements for Reinforced Concrete
8. ACI 347-68 Recommended Practice for Concrete Formwork
9. AISC Specification for the Design, Fabrication and Erection of Structural Steel for Buildings Seventh Edition (February 12, 1969); Supplement No. 1 (November 1, 1970), Supplement No. 2 (December 8, 1971), and Supplement No. 3 (June 12, 1974)

NOTE: AISC Eighth Edition (1980) used in design of containment building enclosure structure

10. AISC Specification for Structural Joints Using ASTM A325 or A490 Bolts (April 18, 1972)
11. ASTM A 36-74 (1977) Specification for Structural Steel NOTE: ASTM A36-77 Used for fabrication of shield doors
12. ASTM A 193-73 Standard Specification for Alloy Steel and Stainless Steel Bolting Materials for High-Temperature Services
13. ASTM A 307-74 Specification for Low Carbon Steel Externally and Internally Threaded Standard Fasteners
14. ASTM A 325-66, 74 Specification for Low Carbon Steel Externally and Internally Threaded Standard Fasteners
15. ASTM A 440-74 Specification for High Strength Structural Steel
16. ASTM A 441-74 Specification for High Strength Low-Alloy Structural Manganese Vanadium Steel 3.8-8 Rev. 30
18. ASTM A 588-79 Specification for High-Strength Low-Alloy Structural Steel with 50,000 psi Minimum Yield Point to 4 Inches Thick
19. ASTM A 615-72 Standard Specification for Deformed Billet Steel Bars for Concrete Reinforcement including Supplement S-1
20. ASTM C 31-69 Making and Curing Concrete Compressive and Flexural Strength Test Specimens in the Field
21. ASTM C 33-71a Standard Specification for Concrete Aggregates (and 1978 Revision)
22. ASTM C 94-71 (1974) Specification for Ready-Mixed Concrete
23. ASTM C 109-1973 Method of Test for Compressive Strength of Hydraulic Mortars (using 2-inch (50 mm) Cube Specimens)
24. ASTM C 143-7 Method of Test for Slump of Portland Cement Concrete
25. ASTM C 150-73 Specification for Portland Cement
26. ASTM C227-71 Test for Potential Reactivity of Cement Aggregate Combinations (Mortar Bar Method)
27. ASTM C 233-69 (1973) Standard Method of Testing Air-entraining Admixtures for Concrete
28. ASTM C 235-68 Test for Scratch Hardness of Coarse Aggregate Particles
29. ASTM C 260-69 (1973, 1974) Air-entraining Admixtures for Concrete
30. ASTM C 289-71 Test for Potential Reactivity of Aggregates (Chemical Method)
31. ASTM C 295-1965 (1973) Recommended Practice for Petrographic Examination of Aggregates for Concrete
32. ASTM C 586-69 Test for Potential Alkali Reactivity of Carbonate Rocks for Concrete Aggregates
33. AWS D1.1-72 Rev. 1-73 (1979) Structural Welding Code NOTE: AWS D1.1-79 used for fabrication of watertight, airtight, pressure resistant, and shield doors.
34. AWS D12.1-61 Recommended Practices for Welding Reinforcing Steel, Metal Inserts and Connections in Reinforced Concrete Construction
35. NRC Regulatory Guides as qualified in Section 1.8 on the following topics:
a. Cadweld Splices 1.8.1.10 3.8-9 Rev. 30
c. Structural Acceptance Testing 1.8.1.18
d. Placement of Concrete 1.8.1.55
e. Design Response Spectra 1.8.1.60
f. Seismic Damping Values 1.8.1.61
36. BOCA Basic Building Code of the Building Officials and Code Administrators International, Inc., 1970
37. State of Connecticut Basic Building Code, 197 1.2.3 Steel Liner and Penetrations re was no applicable code for the design of concrete containment structure liners at the inning of the construction of the Millstone liner. However, ASME Section III, Divisions 1 and nd Section VIII were used as a guide.

ign, materials, fabrication, testing, and inspection, where applicable, conform to the following es, standards, and specifications:

a. ASME Boiler and Pressure Vessel Code Sections II, III, and V, 1971 issue, including addenda up to and including the 1973 Summer addendum, and Section III, Division 2, Subsection CC, 1980 issue, including addenda up to and including Summer 1982.
b. ASME Boiler and Pressure Vessel Welding Qualifications,Section IX, issue in effect at the time of qualification.
c. American National Standards Institute, ANSI N101.4, Quality Assurance for Protective Coatings Applied to Nuclear Facilities, 1972 issue.
d. American Society for Testing and Materials, ASTM-E208, Conducting Drop Weight Test to Determine Nil Ductility Transition Temperature of Ferritic Steels, 1969 issue.
e. American Society for Testing and Materials, ASTM E436, Drop Weight Tear Tests of Ferritic Steels, 1971 issue.
f. American Welding Society Structural Welding Code D1.1, 1972 issue.

3.8-10 Rev. 30

Stations, February 4, 1971 issue.

h. Steel Structures Painting Council, Paint Applications Specification, SSPC-PA1, Shop and Field Maintenance Painting 1964 issue.
i. Steel Structures Painting Council, Surface Preparation Specifications SP1, SP2, SP5, SP6, and SP10, 1963 issue.
j. Steel Structures Painting Council, Surface Preparation Specification, SSPC-PS8.01, Rust Preventive System with Thick Film Compounds, 1964 issue.
k. American Society for Nondestructive Testing, Recommended Practice for Nondestructive Testing Personnel, Qualifications, and Certification, ASNT-TC-1A, 1971 issue.
l. American Society for Testing and Materials (ASTM) Specifications A105, A108, and American Society of Mechanical Engineers (ASME) Material Specifications SA131, SA182, SA193, SA194, SA213, SA240, SA234, SA312, SA320, SA333, SA350, SA376, SA403, SA496, SA508, SA516, and SA537.

lier or subsequent revisions (or reissues) of the referenced codes, standards, and specifications be used insofar as their use does not in any way degrade the safety or strength of any cture or portion of a structure.

tion 3.8.1.6 describes materials and quality control procedures and any exceptions to the ve listed codes.

Protection is described in Section 9.5.1.

1.3 Loads and Loading Combinations following sections discuss the loads and load combinations used in the design of the 1.3.1 Containment Mat, Shell, and Dome loadings applicable to concrete containment design include the following:

a. Those loads encountered during preoperational testing.
b. Those loads encountered during normal plant startup, operation, and shutdown, including dead loads, live loads, and thermal loads due to operating temperature.
c. Those loads to be sustained during severe environmental conditions, including those induced by the design wind and the operating basis earthquake.

3.8-11 Rev. 30

e. Those loads to be sustained during abnormal plant conditions, which include a loss-of-coolant accident (LOCA). The main abnormal plant condition for containment design is the design basis LOCA. Also considered are other accidents involving various high energy pipe ruptures. Loads induced on the containment by such accidents include elevated temperatures and pressures and localized loads such as jet impingement and associated missile impact.
f. Those loads to be sustained after abnormal plant conditions including flooding of the containment subsequent to a LOCA for fuel recovery.

ds generated by the design basis accident (DBA) are described in Section 6.2.1. The design s wind and tornado loadings are discussed in Section 3.3. The loadings for groundwater and d are given in Section 3.4.

sile loads are described in Section 3.5.1.4.

design bases for protection against postulated rupture of piping are described in Section 3.6.

earthquake loadings which include consideration of simultaneous excitation from two hogonal) horizontal and one vertical earthquake motion are described in Section 3.7.

mal operating temperatures are described in Section 9.4.7.

the containment structure shell and its foundation mat, the load combinations given in Section 2 of ACI 318 are replaced by the following:

For service load categories, the structure is designed using stresses well within elastic limits. Specifically, the allowable compressive stress in concrete is 0.45 f'c, the allowable stress for steel is 0.5 fy, and shear stresses are 50 percent of capacity as given in ACI 318.

In effect, this permits stress levels approximately 50 percent of those for the factored load conditions. However, structural members subjected to test pressure, temperature, or wind, when combined with other forces, are designed for the allowable stresses increased by 33 percent.

CONTAINMENT LOAD COMBINATIONS AND FACTORS Case Load Category Load Combination Service:

1 Test 1.OD + 1.OL + 1.0Pt + 1.0Tt 2 Construction 1.OD + 1.OL + 1.OTo 3.8-12 Rev. 30

3 Normal 1.OD + 1.OL + 1.OTo + 1.0 OBE + 1.ORo + 1.0Pv Factored:

4 Extreme 1.OD + 1.OL + 1.OTo + 1.OWt + 1.OPv + 1.ORo environmental 5 Extreme 1.OD + 1.OL + 1.OTo + 1.0SSE + 1.ORo + 1.OPv environmental 6 Abnormal 1.OD + 1.OL + 1.5Pa + 1.OTa + 1.ORa 7 Abnormal/Severe 1.OD + 1.OL + 1.25Pa + 1.OTa + 1.25OBE + 1.ORa environmental 8 Abnormal/Severe 1.OD + 1.OL + 1.25Pa + 1.OTa + 1.25W + 1.ORa environmental 9 Abnormal/Extreme 1.OD + 1.OL + 1.OPa + 1.OTa + 1.0SSE + 1.ORa environmental In Case 3, W is substituted for OBE if its effect is greater.

In Cases 6, 7, 8, and 9, Yr is substituted for Ra if its effect is greater.

menclature is as follows:

D = Dead load of structures and contents, including effects of earth and hydrostatic pressures, buoyancy, ice, and snow loads.

SEE = Safe shutdown earthquake (see note).

OBE = Operating basis earthquake (see note).

L = Live load.

Pa = Pressure load from DBA pressure transient, including the design margin, as defined in Section 6.2.1.

Pt = Test pressure.

Pv = Subatmospheric minimum operational pressure of containment structure (or atmosphere).

Ra = Piping loads due to increased temperature resulting from the DBA.

Ro = Pipe loads during normal operating conditions.

Ta = Load due to temperature gradient through the concrete shell and mat plus load exerted by the liner, based on temperature associated with the DBA. This is that 3.8-13 Rev. 30

To = Load due to temperature gradient through the concrete shell and mat, plus load exerted by the liner based on temperature associated with normal operation.

Tt = Load from test temperature.

W = Wind load (see note).

Wt = Tornado load (see note).

Yr = Reaction from pipe restraint (penetration) due to pipe break (local effects).

TE: Wind or tornado loads are not coincident with earthquake loads.

oad factor of 1.0 is assigned to loads caused by the design tornado because the wind velocity en in Section 3.3.2 is a conservative estimate of the actual tornado wind velocity to be ected, and the probability of a tornado striking the unit is low (Section 2.3.1.3). A load factor

.0 is also used for loads caused by the SSE (Section 3.7.1) since the SSE acceleration is two es the largest acceleration expected at the site.

nado wind effects are determined as described in Section 3.3.

effects of hydrostatic loading due to groundwater and the probable maximum flood level ction 3.4.2) are considered so as to reduce the effect of the dead load (D).

ing construction, dewatering was maintained until the weight of the containment exceeded the tulated hydrostatic uplift forces.

tion 3.8.1.4 gives the design stresses to be used with these loads.

1.3.2 Steel Liner and Penetrations containment structure liner plate, penetrations, brackets and attachments, anchors, and access nings were designed for the load combinations presented in Tables 3.8-1 and 3.8-2. Sleeves of penetrations were considered part of the containment boundary. The maximum containment cture design pressure is 45 psig. The minimum containment structure design pressure is 8.0

. This pressure is equivalent to the minimum operating pressure minus the pressure drop due he maximum, hypothetical containment cooldown situation. During this condition, the tainment atmosphere pressure is assumed to be decreased below normal operational pressure the inadvertent operation of the quench spray system during normal unit operation. The lting total pressure is above the minimum containment design pressure of 8.0 psia. For a her discussion of external pressure, refer to Section 6.2.1.1.3.5.

change in barometric pressure due to tornadoes is not expected to exceed 3 psi and the change to the maximum probable hurricane is approximately 1.1 psi. These pressure changes result in ecrease in the atmospheric pressure, which, in turn, decreases the differential between 3.8-14 Rev. 30

mal, and seismic loads.

ures 3.8-25 through 3.8-55 are scaled load plots of moments, shears, and membrane forces for containment structure. The loading conditions are listed on the plots. Compressive forces are ted as negative values and bending moments are plotted on the tension side.

1.4 Design and Analysis Procedures 1.4.1 Containment Structure containment structure consists of a hemispherical dome, a cylindrical shell, and a circular mat ported by an elastic subgrade. The design, analyses, and construction of the containment cture is similar to and takes full advantage of SWEC experience gained in the designs for ts indicated in Section 3.8.1.1.

er design internal pressure, discontinuity forces exist at the junction of the mat and the ndrical shell and also at the junction of the shell and the dome. The cylindrical shell is of such ngth that the influence of one discontinuity on the other is negligible.

analysis of the mat for axisymmetric loading is accomplished using the program MAT 5 ch is described in Appendix 3A. This program treats the mat as a symmetrically loaded ular plate on an elastic foundation. The general method is described by Zhemochkin et al.,

62). The program is set up so that the foundation stiffness can be formulated either through ssinesqs approach or through a Winkler-type assumption. Zhemochkins method is modified ccount for a finite depth of elastic foundation, i.e., the distance between mat and underlying

k. In addition, the cylindrical containment wall, crane wall, and primary shield wall are sidered as elastic constraints, which are determined by applying compatibility conditions at shell mat interface.

the purpose of calculating the elastic constraint of the containment, the base of the cylinder is med to be completely cracked vertically and cracked horizontally to the neutral axis of the sformed section. At this location, the cylinder has a hoop stiffness of the circumferential rs and the meridional bending stiffness of the transformed section.

hort distance above the mat, the meridional bending moment becomes so small that the entire idional cross-section is in a state of tension. Above this plane, the cylinder is assumed to be pletely cracked horizontally and vertically.

s, the elastic constraint is determined from a cylinder which is divided into a short shell ing the properties of a cross-section completely cracked vertically and partially cracked zontally, and a long shell having the properties of a cross-section completely cracked zontally and vertically.

3.8-15 Rev. 30

g the finite difference computer program SHELL I.

discontinuity forces at the mat-shell junction, calculated as a part of the mat analysis, are used oundary conditions for the analysis of the cylindrical shell which is performed using the gram SHELL I. The seismic analysis of the containment structure provides the accelerations to ch the containment structure would be subjected. These accelerations are used for determining ic loading on the shell. The tangential shear caused by the seismic loading is resisted by the crete or by the concrete and a system of diagonal rebars. The maximum allowable tangential ar stress carried by the concrete, (Vc) is assumed to be 40 psi. Stresses in excess of Vc are sted by diagonal reinforcing bars. The specified design compressive strength of the concrete ying the tangential shear is not less than 3,000 psi with coarse aggregate not smaller than size 67 as given in ASTM C33.

diagonal reinforcement required to supplement the tangential shear capacity of the concrete sists of No. 18 rebars anchored in the mat as shown in Figure 3.8-56. The spacing between e diagonal bars is increased as the design tangential shear decreases at higher wall elevations.

diagonal rebars are required above the spring line.

requirements for diagonal reinforcement necessary for carrying the (Vu-Vc) shear force are rmined by an analysis based on a paper by Prof. M. J. Holley, Jr. (1969). The mesh of vertical, zontal, and diagonal reinforcement is assumed to be a continuum. Forces can be determined ach type of bar for symmetric loadings such as internal pressure, dead load, vertical, and zontal earthquake loads. The diagonal reinforcement is designed to resist force due to metric loadings and the entire tangential shear not carried by the concrete (Vu-Vc).

o temperature conditions are considered in the analysis of the containment:

1. Temperature under normal operating condition.
2. Temperature associated with the DBA, when combined with the coincident internal pressure, produces the most adverse effects on the reactor containment structure.

er normal operating conditions, the temperature gradient to produce the highest stress ltant is used.

effect of the liner temperature increase associated with the DBA condition on the concrete tainment shell is determined by the following procedure. For this condition, it is assumed that temperature of the liner increases and the concrete remains at its ambient temperature. Liner ansion is limited by the concrete shell and a pressure develops between the concrete shell and liner. The equivalent pressure exerted by the liner on the concrete shell is given by the ression:

Pe = LH*hL/RL 3.8-16 Rev. 30

Pe = Equivalent pressure LH = Circumferential liner stress hL = Thickness of liner RL = Radius of liner he junction of the mat and shell, it is assumed that the mat prevents radial movement of the l; therefore, at this location, the circumferential stress in the liner used to determine equivalent sure is:

LH = alpha*Es*delta T re:

alpha = Coefficient of thermal expansion of the liner Es = Young's modulus of the steel liner delta T = Change in temperature due to DBA hort distance above the mat, where the effect of the mat to shell discontinuity is negligible, the r and concrete shell expand due to the DBA pressure and temperature of the liner. Free radial lacement of the liner due to DBA temperature is larger than the displacement of the reinforced crete shell due to the DBA pressure. Thus, the liner is constrained by the concrete shell. The lting pressure exerted by the liner is determined from the expression for equivalent pressure n before. The liner stress is determined by the following procedure:

Expressions for stresses in the shell and liner are written in terms of the meridional and circumferential strains. These are inserted into the force equilibrium equations resulting in an explicit solution for liner strains from which liner stresses are determined.

effects of creep and shrinkage of concrete are not important considerations in the analysis and gn of a nonprestressed concrete containment. Shrinkage results in meridional and radial lacements which are the opposite of the displacements caused by the principal loads, perature and internal pressure. Consequently, the effects of creep and shrinkage can be safely ored.

cking is an important consideration in the analysis and design of a reinforced concrete tainment. For this reason, stiffness of the concrete is adjusted for the extent of cracking present er various design conditions. When the concrete is completely cracked, calculation of the ness of the structure uses only the properties of the reinforcing steel. The steel liner is med to make no contribution to the structural integrity of the containment shell except during structural acceptance test.

3.8-17 Rev. 30

1. 12-inch diameter (nominal) or less No special or additional reinforcing is provided. The principal wall reinforcement is located to avoid interference with the penetration.
2. All piping penetrations larger than 12-inch diameter (nominal).

Reinforcing bars terminated at penetrations are replaced by at least twice the number of bars, half of these being placed on each side of the opening. Diagonal reinforcing bars are also provided around openings to take shear and tension. The anchorage length of the additional bars that frame the openings is determined by using a conservative value for bond stress. In addition, shear assemblies as shown in Figure 3.8-5 are added around 48-inch penetrations. This method is consistent with established practice and pressure-tested at the plants previously listed.

3. Personnel Access and Equipment Access Hatches Penetrations for the equipment and personnel access hatches are analyzed using the 3-dimensional finite element capability of the computer program STRUDL II which is discussed in Appendix 3A.

The thickened ring beam and cylinder wall for both hatches are assumed to be cracked, and to have the extensional stiffness of the reinforcing bars only. The analysis shows that sizable tangential (in plane) shears exist in the wall near the ring beam. These shears are resisted by reinforcing bars which are placed parallel to the typical earthquake shear bars.

The ring beam is designed to resist the axial tension and shears resulting from the loading criteria listed in Section 3.8.1.3. The axial tension is assumed to be resisted by the reinforcing bars only. The shears, including torsional shear, are resisted entirely by stirrups placed radially around the penetrations.

In effect, any concrete resistance to tension and shear is neglected. The principal circumferential and meridional reinforcing bars, as designed, are extended to the inner face of the ring beam, hooked and Cadwelded to each other, thereby providing shear resistance additional to that provided in the design.

The normal pattern of membrane forces and moments (meridional and circumferential) in the containment wall is disrupted in the region of the hatch openings. The redistribution of these forces and moments is provided by the finite element computer program and extra reinforcement is added to areas of marked deviation from the normal pattern.

3.8-18 Rev. 30

sses due to strain compatibility of the liner with the reinforced concrete shell due to various binations of pressure, thermal, self-weight, and seismic load were determined using Stone &

bsters KALNINS program. This is a direct integration program for static analysis of tilayered thin shells of revolution. The stress analysis of a shell subjected to mechanical and mal surface loads and edge loads, is reduced to a boundary value problem governed by a em of nonhomogeneous, linear, partial differential equations. The equations are separable h respect to the meridional and circumferential coordinates of the shell. The solution for each arable component of the loads is obtained by solving a typical two point boundary value blem governed by eight first order linear ordinary differential equations using direct gration. Local stresses due to irregular spacing of liner-headed anchor studs were determined g the ANSYS finite element program and by manual calculations. Analytical evaluation of penetration discontinuities were modeled on the ASAAS program (Asymmetric Stress lysis of Axisymmetric Solids). The method of analysis employed is based on a finite element lization of an axisymmetric solid. Each element is an axisymmetric ring of a constant cross ion. Since such a solid may be loaded and may deform in nonaxisymmetric modes and since properties of the material may vary in all directions (e.g., due to temperature variations), all dependent variables including the material properties are expressed as truncated Fourier series h the circumferential coordinate being the independent variable.

perature profiles for the penetrations were determined using the TAC-2D program. TAC-2D computer program for calculating steady-state and transient temperatures in two dimensional blems by the finite difference method. The configuration of the body to be analyzed is cribed in the rectangular, cylindrical, or circular (polar) coordinate system by orthogonal lines onstant coordinate called grid lines. The grid lines specify an array of nodal elements. Nodal nts are defined as lying midway between the bounding grid lines of these elements. A finite erence equation is formulated for each nodal point in terms of its capacitance, heat generation heat flow paths to neighboring nodal points.

1.5 Structural Acceptance Criteria 1.5.1 Containment Structure containment structure is designed for the loads and load combinations presented in tion 3.8.1.3.1. Allowable stresses, unless otherwise defined, are in accordance with ACI

-71. For the factored load combinations, design of the containment structure meets the broad nt of Article CC-3400 of ASME III Division 2. Details of the design conform to ACI 318-71 the additional requirements discussed in Section 3.8.1.4, rather than the parallel requirements C-3000. Major features of the concrete design are similar using either code.

ept for test conditions, the specific limits for service loads given in CC-3430 are not addressed cceptance criteria for ACI 318-71. However, design of the containment equals or exceeds ACI

-71 requirements for serviceability. Predicted stresses and strains for structural acceptance s are well within the limits stated in CC-3430.

3.8-19 Rev. 30

ential shear stress, Vc, carried by the concrete is assumed to be not greater than 40 psi.

1.5.2 Steel Liner and Penetration containment structure liner plate, brackets, attachments, anchors, and access openings are gned to meet the design allowables presented in Table 3.8-1. Initial penetration sizing is ormed in accordance with Table 3.8-2. The final design verification is in accordance with les 3.8-4 through 3.8-6.

minimize the probabilities of crack propagation as the containment is exposed to the design ditions as indicated in Table 3.8-1, stress levels reached are kept within the limits given by ini and Loss in NRL Report 6900, Integration of Metallurgical and Fracture Mechanics cepts of Transition Temperature Factors Relating to Fracture-Safe Design for Structural ls.

1.6 Materials, Quality Control, and Special Construction Techniques tion 3.8.1.2 contains applicable codes, standards, and specifications.

quality assurance activities required by this section are described in Section 17.1.

1.6.1 Concrete Materials and Quality Control I 301, ACI 347, and ACI 318 form the general basis for the concrete specifications.

I 301 was supplemented, as necessary, with mandatory requirements relating to types and ngths of concrete, including minimum concrete densities, proportioning of ingredients, forcing steel requirements, joint treatments, and testing agency requirements.

mixtures, types of cement, bonding of joints, embedded items, concrete curing, additional test cimens, additional testing services, cement and reinforcing steel mill test report requirements, additional concrete test requirements were specified in detail.

Portland cement conformed to requirements of ASTM C150 for Portland cement, Type II.

alkali Portland cement was used where examinations and tests of aggregates, or of structures taining similar aggregates, indicate potential reactivity with the cement. One exception was calcium aluminate (high-alumina) cement was used for porous concrete enclosed within the erproof membrane under the containment mat. This substitution is made to reduce the sibility of clogging of voids by continued hydration that occurs with Portland cement.

tified copies of mill tests, showing that the cement meets or exceeds the ASTM requirements Portland cement, were furnished by the manufacturer. An independent testing laboratory was 3.8-20 Rev. 30

air-entraining admixture was used in the concrete. The total air content, expressed as a entage by volume, conforms to the requirements in Table 3.4.1 of ACI 301, Total Air Content Various Sizes of Coarse Aggregates for Normal Weight Concrete. This admixture conforms to requirements of ASTM C 260 when tested in accordance with ASTM C 233. The entraining agent was added separately to the batch in solution. The solution was batched by ns of a mechanical dispenser capable of accurate measurement and was subject to periodic cks. Air-entrained cement was not used.

er admixtures to control the rate of set, reduce the water content, or improve the workability cohesiveness of concrete were used in specific instances. Such admixtures were used only r tests were made in combination with the cement and aggregates being used and were cifically approved by the structural engineer. Calcium chloride was not used under any umstance.

ing water and ice were clean and free from injurious amounts of oils, acids, alkalies, salts, anic materials, or other substances which may be deleterious to concrete or steel. The mixing er and water used for making ice are periodically checked and tested for suitability by ASTM 09 Method of Test for Compressive Strength of Hydraulic Cement Mortars (using 2-inch mm) Cube Specimens).

e and coarse aggregates conform to the requirements of ASTM C 33. Coarse aggregates tained not more than 5 percent soft fragments in accordance with ASTM C 235. Aggregates e evaluated for potential chemical alkali reactivity prior to use by ASTM C 295, ommended Practice for Petrographic Examination of Aggregates for Concrete, and ASTM C

, Test for Potential Reactivity of Aggregates (Chemical Method). Also, prior to use of an regate, one of the following tests for alkali reactivity (whichever is appropriate for the regate) was started: ASTM C 227, Test for Potential Reactivity of Cement Aggregate mbinations (Mortar Bar Method), or ASTM C 586, Test for Potential Alkali Reactivity of bonate Rocks for Concrete Aggregates. Test results were evaluated for aggregate suitability h the cement. Aggregates were free from any materials that would be deleteriously reactive in amount sufficient to cause excessive expansion of mortar or concrete. All aggregates were ed for compliance with the above requirements by an independent testing laboratory.

portioning of structural concrete conforms to ACI 301, Chapter 3. In general, concrete mixes e of a 28 day strength of 3,000 psi unless otherwise specified by the Engineers.

crete used for biological shielding has a density of not less than 140 lb/cu ft.

portioning of ingredients in concrete mixes were determined and tests conducted in ordance with the methods detailed in ACI 301 and ACI 211.1 for combinations of materials to stablished by trial mixes.

mp of mass concrete is in accordance with ACI 301, Chapter 3.

3.8-21 Rev. 30

with the exceptions given in Section 1.8.

ching and mixing conformed to ACI 301, Chapter 7. Placing of concrete was by bottom dump kets, chuting, concrete pump, or by conveyor belt. Aluminum pipe was not permitted for ping concrete. The rate of placing concrete was controlled so that concrete was effectively ed and compacted with particular attention given around embedded items and near the forms.

tical drops greater than 5 feet for any concrete was not permitted, except where suitable means e provided to prevent segregation. Placing equipment and methods were reviewed for pliance with the specifications.

ACI and ASTM specifications were supplemented as necessary with mandatory requirements ting to types and strengths of concrete, minimum concrete density, proportioning of edients, reinforcing steel requirements, joint treatments, testing requirements, and quality trol.

ing and protection of freshly deposited concrete conform to Chapter 12 of ACI 301, with the owing supplementary provisions:

1. Concrete cured with water was kept wet by covering with an approved water saturated material, by a system of perforated pipes or mechanical sprinklers, or by any other approved methods which kept surfaces continuously wet.
2. The surfaces on which curing compounds may be used were specified by the structural engineer. Curing compounds whose base is composed of sodium silicate, magnesium fluosilicate, or zinc fluosilicate were used on surfaces to which additional concrete was to be bonded, except those surfaces specifically requiring water curing. Other curing compounds were not used on surfaces to which additional concrete is bonded.

struction joints that were to be bonded or transferring shear through shear friction (per I-318) were properly prepared as follows.

er the initial concrete set had occurred, but before the concrete had reached its final set, the aces of these construction joints are thoroughly cleaned to remove all laitance and to expose n, sound aggregate. After cutting, the surface was washed and rinsed. All excess water which not absorbed by the concrete was removed.

ere, in the opinion of the field engineer, the use of an air-water jet was not advisable, then that ace was roughened by bushhammering, sand blasting, or other satisfactory means to produce requisite clean surface. Horizontal construction joints were covered by a 0.5 inch thick layer and/cement grout and new concrete then placed immediately against the fresh grout.

3.8-22 Rev. 30

gn of concrete placed in any one day.

test specimens for compressive strength were 6 inch diameter by 12 inch long cylinders forming to ASTM C 31.

en required, concrete strength tests were evaluated on a statistical basis by the engineers in ordance with ACI 214, Recommended Practice for Evaluation of Compression Test Results of d Concrete, and Chapter 17 of ACI 301. The strength level of the concrete was considered sfactory if it conformed to Section 4.3.3 of ACI 318.

field tests for slump of Portland cement concrete followed ASTM C 143, Method of Test for mp of Portland Cement. Any batch not meeting specified requirements was rejected. Slump s were made frequently during concrete placement and each time concrete test specimens were e.

p detail drawings for the reactor containment mat, shell, and dome reinforcement were cked by the designer.

Special Construction Techniques special construction techniques were used in the placing of concrete for the containment ctures.

1.6.2 Reinforcing Steel Materials and N18 reinforcing bars are controlled chemistry steel of 50,000 psi minimum yield point.

y conform to Grade 40 of the Standard Specification for Deformed Billet-Steel Bars for crete Reinforcement ASTM A 615, as modified, to meet the following chemical and sical requirements:

bon 0.35 percent max nganese 1.25 percent max con 0.15 to 0.25 percent sphorous 0.05 percent max fur 0.05 percent max ld strength 50,000 psi min ngation 13 percent min in an 8-in test sample sile strength 70,000 psi min 3.8-23 Rev. 30

weld T-Series reinforcing steel splices are full tension splices manufactured by Erico ducts, Inc., Cleveland, Ohio, and are used to splice N14 and N18 reinforcing bars. In restricted s, reinforcing bars were butt welded in a manner conforming to the requirements of AWS

.1. Cadweld splices were made in accordance with the instructions for their use issued by the ufacturer, Erico Products, Inc.

nforcing bars smaller than N14 were generally lap spliced. Where lap splicing was ractical, splicing is accomplished by:

1. Use of mechanical (Cadweld) splices as manufactured by Erico Products, Inc.,

Cleveland, Ohio, or equivalent, using the T-series sleeves that develop the full tensile strength of the reinforcing bars, or

2. Butt welding in accordance with the requirements of AWS D12.1.

Quality Control

1. Reinforcing Bars For the special chemistry N14 and N18 bars used in the containment structure, ingots and billets were traced with identifying heat numbers. Bundles of bars were tagged with a heat number as they come off the rolling mill. special mark was rolled into bars conforming to special chemistry N14 and N18 bar to identify them as possessing the chemical and mechanical qualities specified.

Testing of reinforcing bars for Seismic Category I concrete structures met the requirements of Regulatory Guide 1.15, as described in Section 1.8.

SWEC inspectors witnessed, on a random basis, the pouring of the heats and the physical and chemical tests performed by the manufacturer of the special chemistry reinforcing bars.

Bars containing unacceptable inclusions or failing to conform to the required chemistry and physical requirements were rejected.

Mill tests reports showing actual chemical ladle analysis, tensile properties, bend properties, variations in weight, and conformance of deformations were obtained from the manufacturer for each heat. In addition, confirmatory tests for each 50 tons or fraction thereof of each heat of steel for each bar size were made to determine tensile properties. Further, for the special chemistry bars an actual chemical check analysis of each heat was made in addition to confirm the chemical content.

3.8-24 Rev. 30

2. Cadweld Splices Splicing complied with the requirements of Regulatory Guide 1.10, with the exceptions given in Section 1.8.
3. Butt-Welded Splices The ends of the bars to be joined by butt-welding were prepared by sawing or flame cutting, and dressing by grinding, where necessary. Welders were qualified in accordance with AWS D12.1.

All welds were visually inspected. Any cracks, porosity, or other defects were removed by chipping or grinding until sound metal was reached, and then repaired by welding. Peening was not permitted.

Completed reinforcing steel butt-welded splices were selected on a random basis from the containment structure and tensile tested in accordance with the following frequency for each welder:

a. One out of first 10 splices.
b. One out of next 90 splices.
c. Two out of the next and subsequent units of 100 splices.

In addition, completed reinforcing steel butt-welded splices were selected on a random basis from the containment structure and radiographically inspected to meet the following frequency for each welder:

a. One out of the first 10 splices.
b. One out of the next and subsequent units of 25 splices.

Reinforcing steel bars butt welded to steel plate were tested by sister splice, in accordance with the following schedules:

a. One sister splice out of the first 10 production splices.
b. Four sister splices for the next 90 production splices.
c. Three sister splices for the next and each subsequent units of 100 production splices.

3.8-25 Rev. 30

1. General Placing of reinforcing steel, in general, conformed to the requirements of Chapter 5 of ACI 301 and Chapter 7 of ACI 318.

Section 3.8.1.1 describes the placing of the reinforcing steel for the containment structure.

Tack welding of designed reinforcing steel was not permitted.

2. Special No special construction techniques were used in the installation of reinforcing steel for the containment structure.

1.6.3 Structural Steel Material l specifications invoked for structural framing, brackets, and attachments are discussed in tion 3.8.1.2.2.

Quality Control main members, columns, baseplates, bracing, trusses, girts, and bolts larger than 1 inch in meter are traceable to a specific heat number. Traceability to a specific heat number for all clip les, seats, stiffeners, gusset plates, bolts of 1 inch diameter and smaller, and weld filler metal is firmed in the suppliers shop or upon receipt at the site. The storage and issuance of these erials for construction is controlled in a manner which assures only those items procured as Category I are installed in QA Category I applications.

Construction Techniques ctural steel material, erection, and fabrication tolerances are in accordance with the AISC cification for the Design Fabrication, and Erection of Structural Steel for Buildings.

ding of structural steel is in accordance with AWS D 1.1-72, Revision 1-73.

1.6.4 Waterproofing Membrane aterproofing membrane (Figure 3.8-57) was placed below the containment structure mat and ineered Safety Features Building and carried up the containment wall and Engineered Safety tures Building walls to above groundwater level. Attached to and entirely enveloping the part he containment structure and Engineered Safety Features Building below ground level, the 3.8-26 Rev. 30

ns to a layer of porous concrete directly below the mats and above the membrane. This layer porous concrete serves as a horizontal drain under the entire containment structure and ineered Safety Features Building. The porous layer is drained into the Engineered Safety tures Building. A non safety-related pump is used as necessary to remove the water during mal plant operation, post-LOCA and LNP conditions (see 9.3.3.2.4.1 for additional rmation).

tandpipe assembly has been installed through the waterproofing membrane extending to the r at elevation (-) 34 feet 9 inches in the Engineered Safety Features Building. This assembly sists of a one inch diameter pipe with a ball valve, a pressure indicator and a globe valve. The pose of this assembly is to measure the hydrostatic pressure and sample the water below the mbrane. Essentially, the standpipe and valve assembly replace a small piece of the membrane.

standpipe has been securely grouted and sealed into place to preclude membrane leakage.

e samples have been removed from the high alumina cement porous concrete layer from under Engineered Safety Features Building basemat. The coring process has disturbed the mbrane at these locations. The waterproofing membrane has been replaced with grout at these tions.

surface of the containment structure steel liner in contact with concrete is not subject to osion because of the alkaline nature of the concrete.

1.6.5 Steel Liner and Penetrations Materials er plate up to 1.25 inches inclusive and bridging plate are made from SA 537 Class 2 nched and Tempered, nil-ductility transition temperature (NDTT) test not higher than -10°F, h the exception of dome liner plate which is made from SA 537, Class 2 normalized to Class 1 tice, NDTT not higher than -10°F. All liner insert plates and embedment material greater than inch thick was ultrasonically tested prior to installation for the purpose of detecting possible inations.

ghness tests (Charpy V-notch) were performed on all materials which form part of the tainment structure boundary. Nil-ductility Transition Temperature Tests were also performed all ferritic steel that formed part of the pressure boundary but were not required of backing es, test channels, hatch bolts, and hatch nuts.

etration sleeves are made of SA537 Grade B Q&T, SA516 Grade 60 fine grain, normalized SA333 Grade 6 fine grain normalized, all with a NDTT of -10°F.

tron shield tank embedment base and the carbon steel penetration forgings are SA508 Class 1 h a NDTT of +10°F.

3.8-27 Rev. 30

dging bars are made of SA350 Grade LF1 and SA516 Grade 70 normalized with NDTT of 0°F.

p liners and bellows are made of Type 304 stainless steel SA240. The stainless steel etration forgings are made of types 304 and 316, SA182.

Special Construction Techniques ction of the cylindrical portion of the liner plate followed completion of the concrete mat. The r plates served as the internal form for the concrete containment during construction. All liner ms are double butt welded, except for the lower 31 feet of the cylindrical shell liner, the liner damage repair areas, and the mat, where the plates are welded using backing plates. The liner e is continuously anchored to the concrete shell with steel anchor studs and deformed bars.

maximum difference in cross-sectional diameters of the liner is in accordance with the rules wn in paragraph NB-4221.1 of Section III, ASME Boiler and Pressure Vessel Code, Nuclear er Plant Components, 1971 Edition. The maximum misalignment between liner plates is in ordance with paragraph NB-4232 of the ASME Boiler and Pressure Vessel Code, Nuclear er Plant Components, 1971 Edition. All measurements were taken on parent metal and not at ds. Flat spots or sharp angles were not allowed.

allowable deviation from true circular form does not affect the elastic stability of the tainment liner because of the restraint provided by the anchor studs and deformed bars tying it he reinforced concrete shell.

1.6.6 Backfill Around Containment Structure crete was used to backfill around the containment structure. A compressible material was d between the backfill and the containment structure wall to provide a rattle space.

1.7 Testing and Inservice Surveillance Requirements 1.7.1 Concrete Containment tructural acceptance test of the containment was performed after the liner was completed, the concrete poured, and all penetrations, sleeves, and hatches installed. The test is conducted to firm that the design and construction of the containment are adequate to withstand the loads sed by the loss-of-coolant accident as described in Chapter 6.2. The test conforms to the uirements of Regulatory Guide 1.18, Structural Acceptance Test for Concrete Primary Reactor tainments, dated December 28, 1972. The measuring points may be varied or relocated in ordance with paragraph C.3 of the guide.

structure is surveyed, measured, and inspected for cracks prior to the test. The containment is ject to an internal pressure equal to 115 percent of the design pressure. The pressure test mences at atmospheric pressure and is raised to 115 percent of design pressure in a minimum 3.8-28 Rev. 30

stant for at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at each level before deflections are recorded. Crack patterns are sured and recorded at atmospheric pressure both before and after the test and at the maximum sure level.

ial deflection is measured along 6 meridians at 13 feet 6 inches above the top of mat, at

-height between mat and springline, and at the springline of the dome. The exact locations of measurements are indicated in the design specification. Vertical deflections are measured at springline of the dome and the apex. Radial measurements are made using differential sducers supported from the crane wall. Radial and tangential deflections are recorded around periphery of the equipment hatch. Vertical measurement is made using invar tapes.

pping of cracks is performed on exterior surfaces of the containment at locations selected prior tart of the pressure application. Mapping is on one meridian line at three locations, and one tion around the equipment hatch.

ting is not conducted under extreme weather conditions. The environmental conditions are sured and monitored to permit the evaluation of their contribution to the response of the tainment. The testing sequence is repeated if the test pressure drops for unexpected reasons to elow the next lower pressure level, or if significant modification or repairs are made to the tainment following the test.

anticipated deflections of the containment and equipment hatch are given on Figure 3.8-58.

anticipated deflections are calculated by taking into account the interaction of the liner, forcing and concrete including the effects of concrete cracking.

imit of 1.3 times the anticipated deflections were used for comparison to the measured ections. This is based on a comparison of predicted and allowable stresses in the membrane p reinforcing at the 52 psi test pressure level. The stress in the hoop reinforcing is predicted to 5,600 psi at the 52 psi test pressure. When increased by 30 percent, the stress is 33,000 psi or of yield, which is the allowable reinforcing stress for this loading condition.

cking was expected in the portion of the containment structure shell away from the mat.

rtical cracks at approximately 18 inches on center, and horizontal cracks at the construction ts 6 foot on center and at 2 foot to 3 foot intervals between the construction joints.) This was ed on observations of previous tests.

final test report, Report on Structural Acceptance Test of the Concrete Primary Containment lstone Nuclear Power Station Unit 3, was prepared covering the test performed on July 12, 1985. This report contains the information outlined in Regulatory Guide 1.18, Regulatory ition 13. The report was compiled by Stone and Webster and presented to NUSCO in tember 1985.

liner inservice surveillance requirements and testing, including leak rate testing, see tion 6.2.6.

3.8-29 Rev. 30

s section, as outlined in the NRC Regulatory Guide 1.70, Rev 3, Standard Format and Content Safety Analysis Reports for Nuclear Power Plants, regarding steel containment, is not licable to the Millstone 3 Containment Structure. A steel lined reinforced concrete tainment is being used as described in Section 3.8.1. The equipment hatch, personnel hatch, portions of the penetration sleeves that are not backed by concrete, and normally considered e steel containment, are also described in Section 3.8.1.

3 CONCRETE AND STRUCTURAL STEEL INTERNAL STRUCTURES OF STEEL OR CONCRETE CONTAINMENTS 3.1 Description of Internal Structures containment structure interior arrangement (Figures 3.8-59 and 3.8-60) consists of heavily forced concrete walls and slabs which are designed to support the principal nuclear steam ply equipment. The interior concrete also provides interior biological shielding and protection m missiles resulting from postulated component failure (Section 3.5).

reactor vessel is supported on the reactor vessel support/shield tank located within, and rally supported by, the concrete primary shield wall.

refueling cavity, located above the reactor vessel, and the fuel transfer canal are stainless steel d concrete structures.

pressurizer and each reactor coolant loop, including their steam generator, reactor coolant ps, and associated valves and piping are enclosed within separate concrete cubicles.

40-ton capacity overhead polar crane is supported by the polar crane wall. The crane is vided with earthquake restraints to preclude its dislodgment from the rails during the safe tdown earthquake (SSE).

n steam and steam generator feedwater lines, electrical cable trays and conduits, HVAC ducts, other miscellaneous pipes are supported by structural steel hangers and supports located on interior structure walls and slabs.

inimum clearance, in accordance with the following, is provided at all levels between the rior structural steel work, platforms, pipe supports, conduit supports, and the containment r and its attachments.

Elevation Minimum Clearance Above EL 57 feet 0 inches 1.05 inches.

Between EL 13 feet 0 inches to 57 feet 0 1 inch inches (inclusive) 3.8-30 Rev. 30

Below EL 13 feet 0 inches 0.75 inches new or unique structural features are used in the design of the internal structure of the tainment.

SS component supports are described in Section 3.9.3.

3.2 Applicable Codes, Standards, and Specifications es, standards, specifications, and NRC regulations and Regulatory Guides used in blishing design methods and material properties for concrete and steel internal structures are en in the following sections:

Codes, Specifications, Design Methods, Section 3.8.1.2 and Material Properties Section 3.8.1.6 NRC Regulatory Guides 1.10, 1.15, NRC Regulatory Guide, as qualified in 1.55, 1.60, and 1.61 Section 1.8 tion 3.8.3.6 describes materials and quality control procedures used for containment structure rior.

3.3 Loads and Loading Combinations rior concrete structures and structural steel within the containment are designed to withstand pressure buildup resulting from the loss of coolant accident (LOCA) discussed in tion 6.2.1. The blowdown of a postulated rupture of a main coolant pipe is assumed to be in one of the steam generator cubicles or adjacent to the reactor vessel. Because the volume of h of these cubicles is less than the entire containment structure, initial differential pressures t between the interior and exterior of the cubicle until full pressurization of the containment is ined. Structural components, walls, floors, and beams enclosing these cubicles are designed to hstand these differential pressures.

e rupture may also cause blowdown jet and reactive forces. The magnitude of a blowdown jet ing function resulting from a pipe rupture is dependent upon the geometry and distance of the et from its source. Critical structures and components are protected from, or designed to resist, effects of these forces. Section 3.6 describes the protection provided against the dynamic cts associated with a postulated rupture of piping.

interior concrete structures protect the containment shell from internal missiles generated by ccident. Safety features are physically protected from potential sources of missiles either by sical separation, barriers, or by providing restraints on the potential missiles. Section 3.5 cribes internal missile generation and the design of barriers to resist these hazards.

3.8-31 Rev. 30

h structure as applicable.

3.4 Design and Analysis Procedures interior structures generally comprise a series of frames, box type structures, and assemblies labs. Structural analyses are based on elastic behavior using commonly accepted principles of ineering mechanics appropriate to the geometry of the structure.

erial quality control procedures, as described in Section 3.8.3.6, ensured that material strength uirements were achieved. Over strength of materials is not a factor because the strength hod of design, as described in Section 3.8.3.5, is used in proportioning and reinforcing crete sections.

amount of reinforcing steel required is determined in accordance with the procedures ined in ACI 318 and the principal reinforcement patterns are located in the direction of tensile sses. Bond and anchorage requirements of ACI 318 are complied with, and where biaxial ile fields exist, the development lengths required by Section 12.5 of ACI 318 are increased by inimum of 25 percent.

ctural steel is designed in accordance with the procedures outlined in the AISC Specification the Design, Fabrication, and Erection of Structural Steel for Buildings, the Structural Welding e, AWS D1.1, and the loading combinations given in Section 3.8.3.3.

heat generated in the primary concrete shield wall due to gamma rays and neutrons nating from the reactor core and due to gamma rays arising from neutron interactions in the ary shield wall is negligible and has no effect on the strength of the structure.

mputer programs that have been used in the design and analysis of structural elements are tified in Appendix 3A.

3.5 Structural Acceptance Criteria ign of interior concrete structures follows ACI 318, using the required strength section based he strength design method. The basic criterion for concrete strength design is expressed as:

Required Strength Calculated Strength members and all sections of members are proportioned to meet this criterion. The required ngth is expressed in terms of design loads, or their related internal moments and forces.

ign loads are defined as loads which are multiplied by their appropriate load factors.

culated strength is computed according to the provisions of ACI 318, including the appropriate acity reduction factors. Capacity reduction factors are the same as those given in Section 9.2 CI 318.

3.8-32 Rev. 30

hods of Part 2 of AISC.

tion 3.7.2.8 describes the variations incorporated into the seismic analysis structural model to ount for a cracked and an uncracked containment structure shell, for variations and ertainties in subgrade shear modulus and spring constants, for virtual mass embedment, and contact pressure distribution. Design of the internal structures is based upon the most servative values resulting from these variations in assumptions and design parameters.

tion 3.7.3.6 discusses differential seismic movement relating to interconnected components, ems, and equipment.

ign for horizontal shear forces in the plane of internal structure walls is in accordance with the uirements of Section 11.6 of ACI 318. Section 11.16 incorporates the combined effects of ar and tensile stresses into the nominal permissible shear stress, Vc, allowed to be carried by concrete.

3.6 Materials, Quality Control, and Special Construction Techniques erial and quality controls used for the internal structures are as described in Sections 3.8.1.2 3.8.1.6. The 60 day compressive strength of concrete is specified as 5,000 psi, except for the crete ballast slab covering the floor liner, which is 3,000 psi concrete.

ctural steel material, erection, and fabrication tolerances are in accordance with the AISC cification for the Design, Fabrication, and Erection of Structural Steel for Buildings bruary 12, 1969), Supplement No. 1 (November 1, 1970), Supplement No. 2 (December 8, 1), and Supplement No. 3 (June 12, 1974). In general, steel used for structural framing forms to ASTM A 36, Specification of Structural Steel. In areas where the design indicates a higher strength steel is required, the steel conforms to the requirements of ASTM A 440, TM A 441, or ASTM A 588 as required.

tified copies of mill test reports showing actual chemical and physical properties are furnished each heat of steel used in making Seismic Category I structural steel.

ding of structural steel is in accordance with AWS D 1.1-72 and AWS D1.1-Rev 1-73 except undercut on structural welds is limited to 0.01 inch deep when its direction is transverse to the ary tensile stress in the part that is undercut only for the following cases:

1. Where cyclic fatigue is a design parameter and/or
2. Where thin material is used (i.e., 0.5 inch thick and less).

s requirement was implemented through notation of specific inspection criteria on the ineers drawings. In all other situations, undercut is limited to a maximum depth of 1/32 inch.

3.8-33 Rev. 30

material, installation, and inspection of high strength bolts conform to the requirements of the earch Council on Riveted and Bolted Structural Joint Specification using ASTM A 325 or A bolts.

iation damage to concrete is insignificant for conditions of neutron fluence at least to 1019 nvt and temperatures at least to 120 C (Clark 1958). The neutron fluence and the perature levels in the primary concrete shield wall are both lower than these levels; therefore, tructural damage due to radiation or temperature effects occurs.

3.7 Testing and Inservice Surveillance Requirements testing or inservice surveillance of the reactor containment interior structure is planned.

4 OTHER SEISMIC CATEGORY I STRUCTURES (AND MAJOR NONSAFETY RELATED STRUCTURES) 4.1 Description of the Structures Seismic Category I structures are indicated on the plot plan (Figure 1.2-2). The arrangement wings for major Seismic Category I structures other than the reactor containment are:

Title Figure Containment Enclosure Building 3.8-61 (EA-42)

Auxiliary Building 3.8-62 (EM-6)

Fuel Building 3.8-63 (EM-7)

Control Building 3.8-64 (EE-27E)

Cable Tunnel 3.8-65 (EC-7A)

Emergency Generator Enclosure and Diesel 3.8-66 (EM-13)

Fuel Oil Tank Vault Engineered Safety Features Building 3.8-67 (EM-2)

Main Steam Valve Building 3.8-68 (EM-2)

Circulating and Service Water Pumphouse 3.8-69 (EM-8)

Hydrogen Recombiner Building 3.8-70 (EM-2)

Circulating Water Discharge Tunnel and 3.8-71 (EC-16C)

Discharge Structure (EC-17A)

Railroad Canopy (EC-38W,S,Y,Z) 3.8-34 Rev. 30

Millstone Stack Elevations and Sections 3.8-83 Millstone Stack - Mathematical Model 3.8-84 refueling water storage tank and the demineralized water storage tank are shown on the plot (Figure 1.2-2).

angement drawings for major Non-Category I structures are:

Title Figure Service Building 3.8-72 (EM-5)

Turbine Building 3.8-73 (EM-3)

Waste Disposal Building 3.8-74 (EM-9)

Warehouse 5 and Millstone 2 Condensate 3.8-75 (EM-50)

Polishing Facility Auxiliary Boiler and Condensate Polishing 3.8-76 (EM-38)

Building Miscellaneous Yard Tankage 1.2-2 Plot Plan nt grade is approximately elevation 24 feet 0 inch.

Containment Enclosure Building (Seismic Category I) containment enclosure building, a cylindrical steel framed structure with uninsulated metal ng and builtup roofing over insulated metal roof deck, envelops the containment structure ve grade. It has a diameter of 156 feet, a height above grade of approximately 166 feet, and is ported on the containment structure.

containment structure below grade is surrounded by ribbed fiberglass sheets, which extend m the top of the containment mat to above grade. This arrangement provides vertical tilation channels which vent directly into the containment enclosure building or into the tting buildings. The corrugated fiberglass sheet is covered by a 2 inch thick layer of pressible material, a 40 mil thick waterproofing sheet membrane, and then one more layer of nch thick compressible material. The waterproofing membrane encloses the containment structure and extends beneath the mat and above plant grade. The enclosure building design rporates horizontal and vertical sliding joints to ensure that the integrity of the containment losure building is maintained during maximum possible pressure transients of the containment cture under DBA conditions. It assures the proper performance of the supplementary leak ection and release system, as described in Section 6.5.1.

provide the required degree of air tightness and to maintain a partial vacuum within the tainment enclosure building, the metal siding, metal deck side joints, and end laps have two 3.8-35 Rev. 30

tainment enclosure building wall is sealed in a similar manner to the subgrade waterproofing mbrane.

structural framing, but not the siding or roofing, is designed to remain intact under tornado ing. In addition, the design of the building is based on the structural loads being transferred ugh the structural steel bracing and into the containment structure. The building is considered e round and moderately smooth.

Auxiliary Building (Seismic Category I) auxiliary building is located west of the fuel building, east of the service building, and north he containment structure. The auxiliary building, approximately 102 feet by 177 feet in plan, is ported on a reinforced concrete mat founded on rock. Rock dowels are installed along the rior walls of the structure to resist uplifting during seismic loading. Section 2.5.4 describes e rock dowels. The basement floor is approximately 20 feet below grade. There is one floor at de and two above. The concrete roof is located approximately 69 feet above grade. The structure and superstructure are reinforced concrete construction. Interior shield walls are forced concrete.

motor control center (MCC) and rod control area is located in the southern end of the iliary building. This area is open on the side adjacent to the containment structure for direct ess to electrical penetrations.

Fuel Building (Partially Seismic Category I) fuel building is located east of the auxiliary building, south of the waste disposal building, north of the containment and engineered safety features structures. The building, roximately 112 feet by 92.5 feet, is supported on compacted fill and/or rock.

Seismic Category I portion of the building is as follows:

1. Spent Fuel Pool
2. Spent Fuel Shipping Cask Area
3. Auxiliary-Emergency Safety Features Building Pipe Tunnel
4. Canopy over Fuel Cask Shipping Area
5. Demineralizer Area
6. Spent Fuel Shipping Cask Washdown Area
7. Spent Fuel Pool Cooler Cubicle 3.8-36 Rev. 30
1. New Fuel Handling Area
2. New Fuel Storage Vault Area
3. Equipment Decontamination Area
4. Equipment Storage Area
5. Heat Tracing Room se non-seismic areas are designed to withstand seismic loading so as to prevent their collapse o adjacent Category I areas. The fuel building has a ground floor at grade elevation and a ement 13 feet below grade. The spent fuel pool portion of the building is reinforced concrete struction from approximately 24 feet below to 28 feet above grade. The spent fuel areas are ected from tornado missiles by a reinforced concrete superstructure. The new fuel handling equipment decontamination areas of the building have reinforced concrete walls with a steel ed roof and metal deck covered with 4-ply asphalt and gravel roofing.

spent fuel pool is L shaped, with the bottom of the pool approximately 13 feet below grade.

floor is 8 foot thick reinforced concrete. The spent fuel shipping cask storage area is 14 feet e by 30 feet long. The bottom of this area is stepped from approximately 20 feet below grade foot above grade to 28 feet above grade to limit the lifting height of the spent fuel shipping

k. The walls of the spent fuel pool and the spent fuel shipping cask storage area are 6 foot thick forced concrete with the exception of the east wall of the spent fuel shipping cask storage area ch is 5 feet thick and extends to approximately 28 feet above grade. Concrete dividing walls mit dewatering the spent fuel shipping cask storage area and the fuel transfer canal without atering the entire pool. The interior walls and floor of both the new fuel storage vault and nt fuel pool, the spent fuel shipping cask storage area, and the fuel transfer canal are lined with inch thick stainless steel plate.

new fuel storage vault is 24 feet 0 inch long by 15 feet 9 inches wide by 18 feet 4 inches deep h the bottom approximately 9 feet 6 inches above grade.

Control Building (Seismic Category I) control building, approximately 101 feet by 116 feet, is located north of the turbine building, th of the emergency diesel generator building, east of the technical support center and office ding, and west of the service building; it is supported by a reinforced concrete mat foundation.

ept for the east portion of the top level described below, all exterior walls, floor slabs, and roof s are reinforced concrete with interior framing and columns of structural steel.

basement level approximately 20 feet below grade houses emergency (essential) switchgear, ery chargers, battery rooms, inverters, and the emergency shutdown panel. The second floor is cable spreading room. The third floor is the control room level which contains the control 3.8-37 Rev. 30

ve grade.

east portion of the top level is nonsafety related; however, the steel superstructure is designed ithstand tornado and seismic loads.

Cable Tunnel (Auxiliary Building to Control Building) (Seismic Category I) electrical cable tunnel measures approximately 24 feet by 29 feet in section and runs through basement level of the service building for its full width. The tunnel is contiguous with and vides support for portions of the service building. The tunnel is reinforced concrete struction (Figure 3.8-65).

Emergency Generator Enclosure and Fuel Oil Tank Vault (Seismic Category I) emergency generator enclosure, located north of the control building, is a reinforced concrete losure approximately 65 feet by 72 feet. The roof is located approximately 41 ft above plant de. Its foundation consists of reinforced concrete footings placed upon glacial till. The cture is founded approximately 15 feet below grade. Each combination of emergency erator and diesel engine has its own foundation founded on compacted fill. The roof and walls reinforced concrete, a minimum of 2 feet thick. A reinforced concrete wall separates the two rgency generator units. Removable sections of the east and west walls in front of the diesel s provide for replacement of equipment.

fuel oil tank vault, approximately 32 feet by 65 feet by 23 feet high, is located east of the n structure. The vault is below grade to provide protection from tornado missiles. A 1 foot-6 thick fire wall separates the two tanks.

Engineered Safety Features Building (Seismic Category I) engineered safety features building wraps around the east side of the containment structure.

structure is founded on rock at elevation 0 feet 6 inch and is approximately 140 feet long by feet wide. Four pump shafts extend down to the containment mat for the containment rculation pumps. The structure has three floor levels and extends 32 feet above grade. The re building, including the pump shafts, is reinforced concrete construction.

Main Steam Valve Building (Seismic Category I) main steam valve building, located west of and directly adjacent to the containment structure ects the main steam valves and piping from tornado missiles. The building consists of a forced concrete structure with a 2 foot thick wall and roof supported on the rock. The structure nds from approximately 16 feet below grade to approximately 63 feet above grade.

3.8-38 Rev. 30

circulating and service water pumphouse is located on the shoreline of Niantic Bay west of lstone 3. Approximately 128 feet by 86 feet, it houses the circulating water pumps and the ice water pumps. It is constructed of reinforced concrete founded on bedrock approximately feet below mean low water. The service water pump room and its supporting elements are mic Category I, i.e., designed to withstand tornado and earthquake loads, and are protected inst flooding to 25.5 feet above mean low water.

etaining wall is located on the west side of the circulating and service water pumphouse and is ategory I counterfort type reinforced, concrete retaining wall. The wall is founded on bedrock is an extension of the west wall of the circulating and service water pumphouse. The function he west retaining wall is to protect the Category I service water and electrical lines located ind the wall and to be part of the shoreline protection. For a discussion of the shoreline ection, refer to FSAR Section 2.5.5. The seawall located to the east of the circulating and ice water pumphouse is not a Category I structure and therefore, is not discussed in this ion.

Hydrogen Recombiner Building (Seismic Category I) hydrogen recombiner building is located adjacent to the containment structure, on the theast side, directly below the equipment hatch. The building, approximately 56 feet by 50 feet 7 feet high, houses the hydrogen recombiner equipment and provides access to the equipment h and support for the removable hatch missile shield. It is constructed of reinforced concrete, protection of safety related equipment, and founded on fill concrete.

Circulating Water Discharge Tunnel (Seismic Category I) circulating water discharge tunnel is a reinforced concrete structure founded entirely below de on rock, concrete fill, or till. The portion of the circulating water discharge tunnel nstream of the service water discharge point is designed as a Seismic Category I structure.

Railroad Canopy (Seismic Category I) railroad canopy is located to the east of the fuel building.

canopy structure is approximately 75 feet long by 26 feet wide with buttresses extending out dditional 20 feet. It has a mat which is about 77 feet long by 54 feet wide and 8 feet thick and ounded on concrete fill. The 50.6 and 51.2 line walls and roof are reinforced concrete 2 feet

k. The buttresses, L-line wall and west wall are all reinforced concrete, 3 feet thick.

top of the mat is at grade. A railroad spur enters the east side of the building at ground level.

entire canopy structure is designed for seismic and tornado loads.

building protects the spent fuel pool from tornado missiles.

3.8-39 Rev. 30

cription unlined, free standing, tapered, reinforced concrete stack has the following dimensions:

Overall height above foundation 3856 Height above adjacent grade 375 Inside diameter at top 7 Outside diameter at base 276 Thickness at top 7 stack configuration is shown on Figure 3.1-7. The Millstone stack was originally designed as mic Class I.

ximum stack stresses occur in the first 208 feet above grade, due to load combinations which ude maximum wind.

stack is designed for a maximum exhaust air temperature of 150°F and a minimum ambient perature of 0°F. Design and construction of the stack is in accordance with applicable uirements of ACI Standard Specifications for the Design and Construction of Reinforced crete Chimneys (ACI-505) as follows:

d Pressures normal design conditions, it was designed at specific ACI allowable unit stresses. Wind sures on projected areas for various height zones above ground conformed to the following uirements for a basic wind pressure area of 30 psf.

Height Pressure Design Wind Pressure 0-49 ft 32 psf 50-99 ft 39 psf 100-199 ft 45 psf 200-299 ft 48 psf 300-375 ft 51 psf ulated wind pressures include the reduction for the circular shape of the stack. The tabular e for pressure of ACI-505 was increased for a gust velocity of 140 mph in accordance with agraph 400 of ACI-505.

3.8-40 Rev. 30

k is 154,000 kip feet.

thquake Loading eveloping a mathematical model, for dynamic analysis, the stack was treated as a flexible tilever system with the base fixed at the top of the foundation. (The stack base is neither rigid can it rotate, but is considered rigid as this assumption results in a shorter period for the stack.

s analytical representation gives higher, more conservative response acceleration.) Forty mass nts were considered to be supported my weightless elastic columns. The model is depicted by ure 3.1-8. Subsequent to the formation of mass and stiffness matrices for the cantilever system, periods and mode shapes were calculated, displacement and inertia force time histories were blished and a time history of shears, moments, displacements and accelerations determined.

top of the stack is at elevation 389 feet or 238 feet above the top of the Reactor Building. The k is located 416 feet east of the Reactor Building east wall. Thus, the top of the stack could not ke the operating floor of the Reactor Building, even if toppled intact about its base, because the k height is 385 feet from the base. The ventilation exhaust is brought through breeching which ers the various exhaust ducts together. The stack is provided with a one foot by four foot ess opening at the base, three galvanized steel balconies, an outside ladder for the full stack ht, aviation obstruction lighting, lightning protection and handling of the isokinetic sampler.

Service Building (Partially Seismic Category I) service building is located between the control and the auxiliary buildings. Approximately feet by 80 feet, it is founded on bedrock and is of steel frame construction with metal siding builtup roofing on insulated metal deck. This building consists of one level below grade, one l at grade, and two levels above grade. The roof is located approximately 43 feet above plant de. The below-grade level houses nonsafety related switchgear and the Seismic Category I trical tunnel. The grade level houses offices, change rooms, radiation protection facilities,

-aid room, and other service facilities. The first floor above grade houses the lunch room, rument repair room, and locker area, while the second above-grade level houses mechanical ipment.

structural framing, but not the siding and roofing, is designed to withstand tornado winds and mic forces.

Turbine Building (Nonsafety Related) turbine building, approximately 325 feet by 115 feet, is located west of the containment cture and is supported on spread footings on basal till and compacted select granular fill. The ine building has a basement level 10 feet below grade and a roof 107 feet above grade. The ndation walls are reinforced concrete to grade. The superstructure is steel framed with metal ng and builtup roofing on an insulated metal deck. There is an auxiliary bay of the same 3.8-41 Rev. 30

structural framing, but not the siding and roofing, is designed to withstand tornado winds and mic forces.

Waste Disposal Building (Nonsafety Related) liquid waste disposal building is located directly north of the fuel building and east of the iliary building. It is a reinforced concrete structure approximately 48 feet by 114 feet with a l framed HVAC penthouse. The building is founded on bedrock and basil till with a basement l 20 feet below grade. The penthouse roof is approximately 73 feet above grade.

solid waste disposal building is located directly north of the liquid waste building. The ding is approximately 38 feet by 114 feet and is founded at grade on soil backfill. The erstructure consists of a 24 foot high reinforced concrete shell and a steel framed enclosure h builtup roofing on insulated metal deck approximately 42 feet above grade.

design equals or exceeds the requirements of Regulatory Guide 1.143.

Warehouse Number 5 and Millstone 2 Condensate Polishing Facility (Nonsafety Related) warehouse structure, approximately 98 feet by 211 feet, is located north of the Millstone 2 ine building and south of the Millstone 3 condensate polishing facility and auxiliary boiler ding. The structure consists of three main levels and a penthouse. The condensate polishing lity is located approximately 20 feet below grade with portions of the waste handling ipment extending to 4 feet and 26 feet above grade. Warehouse storage and Millstone 3 afiltration equipment are located 4 feet above grade. Records storage is located 26 feet above de. The penthouse, 50 feet by 46 feet, is located 40 feet above grade in the middle of the cture near the west wall. This enclosure houses the elevator machine room and building tilation equipment.

superstructure is steel framed with the main roof approximately 40 feet above grade and the thouse roof approximately 58 feet above grade.

warehouse structural framing, but not the siding and roofing, is designed to withstand tornado ds and seismic loadings.

Auxiliary Boiler and Condensate Polishing Facility (Nonsafety Related) auxiliary boiler room is located south of the turbine building, east of the condensate polishing lity, and north of the warehouse. The area is approximately 66 feet by 58 feet in plan and is ported on reinforced concrete footings with a floor 0 feet 6 inches above grade. A roof with lation and 4-ply asphalt and gravel is provided at approximately 36 feet above grade.

3.8-42 Rev. 30

cture, approximately 58 feet by 64 feet, designed for radiation protection. The structure is ported by spread footings. The basement floor is approximately 10 feet below grade with the ond level approximately 14 feet above grade.

1. Yard Structures Vacuum Priming Pumphouse (Nonsafety Related)

The vacuum priming pumphouse is a reinforced concrete structure located on top of the outfall structure. The area is approximately 40 feet by 35 feet with a floor 0 feet 6 inches above grade. A roof with insulation and 4-ply asphalt and gravel is provided at approximately 17 feet above grade.

2. Miscellaneous Yard Tankage Boron recovery tanks, primary grade water tanks, demineralized water storage tank, refueling water storage tank, boron and waste test tanks, condensate storage tank, condensate surge tank and water treatment storage tanks are located on concrete pads with oil sand cushion 0 feet 6 inches above grade. The demineralized water storage tank is protected by 2 feet 0 inch thick reinforced concrete walls and roof. The boron recovery tanks are enclosed in a concrete and steel structure.
3. Electrical/Conduit Manholes Electrical manholes are reinforced concrete structures constructed below grade with access through manhole covers at grade.

other nonsafety related structures are located such that their failure does not damage safety ted systems, structures, or components.

4.2 Applicable Codes, Standards, and Specifications es, standards, specifications, and NRC regulatory guides used in establishing design methods material properties for Seismic Category I concrete and steel structures other than the tainment are given in Section 3.8.1.2.

4.3 Loads and Loading Combinations Seismic Category I structures other than the containment structure mat, shell, and dome are gned for the loads and load combinations in Table 3.8-3. Section 3.8.4.5 describes allowable ss levels.

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gn. The historical design of the spent fuel pool walls and mat considered the thermal effects ed on the temperatures indicated in Figures 3.8-79 and 3.8-80. The analysis for classifying a core off load as a normal evolution evaluated the thermal effects based on temperatures cated in Figure 3.8-82. The spent fuel pool walls and mat were also investigated for the sed thermal transient effects due to the storage of higher enrichment fuels as shown in ure 3.8-81. Utilizing the loads and load combinations in Table 3.8-3, the allowable stress ls described in Section 3.8.4.5 were satisfied.

4.4 Design and Analysis Procedures eneral, design and analysis procedures conform to the requirements of the ACI 318-71 Code the AISC specification, 7th Edition, for the Design, Fabrication and Erection of Structural l for Buildings except as noted in Section 3.8.4.3. Structural analyses are based on elastic avior using commonly accepted principles of engineering mechanics appropriate to the metry of the structure.

boundary conditions assumed for structural elements under design are based on the stiffness he elements into which these elements are framed.

erial quality control procedures, as referred to in Section 3.8.4.6, ensure that minimum ngth material requirements are achieved.

mic accelerations on the structures are determined by dynamic analysis as described in tion 3.7.3.4. Forces are determined and then applied statically in the design of the structures.

analytical techniques used to determine the forces are given in Section 3.7.2.

hake space, consistent with building displacements, is provided, above grade, between all ependent Seismic Category I and nonseismic structures so as to prevent their interaction ng a seismic event.

nado loads (described in Section 3.3.2) include wind force loads and the loads from tornado erated missiles. Section 3.5.3 gives tornado missile impact effects.

mputer programs which have been used in the design and analysis of structural elements are tified in Appendix 3A.

4.5 Structural Acceptance Criteria mic Category I structures, as identified in Table 3.2-1, are designed to withstand the loading binations given in Table 3.8-3.

crete structures are designed by the strength design method of ACI 318-71.

basic criterion for strength design is expressed as:

3.8-44 Rev. 30

concrete members and all sections of concrete members are proportioned to meet this erion. The required strength is expressed in terms of design loads, or their related internal ments and forces. Design loads are defined as loads that have been multiplied by their ropriate load factors. Calculated strength is that computed by the provisions of ACI 318, uding the appropriate capacity reduction factors. Capacity reduction factors are taken as given ection 9.2 of ACI 318. Allowable stresses and strains for concrete structures are within the ts specified in Section 9.2 of ACI 318.

l structures, except as noted in this section, are designed in the elastic range to maintain actual sses less than allowable stress given in Part 1 of the AISC Specification for the Design, rication and Erection of Structural Steel for Buildings and also the Structural Welding Code, S D1.1. For members requiring tornado or seismic (SSE) design, allowable stresses are as ows:

Tension Ft = 0.90Fy Shear Fv = 0.60Fy Compression Members owable stresses are 1.6 times the values given by Section 1.5.1.3 (p 5-64) of AISC cification, 7th Edition, for the Design, Fabrication, and Erection of Structural Steel for ldings.

Bending sion and compression for compact, adequately braced members symmetrical about, and ed in, the plane of their minor axis.

Fb = 0.90Fy owable stresses for other members are 1.6 times those values given in AISC Section 1.5.1.4, in no case greater than 0.90 Fy.

Connections Welds - 1.6 times values given in AISC Section 1.5.3, Table 1.5.3 High Strength Bolts - 1.6 times values given in AISC Section 1.5.2, Table 1.5.2.1 3.8-45 Rev. 30

l design.

majority of the Millstone 3 structures have a relatively small height-to-width ratio; therefore, mn drift due to wind is not a problem. The only building with significant height, the turbine ding, has been checked for wind deflections. These deflections are limited such that overall e stability is maintained and deflections under service loads are within common practice for industry.

4.6 Materials, Operating Control, and Special Construction Techniques tions 3.8.1.2 and 3.8.1.6 describe material and quality control. There are no special techniques d in constructing the structures (Section 3.8.4.1).

60 day compressive strength of concrete for the spent fuel pool and the fuel building is cified as 5,000 psi.

4.7 Testing and Inservice Surveillance Requirements re are no special testing or inservice surveillance requirements for Category I structures ide the containment.

4.8 Masonry Walls sonry walls in safety related areas in the plant comply with the requirements in Appendix A to P Section 3.8.4. Locations of walls are given in Figure 3.8-64, Sheet 1 and 3.

5 FOUNDATIONS 5.1 Description of the Foundations ndations for all of the major structures consist of soil or rock supported reinforced concrete s or spread footings as described in Table 2.5.4-14. Figures 2.5.4-1 through 2.5.4-17 show and section views of the major foundations.

provide for independent movement of structures during a seismic event, a minimum of 1 inch k compressible material is provided below grade between all structures. The containment is arated, below grade between all structures, by a minimum 4 inch shake space filled with a pressible material.

izontal shear keys are provided for the control, fuel, and auxiliary buildings and for the ulating and service water pumphouse foundations. Figure 3.8-77 shows the arrangement of e horizontal shear keys and Figure 3.8-78 shows a typical detail. Rock dowels are used in the iliary building foundation (Section 3.8.5.5).

3.8-46 Rev. 30

k dowels are used in the exterior walls of the auxiliary building to resist uplift during seismic ing. Section 2.5.4 describes the rock dowels and the installation program.

containment enclosure building is supported entirely on the containment structure and has no ndations. Figure 3.8-61 shows typical details of the interface between the enclosure and und/adjacent structures.

nificant amounts of groundwater are not expected. Figure 2.5.4-37 shows the design levels for undwater. No Seismic Category I dewatering system is required. However, the following ures have been incorporated to prevent seepage of groundwater into portions of structures w the piezometric surface:

1. All structures, except the containment, have waterstops installed at construction joints below grade.
2. The containment substructure is encased with a waterproof membrane to elevation 25 feet 0 inches or to the bottom or approximate midpoint at slabs abutting the containment structure below elevation 25 feet 0 inches. Such slabs are provided with waterstops or the membrane is continued as an encasement for the abutting structure to preclude seepage at the interface of the slab and the containment wall.

A drainage system is provided on the containment/Engineered Safety Features Building side of the membrane and connects to a sump located in the Engineered Safety Features Building to remove groundwater which leaks through the membrane. A non safety-related pump is used as necessary to remove the water during normal plant operation, post-LOCA and LNP conditions (see 9.3.3.2.4.1 for additional information).

3. The service, control, auxiliary, and Engineered Safety Features Building substructures are encased with a waterproof membrane to elevation 23 feet 6 inches and have drainage systems located under the mat of each building. These run into sumps for collection and then discharge. The coefficient of friction between the membrane and the concrete is equal to or greater than that between the concrete below the membrane and the soil or rock. Sliding stability is therefore not affected by the presence of the membrane.
4. The Technical Support Center, fuel, and waste disposal buildings are provided with perimeter and substructure drains.

5.2 Applicable Codes, Standards, and Specifications tion 3.8.3.2 contains the codes, standards, specifications, and NRC regulatory guides used in blishing design methods and material properties for foundations and concrete supports.

3.8-47 Rev. 30

ndation design is based upon appropriate loading combinations. The loads and loading binations given in Section 3.8.1.3 are used for the containment foundation design. The loads loading combinations given in Table 3.8-3 are used for the design of all other Seismic egory I foundations.

ddition to the above loads and load combinations, the following were used to check against ing and overturning due to earthquakes, winds, tornadoes, and the design basis flood:

1. D + H + OBE
2. D+H+W
3. D + H + SSE
4. D + H + Wt
5. D+F re:

D, OBE, W, SSE, Wt are as defined in Table 3.8-3 H = The lateral earth pressure F = The buoyant and lateral force effects of the design basis flood tion 3.8.5.5 gives stability factors for these conditions.

ere Seismic Category I structures extend below the surface of the finished ground grade, their rnal walls are designed for seismic lateral earth pressure and groundwater effects in addition tatic lateral earth pressures due to soil loads, surcharge loads applied at ground surface, and ral and buoyant force effects from groundwater or flood and loads of adjacent footings or s.

5.4 Design and Analysis Procedures tion 3.8.1.4 describes the design and analysis of the reactor containment foundation mat.

tion 3.8.4.4 describes design and analysis of Seismic Category I foundations other than the tor containment.

tion 3.7.2.14 describes determination of seismic overturning moments.

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ctural design of the reactor containment foundation is in accordance with ACI 318, using the criteria given in Section 3.8.1.3. Structural design of all Seismic Category I foundations other the reactor containment foundation is in accordance with ACI 318, using strength design and load criteria given in Section 3.8.4.3.

basic criterion for strength design is expressed as:

Required Strength Calculated Strength members and all sections of members are proportioned to meet this criterion. The required ngth is expressed in terms of design loads or their related internal moments and forces. Design s are defined as loads that are multiplied by their appropriate load factors. The calculated ngth for all Seismic Category I foundations is that computed by the provisions of ACI 318, uding the appropriate capacity reduction factors. Capacity reduction factors are as given in tion 9.2 of ACI 318.

ing and overturning factors of stability are:

MINIMUM FACTORS OF SAFETY LOADING CONDITION (SECTION 3.8.5.3) OVERTURNING SLIDING FLOTATION

1. Operating Basis Earthquake 1.5 1.5 -
2. Normal Wind 1.5 1.5 -
3. Safe Shutdown Earthquake 1.1 1.1 -
4. Tornado 1.1 1.1 -
5. Design Basis Flood - - 1.1 differential settlement of the reactor containment is anticipated.

5.6 Materials, Quality Control, and Special Construction Techniques ous concrete is used to provide subsurface drainage under and around the containment cture. This type of concrete is formed by the omission of the fine aggregate from a standard crete mix. The mix is designed to have a minimum 28 day compressive strength of 1,000 psi.

osity tests of porous concrete performed at Northeastern University in September 1962, used nch diameter by 12 inch long cylinders. These cylinders were prepared in the laboratory by pacting the material in three layers with a standard tamping rod. Results indicated water osities of from 28 to 47 gpm/sq ft, depending upon the amount of compaction and resulting sity of the cylinders.

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licable conditions.

tion 3.8.4.6 describes the materials and quality control used for Seismic Category I ndations and supports. No special construction techniques are used for Seismic Category I ndations.

5.7 Testing and Inservice Surveillance Requirements entire reactor containment undergoes structural acceptance testing, as described in tion 3.8.1.7. Except for this test, no other testing or inservice surveillance of foundation ems is planned.

re are no special testing or inservice surveillance requirements for Seismic Category I ctures outside the containment.

6 REFERENCES FOR SECTION 3.8 1 Clark, R.G. 1958. Radiation Damage to Concrete, US AEC Report H.W.-56195, Hanford Atomic Products Operation, March 31, 1945.

2 Holley, M.J., Jr. 1969. Provision of Required Seismic Resistance. MIT, Cambridge, Mass.

3 TRW, Inc. 1975. Embedment Properties Headed Studs. TRW Nelson Division, Lorain, Ohio.

4 Zhemochkin, B.N. and Sinitzin, A.P. 1962. Practical Methods for Analysis of Beams and Plates on Elastic Foundations. In Russian, Gosstroiizdar, Moscow.

3.8-50 Rev. 30

Design Allowables (per ASME III Category Load Conditions Nomenclature) ergency D+PD+TD+SSE Pm+Pb+Q < 3Sm st D+1.15P Pm < 0.9Sy Pm+Pb < 1.35Sy

+ CAT curve considerations rmal 100 cycles of P NB-3222.4 (d) or (e) 400 cycles of T 100 cycles of 1/2-SSE vere D+Pmin +Tmin + 1/2-SSE Pm < Sm Without erational temperature Pm+Pb < 1.5Sm Pm+Pb+Q < 3Sm ergency D+Pmin+TD + SSE SC = 0.014 ANCHORS ergency D+PD+TD+SSE Maximum shear < 0.425 Su vere D+Pmin+Tmin+ 1/2-SSE tensile < 0.45 Su erational ergency D+Pmin+TD+SSE Shear a = 0.5 u Tension a = 0.5 u TE:

The normal and test load combinations are producing negligible effects.

ere:

D = Dead load effect of reinforced concrete structure acting on the liner plus dead load of the liner PD = Design pressure (pressure resulting from design basis accident and safety margin) 3.8-51 Rev. 30

TD = Load due to thermal expansion, resulting when the liner is exposed to the design temperature SSE = Stresses in the liner derived from applying the effect of the safe shutdown earthquake P = Differential pressure between operating pressure and atmospheric pressure (100 cycles are assumed on the basis of 2.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> refueling cycles per year on a 40 year span)

T = Load due to thermal expansion, resulting when the liner is exposed to the differential temperature between operating and seasonal refueling temperatures (400 cycles are assumed on the basis of 10 such variations per year, on a 40 year span (100 cycles of 1/2-SSE is an assumed number of cycles for this type of earthquake)

Pmin = Minimum pressure resulting during operation of the containment Tmin = Load due to thermal expansion resulting when the liner is exposed to the minimum pressure Sy = Yield strength of the material Sm = The smaller of 1/3-ultimate strength or 2/3-yield strength Su = Ultimate strength of the stud material a = Allowable displacement for liner anchors, inches u = Ultimate displacement capacity for liner anchors, inches SC = Allowable liner plate compressive strain, inches/inches 3.8-52 Rev. 30

Areas of Analysis See Figure Stress Allowables per ASME III 3.8-15) Category Load Combinations Nomenclature Design Mp or Tp or Jax or Jsh Pm < 0.9 Sy Pm + Pb< 0.9 Sy P concrete bearing < 2,400 psi Emergency Pd + Td + Ro Pm + Pb+ Q < 3 Sm Design(1) Mp or Tp or Jax or Jsh Pm < 0.9 Sy

  • Pm + Pb < 0.9 Sy
  • Design (2) Pg + Tg + Design (1) (P1 + Pb) <1.5 Sm (Pm+ Pb + Q) < 3 Sm Normal Pg + Tg + Re ASME III Table NC-3611.1(b) (3)-1 or Para. NB-3222.4(d) or (e) or the pipe portion, refer to Section 3.7B.3.1 ere:

Mp = Yielding moment = Required bending moment to produce stresses equal to the yield strength of the pipe material Tp = Yielding torque = Required torsional moment to produce stresses equal to the yield strength of the material Jax = Axial jet force = Load equal to the piping design pressure times the inside area of the pipe, acting in the axial direction of the piping Jsh = Shear jet force = Load equal to the piping design pressure times the inside area of the pipe acting transversely to the pipe Pd = Containment design pressure Td = Containment design temperature Pg = Piping design pressure Tg = Piping design temperature Ro = Piping reactions due to normal operation (including SSE effects)

Re = Piping reactions due to normal operation (including 1/2 - SSE)

Design (1) - Applies to the sizing of the sleeve and attachment plate Design (2) - Applies to the evaluation of stresses in the area of Analysis 2 due to the given load combinations 3.8-53 Rev. 30

Concrete Structures (Containment Internal Structures and Category I Structures, other than the containment mat, shell, and dome).

ads and loading combinations are based on ACI 318, and AEC Enclosure 3 - Structural sign Criteria for Evaluating Effects of High Energy Pipe Breaks on Category I Structures tside the Containment, Structural Engineering Branch, Directorate of Licensing.

1. U = 1.4D + 1.7L
2. U = 1.4D + 1.7L + 1.7H 2a. U = 0.9D + 1.7H
3. U = 1.4D + 1.7L + 1.4F 3a. U = 0.9D + 1.4F
4. U = 0.75 (1.4D + 1.7L + 1.7W) 4a. U = 0.90D + 1.3W
5. U = 0.75 (1.4D + 1.7L + 1.7 x 1.1 (1/2 SSE) 5a. U = 0.90D + 1.3 x 1.1 (1/2 SSE)
6. U = 1.1 D + 1.1 L + 1.1 SSE 6a. U = 0.9D + 1.1 SSE
7. U = 1.1 D + 1.1 L + 1.0 Wt 7a. U = 0.9D + 1.0 Wt
8. U = D + L + Ta + Ra + 1.5Pa
9. U = D + L + Ta + Ra + 1.25Pa + 1.25 OBE + 1.0 (Yr + Yj + Ym)
10. U = D + L + Ta + Ra + Pa + SSE + 1.0 (Yr + Yj + Ym) tes - Concrete Structures (1) U is the required section strength based on strength design methods described in ACI 318-71.

(2) In combinations 8, 9, and 10, the maximum values of Pa, Ta, Ra, Yj, Yr, and Ym, including an appropriate dynamic load factor, shall be used unless a time-history analysis is performed to justify otherwise.

(3) For load combinations 9 and 10, local section strengths and stresses may be exceeded under the concentrated loads Yr, Yj, and Ym, provided there will be no loss of function of any safety related system.

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(4) For load combinations 7 and 7a, local section strengths and stresses may be exceeded under the tornado missile load provided there will be no loss of function of any safety related system.

Steel Structures A. Elastic Working Stress Design Service Load Conditions

1. 1.0 S = D + L 1a. 1.5 S = D + L + To + Ro (If To and Ro = 0 use Equation 1)
2. 1.0 S = D + L + OBE 2a. 1.5 S = D + L + To + Ro + OBE (If To and Ro = 0 use Equation 2)
3. 1.33 S = D + L + W 3a. 1.5 S = D + L + W + To + Ro (If To and Ro = 0 use Equation 3)

Factored Load Conditions

4. 1.6 S = D + L + To + Ro + SSE
5. 1.6 S = D + L + To + Ro + Wt
6. 1.6 S = D + L + Ta + Ra + Pa
7. 1.6 S = D + L + Ta + Ra + Pa + 1.0 (Yr+Yj+Ym) + OBE
8. 1.7 S = D + L + Ta + Ra + Pa + 1.0 (Yr+Yj+Ym) + SSE B. Plastic Design Factored Load Conditions
9. 0.9Y = D + L + Ta + Ra + 1.5 Pa 9a. 1.0Y = D + L + Ta + Ra + 1.5 Pa
10. 0.9Y = D + L + Ta + Ra + 1.25 Pa + 1.25 OBE + 1.0 (Yr+Yj+Ym) 10a. 1.0Y = D + L + Ta + Ra + 1.25 Pa + 1.25 OBE + 1.0 (Yr+Yj+Ym) 3.8-55 Rev. 30
11. 0.9Y = D + L + Ta + Ra + Pa + SSE + 1.0 (Yr+Yj+Ym) 11a. 1.0Y = D + L + Ta + Ra + Pa + SSE + 1.0 (Yr+Yj+Ym) tes - Steel Structures (1) S is the required section strength based on the elastic design methods and allowable stresses defined in Part 1 of the AISC, Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings.

(2) Y is the section strength required to resist design loads based on plastic design methods described in Part 2 of the AISC, Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings.

(3) Both cases of L having its full value or being completely absent are checked for load combinations 1, 1a, 2, 2a, 3, 3a, 4, and 5.

(4) In combinations 4 to 8 and 9 to 11, thermal loads are neglected when it can be shown that they are secondary and self-limiting in nature, or where the material is ductile.

(5) In combinations 6, 7, 8, 9, 10, and 11, the maximum values of Pa, Ta, Ra, Yr, Yj, and Ym, including an appropriate dynamic factor, are used unless a time-history analysis is performed to justify otherwise.

(6) Combination 5 shall be satisfied without the tornado missile load. Combinations 7, 8, 10 and 11 shall be first satisfied without Yr, Yj, and Ym. When considering these loads, however, local section strengths may be exceeded under the effect of these concentrated loads, provided there will be no loss of function of any safety related system. Furthermore, in computing the required section strength, S, the plastic section modulus of steel shapes may be used for combinations 7 and 8.

(7) Combinations 1a, 2a, 3a, 9a, 10a, and 11a were added for use in design evaluations since September 1985. These combinations are consistent with NUREG-0800, Section 3.8.4, (Revision 1, July 1981).

ads, Definition of Terms, and Nomenclature

1. Normal Loads - Those loads encountered during normal plant operations and shutdown. They include the following:

D - Dead loads or their related internal moments and forces including any permanent equipment loads L - Live loads or their related internal moments and forces, including any movable equipment loads and other loads which vary with intensity and occurrence F - Lateral and vertical pressure of liquids, or their related internal moments and forces; F is included in D for steel structures.

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H - Lateral earth pressure, or its related internal moments and forces; H is included in L for steel structures To - Thermal loads during normal operating or shutdown conditions, based on the most critical transient or steady-state condition; To is included in D for concrete structures equations 1, 2, 2a, 3, 3a, 4, 4a, 5, 5a, 6, 6a, 7, and 7a.

Ro - Pipe loads during operating or shutdown conditions, based on the most critical transient or steady-state condition; Ro is included in D for concrete structures equations 1, 2, 2a, 3, 3a, 4, 4a, 5, 5a, 6, 6a, 7, and 7a.

2. Severe Environmental Loads - Those loads that could infrequently be encountered during the plant life. They include:

OBE- Loads generated by the operating basis earthquake W - Loads generated by the design wind specified for the plant site (Section 3.3.1)

3. Extreme Environmental Loads - Those loads which are credible but highly improbable. They include:

SSE- Loads generated by the safe shutdown earthquake Wt - Loads generated by the design tornado specified for the plant site (Section 3.3.2)

4. Abnormal Loads. Those loads generated by a postulated high-energy pipe break accident within a building and/or compartment thereof. Included in this category are the following:

Pa = Maximum differential pressure load generated by a postulated break Ta = Thermal loads under accident conditions generated by a postulated break Ra = Pipe and equipment reactions under accident conditions generated by a postulated break Yr = Loads on the structure generated by the reaction on the broken high-energy pipe during a postulated break Yj = Jet impingement load on a structure generated by a postulated break Ym = Missile impact load on a structure generated by or during a postulated break, such as pipe whipping 3.8-57 Rev. 30

RECIRCULATION, AND SAFETY INJECTION PIPING Plant Operating Condition NC 3600 Equations Load Combinations(1) Allowable Stress Design 8 Pd + D Sh Normal/Upset 9 Pp + D + E + H 1.2 Sh 10 (2) T+R+A SA 10a S 3 Sc 11 (2) Pd + D + T + R + A SA + Sh Emergency 9 Pp + D + H + E' + Y 1.8 Sh Faulted 9 Pp + D + H + E' + Y' + A1 2.4 Sh (3)

Pp + D + B' 2.4 Sh (3)

Test 8 Pt + D Sc NOTES:

1. Refer to Table 3.8-7 for definition of loadings.
2. Either the requirements of Equation (10) or (11) must be satisfied.
3. In pipe break exclusion zones, the allowable stress is 1.8 Sh.
4. In break exclusion area, sum of stress given by Equations (9) and (10) should not exceed 0.8 (SA + 1.2 Sh).

3.8-58 Rev

RECIRCULATION SPRAY, AND SAFETY INJECTION SYSTEMS Plant Operating Condition NC 3600 Equations Load Combinations (1) Allowable Stress Design 8 Pd + D Sh Normal/Upset 9 Pp + D + E + H 1.2 Sh 10 (2) T+R+A SA 10a S 3 Sc 11 (2) Pd + D + T + R + A SA + Sh Emergency 9 Pp + D + H + E' + Y 1.8 Sh Faulted 9 Pp + D + H + E' + Y' + A1 1.8 Sh Pp + D + B' 1.8 Sh 10 (2) T + R' + A' + X SA 11 (2) Pd + D + T + R' + A' + X SA + Sh Test 8 Pt + D Sc NOTES:

1. Refer to Table 3.8-7 for definition of loadings.
2. Either the requirements of Equation (10) or (11) must be satisfied.
3. In break exclusion area, sum of stress given by Equations (9) and (10) should not exceed 0.8 (SA + 1.2 Sh).

3.8-59 Rev

Plant Design or Operating Condition ASME Code Reference Load Combinations(1) Allowable Stress(2)

Design Primary Stress Intensity Fig. NE-3221-1 Pd + D + E + H Pm < Sm. PL < 1.5Sm and PL + Pb < 1.5Sm Normal/Upset Primary and Secondary Stress Range NB-3222.2 PO + T + R + A + E + H + L PL + Pb + Pe + Q < 3Sm Peak Stress Range NB-3222.4e PO + T + R + A + E + H + L PL + Pb + Pe + Q + F (3)

Simplified Plastic-Elastic Analysis NB-3228.3(a) PO + D + E + H + L PL + Pb + Q < 3Sm (4)

Expansion Stress Intensity NB-3222.3 T+R Pe < 3Sm Emergency Fig. NB-3224-1 PL + D + E' + H + Y (5)

Faulted Appendix F PL + D + E' + H + Y' + A1 (6)

Appendix F PL + D + E' + H + B' (6)

Appendix F PL + D + E' + X + T + R' + A' (6) (7))

NOTES:

1. Refer to Table 3.8-7 for definition of loadings.
2. Class MC allowables from Table I-1.1 of ASME Code per NE 3131(d).
3. The allowable cumulative usage factor is 1.0.
4. Q excludes the thermal bending stress.
5. Pm <1.2Sm or Sy , PL <1.8Sm or 1.5Sy and PL + Pb <1.8Sm or 1.55y (use the largest value).
6. Pm <Sm , PL <1.5Sm and PL + Pb <1.5Sm 3.8-60 Rev

Inelastic - Elastic Analysis - Use an Sm value equal to the greater of 0.7Su or [Sy + 1/3 (Su

- Sy)](8).

Includes the steady-state load from B' and the stress induced in the sleeve from the liner.

Use 85 percent of these values if Y' or B' exists in the load combination.

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- Sustained mechanical loads, including deadweight of piping, components, contents, and insulation.

- Loads due to thermal expansion of the system in response to average fluid temperature.

- Loads induced in the piping due to the thermal growth of equipment and/or structures to which the piping is connected as a result of plant normal or upset plant conditions.

- Loads induced in the piping due to thermal growth of equipment and/or structures to which the piping is connected as a result of plant faulted plant conditions.

Note that R' includes R.

- Inertia effects of the OBE.

- Inertia effects of the SSE.

- Loads induced in the piping due to inertia effects of LOCA or displacements due to LOCA or displacements due to pipe rupture.

- Loads induced in the piping due to response of the connected equipment and/or civil structures to the OBE (commonly referred to as OBE anchor movements).

- Loads induced in the piping due to response of the connected equipment and/or civil structures to the SSE (commonly referred to as SSE movements).

- Loads induced due to building settlement effects.

- Loads resulting from occasional loads other than seismic. Examples of these loads would be: water hammer, steam hammer, opening and closing of safety relief valves, etc, as defined for the emergency plant condition.

- Effects of pipe striking pipe (pipe whip) or effects of blowdown of an adjacent system (jet impingement loads), as defined for the emergency plant condition.

- Effects of pipe striking pipe (pipe whip) or effects of blowdown of an adjacent system (jet impingement loads), as defined for the faulted plant condition.

- Loads induced in the piping due to pressure response (growth) of the containment during a faulted plant condition.

- Local stress effects in piping and/or piping components due to sudden changes in fluid temperature. These loads are commonly referred to as thermal transient effects.

- Loads on restraints induced by blowdown and subsequent pipe response of a ruptured system for faulted plant conditions.

- Internal pressure loads due to design pressure.

- Internal pressure loads due to peak pressure.

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- Internal pressure loads due to test pressure.

- Internal pressure loads for emergency and faulted conditions as applicable.

- Internal pressure loads due to range of the operating pressure.

3.8-63 Rev. 30

tions whose identification numbers include the letter B contain material within the nce-of-plant (BOP) scope, while sections whose identification numbers include the letter N tain material within the nuclear steam supply system (NSSS) scope.

1 Rev. 16

B.1.1 Design Transients design of the reactor coolant system (RCS), RCS component supports, and reactor internals siders the following five operating conditions, defined in Section III of the ASME Boiler and ssure Vessel Code.

1. Normal Conditions - Any condition in the course of startup, operation in the design power range, hot standby and system shutdown, other than upset, emergency, faulted or testing conditions.
2. Upset Conditions (Incidents of Moderate Frequency) - Any deviations from normal conditions anticipated to occur often enough that design should include a capability to withstand the conditions without operational impairment. The upset conditions include:
a. Transients which result from any single operator error or control malfunction.
b. Transients caused by a fault in a system component requiring its isolation from the system.
c. Transients due to loss of load or power.
d. Abnormal incidents not resulting in a forced outage.
e. Forced outages for which the corrective action does not include any repair of mechanical damage. The estimated duration of an upset condition is included in the design specifications for each component.
3. Emergency Conditions (Infrequent Incidents) - Those deviations from normal conditions which require shutdown for correction of the conditions or repair of damage in the system. The conditions have a low probability of occurrence but are included to provide assurance that no gross loss of structural integrity results as a concomitant effect of any damage developed in the system. The total number of postulated occurrences for such events does not cause more than 25 stress cycles having an S value greater than that for 106 cycles from the applicable fatigue design curves of the ASME Code Section III.
4. Faulted Conditions (Limiting Faults) - Those combinations of conditions associated with extremely low probability postulated events whose consequences are such that the integrity and operability of the plant may be impaired to the extent that consideration of public health and safety are involved. Such considerations require compliance with safety criteria as may be specified by jurisdictional authorities.

3.9B-2 Rev. 30

classified under normal, upset, emergency or faulted conditions.

provide the necessary high degree of integrity for the components in the RCS the transient ditions selected for component fatigue evaluation are based upon a conservative estimate of magnitude and frequency of the temperature and pressure transients resulting from various rating conditions in the plant. The transients provided are representative of operating ditions which prudently should be considered to occur during plant operation and are iciently severe or frequent to be of possible significance to component cyclic behavior. The sients analyzed may be regarded as a conservative representation of transients which, used as sis for component fatigue evaluation, provide confidence that the component is appropriate its application in accordance with the requirements of ASME Section III.

MNPS-3 documents that provide detailed descriptions of design transients used for fatigue luation of RCP and associated Class 1 branch piping are listed in Table 3.9B-1. The applicable gn transients for RPV internals are described in Section 3.9N.1.

B.1.2 Computer Programs Used in Analysis s of computer programs that are used in the design of Seismic Category I components and ng systems within the BOP scope are provided in Appendix 3A. Also included in endix 3A are brief descriptions of each program, the extent of its application, and program fications which demonstrate the applicability and validity of each program.

B.1.3 Experimental Stress Analysis experimental stress analysis methods are employed. Analytical methods for the design of BOP ipment, components, and piping systems are used exclusively.

B.1.4 Consideration for the Evaluation of the Faulted Conditions B.1.4.1 Loading Conditions structural stress analysis performed on the reactor coolant loop piping and other Seismic egory I ASME Code Class 1, 2, and 3 piping consider the loadings and load combinations cified in Tables 3.9B-10, 3.9B-11, and 3.9B-12. The stress limits utilized for the faulted plant dition are also outlined in Tables 3.9B-10 and 3.9B-11.

loading conditions and associated stress limits defined in the above tables are applicable for tic analysis only. Inelastic analysis is not generally used in the qualification of piping and ng components. Where inelastic analysis is utilized in qualifying the piping and/or ponents, the MNPS-3 documents containing details of such analysis are identified in le 3.9B-2. The procedure used for the modeling and analytical methods for evaluating the tor coolant loop and supports for faulted loading conditions are described in tion 3.9B.1.4.2. The procedure used for modeling of remaining Seismic Category I ASME 3.9B-3 Rev. 30

sections NB, NC, ND (1971 through Summer 1973 addenda) and Appendix F.

B.1.4.2 Evaluation of Reactor Coolant Loop and Supports for Faulted Loading Condition faulted loading condition of reactor coolant loop and supports considers loading due to:

1. Internal pressure
2. Weight
3. Safe shutdown earthquake
4. Loss-of-coolant accident (pipe break)
5. Transients Internal Pressure rnal pressure is identified as faulted condition pressure in accordance with ASME Section III uirements.

Weight ght consists of the weight of the piping system, insulation (if any), and contained fluid during mal operating conditions.

Seismic earthquake loads are part of the mechanical loading conditions specified in the design cification for piping systems. Mechanical stresses are evaluated for SSE (safe shutdown hquake) condition consistent with ASME Section III requirements for faulted condition ing evaluation.

Loss of Coolant Accident chanical loads are developed in the broken and unbroken reactor coolant loops and in the tor vessel as a result of transient flow and pressure fluctuations following a postulated pipe k in one of the reactor coolant loops. Structural consideration of dynamic effects of postulated breaks requires postulation of a finite number of break locations. Postulated pipe break tions are given in Section 3.6.

e history dynamic analysis is performed for these postulated break cases. Hydraulic models used to generate time-dependent hydraulic forcing functions used in the analysis of the reactor 3.9B-4 Rev. 30

Transients mmetric pressurization, resulting from the postulated pipe breaks, of the reactor cavity, steam erator, or pressurizer cubicles also imposes unbalanced mechanical loads on the RCS piping equipment. Pressure time histories are developed based on release rates of energy and fluid s from the break, which are compatible with the hydraulic analysis cited above. The hanical response to asymmetric pressure loading is calculated separately and then combined h the response to hydraulic forces.

itional localized loadings due to direct impingement of a fluid jet from the postulated pipe k are also combined. Where the jet target is a support, the effect is assumed to be limited to support and its immediate connection joints. Where the target is a major RCS component, the orce in excess of asymmetric pressure loadings is included in the RCS dynamic analysis ction 6.2.1).

B.1.4.3 Reactor Coolant Loop Models and Methods analytical methods used in obtaining the solution consist of the transfer matrix method and ness matrix formulation for the static structural analysis, the response spectra method or time ory method for seismic dynamic analysis, and time history direct integration method for the

-of-coolant accident dynamic analysis.

integrated reactor coolant loop/supports system model is the basic system model used to pute loadings on components, component supports, and piping. The system model includes stiffness and mass characteristics of the reactor coolant loop piping and components, the ness of supports, the stiffness of auxiliary line piping which affects the system and the ness of piping restraints. The deflection solution of the entire system is obtained for the ous loading cases from which the internal member forces and piping stresses are calculated.

Static reactor coolant loop/supports system model, constructed for the NUPIPE-SW computer gram, is represented by an ordered set of data which numerically describes the physical em. Figure 3.9B-1 shows an isometric line schematic of this mathematical model for one of reactor coolant loops. The steam generator and reactor coolant pump vertical and lateral port members are described in Section 5.4.14.

spatial geometric description of the reactor coolant loop model is based upon the reactor lant loop piping layout and equipment drawings. The node point coordinates and incremental ths of the members are determined from these drawings. Geometrical properties of the piping elbows along with the modulus of elasticity E, the coefficient of thermal expansion, the rage temperature change from ambient temperature T, and the weight per unit length are cified for each element. The primary equipment supports are represented by stiffness matrices 3.9B-5 Rev. 30

hematical model. The thermal growth of the reactor vessel is considered in the construction of model.

model is made up of a number of sections, each having an overall transfer relationship med from its group of elements. The linear elastic properties of the section are used to define stiffness matrix for the section. Using the transfer relationship for a section, the loads required uppress all deflections at the ends of the section arising from the thermal and boundary forces the section are obtained. These loads are incorporated into the overall load vector.

er all the sections have been defined in this manner, the overall stiffness matrix and associated vector to suppress the deflection of all the network points is determined. By inverting the ness matrix, the flexibility matrix is determined. The flexibility matrix is multiplied by the ative of the load vector to determine the network point deflections due to the thermal and ndary force effects. Using the general transfer relationship, the deflections and internal forces then determined at all node points in the system.

static solutions for deadweight and thermal loading conditions are obtained by using the PIPE-SW computer program.

Seismic model used in the static analysis is modified for the dynamic analysis by including the mass racteristics of the piping and equipment. The effect of the equipment motion on the reactor lant loop/supports system is obtained by modeling the mass and the stiffness characteristics of equipment in the overall system model.

steam generator is typically represented by five discrete masses. The lower mass is located r the intersection of the centerlines of the inlet and outlet nozzles of the steam generator. A dle mass is located near the center of the steam generator and a top mass is located at the et of the steam generator. The other two intermediate masses are located at the preheater ion and steam separator section, respectively.

reactor coolant pump is typically represented by a two-discrete-mass model. The lower mass cated at the intersection of the centerlines of the pump suction and discharge nozzles. The er mass is located near the center of gravity of the motor.

component upper and lower lateral supports are passive during plant heatup, cooldown, and mal plant operating conditions. However, these restraints become active under the rapid ions of the reactor coolant loop components that occur from the dynamic loadings.

mponent supports are represented by stiffness matrices and/or individual tension or pression spring members in the dynamic model.

solution for the seismic disturbance employs the response spectra or the time history method.

h methods employ the lumped mass technique, linear elastic properties, and the principle of 3.9B-6 Rev. 30

m the mathematical description of the system, the overall stiffness matrix [K] is developed m the individual element stiffness matrices using the transfer matrix method. After deleting the s and columns representing rigid restraints, the stiffness matrix is revised to obtain a reduced ness matrix [KR] associated with mass degrees of freedom only. From the mass matrix and the uced stiffness matrix, the natural frequencies and the normal modes are determined.

he response spectra method, the modal participation factor matrix is computed and combined h the appropriate response spectra value to give the modal amplitude for each mode. The total dal amplitude is obtained by taking the square root of the sum of the squares of the tributions for each direction. The modal amplitudes are then converted to displacements in the bal coordinate system and applied to the corresponding mass point. From these data the forces, ments, deflections, rotations, support reactions and piping stresses are calculated for all ificant modes. The total seismic response is computed by combining the contributions of the ificant modes by using the methods described in Section 3.7.

the time history method, the seismic model described above is expanded in scope to include our reactor coolant loops, the reactor vessel and its support system, and the containment-ding- model described in Section 3.8.1 with the exception of the base-mat and soil springs.

resulting model is coded for the STARDYNE (Appendix 3A2.5) program system.

ure 3.9B-2 shows an isometric sketch of a typical loop of the STARDYNE model. The center ion model is shown schematically in Figure 3.9B-3. The plan view of the latter figure cates the points of attachment of the four individual loop submodels to the reactor vessel. In ition, all embedment points of steam generator and pump supports are connected to the ding model at the mass point of nearest elevation. This connectivity is shown in the computer s of the complete STARDYNE model (Figure 3.9B-4).

r to extraction of the natural frequencies and mode shapes, a mass condensation is performed educe the number of dynamic degrees-of-freedom (D-O-F) to 350. The other approximately 0 D-O-F are recovered by eigenvector expansion.

Loss-of-Coolant Accident mathematical model used in the time history seismic analyses is modified for the

-of-coolant accident analyses. To represent the severance of the reactor coolant loop piping at stulated guillotine break location, two distinct nodes, each containing six dynamic degrees of dom and located on each side of the break, are included in the mathematical model.

en no nonlinear elements are active, i.e., for a longitudinal split or severance of the SI, RHR, MS, or FW lines, the dynamic structural solution for the loss-of-coolant accident is obtained sing the STARDYNE subprogram DYNRE, as in the time-history seismic analysis. The ral frequencies and eigenvectors are determined from each broken model with the RDYNE subprogram STAR/HQR.

3.9B-7 Rev. 30

hematically by the combination of a gap, a spring, and a viscous damper. The force in each linear element is treated as an externally applied force in an iterative solution technique which verges when the force and displacement are compatible. Multiple nonlinear elements can be lied at the same node, if necessary, or at several nodes. The time-history solution is performed he STARDYNE subprogram DYNRE. The input to this subprogram consists of the system ses and stiffnesses, applied forces, and nonlinear elements.

transient applied forces are described in Section 3.6B.2.

imulate the release of the strain energy in the pipe for a postulated guillotine break, the rnal forces in the system at the break location due to the initial steady state hydraulic forces, mal forces, and weight forces, are determined by a STAR static analysis. The release of the in energy is accounted for by applying the negative of these internal forces as a step function ing concurrent with the hydraulic forces. The initial conditions are equal to zero because the tion is for the transient problem (the dynamic response of the system from the static ilibrium position). To be consistent with this analysis scheme, the hydraulic forces are sted so as to eliminate the initial value which corresponds to the initial static loading dition.

loss of coolant accident displacements of the reactor vessel are applied in time history form nother input to the dynamic analysis of the reactor coolant loop. The loss of coolant accident lysis of the reactor vessel includes all the forces acting on the vessel including internals tions, cavity pressure loads, and loop mechanical loads. The reactor vessel analysis is cribed in Section 3.9.N.1.4.3.

time-history displacement response is used in computing support loads and in performing ss evaluation of the reactor coolant loop piping. The support loads [F] are computed by tiplying the support stiffness matrix [K] and the displacement vector [X] at the support point.

support loads are used in the evaluation of the supports.

time-history displacements of the primary loop piping are used to determine the internal es, deflections, and stresses at each end of the piping elements. For this calculation, the lacements are treated as imposed deflections on the reactor coolant piping. The results of this tion are used in the piping stress evaluation.

part of the Stretch Power Uprate (SPU) evaluation, the LOCA analysis was revised taking into ount the updated pipe break hydraulic forcing functions and LOCA RPV motions. This revised CA analysis was performed using the NUPIPE-SWPC computer program and the leak-before-k (LBB) criteria for primary reactor coolant loop pipe breaks.

B.1.4.4 Primary Component Supports Models and Methods static and dynamic structural analyses employ the matrix method and normal mode theory for solution of lumped-parameter, multimass structural models, or time-history integration 3.9B-8 Rev. 30

luate the individual support member stresses due to the forces imposed upon the supports by loop.

dels for the STARDYNE computer program (Appendix 3A) are constructed for the steam erator lower, steam generator upper lateral, reactor coolant pump lower and pressurizer ports. The reactor vessel support is modeled using the STRUDL computer program. Structure metry, topology and member properties are used in the modeling.

escription of the supports is found in Section 5.4.14. Detailed models are developed using m elements and plate elements, where applicable.

respective computer programs are used with these models to obtain support stiffness matrices member influence coefficients for the steam generator, reactor coolant pump, pressurizer, and tor vessel supports. Unit force along and unit moment about each coordinate axis are applied he models at the equipment vertical centerline joint. Stiffness analyses are performed for each load for each model.

t displacements for applied unit loads are formulated into flexibility matrices. These are rted to obtain support stiffness matrices which are included in the reactor coolant loop model.

ds acting on the supports are obtained from the reactor coolant loop analysis. For each support lyzed, the following is performed:

a. Combine the various types of support plane loads to obtain operating condition loads (Normal, Upset, Emergency or Faulted).
b. Multiply member influence coefficients by operating condition loads to obtain all member internal forces and moments.
c. Solve appropriate stress or interaction equations for the specified operating condition. Maximum normal stress, shear stress, and combined load interaction equation values are printed. ASME Boiler and Pressure Vessel Code Section III, Subsection NF, stress and interaction equations are used and results are compared with limits for the operating condition specified.

reactor vessel support structure is analyzed for all loading conditions using a finite element del. Vertical and horizontal forces delivered to the support structures from the reactor vessel e are applied to the structure, and element stresses and concrete forces are obtained.

B.1.4.5 Equipment and Components elastic analysis techniques described in Section 3.7B.3.1.1 are utilized in the qualification of mic Category I ASME code and noncode equipment within BOP scope. Stress limits utilized the faulted plant condition are as outlined in Section 3.9B.3.1. The design conditions and 3.9B-9 Rev. 30

onstrating maintenance of function and/or structural integrity, will be established prior to lementation.

B.2 DYNAMIC TESTING AND ANALYSIS B.2.1 Preoperational Vibration and Dynamic Effects Testing on Piping reoperational vibration, thermal expansion (in discrete temperature step increments), and amic effects testing program is conducted on:

1. ASME Code Class 1, 2, and 3 piping systems.
2. High energy piping systems inside Seismic Category I structures.
3. High energy portions of systems whose failure could reduce the functioning of any Seismic Category I plant feature to an unacceptable level.
4. Seismic Category I portions of moderate-energy piping systems located outside containment. The purpose of the tests is to confirm that these piping systems, restraints, components, and supports have been designed adequately to withstand the flow-induced dynamic loadings under operational transient and steady state conditions anticipated during service and to confirm that normal thermal motion is not restrained.

st of the systems and the types of tests being conducted is contained in Table 3.9B-3. The erent flow modes of operation and transients to which each system was subjected during the s are contained in Chapter 14. The test titles, test prerequisites, test objectives, and summary esting are also described in Chapter 14. For each system defined in items 1 through 4, all flow des of operation that the systems are subjected to during the tests are visually observed where essible. In addition, systems that were stress analyzed for fluid flow instabilities have rumented measurements at selected locations for the specific flow modes analyzed. The sured results were compared to the analytically predicted values. Instrumented measurements conducted (as needed) for all other systems and conditions. For ASME Code Class 1, 2, and 3 ng systems, design and supervision of the tests, definition of acceptance criteria, evaluations est results, and the making of any changes in the piping system necessary to ensure that the ng is adequately designed and supported, were performed as required by Section III of the ME code.

erved vibrations which from visual examination appeared to be excessive in the opinion of erienced engineers who supervised, conducted, and witnessed the various tests, resulted in er:

3.9B-10 Rev. 30

ASME code stress and fatigue limits to be exceeded.

2. The cause of vibration is eliminated.
3. A corrective support system is designed and installed and the effect of the modification incorporated in the Pipe Stress Analysis.

ddition to the above, thermal expansion effects of piping are observed during the testing.

ations for monitoring thermal expansion are chosen based on:

1. Expected large movements
2. Areas with tight clearances
3. Snubber locations to verify correct and unrestricted motion B.2.2 Seismic Qualification Testing of Safety Related Mechanical Equipment methods and procedures used in the design and qualification of Seismic Category I hanical equipment within balance-of-plant (BOP) scope are outlined in Sections 3.7B.3.1.1, B.3, and 3.10B. Loading combinations include operating as well as earthquake loading for lification by testing and/or analytical methods.

ety related mechanical equipment (Seismic Category I), not covered by the ASME Boiler and ssure Vessel Code, are seismically qualified in accordance with the procedures of tion 3.7B.3.1.1. Non-ASME Code equipment typically include diesel generators, fans, lers, and emergency ventilation equipment. Cranes are seismically qualified in accordance h criteria that preclude the possibility of the crane being dislodged by a seismic disturbance.

ept as noted elsewhere in the FSAR, if codes are used in the design of a component, the delines generally require the addition of operating loads to the operating base earthquake E) (1/2 Safe Shutdown Earthquake (SSE)) load with no increase in code allowable stress. If odes are used, the stress level under the combined loading is limited to 75 percent of the imum yield strength of the material per the ASTM specification. The general criteria for lysis of the SSE, pipe rupture (if applicable), and operating loads require that deformation of ponents be allowed only with no loss of safety function. Stresses under combined loadings generally limited to the smaller of 100 percent of the minimum yield strength, or 70 percent of minimum ultimate tensile strength, of the material (at temperature) per the ASTM, or ivalent specification for the material.

3.9B-11 Rev. 30

B.3.1 Loading Combinations, Design Transients, and Stress Limits B.3.1.1 ASME III Class 1 Components mponents and systems which are Seismic Category I and which are within the jurisdiction of ME Section III, Subsection NB (Class 1), have the following combinations and categorizations oad with relevant stress limits in accordance with NB-3000:

1. Design condition - Design pressure and temperature; design mechanical loads.
2. Normal condition - Normal operating loads (deadweight, pressure, thermal, etc.).
3. Upset condition - Upset plant condition loads plus the operating basis earthquake (OBE).
4. Emergency condition - Emergency plant condition loads.
5. Faulted condition - Dynamic system loads associated with the faulted plant condition plus the safe shutdown earthquake (SSE).
6. Testing condition - Hydrostatic tests, pneumatic tests, leak tests, as applicable.

the conditions specified, the allowable stress limits defined in Tables 3.9B-5 and 3.9B-10 are licable to stress results obtained by elastic analysis techniques and are compared with ulatory Guide 1.48 in Table 3.9B-6. The analysis techniques described in Section 3.7B.3.1 are zed in implementing these criteria. Design transients included in the above category are ned in Section 3.9B.1.

general extent of compliance with Regulatory Guide 1.48 is discussed in Section 1.8.

B.3.1.2 ASME III Class 2 and 3 Components les 3.9B-7 and 3.9B-11 provide loading conditions and stress limits for ASME Code Class 2 3 components of Seismic Category I fluid systems which are constructed in accordance with ME III Subsections NC and ND. The conditions generally relate to ASME Section III, Code ss 1 requirements, and include combinations as follows:

1. Design Condition I - includes the specified design loads (temperature, pressure, etc.) plus the OBE loads.
2. Design Condition II - includes the specified design loads (as above) plus SSE loads, or pipe rupture loads (if applicable).

3.9B-12 Rev. 30

ign Condition II. The stress limits for these design conditions are presented in Tables 3.9B-7 3.9B-11. Since design temperature and pressure exceed those associated with upset, rgency, and faulted conditions, satisfaction of primary stress limits is assured.

se requirements are intended to be consistent with the present code format and philosophy.

en implemented, they were a supplement to the requirements of ASME III, Subsection NC ND.

stress limits and design conditions presented in Table 3.9B-7 are intended to ensure that no ss deformation of the component occurs. Table 3.9B-8 compares Table 3.9B-7 with ulatory Guide 1.48. These limits are applicable for an elastic system (and component) lysis. Inelastic deformation is not performed on any ASME Code Class 2 and 3 Components.

nd when, inelastic analysis is contemplated, detailed design bases demonstrating maintenance ither function and/or structural integrity will be proposed prior to implementation. The lysis techniques of Section 3.7B.3.1 are utilized in implementing these criteria.

general extent of compliance with Regulatory Guide 1.48 is discussed in Section 1.8.

B.3.2 Pump and Valve Operability Assurance ps and valves installed in Seismic Category I piping systems are designed in accordance with requirements of ASME III, NB, NC, and ND. Inactive pumps and valves are designed for the ing combinations of Sections, 3.9B.3.1 and 3.9B.3.2, and for the stress limits indicated in le 3.9B-7.

ive components are those that must perform a mechanical motion during the course of omplishing its safety function.

tive components are those for which mechanical movement does not occur in order for the ponent to accomplish its intended safety function.

rability of active pumps and valves is assured by satisfying the requirements of various grams. Safety related valves are qualified by prototype testing and analysis, and safety related ve pumps by analysis with suitable stress limits and nozzle loads. The content of these grams is detailed below.

B.3.2.1 Pump Operating Program ive pumps are qualified for operability by being subjected to rigid tests both prior to and after allation in the plant. The in-shop tests include:

1. Hydrostatic tests to ASME Section III requirements.
2. Seal leakage tests at the same pressure used in the hydrostatic tests.

3.9B-13 Rev. 30

requirements, and other pump/motor parameters.

o monitored during these operation tests are bearing temperatures and vibration levels, which shown to be below appropriate limits specified to the manufacturer for design of each active p.

er the pump is installed in the plant, it undergoes the cold hydro tests, hot functional tests, and required periodic inservice inspection and operation as applicable. These tests demonstrate ability of the pump for the design life of the plant.

ddition to these tests, the safety related active pumps are qualified for operability during an condition by assuring that:

1. The pump is not damaged during the seismic event.
2. The pump continues operating when subjected to the SSE loads.

pump manufacturer is required to show that the pump operates normally when subjected to maximum applicable amplified seismic (floor) accelerations, attached piping nozzle loads, dynamic system loads associated with the faulted operating condition. Analysis and/or testing cedures are utilized in accordance with those outlined in Section 3.7B.3.1.1. Natural frequency ulations are performed in order to determine maximum seismic accelerations based on licable amplified (floor) response spectra.

avoid damage during the faulted condition, the stresses caused by the combination of normal rating loads, SSE, and dynamic system loads are limited as indicated in Table 3.9B-7. The rage membrane stress (Pm) for the faulted condition loads is maintained at 1.2 S, or roximately 0.75 Sy (Sy = yield stress) and the maximum stress in local fibers (Pm + bending ss Pb) is limited to 1.8 S, or approximately 1.1 Sy. In addition, the pump stresses caused by the imum seismic nozzle loads are limited to the stresses outlined in Table 3.9B-7. The maximum mic nozzle loads are also considered in an analysis of the pump supports to assure that a em misalignment cannot occur. A static shaft deflection analysis of the rotor is performed h horizontal and vertical accelerations based on floor response levels. The deflections rmined from the static shaft analysis are compared to the allowable rotor clearances. The re of seismic disturbances dictates that the maximum contact (if it occurs) is of short duration.

ss 1 pumps are designed/analyzed according to the rules of ASME Section III, Subsection NB 0.

ormance of these analyses with the conservative loads stated and, with the restrictive stress ts of Table 3.9B.5 as allowables, assures that critical parts of the pump are not damaged ng the short duration of the faulted condition and that the reliability of the pump for post-ted condition operation is not impaired by the seismic event.

3.9B-14 Rev. 30

ed under all conditions unless the rotor becomes completely seized (i.e., no rotation).

ically, the rotor can be seized 5 full seconds before a circuit breaker trips to prevent damage to motor. However, the high rotary inertia in the operating pump rotor and the nature of the dom, short duration loading characteristics of the seismic event prevent the rotor from oming seized. In actuality, the seismic loadings cause only a slight increase, if any, in the ue (i.e., motor current) necessary to drive the pump at the constant design speed. Therefore, pump does not shut down during the SSE and operates at the design despite the SSE loads.

complete the seismic qualification procedures, the pump motor is independently qualified for ration during the maximum seismic event. Any auxiliary equipment which is vital to the ration of the pump or pump motor, and which is not qualified for operation during the pump lysis or motor qualifications, is also separately qualified for operation at the accelerations urring at its mounting. The pump motor and vital auxiliary equipment are qualified by meeting requirements of IEEE Std 344-1975. If the testing option is chosen, sinusoidal or sine-beat ing is justified by satisfying one or more of the following requirements to demonstrate that the ti-frequency response is negligible or the input is of sufficient magnitude to conservatively ount for this effect:

1. The equipment response is basically due to one mode.
2. The sinusoidal or sine-beat response spectra envelop the floor response spectra in the region of significant response.
3. The floor response spectra consist of one dominant mode and have a peak at this frequency.

eneral, the degree of coupling in the equipment determines if a single or multi-axis test is uired. Multi-axis testing is required if there is considerable cross coupling. If coupling is very t, then single axis testing is justified. If the degree of coupling can be determined, then single testing can be used with the input sufficiently increased to include the effect of coupling on response of the equipment.

m previous arguments, the safety related pump/motor assemblies are not damaged and tinue operating under SSE loading and, therefore, perform their intended functions. These posed requirements take into account the complex characteristics of the pump and are icient to demonstrate and assure the seismic operability of the active pumps. The functional ity of active pumps after a faulted condition is assured, since only normal operating loads and dy state nozzle loads exist. Since it is demonstrated that the pumps are not damaged during the ted condition, the post-faulted condition operating loads are identical to the normal plant rating loads. This is assured by requiring that the imposed nozzle loads (steady state loads) for mal conditions and post-faulted conditions are limited by the magnitudes of the normal dition nozzle loads. The post-faulted condition ability of the pumps to function under these lied loads is proven during the normal operating plant conditions for active pumps.

3.9B-15 Rev. 30

ety related active valves must perform their mechanical motion during the course of orming their safety function. Assurance that these valves will operate during a seismic event t be supplied. Qualification tests accompanied by analyses are conducted for all active valve mblies.

ves without significant extended structure are proven seismically adequate by analysis of ng seismic adequacy. For valves with operators having significant extended structures, and if e structures are essential to maintaining pressure integrity, analysis is performed based upon ic forces resulting from equivalent earthquake accelerations acting at the centers of gravity of extended masses. For active valves, this requirement for analysis is extended to the hanical (nonpressure boundary) components of valve top-works to ensure operability.

safety related valves are subjected to a series of stringent tests prior to service and during t life. Prior to installation, the following tests are performed:

1. Shell hydrostatic test to ASME III requirements.
2. Backseat and main seat leakage tests.
3. Disc hydrostatic test.
4. Functional tests to verify that the valve will open and close within the specified time limits when subjected to the design differential pressure.
5. Operability qualification of motor operators for the environmental conditions over the installed life (i.e., aging, radiation, accident environmental simulation, etc.)

according to IEEE Std 382-1972.

d hydro qualification tests, hot functional qualification tests, periodic inservice inspections, periodic inservice operation are performed in situ to verify and assure the functional ability of valve. These tests guarantee reliability of the valve for the design life of the plant. The valves designed using either stress analyses or the pressure containing minimum wall thickness uirements. On all active valves, an analysis of the extended structure is also performed for ic equivalent seismic SSE loads supplied at the centers of gravity of the extended structure.

maximum stress limits allowed in these analyses show structural integrity. The limits that are d for Class 2 and 3 active valves are shown in Table 3.9B-7. Class 1 valves are designed/

lyzed according to the rules of ASME Section III, Section NB-3500.

ddition to these tests and analyses, representative valves of each design type are tested for fication of operability during a simulated seismic event by demonstrating operational abilities within the specified limits. The testing procedures are described below.

valve is mounted in a manner which conservatively represents a typical valve installation.

valve includes the operator and all appurtenances normally attached to the valve in service.

3.9B-16 Rev. 30

1. Active valves are designed to have a first natural frequency greater than 33 Hz.

This may be shown by suitable test or analysis.

2. The actuator and yoke of the valve system are statically loaded an amount greater than that determined by an analysis, as representing SSE accelerations applied at the center of gravity of the operator alone in the direction of the weakest axis of the yoke. The design pressure of the valve is simultaneously applied to the valve during the static deflection tests.
3. The valve is then operated while in the deflected position (i.e., from the normal operating mode to the faulted mode). The valve must perform its safety related function within the specified operating time limits.
4. Motor operators and other electrical appurtenances necessary for operation are qualified as operable during an SSE by appropriate IEEE Seismic Qualification Standards, such as IEEE Std 382-1972 and IEEE Std 344-1975, prior to their installation on the valve.

accelerations used for static valve qualification are 3.0 g horizontal and 2.0 g vertical applied h the valve in its proper orientation. Where it is necessary to allow for random orientation of a e, a 3.0 g horizontal and a 3.0 g vertical are applied. The piping designer maintains the motor rator accelerations to these levels with an adequate margin of safety.

e frequency of the valve, by test or analysis, is less than 33 Hz, the valve system is analyzed to rmine the equivalent acceleration applied during the static test. The analysis provides the lification of the input acceleration considering the natural frequency of the valve and the uency content of the applicable plant floor response spectra. The adjusted acceleration is then d in the static analysis and valve operability is assured by the methods outlined in steps (2) to above, using the modified acceleration input.

above testing program applies only to valves with overhanging structures (i.e., the motor rator). The testing is conducted on a representative number of valves. Valves from each of the ary safety related design types (e.g., motor-operated gate valve) are tested. Valve sizes which er the range of sizes in service are qualified by the tests and the results are used to qualify all es within the intermediate range of sizes. Stress and deformation analysis is used to support extrapolation.

ves which are safety related, but can be classified as not having an overhanging structure, such heck valves and safety-relief valves, are considered separately.

ck valves are characteristically simple in design and their operation is not affected by seismic elerations or the maximum applied nozzle loads. The check valve design is compact and there no extended structures or masses whose motion could cause distortions which could restrict ration of the valve. The nozzle loads due to maximum seismic excitation do not affect the 3.9B-17 Rev. 30

ricted due to any casing distortions caused by nozzle loads. Therefore, the design of these es is such that once the structural integrity of the valve is assured using standard design or lysis methods, the ability of the valve to operate is assured by the design features. In addition hese design considerations, the valve also undergoes the following tests and analysis:

1. Stress analysis including the SSE loads.
2. In-shop hydrostatic test.
3. In-shop seat leakage test.
4. Periodic in situ valve exercising and inspection to assure the functional ability of the valve.

ety and relief valves are subjected to tests and analyses similar to check valves; stress and ormation analyses for SSE loads, in-shop hydrostatic and seat leakage tests, and periodic in valve inspection. In addition, a static load equivalent to the SSE is applied to the top of bonnet the pressure is increased until the valve mechanism is activated. Successful actuation within gn requirements assures its overpressurization safety capabilities during a seismic event.

ng the methods described, all the safety related valves in the system are qualified for rability during a seismic event. These methods conservatively simulate the seismic event and ure that the active valves perform their safety related function when necessary.

ernative valve operability testing, such as dynamic vibration testing, is allowed if it is shown to quately assure the faulted condition functional ability of the valve system.

B.3.3 Design and Installation Details for Mounting of Pressure Relief Devices design criteria for all safety and relief valves are in accordance with the rules in Subarticles

-3677 and NC-3677 of the ASME Boiler and Pressure Vessel Code,Section III (see mponent Description for relief valves Letdown Line Downstream of Low Pressure Letdown ves, Sealwater Return Line, and Letdown Reheat Heat Exchanger section 9.3.4.2.5), and the s of Code Case 1569, applicable to the classification of the piping component under stigation. For open relief systems, the design criteria and the analyses used to calculate imum stresses and stress intensities are in accordance with Subarticles NB-3600 and

-3600 of Section III. The maximum stresses are calculated based upon the full discharge loads, uding the effects of the system dynamic response, and the system design internal pressure.

sses are determined for all significant points in the piping system including the safety valve t pipe nozzle and the nozzle to shell juncture.

3.9B-18 Rev. 30

total stead-state discharge thrust load for an open system discharge is expressed as the sum of pressure and momentum forces as follows:

F = 144PA + [AV2]/g (3.9B-1) re:

F = Total reaction force (lb)

A = Exit flow area (sq ft)

P = Exit pressure (psig)

V = Exit fluid velocity (fps)

P = 32.2 fps2 ensure consideration of the effects of the suddenly applied load at the junction of the valve zle and pipe run, a dynamic load factor is computed. The calculation of dynamic load factor is ed on modeling the valve and nozzle as a single degree of freedom dynamic system. The ped mass of this system corresponds to the weight of the valve and nozzle and is assumed to t the valve center of gravity. The rotational degree of freedom of this system is considered to n the direction that causes maximum bending stress in the nozzle at the junction of the nozzle run-pipe. Rotational flexibility of the system is computed by a series combination of nozzle ibility and local run-pipe flexibility (at the junction of the nozzle and run-pipe).

rise time of the discharge force at the outlet of the safety valve elbow is assumed to be the imum valve opening time, and the discharge force is assumed to rise linearly with time and ain thereafter constant value. The ratio of maximum dynamic rotations predicted by this le degree of freedom system to the corresponding static rotation caused by the steady state harge force represents the dynamic load factor.

ensure the consideration of the effects of the suddenly applied loads on the pipe system, a amic time-history analysis is performed on the piping system. The forcing function applied at point of discharge is a linear force change from zero to the value of F that is determined in the ve equation over a time period, t, that corresponds to the valve opening time which is provided he valve manufacturer. After time t has been reached, the force remains at the value of F until conclusion of the time-history integration. The lumped mass model that represents the piping em includes the safety/relief valves.

ere more than one valve is mounted on a common header, two cases are computed. In the first, discharge of all valves is assumed to occur simultaneously. In the second, the forcing ctions are applied to a combination of valves that yields the worst-load case. This worst-load is first verified by trial through a series of static load cases.

3.9B-19 Rev. 30

relief valves discharging into a closed system, an analytical model of one-dimensional sient flow characteristics following the blow-off of the upstream safety/relief valve into the harging piping system is established. The time-dependent pressure, temperature, density, city, and hence the momentum of the downstream pipe flow are then computed from this servative hydrodynamic/thermodynamic flow model. The effects such as flow restrictions and tional resistance are considered.

unbalanced transient hydraulic forcing function acting on the piping system computed from flow model is then used to determine the transient dynamic responses of the piping structural del. Adapting the lumped-mass method incorporated with the modal analysis of piping system, time-history modal responses are computed. Computations of maximum stress intensities for ME Code Class 1 piping or maximum stress levels for ASME Code Class 2 and 3 piping are ed on the dynamic analysis of the system.

B.3.4 Component Supports mponent supports which are Seismic Category I and which are within the jurisdiction of ME Section III, Subsection NF, utilize applicable loading conditions outlined in Sections B.3.1.1 and 3.9B.3.1.2 for ASME Class 1, 2, and 3 components. Stress limits and loading binations to be utilized for component support evaluation are as outlined in Tables 3.9B-9 and B-9A, respectively. The stress limits are used in conjunction with the applicable analytical cedures outlined in Section 3.7B.3 to assure integrity of supports and mounted components.

mponent supports which do not come under the jurisdiction of ASME Section III, Subsection are summarized in Section 5.4.14.

loads considered in the analysis of the supports are normal operating, seismic (OBE and

), and pipe rupture loads. The latter includes the effects of the following:

1. The blowdown forces in the primary loop.
2. Asymmetric pressurization of the reactor cavity and the SG and RCP cubicles.
3. Fluid jet impinging on the supports or components, as applicable.

amic analyses were performed to determine the stresses in the component supports as well as pipe rupture restraints. The results of the analysis are shown in the following tables.

1. Table 3.9B-15 gives the location of the postulated breaks in the primary loop system.
2. Tables 3.9B-16 and 3.9B-17 give the embedment loads of the SG and RCP.
3. Table 3.9B-18 gives the stresses in the component supports of the SG and RCP.

3.9B-20 Rev. 30

5. Table 3.9B-20 gives the stresses in the component support of the pressurizer safety valve support.
6. Table 3.9B-21 gives the stresses in the primary members of the RPV support.
7. Table 3.9B-22 gives the stresses in the primary loop bumper sections.

3.9B-21 Rev. 30

N.1 SPECIAL TOPICS FOR MECHANICAL COMPONENTS N.1.1 Design Transients following five operating conditions, as defined in Section III of the ASME B&PV Code, are sidered in the design of the reactor coolant system (RCS) components and reactor internals.

1. Normal Conditions Any condition in the course of startup, operation in the design power range, hot standby and system shutdown, other than upset, emergency, faulted or testing conditions.
2. Upset Conditions (Incidents of Moderate Frequency)

Any deviations from normal conditions anticipated to occur often enough that design should include a capability to withstand the conditions without operational impairment. The upset conditions include those transients which result from any single operator error or control malfunction, transients caused by a fault in a system component requiring its isolation from the system and transients due to loss of load or power. Upset conditions include any abnormal incidents not resulting in a forced outage and also forced outages for which the corrective action does not include any repair of mechanical damage.

3. Emergency Conditions (Infrequent Incidents)

Those deviations from normal conditions which require shutdown for correction of the conditions or repair of damage in the system. The conditions have a low probability of occurrence but are included to provide assurance that no gross loss of structural integrity results as a concomitant effect of any damage developed in the system. The total number of postulated occurrences for such events does not cause more than 25 stress cycles having an S value greater than that for 106 cycles from the applicable fatigue design curves of the ASME Code Section III.

4. Faulted Conditions (Limiting Faults)

Those combinations of conditions associated with extremely low probability, postulated events whose consequences are such that the integrity and operability of the nuclear energy system may be impaired to the extent that consideration of public health and safety are involved. Such considerations require compliance with safety criteria as may be specified by jurisdictional authorities.

5. Testing Conditions 3.9N-27 Rev. 30

or faulted conditions as appropriate.

provide the necessary high degree of integrity for the equipment in the RCS the transient ditions selected for equipment fatigue evaluation are based upon a conservative estimate of magnitude and frequency of the temperature and pressure transients resulting from various rating conditions in the plant. To a large extent, these transients are based upon engineering gment and experience, and are considered to be of such magnitude and/or frequency as to be ificant in the component design and fatigue evaluation. The transients selected may be rded as a conservative representation of transients which, used as a basis for component gn and evaluation, provide confidence that the component is appropriate for its application r the design life of the plant.

design transients and the number of cycles of each that are used for fatigue evaluations are wn in Table 3.9N-1. In accordance with ASME III, emergency and faulted conditions are not uded in fatigue evaluations.

Normal Conditions following primary system transients are considered normal conditions:

1. Heatup and cooldown at 100°F per hour
2. Unit loading and unloading at 5 percent of full power/per minute
3. Step load increase and decrease of 10 percent of full power
4. Large step load decrease with steam dump
5. Steady state fluctuations
a. Initial
b. Random
6. Feedwater cycling at hot shutdown
7. Not Used.
8. Unit loading and unloading between 0 and 15 percent of full power
9. Boron concentration equalization
10. Refueling 3.9N-28 Rev. 30
12. Reactor coolant pumps startup and shutdown
13. Turbine roll test
14. Primary side leak test
15. Secondary side leak test
16. Tube leakage test
17. Heaters out of service Heatup and Cooldown at 100°F per Hour design heatup and cooldown cases are conservatively represented by continuous operations ormed at a uniform temperature rate of 100°F per hour. (These operations can take place at er rates approaching the minimum of 0°F per hour. The expected normal rates are 50°F per r).

these cases, the heatup occurs from ambient (assumed to be 120°F) to the no-load temperature pressure condition and the cooldown represents the reverse situation. In actual practice, the of temperature change of 100°F per hour will not be attained because of other limitations h as:

a. Material ductility considerations which limit temperature rates of change, as functions of plant pressure and temperature.
b. Slower initial heatup rates when using pump energy only.
c. Interruptions in the heatup and cooldown cycles due to such factors as pressurizer steam bubble formation, rod withdrawal, sampling, water chemistry and gas adjustments.

number of such complete heatup and cooldown operations is specified as 200 each, for the ear plant design life.

Unit Loading and Unloading at 5 Percent of Full Power per Minute unit loading and unloading operations are conservatively represented by continuous and orm ramp power change of 5 percent per minute between 15 percent load and full load. This swing is the maximum possible consistent with operation under automatic reactor control.

reactor temperature will vary with load as prescribed by the reactor control system. The ber of loading and unloading operations is defined as 13,200 during the 60 year design life of plant.

3.9N-29 Rev. 30

+/-10 percent step change in load demand results from disturbances in the electrical network which the plant output is tied. The reactor control system is designed to restore plant ilibrium without reactor trip following a 10 percent step change in turbine load demand ated from nuclear plant equilibrium conditions in the range between 15 percent and 100 ent of full load, the power range for automatic reactor control. In effect, during load change ditions, the reactor control system attempts to match turbine and reactor outputs in such a ner that peak reactor coolant temperature is minimized and reactor coolant temperature is ored to its programmed setpoint at a sufficiently slow rate to prevent excessive pressurizer sure decrease.

owing a step decrease in turbine load, the secondary side steam pressure and temperature ally increase since the decrease in nuclear power lags behind the step decrease in turbine load.

ing the same increment of time, the RCS average temperature and pressurizer pressure also ally increase. Because of the power mismatch between the turbine and reactor and the ease in reactor coolant temperature, the control system automatically inserts the control rods educe core power. With the load decrease, the reactor coolant temperature will ultimately be uced from its peak value to a value below its initial equilibrium value at the inception of the sient. The reactor coolant average temperature setpoint change is made as a function of ine-generator load as determined by first stage turbine pressure measurement. The pressurizer sure will also decrease from its peak pressure value and follow the reactor coolant decreasing perature trend. At some point during the decreasing pressure transient, the saturated water in pressurizer begins to flash which reduces the rate of pressure decrease. Subsequently the surizer heaters come on to restore the plant pressure to its normal value.

owing a step increase in turbine load, the reverse situation occurs, i.e., the secondary side m pressure and temperature initially decrease and the reactor coolant average temperature and sure initially decrease. The control system automatically withdraws the control rods to ease core power. The decreasing pressure transient is reversed by actuation of the pressurizer ters and eventually the system pressure is restored to its normal value. The reactor coolant rage temperature will be raised to a value above its initial equilibrium value at the beginning he transient.

number of each operation is specified at 2,000 times for the 60 year plant design life.

Large Step Load Decrease with Steam Dump s transient applies to a step decrease in turbine load from full power, of such magnitude that resultant rapid increase in reactor coolant average temperature and secondary side steam sure and temperature will automatically initiate a secondary side steam dump that will vent both reactor trip and lifting of steam generator and pressurizer safety valves. This plant is gned to accept a step decrease of 50 percent from full power, and the steam dump system vides the heat sink to accept the difference in allowable unloading rates between the turbine the reactor coolant system.

3.9N-30 Rev. 30

Steady State Fluctuations assumed that reactor coolant pressure and temperature and pressure at any point in the system y around the nominal (steady state) values. For design purposes two cases are considered:

a. Initial Fluctuations - These are due to control rod cycling during the first 20 full-power months of reactor operation. Temperature is assumed to vary +/-3°F and pressure by +/-25 psi, once during each 2 minute period. The total number of occurrences is limited to 1.5 x 105. The fluctuations are assumed to occur consecutively, and not simultaneously with the random fluctuations.
b. Random Fluctuations - Temperature is assumed to vary by +/-0.5°F and pressure by

+/-6 psi, once every 6 minutes. With a 6 minute period, the total number of occurrences during the plant design life does not exceed 3.0 x 106.

Feedwater Cycling at Hot Shutdown se transients can occur when the plant is at no load conditions, during which intermittent ing of 32°F feedwater into the steam generators is assumed. Due to fluctuations arising from mode of operation, the reactor coolant average temperature decreases to a lower value and immediately begins to return to normal no-load temperature. This transient is assumed to ur 2,000 times over the life of the plant.

Not Used Unit Loading and Unloading Between 0 and 15 Percent of Full Power unit loading and unloading cases between the zero and 15 percent power are represented by tinuous and uniform ramp power changes, requiring 30 minutes for loading and 5 minutes for oading. During loading, reactor coolant temperatures are increased from the no load value to normal load program temperatures at the 15 percent power level. The reverse temperature nge occurs during unloading.

r to loading, it is assumed that the plant is at hot shutdown conditions, with 32°F feedwater ling. During the 2-hour period following the beginning of loading, the feedwater temperature eases from 32°F to 300°F due to steam dump and turbine startup heat input to the feedwater.

sequent to unloading, feedwater heating is terminated, steam dump is reduced to residual heat oval requirements, and feedwater temperature decays from 300°F to 32°F.

number of these loading and unloading transients is assumed to be 500 each during the ear plant design life.

3.9N-31 Rev. 30

owing any large change in boron concentration in the RCS, spray is initiated in order to alize concentration between the loops and the pressurizer. This can be done by manually rating the pressurizer backup heaters, thus causing a pressure increase, which will initiate y at a compensated pressurizer pressure of approximately 2,275 psia. The proportional sprays rn the pressure to 2,250 psia and maintain this pressure by matching the heat input from the kup heater until the concentration is equalized. For design purposes, it is assumed that this ration is performed once after each load change in the design load follow cycle. The total ber of load change occurrences is 26,400 over the 60 year design life.

Refueling he end of the plant cooldown, the temperature of the fluid in the RCS is less than 125°F. At time, the vessel head is removed and the refueling canal is filled. This is done by pumping er from the refueling water storage tank, which is outside and conservatively assumed to be at F, into the loops by means of the low head safety injection pumps. The refueling water flows ctly into the reactor vessel via the accumulator connections and cold legs.

s operation is assumed to occur 80 times over the life of the plant.

Reduced Temperature Return to Power reduced temperature return to power operation is designed to improve the spinning reserve abilities of the plant during load following operations without part length rods. The transient normally begin at the ebb (50 percent) of a load follow cycle and will proceed at a rapid itive rate (typically 5 percent per minute) until the abilities of the control rods and the coolant perature reduction (negative moderator coefficient) to supply reactivity are exhausted. At that nt, further power increases are limited to approximately one percent per minute by the ability he boron system to dilute the reactor coolant. The reduction in primary coolant temperature is ted by the protection system to about 20°F below the programmed value.

reduced temperature return to power operation is not intended for daily use. It is designed to ply additional plant capabilities when required because of network fault or upset condition.

ce, this mode of operation is expected to occur 2,000 times in 60 years.

Reactor Coolant Pump (RCP) Startup and Shutdown reactor coolant pumps are started and stopped during routine operations such as RCS venting, t heatup and cooldown, and in connection with recovery from certain transients such as loss ower. Other (undefined) circumstances may also require pump starting and stopping.

he spectrum of RCS pressure and temperature conditions under which these operations may ur, three conditions have been selected for defining transients:

a. Cold condition - 70°F and 400 psig 3.9N-32 Rev. 30
c. Hot condition - 557°F and 2235 psig RCP starting and stopping operations, it is assumed that variations in RCS primary side perature and in pressurizer pressure and temperature are negligible and that the steam erator secondary side is completely unaffected. The only significant variables are the primary em flow and the pressure changes resulting from the pump operations. Occurrences for the p starting/stopping conditions are given in Table 3.9N-1.

Turbine Roll Test s transient is imposed upon the plant during the hot functional test period for turbine cycle ckout. Reactor coolant pump power is used to heat the reactor coolant to operating perature (no-load conditions) and the steam generated is used to perform a turbine roll test.

wever, the plant cooldown during this test exceeds the 100°F per hour design rate.

number of such test cycles is specified at 20 times, to be performed at the beginning of plant rating life prior to reactor operation.

Primary Side Leakage Test sequent to each time the primary system has been opened, a leakage test will be performed.

ing this test the primary side pressure may be raised to a maximum of 2500 psia, in ordance with the system temperature limitations imposed y the reactor vessel material ductility uirements, while the system is checked for leaks. Normally, to prevent the pressurizer safety es from lifting during the leak test, the primary system will be pressurized to approximately 5 psig, as measured at the pressurizer.

ing this leakage test, the secondary side of the steam generator must be pressurized so that the sure differential across the tube sheet does not exceed 1,600 psi. This is accomplished with steam, feedwater, and blowdown lines closed off. For design purposes it is assumed that 200 les of this test will occur during the 60 year life of the plant.

Secondary Side Leakage Test ing the life of the plant, it may be necessary to check the secondary side of the steam generator ticularly, the manway closure) for leakage. For the design purposes, it is assumed that the m generator secondary side is pressurized to just below its design pressure, to prevent the ty valves from lifting. In order not to exceed a secondary side to primary side pressure erential of 670 psi, the primary side must also be pressurized. The primary system must be ve the minimum temperature imposed by reactor vessel material ductility requirements. It is med that this test is performed 80 times during the 60 year life of the plant.

Tube Leakage Test 3.9N-33 Rev. 30

) of the tube sheet for water leakage, with the secondary side pressurized. Tube leakage tests performed during plant cold shutdowns.

these tests the secondary side of the steam generator is pressurized with water, initially at a tively low pressure, and the primary system remains depressurized. The underside of the tube et is examined visually for leaks. If any are observed, the secondary side is then depressurized repairs made by tube plugging. The secondary side is then repressurized (to a higher pressure) the underside of the tube sheet is again checked for leaks. This process is repeated until all the s are repaired. The maximum (final) secondary side test pressure reached is 840 psig.

total number of tube leakage test cycles is defined as 800 during the 60 year life of the plant.

owing is a breakdown of the anticipated number of occurrences at each secondary side test sure:

Test Pressure (psig) Number of Occurrences 200 400 400 200 600 120 840 80 h the primary and secondary sides of the steam generators will be at ambient temperatures ng these tests.

Heaters Out of Service se transients occur when one or more feedwater heaters are taken out of service. During the od of time that the heaters are out of service, it is desirable to maintain the plant at full rated mal load. To accomplish this, first the steam flow is reduced to the amount that will maintain plant at full rated thermal load when the heater(s) is taken out of service. It takes roximately 10 minutes for plant conditions to reach a new steady state. Then the heater(s) is n out of service.

two cases considered here are one heater out of service and one bank of heaters out of service.

design purposes, it is assumed that each of these transients occurs 120 times over the life of plant.

Upset Conditions following primary system transients are considered upset conditions:

1. Loss of load (without immediate reactor trip) 3.9N-34 Rev. 30
3. Partial loss of flow
4. Reactor trip from full power
5. Inadvertent reactor coolant system depressurization
6. Control rod drop
7. Inadvertent safety injection actuation
8. Operating basis earthquake
9. Excessive feedwater flow
10. RCS Cold Overpressurization Loss of Load (without immediate reactor trip) s transient applies to a step decrease in turbine load from full power (turbine trip) without ediately initiating a reactor trip and represents the most severe pressure transient on the RCS er upset conditions. The reactor eventually trips as a consequence of a high pressurizer level initiated by the reactor protection system (RPS). Since redundant means of tripping the tor are provided as part of the RPS, transients of this nature are not expected, but is included nsure a conservative design.

number of occurrences of this transient is specified at 80 times for the 60 year plant design Loss of Power s transient applies to a blackout situation involving the loss of offsite electrical power to the ion, assumed to be operating initially at 100-percent power, followed by reactor and turbine

s. Under these circumstances, the reactor coolant pumps are deenergized and, following stdown of the reactor coolant pumps, natural circulation builds up in the system to some ilibrium value. This condition permits removal of core residual heat through the steam erators which at this time are receiving feedwater, assumed to be at 32°F, from the auxiliary water system operating from diesel generator power. Steam is removed for reactor cooldown ugh atmospheric relief valves provided for this purpose.

number of occurrences of this transient is specified at 40 times for the 60 year plant design Partial Loss of Flow 3.9N-35 Rev. 30

nt are a reactor trip, on low reactor coolant flow, followed by turbine trip and automatic ning of the steam dump system. Flow reversal occurs in the affected loop which causes reactor lant at cold leg temperature to pass through the steam generator and be cooled still further.

s cooler water then flows through the hot leg piping and enters the reactor vessel outlet zles. The net result of the flow reversal is a sizable reduction in the hot leg coolant temperature he affected loop.

number of occurrences of this transient is specified at 80 times for the 60 year plant design Reactor Trip From Full Power eactor trip from full power may occur from a variety of causes resulting in temperature and sure transients in the RCS and in the secondary side of the steam generator. This is the result ontinued heat transfer from the reactor coolant in the steam generator. The transient continues l the reactor coolant and steam generator secondary side temperatures are in equilibrium at power conditions. A continued supply of feedwater and controlled dumping of steam remove core residual heat and prevent the steam generator safety valves from lifting. The reactor lant temperature and pressure undergo a rapid decrease from full power values as the RPS ses the control rods to move into the core. For design purposes, reactor trip is assumed to ur a total of 400 times over the life of the plant.

severity of the cooldown transients associated with reactor trips depends on the extent of ondary side cooling. Three cooldown cases are considered:

a. Reactor trip with no inadvertent cooldown - 230 occurrences
b. Reactor trip with cooldown but no safety injection - 160 occurrences
c. Reactor trip with cooldown actuating safety injection - 10 occurrences Inadvertent Reactor Coolant System Depressurization eral events can be postulated as occurring during normal plant operation which will cause d depressurization of the RCS. These include:
a. Actuation of a single pressurizer safety valve.
b. Inadvertent opening of one or both pressurizer power-operated relief valve due either to equipment malfunction or operator error.
c. Malfunction of a single pressurizer pressure controller causing two pressurizer spray valves to open.

3.9N-36 Rev. 30

e. Inadvertent auxiliary spray. A lockout feature on auxiliary spray valve prevents inadvertent spray of unheated water in the pressurizer.

hese events, the pressurizer safety valve actuation causes the most severe transients, and is d as an umbrella case to conservatively represent the reactor coolant pressure and perature variations arising from any of them.

en a pressurizer safety valve opens, and remains open, the system rapidly depressurizes, the tor trips, and the safety injection system is actuated. Also, the passive accumulators of the SIS actuated when pressure decreases by approximately 1,600 psi. The depressurization and ldown are eventually terminated by operator action. All of these effects are completed within roximately 18 minutes. It is conservatively assumed that none of the pressurizer heaters are rgized.

h pressure constant and safety injection in operation, boil off of hot leg liquid through the surizer and open safety valve will continue.

design purposes this transient is assumed to occur 20 times during the 60 year design life of plant.

Control Rod Drop s transient occurs if a bank of control rods drops into the fully inserted position due to a single ponent failure. The reactor is assumed to be tripped on OTT, OPT or low pressurizer sure. It is assumed that this transient occurs 80 times over the life of the plant.

Inadvertent Safety Injection Actuation design purposes, no credit is taken for the P-19 cold leg injection permissive for the vertent safety injection actuation. A spurious safety injection signal results in an immediate tor trip followed by actuation of the high head centrifugal charging pumps. Without crediting 9, these pumps deliver borated water to the RCS cold legs. The initial portion of this transient milar to the reactor trip from full power with no cooldown. Controlled steam dump and iliary feedwater flow after trip removes core residual heat. Reactor coolant temperature and sure decrease as the control rods move into the core.

er in the transient, the injected water causes the RCS pressure to increase to the pressurizer er operated relief valve (PORV) setpoint and the primary and secondary temperatures to rease gradually. Operator action is credited to verify that the PORV block valves are open hin 10 minutes of initiation of the event. Water relief through the safety valves will not occur vided at least one PORV actuates on demand. The transient continues until the operator stops charging pumps. It is assumed that the plant is then returned to no load conditions, with sure and temperature changes controlled within normal limits.

3.9N-37 Rev. 30

Operating Basis Earthquake operating basis earthquake is that earthquake which can reasonably be expected to occur ng the plant life.

Excessive Feedwater Flow excessive feedwater flow transient is conservatively defined as an umbrella case to cover urrence of several events of the same general nature. The postulated transient results from vertent opening of a feedwater control valve while the plant is at the hot standby or no load dition, with the feedwater, condensate, and heater drain systems in operation.

assumed that the stem of a feedwater control valve fails and the valve immediately reaches full open position. In the steam generator directly affected by the malfunctioning valve iled loop), the feedwater flow step increases from essentially zero flow to the value rmined by the system resistance and the developed head of all operating feedwater pumps.

m flow is assumed to remain at zero and the temperature of the feedwater entering the steam erator is conservatively assumed to be 32°F. Feedwater flow is isolated on a reactor coolant Tavg signal; a low pressurizer pressure signal actuates the safety injection system. Auxiliary water flow, initiated by the safety injection signal, is assumed, to continue with all pumps harging into the affected steam generator. It is also assumed, for conservatism in the ondary side analysis, that auxiliary feedwater flows to the steam generators not affected by the functioned valve, in the unfailed loops. Plant conditions stabilize at the values reached in seconds, at which time auxiliary feedwater flow is terminated. The plant is then either taken old shutdown, or returned to the no load condition at a normal heatup rate with the auxiliary water system under manual control.

design purposes, this transient is assumed to occur 30 times during the life of the plant.

RCS Cold Overpressurization S cold overpressurization can occur during startup and shutdown conditions at low perature, with or without the existence of a steam bubble in the pressurizer, and is especially ere when the reactor coolant system is in a water-solid configuration. The event is inadvertent, usually generated by any one of a variety of malfunctions or operator errors. All events which e occurred to date may be categorized as belonging to either events resulting in the addition of s (mass input transient), or events related to the addition of heat (heat input transient). All e possible transients are represented by composite umbrella design transients, referred to as S cold overpressurization.

design purposes, this transient is assumed to occur 10 times during the 60 year design life of plant.

3.9N-38 Rev. 30

following primary system transients are considered emergency conditions:

1. Small loss-of-coolant accident
2. Small steam line break
3. Complete loss of flow Small loss of coolant accident design transient purposes the small loss of coolant accident is defined as a break equivalent to severance of a 1 inch inside diameter branch connection. (Liquid breaks are limited to less 0.375 inch inside diameter and pressurizer steam space breaks are limited to 0.25 inch by the allation of an orifice. This ensures that the break can be handled by the normal makeup system produce no significant fluid systems transients). Breaks which are much larger than one inch cause accumulator injection soon after the accident and are regarded as faulted conditions.

design purposes it is assumed that this transient occurs five times during the life of the plant.

ould be assumed that the safety injection system is actuated immediately after the break urs and subsequently delivers water at a minimum temperature of 32°F to the RCS.

Small Steam Line Break design transient purposes, a small steam break is defined as a break equivalent in effect to a m safety valve opening and remaining open. This transient is assumed to occur 5 times during life of the plant. The following conservative assumptions are used in defining the transients:

a. The reactor is initially in a hot, zero-power condition.
b. The small steam break results in immediate reactor trip and safety injection actuation.
c. A large shutdown margin, coupled with no feedback or decay heat, prevents heat generation during the transient.
d. The safety injection system operates at a design capacity and repressurizes the RCS within a relatively short time.

Complete Loss of Flow s accident involves a complete loss of flow from full power resulting from simultaneous loss ower to all reactor coolant pumps. The consequences of this incident are a reactor trip and ine trip on undervoltage followed by automatic opening of the steam dump system. For design poses this transient is assumed to occur five times during the plant lifetime.

3.9N-39 Rev. 30

following primary system transients are considered faulted conditions. Each of the following dents should be evaluated for one occurrence:

1. Reactor coolant pipe break (Large loss-of-coolant accident)
2. Large steam line break
3. Feedwater line break
4. Reactor coolant pump locked rotor
5. Control rod ejection
6. Steam generator tube rupture
7. Safe shutdown earthquake Reactor Coolant Pipe Break (Large Loss-of-Coolant Accident) owing postulated rupture of a reactor coolant pipe resulting in a large loss of coolant, the ary system pressure decreases causing the primary system temperature to decrease. Because he rapid blowdown of coolant from the system and the comparatively large heat capacity of metal sections of the components, it is likely that the metal will remain at or near the operating perature by the end of the blowdown. It is conservatively assumed that the safety injection em is actuated to introduce water at a minimum temperature of 32°F into the RCS. The safety ction signal will also result in reactor and turbine trips.

Large Steam Line Break transient is based on the postulated complete severance of the largest steam line. The owing conservative assumptions are made:

a. The reactor is initially in a hot, zero-power condition.
b. The large steam line break results in immediate reactor trip and in actuation of the safety injection system.
c. A large shutdown margin, coupled with no feedback or decay heat, prevents heat generation during the transients.
d. The safety injection system operates at design capacity and repressurizes the reactor coolant system within a relatively short time.

Feedwater Line Break 3.9N-40 Rev. 30

others. The blowdown is completed in approximately 27 seconds. Conditions were servatively chosen to give the most severe primary side and secondary side transients. All iliary feedwater flow exits at the break. The incident is terminated when the operator manually igns the auxiliary feedwater system to isolate the break and to deliver auxiliary feedwater to intact steam generators.

Reactor Coolant Pump Locked Rotor s accident is based on the instantaneous seizure of a reactor coolant pump with the plant rating at full power. The locked rotor can occur in any loop. Reactor trip occurs almost ediately, as the result of low coolant flow in the affected loop.

Control Rod Ejection s accident is based on the single most reactive control rod being instantaneously ejected from core. This reactivity insertion in a particular region of the core causes a severe pressure ease in the RCS such that the pressurizer safety valves will lift and also causes a more severe perature transient in the loop associated with the affected region than in the other loops. For servatism the analysis is based on the reactivity insertion and does not include the mitigating cts (on the pressure transient) of coolant blowdown through the hole in the vessel head ated by the ejected rod.

Steam Generator Tube Rupture s accident postulates the double-ended rupture of a steam generator tube resulting in a rease in pressurizer level and reactor coolant pressure. Reactor trip will occur due to the lting safety injection signal. In addition, safety injection actuation automatically isolates the water lines, by tripping all feedwater pumps and closing the feedwater isolation valves. When accident occurs, some of the reactor coolant blows down into the affected steam generator sing the shell side level to rise. The primary system pressure is reduced below the secondary ty valve setting. Subsequent recovery procedures call for isolation of the steam line leading m the affected steam generator. The recovery actions may involve the use of unheated auxiliary y into the pressurizer. Such an action will require override of the auxiliary spray valve kout.

Safe Shutdown Earthquake safe shutdown earthquake is defined as the maximum vibratory ground motion which can onably be predicted from geologic and seismic evidence.

Test Conditions following primary system transients under test conditions are discussed:

3.9N-41 Rev. 30

2. Secondary Side Hydrostatic test Primary Side Hydrostatic Test pressure tests include both shop and field hydrostatic tests which occur as a result of ponent or system testing. This hydro test is performed at a water temperature which is patible with reactor vessel material ductility requirements and a test pressure of 3,107 psig 5 times design pressure). In this test, the RCS is pressurized to 3,107 psig coincident with m generator secondary side pressure of 0 psig. The RCS is designed for 10 cycles of these rostatic tests, which are performed prior to plant startup. The number of cycles is independent ther operating transients.

itional hydrostatic tests may be performed to meet the inservice inspection requirements of ME Section XI subarticle IS5-20. A total of four such tests is expected. The increase in the gue usage factor caused by these tests is easily covered by the conservative number (200) of ary side leakage tests that are considered for design.

Secondary Side Hydrostatic Test secondary side of the steam generator is pressurized to 1.25 design pressure with a minimum er temperature of 120°F coincident with the primary side at 0 psig.

design purposes it is assumed that the steam generator will experience 10 cycles of this test.

se tests may be performed either prior to plant startup, or subsequently following shutdown major repairs, or both.

N.1.2 Computer Programs Used in Analyses following computer programs have been used in the analysis of Seismic Category I ponents and equipment within the NSSS scope.

1. ANSYS - General purpose finite element code used or problems of structural dynamic response in addition to static elastic and inelastic analysis. The code is widely used in the public domain.

N.1.3 Experimental Stress Analysis experimental stress analysis methods are used for Category I systems or components.

wever, for reactor internals, Westinghouse makes extensive use of measured results from otype plants and various scale model tests as discussed in Section 3.9N.2.

3.9N-42 Rev. 30

N.1.4.1 Loading Conditions structural evaluations performed on the reactor coolant system consider the loadings cified in Table 3.9N-2. These loads result from thermal expansion, pressure, weight, operating s earthquake (OBE), safe shutdown earthquake (SSE), design basis loss-of-coolant accident, plant operational thermal and pressure transients.

N.1.4.2 Analysis of Primary Components ipment which serves as part of the pressure boundary in the reactor coolant loop includes the m generators, the reactor coolant pumps, the pressurizer, and the reactor vessel. This ipment is ANS Safety Class 1 and the pressure boundary meets the requirements of the ASME ler and Pressure Vessel Code,Section III, Subsection NB.

ds are applied to the RCS components for all loading conditions on an umbrella load basis.

t is, on the basis of previous plant analyses, a set of loads are determined which should be er than those seen in any single plant analysis. The umbrella loads represent a conservative ns of allowing detailed component analysis prior to the completion of the system analysis.

results of the reactor coolant loop analysis are used to determine the actual loads acting on the ipment nozzles and the support/component interface locations. Upon completion of the system lysis, conformance is demonstrated between the actual plant loads and the loads used in the lyses of the components. Any deviations where the actual load is larger than the umbrella load handled by individualized analysis.

mic analyses are performed individually for the reactor coolant pump, the pressurizer, and the m generator. Detailed and complex dynamic models are used for the dynamic analyses. The onse spectra corresponding to the building elevation at the highest component/building chment elevation are used for the component analysis. The reactor pressure vessel is qualified tatic analysis based on loads derived from dynamic system analysis.

pressure boundary portions of Class 1 valves in the RCS are designed and analyzed according he requirements of NB-3500 of ASME III.

ves in sample lines connected to the RCS are not considered to be ANS Safety Class 1 nor ME Class 1. This is because the nozzles, where the lines connect to the primary system piping, orificed to a 3/8-inch hole. This hole restricts the flow such that loss through a severance of of these lines can be made up by normal charging flow.

3.9N-43 Rev. 30

N.1.4.3.1 Introduction s section presents the method of computing the reactor pressure vessel response to a postulated

-of-coolant accident (LOCA). The resulting structural analysis considers loads on the reactor sel resulting from the reactor coolant loop mechanical loads, internal hydraulic pressure sients, and reactor cavity pressurization (for postulated breaks in the reactor coolant pipe at vessel nozzles). The vessel is restrained by reactor vessel support pads and shoes beneath 4 of reactor vessel nozzles, and the reactor coolant loops with the primary supports of the steam erators and the reactor coolant pumps.

the limiting break in the main loop piping, pipe displacement restraints installed in the ary shield wall limit the break opening area of the vessel nozzle pipe breaks to less than 144 are inches. This break area was determined to be an upper bound by using worst case vessel pipe relative motions based on similar plant analyses. Detailed studies have shown that pipe ks at the hot or cold leg reactor vessel nozzles, even with a limited break area, would give the hest reactor vessel support loads and the highest vessel displacements, primarily due to the uence of reactor cavity pressurization. By considering these breaks, the most severe reactor sel support loads are determined. In addition, two breaks outside the shield wall, for which e is no cavity pressurization, were also analyzed; specifically, the pump outlet nozzle pipe k and the steam generator inlet nozzle pipe break were considered.

ce the plant has been approved for the leak-before-break (LBB) analysis methodology, the CA analyses of the reactor vessel system are not required to include the dynamic effects of the n RCS pipe breaks, in accordance with GDC-4. The most limiting breaks to be considered are branch line breaks. These breaks consist of breaks in the accumulator line in cold leg; the surizer surge line in the hot leg; and the residual heat removal (RHR) line in the hot leg. These e breaks were considered for the dynamic effects of the reactor vessel system. For the lyses of the branch line breaks, the reactor cavity pressurization loads described in the next ion are not utilized in the analysis.

results of the LOCA analysis of the reactor pressure vessel were determined using separate lyses. Westinghouse performed analyses to determine the reactor vessel displacement due to sel internal hydraulic loads. RPV support and reactor coolant loop stiffnesses, supplied by EC, were included in the RPV structural model. The results of the Westinghouse analyses e dynamic displacement data for the RPV.

EC determined the reactor vessel displacements for the remaining LOCA loads. By bining these displacements with the displacements from the Westinghouse analyses, the total bined RPV displacements were determined for the postulated pipe ruptures.

3.9N-44 Rev. 30

owing a postulated pipe rupture at the reactor vessel nozzle, the reactor vessel is excited by e-history forces. As previously mentioned, these forces are the combined effect of three nomena:

1. Reactor Coolant Loop Mechanical Loads The reactor coolant loop mechanical forces are derived from the elastic analysis of the loop piping for the postulated break.
2. Reactor Cavity Pressurization Forces Reactor cavity pressurization forces arise for the pipe breaks at the vessel nozzles from the steam and water which is released into the reactor cavity through the annulus around the broken pipe. The reactor cavity is pressurized asymmetrically with higher pressure on the side of the broken pipe resulting in horizontal forces applied to the reactor vessel. Smaller vertical forces arising from pressure on the bottom of the vessel and the vessel flanges are also applied to the reactor vessel.

The internals reaction forces develop from asymmetric pressure distributions inside the reactor vessel. For a vessel inlet nozzle break, the depressurization wave path is through the broken loop inlet nozzle and into the downcomer annulus between the core barrel and reactor vessel. The initial waves propagate up, down, and around the downcomer annulus and up through the fuel. In the case of an RPV outlet nozzle break and steam generator inlet nozzle break, the wave passes through the RPV outlet nozzle and directly into the upper internals region, depressurizes the core, and enters the downcomer annulus from the bottom of the vessel. Thus, for an outlet nozzle break, the downcomer annulus is depressurized with much smaller differences in pressure horizontally across the core barrel than for the inlet break. For both the inlet and outlet nozzle breaks, the depressurization waves continue their propagation by reflection and translation through the reactor vessel fluid but the initial depressurization wave has the greatest effect on the loads.

3. Reactor Internal Hydraulic Forces The reactor internals hydraulic pressure transients were calculated including the assumption that the structural motion of the core barrel is coupled with the pressure transients. This phenomenon is known as hydro-elastic coupling or fluid-structure interaction. The hydraulic analysis considers the fluid-structure interaction of the core barrel by accounting for the deflections of constraining boundaries which are represented by masses and springs. The dynamic response of the core barrel in its beam bending mode responding to blowdown forces compensates for internal pressure variation by increasing the volume of the more 3.9N-45 Rev. 30

N.1.4.3.3 Reactor Vessel and Internals Modeling mathematical model of the RPV is a three-dimensional nonliner finite element model which esents the dynamic characteristics of the reactor vessel and its internals in the six geometric rees of freedom. The RPV three-dimensional nonlinear finite element model is shown in ure 3.9N-1. The model consists of three concentric structural submodels connected by linear impact elements and stiffness matrices, which is connected to a submodel of the DMs and CRDM seismic platform, tie rods, and lifting legs. The first submodel, represents the tor vessel shell and associated components. The reactor vessel is restrained by four reactor sel supports (situated beneath alternate nozzles) and by the attached primary coolant piping.

h reactor vessel support is modeled by a linear horizontal stiffness and a vertical impact ment. The attached piping is represented by a stiffness matrix.

second submodel, represents the reactor core barrel (RCB), neutron panels, lower support e, tie plates, and secondary core support components. This submodel is physically located de the first, and is connected to it by a stiffness matrix at the internals support ledge. Core el to vessel shell impact is represented by nonlinear elements at the core barrel flange, core el nozzle, and lower radial support locations.

third and innermost submodel, represents the upper support plate, guide tubes, support mns, upper and lower core plates, and fuel. The third submodel is connected to the first and ond by stiffness matrices and nonlinear elements.

d-structure or hydro-elastic interaction is included in the reactor pressure vessel model for mic evaluation. The horizontal hydro-elastic interaction is significant in the cylindrical fluid region between the core barrel and reactor vessel (the downcomer). Mass matrices with off-onal terms (horizontal degrees-of-freedom only) attach between nodes on the core barrel and tor vessel shell.

diagonal terms of the mass matrix are similar to the lumping of water mass to the vessel shell core barrel. The off-diagonal terms reflect the fact that all the water does not participate when e is no relative motion of the vessel and core barrel. It should be pointed out that the rodynamic mass matrix has no artificial virtual mass effect and is derived in a straight-ward, quantitative manner.

matrices are a function of the properties of two cylinders with a fluid in the cylindrical ulus, specifically; inside and outside radius of the annulus, density of the fluid and length of cylinders. Vertical segmentation of the RCB allows inclusion of radii variations along the B height and approximates the effects of RCB beam deformation. These mass matrices were rted between selected nodes on the core barrel and reactor vessel shell.

he finite element approach, the structure is divided into a finite number of members or ments. Nodal displacements and impact forces are stored for post-processing.

3.9N-46 Rev. 30

rnate reactor vessel nozzles. The RPV support model (supplied by SWEC) is shown in Figure N-3. All support spring elements are linear, double-acting elements and represent the restraint vided by the attached piping and the vessel supports, including the primary shield wall.

N.1.4.3.4 Analytical Methods time history effects of the internals hydraulic loads are applied to the appropriate nodes of the hematical model of the reactor vessel and internals. The analysis is performed by numerically grating the differential equations of motion to obtain the transient response. The output of the lysis is the time history displacements of the reactor vessel system. The output from the lysis is used for detailed component evaluation.

N.1.4.4 Stress Criteria for Reactor Coolant System Components RCS components are designed and analyzed for the design, normal, upset, and emergency ditions to the rules and requirements of the ASME Code Section III. The analysis methods and ciated stress or load allowable limits used in evaluation of faulted conditions are those that defined in Appendix F of the ASME Code.

ding combinations and allowable stresses for RCS components are given in Tables 3.9N-2 and N-3, respectively. For faulted condition evaluations, the effects of the safe shutdown hquake (SSE) and loss-of-coolant accident (LOCA) are combined using the square root of the of the squares (SRSS) method.

N.2 DYNAMIC TESTING AND ANALYSIS N.2.1 Preoperational Vibration and Dynamic Effects Testing on Piping Section 3.9B.2.1.

N.2.2 Seismic Qualification Testing of Safety Related Mechanical Equipment Section 3.10.

N.2.3 Dynamic Response Analysis of Reactor Internals Under Operational Flow Transients and Steady State Conditions vibration characteristics and behavior due to flow induced excitation are very complex and readily ascertained by analytical means alone. Reactor components are excited by the flowing lant which causes oscillatory pressures on the surfaces. The integration of these pressures over applied area should provide the forcing functions to be used in the dynamic analysis of the ctures. In view of the complexity of the geometries and the random character of the pressure llations, a closed form solution of the vibratory problem by integration of the differential ation of motion is not always practical and realistic. The determination of the forcing 3.9N-47 Rev. 30

racteristics of the forcing functions that essentially determine the response of the structures.

studying the dynamic properties of the structure from previous analytical and experimental k, the characteristics of the forcing function can be deduced. These studies indicate that the t important forcing functions are flow turbulence, and pump-related excitation. The relevance uch excitations depends on many factors such as type and location of component and flow ditions. The effects of these forcing functions have been studied from test runs on models, otype plants and in component tests, (WCAP-8303-P-A, WCAP-8517; Trojan Final Safety lysis Report, Appendix A-12; WCAP 8780, WCAP-9945).

Indian Point No. 2 plant (Docket No. 50.247) has been established as the prototype for a

-loop plant internals verification program and was fully instrumented and tested during hot ctional testing. In addition, the Trojan plant (Docket No. 50-344) has provided, and the uoyah No. 1 plant (Docket No. 50-327) has provided, prototype data applicable to Millstone WCAP-8517; Trojan Final Safety Analysis Report, Appendix A-12; WCAP-9945).

lstone 3 is similar to Indian Point No. 2; the only significant differences are the modifications lting from the use of 17 x 17 fuel, replacement of the annular thermal shield with neutron lding panels, and the change to the UHI-style inverted top hat support structure configuration.

se differences are addressed below.

1. 17 x 17 fuel The only structural changes in the internals resulting from the design change from the 15 x 15 to the 17 x 17 fuel assembly is the guide tube. The new 17 x 17 guide tubes are stronger and more rigid, hence they are less susceptible to flow induced vibration. The fuel assembly itself is relatively unchanged in mass and spring rate, and thus no significant deviation of internal vibration is expected from the vibration with the 15 x 15 fuel assemblies vibration characteristics.
2. Neutron shield panel lower internals The primary cause of core barrel excitation is flow turbulence, which is not affected by the upper internals (WCAP-8517). The vibration levels due to core barrel excitation for Trojan and Millstone 3 both having neutron shield panels are expected to be similar. The coolant inlet density of Millstone 3 is slightly lower than Trojan 1 and the flowrate is slightly higher. Scale model tests show that the core barrel vibration varies as velocity, raised to a small power (WCAP-8317). The difference in fluid density and flowrate result in approximately 3 percent higher core barrel vibration for Millstone 3, than for Trojan 1. However, scale model test results (WCAP-8317-A) and results from Trojan (WCAP-8780), show that core barrel vibration of plants with neutron shielding pads is significantly less than that of plants with thermal shields. This information and the fact that low core barrel stresses with large safety margins were measured at Indian Point No. 2 (thermal shield configuration) lead to the conclusion that stresses approximately equal to 3.9N-48 Rev. 30
3. UHI-style inverted top hat upper support configuration The components of the upper internals are excited by turbulent forces due to axial and crossflows in the upper plenum (WCAP-8517) and by pump-speed related excitations (WCAP-8517; WCAP-8780). Sequoyah and Millstone 3 have the same basic upper internals configuration, therefore, the general vibration behavior is not changed.

Results from Sequoyah Unit 1 plant testing (Altman et al., 1981) show high factors of safety for upper internals components. These results, which are supported by scale model test results and analytical work, can be used to determine the adequacy of the Millstone 3 new upper internals.

original test and analysis of the four-loop configuration is augmented by WCAP-8317-A; AP-8517; Trojan Final Safety Analysis Report, Appendix A-12, WCAP-8780; and AP-9945, supported by scale model and analytical work to cover the effects of successive dware modifications.

N.2.4 Preoperational Flow-Induced Vibration Testing of Reactor Internals ause the Millstone 3 reactor internals design configuration is well-characterized, as was ussed in Section 3.9N.2.3, it is not considered necessary to conduct instrumented tests of the lstone Unit 3 hardware. The recommendations of Regulatory Guide 1.20 are met by ducting the confirmatory preoperational testing examination for integrity per Section D, of ulatory Guide 1.20, Regulations for Reactor Internals Similar to the Prototype Design. This mination will include some 30 points with special emphasis on the following areas:

1. All major load-bearing elements of the reactor internals relied upon to retain the core structure in place
2. The lateral, vertical and torsional restraints provided within the vessel
3. Those locking and bolting devices whose failure could adversely affect the structural integrity of the internals
4. Those other locations on the reactor internal components which are similar to those examined on the prototype designs
5. The inside of the vessel will be inspected before and after the hot functional test, with all the internals removed, to verify that no loose parts or foreign materials are in evidence.

3.9N-49 Rev. 30

1. Lower internals
a. Upper barrel to flange girth weld
b. Upper barrel to lower barrel girth weld
c. Upper core plate aligning pin. Examine bearing surfaces for any shadow marks, burnishing, buffing or scoring. Inspect welds for integrity.
d. Irradiation specimen guide screw locking devices and dowel pins. Check for lockweld integrity.
e. Baffle assembly locking devices. Check for lockweld integrity.
f. Lower barrel to core support girth weld
g. Neutron shield panel screw locking devices and dowel pin lockwelds.

Examine the interface surfaces for evidence of tightness. Check for lockweld integrity

h. Radial support key welds
i. Insert screw locking devices. Examine soundness of lockwelds.
j. Core support columns and instrumentation guide tubes. Check the joints for tightness and soundness of the locking devices.
k. Secondary core support assembly weld integrity
l. Lower radial support keys and inserts. Examine bearing surfaces for shadow marks, burnishing, buffing or scoring. Check the integrity of the lockwelds. These members supply the radial and torsional constraint of the internals at the bottom relative to the reactor vessel while permitting axial and radial growth between the two. Subsequent to the hot functional testing, the bearing surfaces of the key and keyway will show burnishing, buffing or shadow marks which indicate pressure loading and relative motion between these parts. Minor scoring of engaging surfaces is also possible and acceptable.
m. Gaps and baffle joints. Check gaps between baffle-to-baffle joints.
2. Upper internals 3.9N-50 Rev. 30
b. Guide tube, support column, orifice plate, and thermocouple assembly locking devices
c. Support column and thermocouple conduit assembly clamp welds
d. Upper core plate alignment inserts. Examine bearing surfaces for shadow marks, burnishing, buffing or scoring. Check the locking devices for integrity of lockwelds.
e. Thermocouple conduit fitting locktab and clamp welds
f. Guide tube enclosure and card welds eptance standards are the same as required in the shop by the original design drawings and cifications.

ing the hot functional test, the internals will be subjected to a total operating time greater than mal full-flow conditions (four pumps operating) of at least 240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br />. This provides a cyclic ing of approximately 107 cycles on the main structural elements of the internals. In addition e will be some operating time with only one, two and three pumps operating.

and post-hot functional inspection results serve to confirm predictions that the internals are quately designed for flow induced vibrations. When no signs of abnormal wear, no harmful ations are detected and no apparent structural changes take place, the four-loop core support ctures are considered to be structurally adequate and sound for operations.

N.2.5 Dynamic System Analysis of the Reactor Internals Under Faulted Conditions reactor internals analysis under faulted events considers the following conditions:

LOCA SSE (Safe Shutdown Earthquake) criteria for acceptability in regard to mechanical integrity analyses are that adequate core ling and core shutdown must be assured. This implies that the deformation of the reactor rnals must be sufficiently small so that geometry remains substantially intact. Consequently, limitations established for the internals are concerned with the deflections and stability of the s in addition to stress criteria to assure integrity of the components.

MULTIFLEX digital computer program (WCAP-8708 and WCAP-8709) which was eloped for the purpose of calculating local fluid pressure, flow, and density transients that ur in pressurized water reactor coolant systems during a loss-of-coolant accident is applied to subcooled, transition, and saturated two-phase blowdown regimes. This is in contrast to 3.9N-51 Rev. 30

nges in the sonic velocities and fluid densities take place. This blowdown code is based on the hod of characteristics wherein the resulting set of ordinary differential equations, obtained m the laws of conservation of mass, momentum, and energy, are solved numerically using a d mesh in both space and time.

hough spatially one-dimensional conservation laws are employed, the code can be applied to cribe three-dimensional system geometries by use of the equivalent piping networks. Such ng networks may contain any number of pipes of channels of various diameters, dead ends, nches (with up to 6 pipes connected to each branch), contractions, expansions, orifices, pumps free surfaces (such as in the pressurizer). System losses such as friction, contraction, ansion, etc are considered.

MULTIFLEX code evaluates the pressure and velocity transients at various locations ughout the system. These pressure and velocity transients are stored as a permanent tape file are made available to the programs LATFORC and FORCE2 which utilize a detailed metric description in evaluating the loadings on the reactor internals. The LATFORC puter code calculates the lateral hydraulic loads on the reactor vessel wall and core barrel, le the FORCE2 code calculates the vertical hydraulic loads on the reactor vessel internals.

h reactor component for which LATFORC and FORCE2 calculations are performed is gnated as an element and assigned an element number. Forces acting upon each of the ments 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 ut to the codes, in addition to the blowdown pressure and velocity transients, includes the ctive area of each element on which the force acts due to the pressure differential across the ment, a coefficient to account for flow stagnation and unrecovered orifice losses, and the total of the element along which the shear forces act.

reactor internals analysis has been performed using conservative assumptions. Some of the e significant assumptions are:

1. The mechanical and hydraulic analyses have considered the effect of hydroelasticity.
2. The reactor internals are represented by concentric pipes, 3-D beams and a multi-mass system connected with springs and dashpots simulating the elastic response and the viscous damping of the components.

3.9N-52 Rev. 30

vertical directions.

pressure waves generated within the reactor are highly dependent on the location and nature he postulated pipe failure. In general, the more rapid the severance of the pipe, the more severe imposed loadings on the components. A one millisecond severance time is taken as the ting case.

he case of the hot leg break, the vertical hydraulic forces produce an initial upward lift of the

. A rarefaction wave propagates through the reactor hot leg nozzle into the interior of the er core barrel. Since the wave has not reached the flow annulus on the outside of the barrel, upper barrel is subjected to an impulsive compressive wave. Thus, dynamic instability ckling) or large deflections of the upper core barrel, or both, are possible responses of the el during hot leg break results in transverse loading on the upper core components as the fluid s the hot leg nozzle.

he case of the cold leg break, a rarefaction wave propagates along a reactor inlet pipe, arriving at the core barrel at the inlet nozzle of the broken loop. The upper barrel is then subjected to a

-axisymmetric expansion radial impulse which changes as the rarefaction wave propagates h around the barrel and down the outer flow annulus between vessel and barrel. After the cold break, the initial steady state hydraulic lift forces (upward) decrease rapidly (within a few iseconds) and then increase in the downward direction. These cause the reactor core and er support structure to move initially downward.

simultaneous seismic event with the intensity of the SSE is postulated with the loss-of-coolant dent, the imposed loading on the internals component may be additive in certain cases and efore the combined loading must be considered. In general, however, the loading imposed by earthquake is small compared to the blowdown loading.

summary of the mechanical analysis follows:

Transverse Excitation Model for Blowdown ious reactor internal components are subjected to transverse excitation during blowdown.

cifically, the barrel, guide tubes, and upper support columns are analyzed to determine their onse to this excitation.

e Barrel - For the hydraulic analysis of the pressure transients during hot leg blowdown, the imum pressure drop across the barrel is a uniform radial compressive impulse.

barrel is then analyzed for dynamic buckling using the following conservative assumptions:

1. The effect of the fluid environment is neglected.
2. The shell is treated as simply supported.

3.9N-53 Rev. 30

r flow annulus between vessel and barrel.

analysis of transverse barrel response to cold leg blowdown is performed as follows:

1. The core barrel is analyzed as a shell with two variable sections to model the support flange and core barrel.
2. The barrel with the core and neutron shield panels, is analyzed as a beam supported at the top and supported at bottom the lower radial support and the dynamic response is obtained.

de Tubes - The guide tubes in closest proximity to the outlet nozzle of the ruptured loop are most severely loaded during a blowdown. The transverse guide tube forces decrease with easing distance from the ruptured nozzle location.

of the guide tubes are designed to maintain the function of the control rods for a break size of in2 and smaller. No credit for the function of the control rods is assumed for break size areas ve 144 in2. However, the design of the guide tube will permit control rod operation in all but control rod positions, which is sufficient to maintain the core in a subcritical configuration, break sizes up to a double-ended hot leg break. This double-ended hot leg break imposes the ting lateral guide tube loading.

er Support Columns - Upper support columns located close to the broken nozzle during hot break will be subjected to transverse loads due to cross flow. The loads applied to the columns computed with a method similar to the one used for the guide tubes, i.e., by taking into sideration the increase in flow across the column during the accident. The columns are studied eams with variable section and the resulting stresses are obtained using the reduced section dulus and appropriate stress risers for the various sections.

effects of the gaps that could exist between vessel and barrel, between fuel assemblies, and ween fuel assemblies and baffle plates are considered in the analysis. The stresses due to the shutdown earthquake (vertical and horizontal components) were combined in the most avorable manner with the blowdown stresses in order to obtain the largest principal stress and ection.

reactor internals components were found to be within acceptable stress and deflection limits both hot leg and cold leg loss of coolant accidents occurring simultaneously with the safe tdown earthquake.

se results indicate that the maximum deflections and stress in the critical structures are below established allowable limits. For the transverse excitation, it is shown that the upper barrel s not buckle during a hot leg break and that it has an allowable stress distribution during a cold break.

3.9N-54 Rev. 30

insertion. For the guide tubes deflected above the no loss of function limit, it must be assumed the rods will not drop. However, the core will still shut down due to the negative reactivity rtion in the form of core voiding. Shutdown will be aided by the great majority of rods that do

p. Seismic deflections of the guide tubes are generally negligible by comparison with the no of function limit.

N.2.6 Correlations of Reactor Internals Vibration Tests with the Analytical Results stated in Section 3.9N.2.3, it is not considered necessary to conduct instrumented tests of the tor vessel internals. WCAP-8516-P, WCAP-8517 (1975) and the Trojan Final Safety Analysis ort, Appendix A12 describe predicted vibration behavior based on studies performed prior to plant tests. These studies, which utilize analytical models, scale model test results, component s, and results of previous plant tests, are used to characterize the forcing functions and blish component structural characteristics so that the flow induced vibratory behavior and onse levels for reactor internals are estimated. These estimates are supported by values uced from plant test data obtained from the Sequoyah and Trojan internals vibration surement programs. Adequacy of the Millstone 3 internals will be verified by the use of uoyah and Trojan results supported by scale model tests and analytical work.

N.3 ASME CODE CLASS 1, 2, AND 3 COMPONENTS, COMPONENT SUPPORTS, AND CORE SUPPORT STRUCTURES ASME Code Class components are constructed in accordance with the ASME Code, Section etailed discussion of ASME Code Class 1 components is provided in Section 3.9N.1. For core port structures, design loading conditions are given in Section 3.9N.2.3. Loading conditions discussed in Section 4.2.

hod of analysis and testing for core support structures are discussed in Sections 3.9N.2.3, N.2.5; and 3.9N.2.6.

N.3.1 Loading Combinations, Design Transients, and Stress Limits for Class 2 Components N.3.1.1 Design Loading Combinations design loading combinations for ASME Code Class 2 and 3 components and supports are n in Table 3.9N-4.

N.3.1.2 Design Stress Limits design stress limits established for the components are sufficiently low to assure that ation of the pressure retaining boundary will not occur. These limits, for each of the loading binations, are component oriented and are presented in Tables 3.9N-5 through 3.9N-10 for 3.9N-55 Rev. 30

design of component supports is discussed in Section 3.9N.3.4.

N.3.2 Pumps and Valve Operability Assurance ipment for Millstone 3 is designed to comply with the intent of Regulatory Guide 1.48; i.e., it esigned/analyzed to ensure structural integrity and operability. However, the load binations and stress limits that were used reflect NRC requirements when the components e purchased and subsequently designed. The codes and procedures which were available when components were purchased are based on conservative design requirements. These codes and cedures have been used by the nuclear industry for the design of components that are installed lants that are presently operating. A list of active pumps and valves is presented in Tables N-11 and 3.9N-12, respectively.

N.3.2.1 ASME Code Class Pumps ive pumps were designed in accordance with Section III of the ASME Code. The stress levels he pumps do not exceed those provided in Tables 3.9N-7 and 3.9N-10. Forces resulting from mic accelerations in the horizontal and vertical directions are included in the analysis of the ps and their supports. To eliminate any amplification of the seismic floor accelerations in the p support structure, the supports were designed to have natural frequencies in excess of 33 pumps are subjected to a series of tests prior to installation and after installation in the plant.

hop tests include hydrostatic tests to 150 percent of the design pressure, seal leakage tests, net itive suction head (NPSH) tests to qualify the pumps for the minimum available NPSH, and ctional performance tests. For the NPSH and functional performance tests, the pumps are ed in a test loop and subjected to operating conditions. After installation, the pumps undergo hydrostatic tests, hot functional tests to verify operation, and periodic inservice inspection operation.

N.3.2.2 Valve Operability ety related valves are subjected to a series of tests prior to service and during the plant life.

r to installation, the following tests are performed: shell hydrostatic test to ASME Code, tion III requirements, backseat and main seat leakage tests, disc hydrostatic test, and rational tests to verify that the valve will open and close. Qualification of motor operators for ironmental conditions is discussed in Sections 3.11 and 1.8 (in the discussion of Regulatory de 1.73, Qualification Tests of Electric Valve Operators Installed Inside the Containment of lear Power Plants). Cold hydrostatic tests, hot functional tests, periodic inservice inspections, periodic inservice operations are performed in situ to verify the functional ability of the valve.

Active components are those whose operability is relied upon to perform a safety function (as well as reactor shutdown function) during the transients or events considered in the respective operating condition categories.

3.9N-56 Rev. 30

ss 2 and 3 valves are shown in Table 3.9N-8. Active valves are designed to have a first natural uency which is equal to or greater than 33 hertz and an analysis of the extended structure is ormed for static equivalent safe shutdown earthquake loads applied at the center of gravity of extended structure.

ddition to these tests and analyses, representative valves of each design type are tested during mulated plant faulted condition event by demonstrating operational capabilities within the cified limits. These tests verify the analysis methods described above. The testing procedures described below.

valve is mounted in a manner which conservatively represents typical valve installations. The e includes the operator, solenoid valves, and limit switches when such are normally attached he valve in service. The operability of the valve during a faulted condition is demonstrated by sfying the following criteria:

1. The actuator and yoke of the valve are statically deflected an amount equal to the deflection caused by the faulted condition accelerations applied at the center of gravity of the extended structure in the direction of the weakest axis of the yoke.

The design pressure of the valve is applied to the valve.

2. The valve is cycled while in the deflected position. The time required to open or close the valve in the deflected position will be compared to similar data taken in the undeflected condition to evaluate the significance of any change.
3. Motor operators, external limit switches, and solenoid valves necessary for operation are qualified in accordance with IEEE Standard 344-1975.

piping designer must limit the inputs to the valve to those levels for which the valve is lified.

above operability program description applies to valves with extended structures. Valves ch are safety related but can be classified as not having an extended structure, such as swing ck valves and safety valves, are considered separately. Valves are qualified by analysis and ing verifies the method of analysis.

ssurizer safety valves are qualified for operability in the same manner as valves with extended ctures as described above. The qualification methods include analysis of the bonnet for static ivalent safe shutdown earthquake loads, in shop hydrostatic and seat leakage tests, and odic in situ valve inspection. Additionally, representative pressurizer safety valves are tested erify analysis methods. This test is described as follows:

1. The safety valve is mounted to represent the worst case installation.
2. The valve body is pressurized to its normal system pressure.

3.9N-57 Rev. 30

4. The pressure is increased until the valve actuates.
5. Actuation of the valve at its setpoint ensures its operability during the faulted condition.

ng these methods, all the safety related valves in the systems are qualified for operability ng a faulted event. These methods outlined above conservatively simulate the seismic event ensure that the active valves will perform their safety related function when necessary.

N.3.2.3 Pump Motor and Valve Operator Qualification ive pump motors and active valve motor operators (and limit switches and solenoid valves) seismically qualified in accordance with IEEE Standard 344-1975.

lification is accomplished by analysis, testing, or by a combination of analysis and testing.

en analysis is used, such methods can be justified by:

1. Demonstrating that equipment being qualified is amenable to analysis, and
2. That the analysis can be correlated with test or be performed using standard analysis techniques.

N.3.3 Design and Installation Details for Mounting of Pressure-Relieving Devices er to Section 3.9B.3.3 N.3.4 Component Supports criteria for Westinghouse supplied supports for ASME Code Class 1 mechanical equipment is ented in Section 3.9N.1.

criteria for Westinghouse supplied supports for ASME Code Class 2 and 3 mechanical ipment is presented in Section 3.9N.3.4.1.

N.3.4.1 Component Supports for Tanks and Heat Exchangers mponent supports for Class 2 and 3 tanks and heat exchangers are of two types: linear and, for most part, plate and shell type supports. The supports for these tanks and heat exchangers are gned and analyzed to the rules and requirements of Subsection NF of Section III of the ASME

e. The design analyses and associated stress or load allowable limits for faulted conditions are e defined in Appendix F of the ASME Code. The only exception to the above is the supports the volume control tank which, because the procurement date of the tank predates Subsection are designed to the requirements of the AISC Code.

3.9N-58 Rev. 30

mponent supports for Class 2 and 3 auxiliary pumps are of two types: plate and shell and, for most part, linear type supports. These supports, with the exception of the supports for the rging and safety injection pumps, are designed by the pump manufacturer to pressure ndary stress limits, but in no case is yield stress of the material exceeded. The supports for the rging and safety injection pumps meet the requirements of Subsection NF of Section III of the ME Code.

N.4 CONTROL ROD DRIVE SYSTEM (CRDS)

N.4.1 Descriptive Information of CRDS Control Rod Drive Mechanism trol rod drive mechanisms are located on the dome of the reactor vessel head. They are pled to rod cluster control assemblies (RCCAs) which have neutron absorber material over the re length of the control rods and derive their name from this feature. The control rod drive hanism is shown in Figure 3.9N-4 and schematically in Figure 3.9N-5.

primary function of the control rod drive mechanism is to insert, withdraw or hold stationary, CAs within the core to control average core temperature and to shutdown the reactor.

control rod drive mechanism is a magnetically-operated jack. A magnetic jack is an ngement of three electromagnets which are energized in a controlled sequence by a power ler to insert or withdraw rod cluster control assemblies in the reactor core in discrete steps.

id insertion of the rod cluster control assemblies occurs when electrical power is interrupted.

control rod drive mechanism consists of four separate subassemblies. They are the pressure sel, coil stack assembly, latch assembly, and the drive rod assembly.

1. The pressure vessel assembly includes a latch housing and a rod travel housing which are connected by a threaded, seal welded, maintenance joint which facilitates replacement of the latch assembly. The closure at the top of the rod travel housing is a threaded cap with a canopy seal weld for pressure integrity.

Seismic support of the control rod drive mechanism is attained by the spacer plates of the rod position indicator coil stack assembly and the seismic support ring.

The latch housing is the lower portion of the pressure vessel and encloses the latch assembly. The rod travel housing is the upper portion of the pressure vessel and provides space for the drive rod assembly during its upward movement as the control rods are withdrawn from the core.

2. The coil stack assembly includes the coil housings, an electrical conduit and connector, and three operating coils:

3.9N-59 Rev. 30

2. The movable gripper coil
3. The lift coil The coil stack assembly is a separate unit which is installed on the control rod drive mechanism by sliding it over the outside of the latch housing. It rests on the base of the latch housing without mechanical attachment.

Energizing the operating coils causes movement of the pole pieces and latches in the latch assembly.

3. The latch assembly includes the guide tube, stationary pole pieces, movable pole pieces, and two sets of latches:
1. The movable gripper latches
2. The stationary gripper latches The latches engage grooves in the drive rod assembly. The movable gripper latches are moved up or down in 5/8-inch steps by the lift pole to raise or lower the drive rod assembly. The stationary gripper latches hold the drive rod assembly while the movable gripper latches are repositioned for the next 5/8-inch step.
4. The drive rod assembly includes a coupling, a drive rod, a disconnect button, a disconnect rod, and a locking button.

The drive rod is machined with external grooves on a 5/8-inch pitch which receives the latches during holding or moving of the drive rod assembly. The coupling is attached to the drive rod and provides the means for coupling to the RCCA.

The disconnect button, disconnect rod assembly, and locking button provide positive locking of the coupling to the RCCA and permits remote disconnection of the drive rod assembly.

control rod drive mechanism can be tripped during any part of the power cycler sequencing if trical power to the coils is interrupted, thereby releasing the drive rod assembly and inserting RCCA.

control rod drive mechanism is threaded and seal welded on a head adaptor on top of the tor vessel head. The drive rod assembly is coupled to the rod cluster control assembly directly w.

3.9N-60 Rev. 30

rgizing the magnetic coils, and insertion is by gravity.

mechanism internals are designed to operate in 650°F reactor coolant. The pressure vessel mbly is designed to retain reactor coolant at 650°F and 2,500 psia. The three operating coils designed to operate at an internal coil temperature of 392°F with forced air cooling required to ntain that temperature.

control rod drive mechanism shown schematically in Figure 3.9N-5 withdraws and inserts an CA as shaped electrical pulses are received by the operating coils. An ON or OFF sequence, ated by silicon controlled rectifiers in the power programmer, causes either withdrawal or rtion of the control rod. Position of the control rod is measured by 42 discrete coils mounted he rod position indicator coil stack assembly surrounding the rod travel housing. Each coil netically senses the entry and presence of the top of the ferromagnetic drive rod assembly as oves through the coil center line.

ing plant operation the stationary gripper coil of the control rod drive mechanism holds the cluster control assembly in a static position until a stepping sequence is initiated at which time movable gripper coil and lift coil are energized sequentially.

Drive Rod Assembly - RCCA Withdrawal drive rod assembly along with a coupled RCCA is withdrawn by repetition of the following uence of events (Figure 3.9N-5).

1. The drive rod assembly is held in a stationary position by the stationary gripper (SG) latches with the SG coil energized.
2. Movable Gripper Coil - ON The latch locking plunger raises the swings the movable gripper latches into the drive rod assembly groove. A nominal 0.059-inch axial clearance exists between the latch tips and the drive rod.
3. Stationary Gripper Coil - OFF The force of gravity acting upon the drive rod assembly and attached control rod, causes the stationary gripper latches and plunger to move downward 0.059 inch until the load of the drive rod assembly and attached control rod is transferred to the movable gripper latches. The plunger continues to move downward and swings the stationary gripper latches out of the drive rod assembly groove.
4. Lift Coil - ON 3.9N-61 Rev. 30
5. Stationary Gripper Coil - ON The plunger raises and closes the gap below the stationary gripper pole. The three links, pinned to the plunger, swing the stationary gripper latches into a drive rod assembly groove. The latches contact the drive rod assembly and lift it (and the attached control rod) 0.059-inch. The nominal 0.059 inch vertical drive rod assembly movement transfers the drive rod assembly load from the movable gripper latches to the stationary gripper latches.
6. Movable Gripper Coil - OFF The latch locking plunger separates from the movable gripper pole under the force of a spring and gravity. Three links, pinned to the plunger, swing the three movable gripper latches out of the drive rod assembly groove.
7. Lift Coil - OFF The gap between the movable gripper pole and lift pole opens. The movable gripper latches drop 5/8 inch to a position adjacent to a drive rod assembly groove.
8. Repeat Step 1 The sequence described above (Items 1 through 7) is defined as one step or one cycle. The rod cluster control assembly moves 5/8 inch for each step or cycle. The sequence is repeated at a rate of up to 72 steps per minute and the drive rod assembly (which has a 5/8-inch groove pitch) is raised 72 grooves per minute. The rod cluster control assembly is thus withdrawn at a rate up to 45 inches per minute.

Drive Rod Assembly/RCCA Insertion sequence for RCCA insertion is similar to that for control rod withdrawal, except the timing ft coil ON and OFF is changed to permit lowering the RCCA. The sequence begins with the energized in hold mode.

1. Lift Coil - ON The 5/8-inch gap between the movable gripper and lift pole closes. The movable gripper latches are raised to a position adjacent to a drive rod assembly groove.
2. Movable Gripper Coil - ON 3.9N-62 Rev. 30

the latch tips and the drive rod assembly.

3. Stationary Gripper Coil - OFF The force of gravity, acting upon the drive rod assembly and attached RCCA, causes the stationary gripper latches and plunger to move downward 0.059 inch until the load of the drive rod assembly and attached RCCA is transferred to the movable gripper latches. The plunger continues to move downward and swings the stationary gripper latches out of the drive rod assembly groove.
4. Lift Coil - OFF The force of gravity and spring force separates the movable gripper pole from the lift pole and the drive rod assembly and attached RCCA drop down 5/8 inch.
5. Stationary Gripper Coil - ON The plunger raises and closes the gap below the stationary gripper pole. The three links, pinned to the plunger, swing the three stationary gripper latches into a drive rod assembly groove. The latches contact the drive rod assembly and lift it (and the attached control rod) 0.059 inch. The nominal 0.059-inch vertical drive rod assembly movement transfers the drive rod assembly load from the movable gripper latches to the stationary gripper latches.
6. Movable Gripper Coil - OFF The latch locking plunger separates from the movable gripper pole under the force of a spring and gravity. Three links, pinned to the plunger, swing the three movable gripper latches out of the drive rod assembly groove.
7. Repeat Step 1 The sequence is repeated, as for rod cluster control assembly withdrawal, up to 72 times per minute which gives an insertion rate of 45 inches per minute.

Holding and Tripping of the Control Rods ing most of the plant operating time, the control rod drive mechanisms hold the RCCAs hdrawn from the core in a static position. In the holding mode, only one coil, the stationary per coil, is energized on each mechanism. The drive rod assembly and attached RCCAs hang pended from the three latches.

ower to the stationary gripper coil is cut off, the stationary gripper return spring combined h the weight of the drive rod assembly and the RCCA is sufficient to move latches out of the 3.9N-63 Rev. 30

N.4.2 Applicable CRDS Design Specifications those components in the control rod drive system comprising portions of the reactor coolant sure boundary, conformance with the General Design Criteria and 10CFR50, Section 50.55a iscussed in Sections 3.1 and 5.2 conformance with regulatory guides pertaining in Section 4.5 5.2.3 and Section 1.8.

Design Bases es for temperature, stress and structural members, and material compatibility are imposed on design of the reactivity control components.

Design Transient and Loading Combinations CRDS is designed to withstand stresses originating from various operating design transients marized in Table 3.9N-1. Structural evaluation performed on CRDM pressure retaining ponents consider the loading combinations specified in Table 3.9N-2.

Allowable Stresses owable stresses for CRDM pressure retaining components are given in Table 3.9N-3.

Dynamic Analysis cyclic stresses due to dynamic loads and deflections are combined with the stresses imposed oads from component weights, hydraulic forces and thermal gradients for the determination of total stresses of the CRDS.

Control Rod Drive Mechanisms control rod drive mechanism (CRDM) pressure housings are Class 1 components designed to t the stress requirements for normal operating conditions of Section III of the ASME Boiler Pressure Vessel Code. Both static and alternating stress intensities are considered. The sses originating from the required design transients are included in the analysis.

ynamic seismic analysis is required on the CRDMs when a seismic disturbance has been tulated to confirm the ability of the pressure housing to meet ASME Code,Section III wable stresses and to confirm its ability to trip when subjected to the seismic disturbance.

Control Rod Drive Mechanism Operational Requirements basic operational requirements for the CRDM's are:

3.9N-64 Rev. 30

2. 144-inch travel (nominal)
3. 360 pound maximum load
4. Step in or out of 45 inches/minute (72 steps/minute)
5. Electrical power interruption shall initiate release of drive rod assembly/RCCA
6. Trip delay time of less than 150 milliseconds or less - Free fall of drive rod assembly shall begin less than 150 milliseconds after power interruption no matter what holding or stepping action is being executed with any load and coolant temperature of 100°F to 550°F
7. 60 year design life with normal refurbishment N.4.3 Design Loads, Stress Limits, and Allowable Deformations N.4.3.1 Pressure Vessel Assembly pressure retaining components are analyzed for loads corresponding to normal, upset, rgency and faulted conditions. The analysis performed depends on the mode of operation er consideration.

scope of the analysis requires many different techniques and methods, both static and amic.

e of the loads that are considered on each component where applicable are as follows:

1. Control rod trip (equivalent static load)
2. Differential Pressure
3. Spring preloads
4. Coolant flow forces (static)
5. Temperature gradients
6. Differences in thermal expansion
a. Due to temperature differences
b. Due to expansion of different materials 3.9N-65 Rev. 30
8. Vibration (mechanically or hydraulically induced)
9. All operational transients listed in Table 3.9N-1
10. Pump overspeed
11. Seismic loads (operational basis earthquake and safe shutdown earthquake)
12. Blowdown forces (due to cold and hot leg break) main objective of the analysis is to satisfy allowable stress limits, to assure an adequate gn margin, and to establish deformation limits which are concerned primarily with the ctioning of the components. The stress limits are established not only to assure that peak sses will not reach unacceptable values, but also to limit the amplitude of the oscillatory stress ponent in consideration of fatigue characteristics of the materials. Standard method of ngth of materials are used to establish the stresses and deflections of these components. The amic behavior of the reactivity control components has been studied using experimental test and experience from operating reactors.

N.4.3.2 Drive Rod Assembly postulated failures of the drive rod assemblies either by fracture or uncoupling lead to a uction in reactivity. If the drive rod assembly fractures at any elevation, that portion remaining pled falls with, and is guided by, the rod cluster control assembly. This always results in tivity decrease for the control rods.

N.4.3.3 Latch Assembly and Coil Stack Assembly Results of Dimensional and Tolerance Analysis h respect to the control rod drive mechanism system as a whole, critical clearances are present he following areas:

1. Latch assembly (diametral clearances)
2. Latch arm-drive rod clearances
3. Coil stack assembly-thermal clearances
4. Coil fit in coil housing following defines clearances that are designed to provide reliable operation in the CRDM in e four critical areas. These clearances have been proven by life tests and actual field ormance at operating plants.

3.9N-66 Rev. 30

magnetic jack has several clearances where parts made of Type 410 stainless steel fit over s made from Type 304 stainless steel. Differential thermal expansion is therefore important.

imum clearances of these parts at 68°F is 0.011 inches. At a maximum design temperature of

°F minimum clearance is 0.0045 inches and at the maximum expected operating temperatures 50°F is 0.0057 inches.

Latch Arm - Drive Rod Clearances CRDM incorporates a load transfer action. The movable or stationary gripper latch are not er load during engagement, as previously explained, due to load transfer action.

ure 3.9N-6 shows latch clearance variation with the drive rod as a result of minimum and imum temperatures. Figure 3.9N-7 shows clearance variations over the design temperature ge.

Coil Stack Assembly - Thermal Clearances assembly clearances of the coil stack assembly over the latch housing was selected so that the mbly could be removed under all anticipated conditions of thermal expansion.

70°F, the inside diameter of the coil stack is 7.428/7.438 inches. The outside diameter of the h housing is 7.390/7.380 inches.

rmal expansion of the mechanism due to operating temperature of the control rod drive hanism results in minimum inside diameter of the coil stack being 7.440 inches at 222°F and maximum latch housing diameter being 7.426 inches at 650°F.

er the extreme tolerance conditions listed above it is necessary to allow time for a 70°F coil k assembly to heat during a replacement operation.

r similar style coil stack assemblies were removed from four hot control rod drive hanisms mounted on 11.035-inch centers on a 550°F test loop, allowed to cool, and then aced without incident as a test to prove the preceding.

Coil Fit in Core Housing DM coils and coil housing clearances are selected so that coil heat up results in a close to tight This is done to facilitate thermal transfer and coil cooling in a hot CRDM.

3.9N-67 Rev. 30

Evaluation of Materials Adequacy ability of the pressure housing components to perform throughout the design lifetime as ned in the design specification is confirmed by the stress analysis report required by the ME Code,Section III.

rnal components subjected to wear have withstood a minimum of 2,500,000 steps without rbishment as confirmed by life tests (Cooper 1974).

confirm the mechanical adequacy of the fuel assembly, the CRDM, and rod cluster control mbly (RCCA), functional test programs have been conducted on a full scale 12-foot control The 12-foot prototype assembly was tested under simulated conditions of reactor perature, pressure, and flow for approximately 1,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. The prototype mechanism umulated about 2,500,000 steps and 600 trips. At the end of the test the CRDM was still rating satisfactorily. A correlation was developed to predict the amplitude of flow-excited ation of individual fuel rods and fuel assemblies. Inspection of the drive line components did reveal significant fretting.

se tests include verification that the trip time achieved by the CRDM's meet the design uirement of 2.2 seconds from start of RCCA motion to dashpot entry. This trip time uirement will be confirmed for each CRDM prior to initial reactor operation and at periodic rvals after initial reactor operation as required by the Technical Specifications.

re are no significant differences between the prototype CRDMs and the production units.

ign materials, tolerances and fabrication techniques (Section 4.2.3.3.2) are the same.

se tests have been reported in WCAP-8446 and WCAP-8449.

expected that all CRDM's will meet specified operating requirements for the duration of plant with normal refurbishment.

n RCCA cannot be moved by its mechanism, adjustments in the boron concentration ensure adequate shutdown margin would be achieved following a trip. Thus, inability to move one CA can be tolerated. More than one inoperable RCCA could be tolerated, but would impose itional demands on the plant operator. Therefore, the number of inoperable RCCA's has been ted to one as discussed in the Technical Specifications (Chapter 16).

rder to demonstrate proper operation of the CRDM and to ensure acceptable core power ributions during operation, RCCA partial-movement checks are performed (Technical cifications). In addition, periodic drop tests of the RCCA are performed at each refueling tdown to demonstrate continued ability to meet trip time requirements, to ensure core criticality after reactor trip, and to limit potential reactivity insertions from a hypothetical CA ejection. During these tests the acceptable drop time of each assembly is not greater than 3.9N-68 Rev. 30

ual experience in operating Westinghouse plants indicates excellent performance of CRDM's.

units are production tested prior to shipment to confirm ability of the CRDM to meet design cification-operation requirements.

h production CRDM undergoes a production test as listed below:

Test Acceptance Criteria Cold (ambient) hydrostatic ASME Section III pressure test Step Length:

Confirm step length and load transfer (stationary gripper to 5/8 +/- 0.015 inches axial movement movable gripper or movable Load Transfer:

gripper to stationary gripper) 0.059 inches nominal axial movement Operating Speed:

45 inches/minute Cold (ambient) performance Trip Delay:

Test at design load -5 full travel Free fall of drive rod assembly to begin excursions within 150 milliseconds as verified by normal gripper latch opening times recorded during CRDM performance tests N.5 REACTOR VESSEL INTERNALS N.5.1 Design Arrangements reactor vessel internals are described as follows:

The components of the reactor internals are divided into three parts consisting of the lower core support assembly (including the entire core barrel and neutron shield pad assembly),

the upper core support assembly and the incore instrumentation support structure. The reactor internals support the core, maintain fuel alignment, limit fuel assembly movement, maintain alignment between fuel assemblies and CRDMs, direct coolant flow past the fuel elements, direct coolant flow to the pressure vessel head, provide gamma and neutron shielding, and provide guides for the incore instrumentation. The coolant flows from the vessel inlet nozzles down the annulus between the core barrel and the vessel wall and then into a plenum at the bottom of the vessel. It then reverses and flows up through the core support and through the lower core plate. The lower core plate is sized to provide the 3.9N-69 Rev. 30

outlet nozzles and directly through the vessel outlet nozzles. A small portion of the coolant flows between the baffle plates and the core barrel to provide additional cooling of the barrel. Similarly, a small amount of the entering flow is directed into the vessel head plenum and exits through the vessel outlet nozzles.

Lower Core Support Assembly The major containment and support member of the reactor internals is the lower core support assembly, shown in Figure 3.9N-8. This assembly consists of the core barrel, the core baffle, the lower core plate and support columns, the neutron shield pads, and the core support which is welded to the core barrel. All the major material for this assembly is Type 304 stainless steel. The lower core support assembly is supported at its upper flange from a ledge in the reactor vessel flange and its lower end is restrained in its transverse movement by a radial support system attached to the vessel wall. Within the core barrel are an axial baffle and a lower core plate, both of which are attached to the core barrel wall and form the enclosure periphery of the assembled core. The lower core support assembly and principally the core barrel serve to provide passageways and control for the coolant flow. The lower core plate is positioned at the bottom level of the core below the baffle plates and provides support and orientation for the fuel assemblies.

The lower core plate is a member through which the necessary flow distribution holes for each fuel assembly are located. Fuel assembly locating pins (two for each assembly) are also inserted into this plate. Columns are placed between this plate and the core support of the core barrel in order to provide stiffness and to transmit the core load to the core support. Adequate coolant distribution is obtained through the use of the lower core plate and core support.

The neutron shield pad assembly consists of four pads that are bolted and pinned to the outside of the core barrel. These pads are constructed of Type 304 stainless steel and are approximately 48 inches wide by 148 inches long by 2.8-inches thick. The pads are located azimuthally to provide the required degree of vessel protection. Specimen guides in which material surveillance samples can be inserted and irradiated during reactor operation are attached to the pads. The samples are held in the guides by a preloaded spring device at the top and bottom to prevent sample movement. Additional details of the neutron shield pads and irradiation specimen holders are given in WCAP-7870 (1972).

Vertically downward loads from weight, fuel assembly preload, control rod dynamic loading, hydraulic loads, and earthquake acceleration are carried by the lower core plate partially into the lower core plate support flange on the core barrel shell and partially through the lower support columns to the core support and thence through the core barrel shell to the core barrel flange supported by the vessel flange. Transverse loads from earthquake acceleration, coolant cross flow, and vibration are carried by the core barrel shell and distributed between the lower radial support to the vessel wall and to the vessel flange. Transverse loads of the fuel assemblies are transmitted to the core barrel shell by 3.9N-70 Rev. 30

The main radial support system of the lower end of the core barrel is accomplished by key and keyway joints to the reactor vessel wall. At equally spaced points around the circumference, an Inconel clevis block is welded to the vessel inner diameter. Another Inconel insert block is bolted to each of these blocks and has a keyway geometry.

Opposite each of these is a key which is attached to the internals. At assembly, as the internals are lowered into the vessel, the keys engage the keyways in the axial direction.

With this design, the internals are provided with a support at the furthest extremity, and may be viewed as a beam supported at the top and bottom.

Radial and axial expansions of the core barrel are accommodated but transverse movement of the core barrel is restricted by this design. With this system, cyclic stresses in the internal structures are within the ASME Section III limits. In the event of an abnormal downward vertical displacement of the internals following a hypothetical failure, energy absorbing devices limit the displacement after contacting the vessel bottom head. The load is then transferred through the energy absorbing devices of the internals to the vessel.

The energy absorbers base plate is contoured on its bottom surface to the reactor vessel bottom geometry. Assuming a downward vertical displacement the potential energy of the system is absorbed mostly by the strain energy of the energy absorbing devices.

Neutron Shield Panel Design for Millstone 3 The neutron shielding panel design of Millstone 3 consists of four sets of stainless steel plates strategically placed on the core barrel in areas of peak fast neutron flux on the reactor pressure vessel. See Figures 3.9N-11 and 3.9N-12. Attachment of each of the pad sections to the core barrel is accomplished through a series of sixteen 7/8-inch stainless steel bolts and three 2-3/8-inch stainless steel pins. The bolts are designed to resist most of the primary flow-induced and seismic normal loads; however, the pins carry all of the weight and are designed to resist the accident loads. In addition, since the pins are press fit into position, they retain a high compression-induced friction force in the pressure vessel radial direction. This enables the pins to also function as a redundant support in the pressure vessel radial direction.

The pads are divided into two sections to reduce the effects of vertical relative thermal expansion between the barrel and the pads. Six specimen baskets are utilized in this design and are positioned on the neutron pads in both tandem and single configurations. The baskets are attached to the neutron pads by sets of eight 3/4-inch stainless steel bolts and two 7/8-inch pins.

Substitution of the neutron pads for the thermal shield results in some significant design advantages, specifically:

3.9N-71 Rev. 30

2. The velocity in the downcomer region is reduced by approximately 15 percent.
3. There is a net reduction in weight of 50,000 pounds. The thermal shield weighs approximately 75,000 pounds, while the weight of the neutron pad assembly is approximately 25,000 pounds.
4. The peak fast neutron flux is slightly reduced since the steel is closer to the core.

The smaller pressure drop and lower velocity leads to a reduction in the magnitude of the exciting forces on the internals.

The lower weight increases the natural frequency of the lower internals system. This is also an aid in reducing the tendency for induced vibrations. The adequacy of the design was confirmed by both analysis and test. Since the design satisfiesSection III of the ASME Code, there is assurance that sufficient margin exists in the design.

Fatigue tests were performed on the bolts to simulate the effect of loading placed on the bolts resulting from the relative thermal deflection between the neutron shielding pads and the core support barrel. The tests performed on the bolts indicate that the bolts satisfy Appendix 1-10 of ASME Section III failure criteria of 20 times expected number of cycles (18,300) or 2 times expected stress. Actually, the bolts were subjected to 370,000 cycles at double the expected operating amplitude.

To confirm that there were no deleterious effects from flow-induced vibration, flow tests were performed on a 1/24 scale flow model of the internals with neutron pads, and the results compared to similar tests with a thermal shield. The results indicated extremely low levels of vibration.

In summary, the neutron shield panel design of Millstone 3 has been shown to be structurally adequate by the following:

1. An in-depth analysis, considering loadings for all plant operating conditions, indicates that the design satisfies all the criteria of Section NG 3000 of Section III of the ASME Code.
2. Tests performed on bolts indicate that the bolts satisfy the failure criteria of 20 times expected number of fatigue cycles (18,300) or double the expected stress levels. Actually, the test data indicated that the safety factors are approximately 3 on stress and 2 x 105 on cycles.
3. Flow tests were performed on a 1/24 scale flow model of the internals with neutron pads and the results compared to similar tests with a thermal shield. The levels of vibration for the neutron shield pads were negligible.

3.9N-72 Rev. 30

The upper core support assembly, shown in Figures 3.9N-9 and 3.9N-10 consists of the upper support, the upper core plate, the support columns, and the guide tube assemblies.

The support columns establish the spacing between the upper support and the upper core plate. They are fastened at top and bottom to these plates. The support columns transmit the mechanical loadings between the two plates and serve the supplementary function of supporting thermocouples. The guide tube assemblies, sheath and guide the control rod drive shafts and control rods. They are fastened to the upper support and are restrained by pins in the upper core plate for proper orientation and support.

The upper core support assembly is positioned in its proper orientation with respect to the lower core support assembly by flat-sided pins in the core barrel flange. At an elevation in the core barrel where the upper core plate is positioned, four equally spaced flat-sided pins are located. Four mating sets of inserts are located in the upper core plate at the same positions. As the upper support assembly is lowered into the lower support assembly, the inserts engage the flat-sided pins in the axial direction. Lateral displacement of the plate and the upper support assembly is restricted by this design. Fuel assembly locating pins protrude from the bottom of the upper core plate and engage the fuel assemblies as the upper assembly is lowered into place. Proper alignment of the lower core support assembly, the upper core support assembly, the fuel assemblies and control rods are thereby ensured by this system of locating pins and guidance arrangement. The upper and lower core support assemblies are restrained from any axial movements by a large circumferential spring which rests between the upper barrel flange and the upper core support assembly and is compressed by installation of the reactor vessel head.

Vertical loads from weight, earthquake acceleration, hydraulic loads and fuel assembly preload are transmitted through the upper core plate via the support columns to the upper support and then into the reactor vessel head. Transverse loads from coolant cross flow, earthquake acceleration, and possible vibrations are distributed by the support columns to the upper support and upper core plate. The upper support is particularly stiff to minimize deflection.

Incore Instrumentation Support Structures The incore instrumentation support structures consist of an upper system to convey and support thermocouples penetrating the vessel through the head and a lower system to convey and support flux thimbles penetrating the vessel through the bottom (Figure 7.7-9 shows the basic flux-mapping system).

The upper system utilizes the reactor vessel head penetrations. Instrumentation port columns are slip-connected to inline columns that are in turn fastened to the upper support.

These port columns protrude through the head penetrations. The thermocouples are carried through these port columns and the upper support at positions above their readout locations. The thermocouple conduits are supported from the columns of the upper core support system. The thermocouple conduits are stainless steel tubes.

3.9N-73 Rev. 30

thimbles that are pushed upward into the reactor core. Conduit tubes extend from the bottom of the reactor vessel down through the concrete shield area and up to a thimble seal table. The minimum bend radii are approximately 144 inches and the trailing ends of the thimbles (at the seal table) are extracted approximately 15 feet during refueling the reactor. The Conduit Tubes are classified as ASME Section III Class 1 and designed in accordance with Subsection NC. The thimbles are closed at the leading ends and serve as the pressure barrier between the reactor pressurized water and the containment atmosphere.

Mechanical seals between the retractable thimbles and conduits are provided at the seal table. During normal operation, the retractable thimbles are stationary. They can move only during refueling or for maintenance, at which time a space of approximately 15 feet above the seal table is cleared for the retraction operation.

The incore instrumentation support structure is designed for support of instrumentation during reactor operation and is rugged enough to resist damage under the conditions imposed during the refueling sequence.

N.5.2 Design Loading Conditions Normal and Upset Conditions normal and upset loading conditions that provide the basis for the design of the reactor rnals are:

1. Fuel and reactor internals weight
2. Fuel and core component spring forces, including spring preloading forces
3. Differential pressure and coolant flow forces
4. Temperature gradients
5. Vibratory loads including OBE seismic loads
6. Normal and upset operational thermal transients listed in Table 3.9N.1
7. Control rod trip (equivalent static load)
8. Loads due to loop(s) out of service
9. Loss of load/pump overspeed Emergency Conditions 3.9N-74 Rev. 30
1. Small LOCA
2. Small steam break
3. Complete loss of flow Faulted Conditions faulted loading conditions that provide the basis for the design of the reactor internals are:
1. Branch line breaks, as determined from leak-before-break (LBB) analysis
2. SSE N.5.3 Design Loading Categories combination of design loadings fit into the normal, upset, emergency or faulted conditions as ned in the ASME Code,Section III.

ds and deflections imposed on components due to shock and vibration are determined lytically and experimentally in both scaled models and operating reactors. The cyclic stresses to these dynamic loads and deflections are combined with the stresses imposed by loads from ponents weights, hydraulic forces and thermal gradients for the determination of the total sses of the internals.

reactor internals are designed to withstand stresses originating from various operating ditions as summarized in Table 3.9N-1.

scope of the stress analysis problem is very large requiring many different techniques and hods, both static and dynamic. The analysis performed depends on the mode of operation er consideration.

Allowable Deflections normal operating conditions, downward vertical deflection of the lower core support plate is ligible.

the loss of coolant accident plus the safe shutdown earthquake condition, the deflection eria of critical internal structures are the limiting values given in Table 3.9N-13. The esponding no loss of function limits are included in Table 3.9N-13 for comparison purposes h the allowed criteria.

criteria for the core drop accident are based upon analyses which have to determine the total nward displacement of the internal structures following a hypothesized core drop resulting 3.9N-75 Rev. 30

. An additional displacement of approximately 3/4 inch would occur due to strain of the rgy absorbing devices of the secondary core support; thus the total drop distance is about 4 inches which is insufficient to permit the tips of the rod cluster control assembly to come of the guide thimble in the fuel assemblies.

cifically, the secondary core support is a device which will never be used, except during a othetical accident of the core support (core barrel, barrel flange, etc.). There are 4 supports in h reactor. This device limits the fall of the core and absorbs much of the energy of the fall ch otherwise would be imparted to the vessel. The energy of the fall is calculated assuming a plete and instantaneous failure of the primary core support and is absorbed during the plastic ormation of the controlled volume of stainless steel, loaded in tension.

maximum deformation of this austenitic stainless steel piece is limited to approximately 15 ent, after which a positive stop is provided to ensure support.

N.5.4 Design Bases design bases for the mechanical design of the reactor vessel internals components are as ows:

1. The reactor internals in conjunction with the fuel assemblies shall direct reactor coolant through the core to achieve acceptable flow distribution and to restrict bypass flow so that the heat transfer performance requirements are met for all modes of operation. In addition, required cooling for the pressure vessel head shall be provided so that the temperature differences between the vessel flange and head do not result in leakage from the flange during reactor operation.
2. In addition to neutron shielding provided by the reactor coolant, a separate neutron pad assembly is provided to limit the exposure of the pressure vessel in order to maintain the required ductility of the material for all modes of operation.
3. Provisions shall be made for installing incore instrumentation useful for the plant operation and vessel material test specimens required for a pressure vessel irradiation surveillance program.
4. The core internals are designed to withstand mechanical loads arising from operating basis earthquake, safe shutdown earthquake and pipe ruptures and meet the requirement of Item 5 below.
5. The reactor shall have mechanical provisions which are sufficient to adequately support the core and internals and to assure that the core is intact with acceptable heat transfer geometry following transients arising from abnormal operating conditions.

3.9N-76 Rev. 30

specified limits. This implies that the deformation of certain critical reactor internals must be kept sufficiently small to allow core cooling.

functional limitations for the core structures during the design basis accident are shown in le 3.9N-13. To ensure no column loading or rod cluster control guide tubes, the upper core e deflection is limited to not exceed the value shown in Table 3.9N-13.

ails of the dynamic analyses, input forcing functions, and response loadings are presented in tion 3.9N.2.

basis for the design stress and deflection criteria is identified below:

Allowable Stresses normal operating conditions the intent of Section III of the ASME Code is used as a basis for luating acceptability of calculated stresses. Both static and alternating stress intensities are sidered.

ould be noted that the allowable stresses in Section III of the ASME Code are based on radiated material properties. In view of the fact that irradiation increases the strength of the e 304 stainless steel used for the internals, although decreasing its elongation, it is considered use of the allowable stresses in Section III is appropriate and conservative for irradiated rnal structures.

allowable stress limits during the design basis accident used for the core support structures based on the 1974 Edition of the ASME Code for Core Support Structures, Subsection NG, the Criteria for faulted conditions.

N.6 INSERVICE TESTING OF PUMPS AND VALVES er to Section 3.9.7.

3.9N-77 Rev. 30

st program has been developed to ensure that all safety related pumps and valves will be in a e of operational readiness throughout plant life.

ASME Code,Section XI, 1980 Edition through Winter 1980 Addendum, provided the basic s used to identify applicable pumps and valves and to develop test requirements.

7.1 Inservice Testing of Pumps rvice testing is required for all Class 1, 2, and 3 pumps (both centrifugal and displacement s) that are provided with an EMERGENCY POWER SOURCE. Drivers are excluded except n the pump and driver form an integral unit and the pump bearings are in the driver.

tests and examination procedures required by the ASME code for Operation and Maintenance uclear Power Plants including schedules, reference values, the location, and type of surement for each of the required test quantities, records of the results, and all corrective on taken are defined in the Inservice Test Pump and Valve Program and performed by the er for Millstone 3.

7.2 Inservice Testing of Valves s section describes the Class 1, 2, and 3 valves required to be exercised and tested to verify rational readiness. These valves (with their actuating and position indicating devices) are uired to perform a special function in bringing a reactor to cold shutdown condition or in gating the consequences of an accident.

ves used for operating convenience only such as manual vent, drain, instrument and test es, and valves used for maintenance only do not require inservice testing. The following gories of valves are subject to inservice testing:

1. Category A - Valves for which seat leakage is limited to a specific maximum amount in the closed position for fulfillment of their function.
2. Category B - Valves for which seat leakage in the closed position is inconsequential for fulfillment of their function.
3. Category C - Valves which are self-actuating in response to some system characteristic, such as pressure (relief valves) or flow direction (check valves).
4. Category D - Valves which are actuated by an energy source capable of only one operation, such as rupture disks or explosive actuated valves.

t and examination procedures required by Subsection IWV, including schedules and the ting values of observed parameters are defined in the Inservice Test Pump and Valve Program performed by the licensee for Millstone 3.

3.9-90 Rev. 30

1 Bohm, G.J. and LaFaille, J.P. 1971. Reactor Internals Response Under a Blowdown Accident, Procedures, First Intl. Conf. on Structural Mech. in Reactor Technology. Berlin, September 20-24, 1971.

2 DeSalvo, G.J. and Swanson, J.A. 1972. ANSYS Users Manual. Engineering Analysis Systems Report, October 1, 1972.

3 Fabic, S. 1967. Computer Program WHAM for Calculation of Pressure, Velocity, and Force Transients in Liquid Filled Piping Networks. Kaiser Engineers Report No.

67-49-R.

4 Trojan Final Safety Analysis Report, Appendix A-12 (Docket No. 50-344).

5 WCAP-7870, 1972, Kraus, S. Neutron Shielding Pads.

6 WCAP-8252, Revision 1, 1977, Documentation of Selected Westinghouse Structural Analysis Computer Codes.

7 WCAP-8303-P-A (Proprietary) and WCAP-8317-A (Non-Proprietary), 1975, Prediction of the Flow-Induced Vibration of Reactor Internals by Scale Model Tests.

8 WCAP-8446 (Proprietary) and WCAP-8449 (Non-Proprietary), 1974, Cooper, F.W. Jr.

1974. 17x17 Drive Line Components Tests - Phase IB, II, III, D-Loop Drop and Deflection.

9 WCAP-8516-P (Proprietary) and WCAP-8517 (Non-Proprietary), 1975. Bloyd, C.N. and Singleton, N.R. UHI Plant Internals Vibration Measurement Program and Pre and Post Hot Functional Examinations.

10 WCAP-8708-P-A (Proprietary) and WCAP-8709-A (Non-Proprietary), 1977. Takeuchi, K. MULTIFLEX-A Fortran-IV Computer Program for Analyzing Thermal - Hydraulic -

Structure System Dynamics.

11 WCAP-8780, 1976, Bloyd, C.N.; Ciaramitaro, W.; and Singleton, N.R., Verification of Neutron Pad and 17x17 Guide Tube Designs by Preoperational Tests on the Trojan 1 Power Plant.

12 WCAP-8929, 1977, Benchmark Problem Solutions Employed for Verification WECAN Computer Program.

13 WCAP-9945, 1981. Altman, D.A. et al., Verification of Upper Head Injection Reactor Vessel Internals by Preoperational Tests on the Sequoyah Unit 1 Power Plant.

3.9-91 Rev. 30

FOR FATIGUE ANALYSIS OF RCP AND ASSOCIATED CLASS 1 PIPING SWEC (MPS-3)

SWEC (MPS-3) Piping Stress Piping System MPS-3 Piping Design Transient Analysis Problem Description Drawing Series Report Number actor coolant loops EP 70 TR 2658-2 7000 cluding surge 7001 achments to coolant 7002 ps) 7003 sidual heat removal EP 71 TR 2658-3 7000 ing 7001 emical volume and EP 74 and part of TR 2658-4 7419 ntrol system EP107 & 108 7420 7422 7423 7425 7427 w pressure safety EP 82 TR 2658-5 7000 ection system 7001 7002 70013 nulus piping EP 107 TR 2658-6 10700 ntains portions of 10701 h pressure safety 10702 ection, auxiliary 10703 ssurizer spray and 10704 emical volume and 10706 ntrol) 10707 10717 10729 bicle piping EP 108 TR 2658-7 10800 cludes letdown) 10802 10803 ssurizer spray, EP 109 TR 2658-8 10900 ety and relief 10901 3.9-92 Rev. 30

REQUIRING INELASTIC ANALYSIS Component Associated MPS-3 PPG System & Document Providing Details of Description Component Location Analysis nch Charging RC Loop 1 & Teledyne Engineering Services zzle on RC Loop 3RCS-003-149-1 EP 74B (H-8) Report TR 2658-21 nch Charging RC Loop 4 & Teledyne Engineering Services zzle on RC Loop 3RCS-003-145-1 EP 74B (E-8) Report TR 2658-21 nch Safety RC Loop 1 & Teledyne Engineering Services ection Nozzle on 3RCS-003-121-1 EP 108A (I-4) Report TR 2658-22 Loop nch Safety RC Loop 2 & Teledyne Engineering Services ection Nozzle on 3RCS-003-133-1 EP 108B (G-7) Report TR 2658-22 Loop nch Safety RC Loop 3 & Teledyne Engineering Services ection Nozzle on 3RCS-003-139-1 EP 108C (C-7) Report TR 2658-22 Loop nch Safety RC Loop 4 & Teledyne Engineering Services ection Nozzle on 3RCS-003-147-1 EP 108D (C-5) Report TR 2658-22 Loop sets of Circumferential Butt Welds (as welded) s 1 through 3 Low Pressure Safety Injection Piping SWEC Calculation to RC Loop 12179-NP(B)-315-XI 10 inch Schedule 140 Butt Welds s 4 & 5 Boron Injection System SWEC Calculation 1.5 inch Schedule 160 Butt Welds 12179-NP(B)-317-XI s 6 through 10 RCS Drains Loops 1, 2, 3 & 4 SWEC Calculation 2 inch Schedule 160 Butt Welds 12179-NP(B)-299-XI 11 Pressurizer Spray Nozzle C-1 Weld SWEC Calculation 3RCS*TK1 12179-NP (F)-316-XI 3RCS-004-224-1 12 Bimetallic Welds between Steam SWEC Calculation Generator and RCS Loop 12179-NP(B)-4019-XI 3.9-93 Rev. 30

Types of Tests (2)

Reg. Guide stem 1.68, Rev. 2 Thermal Transient Steady State ode System Title Classification (3) Expansion Vibrations Vibrations S Primary Grade Water A, D NR NR (4) V (5)

S Instrument Air A, D NR NR NR N Nitrogen System A, B NR NR V E Charging Pump A NR NR V Cooling P Service Water A, D NR NR (6) V TC Water Treating - A NR NR NR Chlorination S Chemical and Volume A, B, A1 V & I (7) V V Control P Reactor Plant A, D NR NR V Component Cooling F Emergency Diesel A, A1 NR NR V Fuel S Chilled Water A, D NR NR V A Air Startup A V NR V Emergency Diesel D Emergency Generator A V NR V Exhaust and Combustion Air S Emergency Diesel A NR NR V Jacket and Intercooler Water SS Main Steam A, B V&I V&I V M Turbine Plant A V NR V Miscellaneous Drains S Quench Spray A, D NR NR V (8)

I Safety Injection A, A1 NR NR V Pump Cooling 3.9-94 Rev. 30

Types of Tests (2)

Reg. Guide stem 1.68, Rev. 2 Thermal Transient Steady State (3) ode System Title Classification Expansion Vibrations Vibrations C Air Conditioning - A NR NR NR Control Building K Chilled Water - A NR NR V Control Building H Safety Injection - A, B NR V V High Pressure S Feedwater A, B V&I V&I V A Auxiliary Feedwater A, B, A1 NR NR V and Recirculation S Residual Heat A, B, D V&I NR V Removal L Safety Injection - A, B V&I NR V Low Pressure S Reactor Coolant Main A V&I V&I V Loops G Steam Generator A, B V&I V V Blowdown S Reactor Plant Aerated A, D V NR V Drains F Steam Generator - A NR NR V Chemical Feed S Containment A, B, C NR NR V (8)

Recirculation Spray R Sampling System - A V NR NR Reactor Plant U Ventilation - A NR NR NR Containment Structure C Fuel Pool Cooling A, D NR NR V and Purification 3.9-95 Rev. 30

Types of Tests (2)

Reg. Guide stem 1.68, Rev. 2 Thermal Transient Steady State (3) ode System Title Classification Expansion Vibrations Vibrations S Hydrogen A, A1 NR NR V Recombiner P Sampling System - A V NR NR Post Accident WS Radioactive Gaseous A V NR NR Waste (9)

S Reactor Plant A, D V NR NR Gaseous Vents S Containment Leakage A NR NR NR Monitoring S Reactor Plant A V NR NR Hydrogenated Drains S Containment A NR NR NR Atmosphere Monitoring S Containment Vacuum A, B, D V NR V I Incore Instrument A, B NR NR NR Lines S Pressurizer Safety A, B V&I V&I V and Relief System S Pressurizer Spray A, B V&I NR V System W Fire Protection Water D NR NR NR WS/ Domestic Water/ D NR NR NR S Sanitary System S Service Air D NR NR NR Containment Service Air V Main Steam Safety B V NR NR Valve Steam Vents and Drains 3.9-96 Rev. 30

Types of Tests (2)

Reg. Guide stem 1.68, Rev. 2 Thermal Transient Steady State (3) ode System Title Classification Expansion Vibrations Vibrations SS Radioactive Solid D NR NR NR Waste S Nuclear Aerated D NR NR NR Vents S Radioactive Liquid B V NR NR Waste S Boron Recovery B, D V NR V S Neutron Shield Tank A NR NR NR Cooling System S Auxiliary Steam B V NR V TES:

The detailed test plans shall identify those portions of each system to be tested.

Type of tests reflect the graded approach.

NRC Regulatory Guide 1.68, Revision 2 Classifications:

A= ASME III, High Energy Piping, Classes 1, 2, and 3.

A1= ASME III, Moderate Energy Piping, Classes 1, 2, and 3.

B = Other ASME III, High Energy Piping Inside Seismic Category I Structure.

C = Other Non-ASME III, High Energy Outside Seismic Category I Structures whose failure could reduce the functioning of any seismic Category I plant feature to any unacceptable level. Since there are no seismic category plant features outside Seismic Category I Structures, this classification does not exist for Millstone 3.

D = Seismic Category I portions of Moderate Energy Piping located outside containment structures.

NR = Testing not specifically required. Temperature < 200°F and/or significant vibrations are not expected; therefore testing testing will not be practical. Systems in this category will be observed by testing personnel for any evidence of concern for steady-state vibration.

V = Visual. When V appears in the table, hand held instruments may be required to perform observations.

I = Instrumented measurements. When V & I appears in the table, only a portion of the system requires instrument observations.

3.9-97 Rev. 30

startup testing program.

Only that portion of system isolatable from the spray rings.

Radioactive Gaseous Waste (GWS) System is non-seismic for design and analysis purposes.

3.9-98 Rev. 30

TABLE 3.9B-4 OMITTED 3.9-99 Rev. 30

(ELASTIC ANALYSIS)

Pressure Vessels, Pumps and Valve Bodies - Pressure Boundary - Designed by Analysis (1)

Reference Paragraph Condition of ASME Section Design III Primary Stress Limits Secondary Stress Limits Peak Stress Limits Pm PL PL + Pb Pe PL + Pb + Pe + Q PL + Pb + Pe + Q +

Design (2) NB-3221 Sm 1.5 Sm 1.5 Sm Not Required Not Required Normal (3) NB-3222 (4) (4) (4) 3 Sm 3 Sm Sa NB-3223 (4) (4) (4) 3 Sm 3 Sm Sa Upset (3)

Emergency NB-3224 Greater of Greater of Greater 1.8Sm Not Required Not Required (3) (5) 1.2Sm or 1.8Sm or or 1.5Sy 1.0Sy 1.5Sy Faulted (3) (4) NB-3225, Lesser of Lesser of Lesser of Not Required Not Required 2.4Sm or 3.6Sm or 3.6Sm or NB-3221, 0.7Su 1.05Su 1.05Su App. F F 1323.1 Testing (6) NB-3226 0.90 Sy 1.35Sy Not Required Not Required NOTES:

1. The nomenclature, conditions, and applications of the above allowables are in accordance with ASME Section III. Stress limits above apply to design by elastic analysis. Limit and plastic analysis is allowed in accordance with ASME Section III criteria.

Special stress limits of NB 3227 as applicable. Stress limits of Subsection NF are used for the design of supports as applicable.

3.9-100 Rev

Use operating loads.

Primary stresses often evaluated and combined with secondary effects.

Use above limits for materials of Table 1-1.2 (ASME Section III). Use 0.7Su for materials of Table 1-1.1 (ASME Section III).

Use test loads (pressure, temperature).

3.9-101 Rev. 30

1.48 VS. TABLES 3.9B-5 AND 3.9B-10 Normal or Normal + Comparison Upset + Faulted + Regulatory With Regulatory Component OBE Emergency SSE Position Position ssels(1) NB-3223 NB-3224 NB-3225 C.1 Agree e (1) NB-3654 NB-3655 NB-3656 C.1 Agree nactive Pumps, NB-3223 NB-3224 NB-3225 C.2 Agree lves (Design by alysis) (2) nactive Valves 1.1 Pr 1.2 Pr 1.5 Pr C.3 Agree esign by ndard or ernate Design les) tive Pumps, NB-3222 NB-3222 NB-3222 C.4 Alternate lves (Design by acceptable basis, alysis) (2)(3) exception to footnotes (3) and (4) tive Valves 1.0 Pr 1.0 Pr 1.0 Pr C.5 Agree with esign by pressure rating ndard or factors, exception ernate Design to footnote (3) les) (3)

TES: (from Regulatory Guide 1.48, pages 1.48-7 and -8)

Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code uding 1972 Winter Addenda thereto.

The requirements of Case 1552 (Interpretations of ASME Boiler and Pressure Vessel Code) uld be met for all sizes of Code Class 1 valves designed by analysis.

In addition to compliance with design limits specified, assurance of operability under all gn loading combinations should be provided by an appropriate combination of the following gested measures:

itu testing (e.g., preoperational testing after the component is installed in the plant)

(a) (a) Full-scale prototype testing 3.9-102 Rev. 30

(c) (c) Detailed stress and deformation analyses (includes experimental stress and deformation analyses)

In the performance of tests or analyses to demonstrate operability, the structural interaction of the entire assembly (e.g., the valve-operator assembly and pump-motor assembly) should be considered. If superposition of test results for other than the combined loading condition is proposed, the applicability of such a procedure should be demonstrated. The design limits for nonactive pumps and valves designed by analysis for applicable loading combinations if assurance is provided by detailed stress and deformation analyses that operability is not impaired when designed to these limits.

Similarly, the primary-pressure ratings P for nonactive valves designed by standard or alternative design rules may be used for the applicable loading combinations if appropriate testing demonstrates that operability is not impaired when the valve is so rated.

Secondary affects (stresses and deformations) should be evaluated for the loading binations designated by Regulatory Positions 4.a.(2) and 4.a.(3). Local affects (peak stresses) d not be considered for these loading combinations.

Applies to all components (vessels, piping, pumps, and valves) that are relied upon to cope h the effects of specified plant conditions.

Identification of the specific transients or events to be considered under each plant condition be addressed in a future regulatory guide.

The provisions of NB-3411 and NB-3413 may be applied for all sizes of Code Class 1 pumps gned by analysis.

Table 1.3-0, Permanent Strain Limiting Factors of Appendix I of the ASME Boiler and ssure Vessel Code,Section III, may be used as an aid in determining the relationship between gn stress and deformation (see Note 2 to Table I-1.2 of Section III of the ASME Code).

3.9-103 Rev. 30

COMPONENTS (ELASTIC ANALYSIS)

Primary Stress Limits (1)

Membrane Plus Design Condition ASME III Class Code Membrane (Pm) Bending (Pm + Pb) ssure Vessels I 2(NC3300) or 1.1 S 1.65 S II 3(ND3300) 2.0 S 2.40 S I (2) 2(NC3200) 1.1 Sm 1.65 Sm II (3) 2.0 Sm 2.40 Sm mps (4) (5), Inactive 2(NC3400) or 1.1 S 1.65 S 3(ND3400) 2.0 S 2.40 S mps (4)(5), Active 2(NC3400) or 1.0 S 1.50 S 3(ND3400) 1.2 S 1.80 S lves (5)(6), Active and ctive 2(NC3500) or 1.1 S 1.65 S 3(ND3500) 2.0 S 2.40 S nks (5) (Steel) 2(NC38-3900) or 1.1 S 1.65 S 3(ND38-3900) 2.0 S 2.40 S TES:

S - Allowable stress values at design temperature from ASME Section III, Appendix I, as allowed by class Sm - Design stress intensity values at design temperature from ASME Section III, Appendix I, as allowed by class Fatigue analysis may be required with operating conditions, reference paragraph NC-3219 and Appendix XIV of ASME Section III, Subsection NC.

When a complete analysis is performed in accordance with NC 3211.1(c), the faulted stress limits of Appendix F shall apply.

3.9-104 Rev. 30

Stress limits of ASME Section III, Subsection NF, are used for the design of supports as applicable.

The standard or alternative design rules of NC-3500 and ND-3500 may be used in conjunction with the stress limits specified.

Valve nozzle (piping load) stress analysis is not required when both the following conditions are satisfied by calculation:

(a) Section modulus and area at the plane normal to the flow passage through the region at the valve body crotch is at least 110 percent of that for the piping connected (or joined) to the valve body inlet and outlet nozzles; and, (b) Code allowable stress, Svlv, for valve body material, is equal to or greater than the code allowable stress, Spip, of connected piping material. If the valve body material allowable stress is less than that of the connected piping, the valve section modulus and area as calculated in (a) above shall be multiplied by the ratio of the allowable stress for the pipe divided by the allowable stress of the valve.

If unable to comply with these requirements, the design by analysis procedure of NB-3545.2 is an acceptable alternative method.

Casting quality factor of 1.0 shall be used.

Design requirements listed in this table are not applicable to valve discs, stems, cast rings, or other parts of valves which are contained within the confines of the body and bonnet.

3.9-105 Rev. 30

Table 3.9B-7 Regulatory Guide 1.48 Loading Regulatory Loading Components Combinations Design Limits Position Combinations Design Limits Comparison Pm Pm(or PL)+Pb Pm Pm(or P1)+Pb Pressure Normal 1.1 S 1.65 S C.6a Design+OBE 1.1 S 1.65 S Acceptable alterna Vessels Classes Upset+OBE Code Case 1607, 1.1 S 1.65 S C.6a 2 and 3 Regulatory Guide Emergency 1.1 S 1.65 S C.6a Design+SSE 2.0 S 2.40 S 1.84 Faulted + SSE 1.5 S 2.25 S C.6b Piping Classes Normal Normal Acceptable alterna 2 and 3 Code Case 1606, Upset+OBE NC- 1.2 Sh C.8.a Upset +OBE NC- 1.2 Sh Regulatory Guide 3611.1 3611.1 1.84 (b)(4)(c) (b)(4)(c)

(b)(1) (b)(1)

Emergency Emergency NC- 1.8 Sh 3611.1 (b)(4)(c)

(b)(2)

Faulted+SSE NC- 1.8 Sh C.8.b Faulted+SSE 2.4 Sh 3611.1 (b)(4)(c)

(b)(2) 3.9-106 Rev

Table 3.9B-7 Regulatory Guide 1.48 Loading Regulatory Loading Components Combinations Design Limits Position Combinations Design Limits Comparison Pm Pm(or PL)+Pb Pm Pm(or P1)+Pb Pumps Classes Acceptable alterna 2 and 3 Code Case 1636, Regulatory Guide Inactive Normal 1.1 S 1.65 S Design+OBE 1.1 S 1.65 S 1.84 Upset+OBE 1.1 S 1.65 S C.9.a Emergency 1.1 S 1.65 S Design+SSE 2.0 S 2.40 S Faulted+SSE 1.2 S 1.8 S C.9.b Active Normal Acceptable alterna program Upset+OBE S 1.5 S C.10.a Design+OBE 1.0 S 1.5 S Emergency Faulted+SSE Design+SSE 1.2 S 1.8 S Valves Classes Acceptable alterna 2 and 3 Code Case 1635, Regulatory Guide Inactive Normal Design+OBE 1.1 S 1.65 S 1.84 Upset+OBE 1.1 Pr C.11.a Emergency Design+SSE 2.0 S 2.4 S Faulted+SSE 1.2 Pr C.11.b including Pr Active All Conditions 1.0 Pr C.12 a Same as for inactive Acceptable alterna program 3.9-107 Rev

For loadings designated in Regulatory Position 8.a(2), only Equation 9 of NC-3651 need be In addition to compliance with the design limits specified, assurance of operability under all gn loading combinations should be provided by any appropriate combination of the following gested measures:

(a) In situ testing (e.g., preoperational testing after the component is installed in the plant)

(b) Full-scale prototype testing (c) Reduced-scale prototype testing (d) Detailed stress and deformation analyses (includes experimental stress and deformation analyses) he performance of tests or analyses to demonstrate operability, the structural interaction of the re assembly (e.g., valve-operator and pump-motor assembly) should be considered. If erposition of test results for other than the combined loading condition is proposed, the licability of such a procedure should be demonstrated. The design limits for nonactive pumps alves may be used for the applicable loading combinations if appropriate analyses and/or ing confirms that operability is not impaired when designed to these limits.

3.9-108 Rev. 30

Condition of Design Plate and Shell Linear Type Component Standard Type Class 1 Class 2 Class 3 Class 1 Class 2 Class 3 Class 1 Class 2 Class 3 (NF3220) (NF3320) (NF3400) (NF3230) (NF3330) (NF3400) (NF3240) (NF3340) (NF3400)

Design NF3221 NF3321.1 NF3321.1 NF3231.1a NF3231.1a NF3231.1a NF3221 NF3321.1 NF3321.1 or or or NF3226 App. XVII App. XVII App. XVI NF3231.1a NF3231.1a NF3231.1a App. XVII App. XVII App. XVII Normal NF3222 NF3221.2 NF3321.2 NF3231.1a NE3231.1a NF3231.1a NF 3222 NF3321.2 NF3321.2 or or or NF3226 NF3321.1 NF3321.1 App. XVII App. XVII App. XVII NF3231.1a NF3231.1a NF3231.1a App. XVII App. XVII App. XVII Upset NF3223 NF3321.2 NF3321.2 NF3231.1a NF3231.1a NF3231.1a NF3223 NF3321.2 NF3321.2 or or or NF3222 NF3321.1 NF3321.1 App. XVII App. XVII App. XVII NF3231.1a NF3231.1a NF3231.1a NF3226 App. XVII App. XVII App. XVII Emergency NF3224 NF3321.3 NF3321.3 NF3231.1b NF3231.1b NF3231.1b NF3224 NF3321.3 NF3321.3 or or or NF3231.1b NF3231.1b NF3231.1b Faulted NF3225 NF3321.4 NF3321.4 NF3231.1c NF3231.1c NF3231.1c NF3231.1c NF3321.4 NF3321.4 App. App. App. App. App. or or F1320 F1370 F1370 F1370 F1370 3.9-109 Rev

Condition of Design Plate and Shell Linear Type Component Standard Type Class 1 Class 2 Class 3 Class 1 Class 2 Class 3 Class 1 Class 2 Class 3 (NF3220) (NF3320) (NF3400) (NF3230) (NF3330) (NF3400) (NF3240) (NF3340) (NF3400)

App. or NF3225 NF3231.1c NF3231.1c F1370 App. App.

F1370 F1370 NOTE:

  • The nomenclature, conditions, and applications of the above referenced allowables are in accordance with ASME III requireme Bolts are qualified in accordance with Appendix XVII as directed by NF3280.

Welded joints are designed in accordance with NF3290.

3.9-110 Rev

3 COMPONENT SUPPORTS Plant Operating Condition Load Combination rmal D + T + P (d + r) set D + T + E + P (d + r + e + a + h) ergency D + T + E' + P (d + r + e' + a' + h) ulted D + T + E' + Al + P (d + r + e' + a' + h + al) dings Applicable to Component Supports Sustained mechanical loads, including deadweight of equipment and contents Loads on supports due to thermal expansion (constraint of free-end displacement) of components Inertia effects of the OBE Inertia effects of the SSE Loads resulting from primary loop pipe rupture asymmetric pressure effects (see FSAR Section 3.9B.1.4)

Piping associated loads as follows:

d - sustained deadweight of piping contents and insulation r - loads induced on component supports due to thermal and pressure growth of piping for appropriate plant condition e - inertia effects of OBE e'- inertia effects of SSE a- loads induced in component supports due to response of civil structure for OBE (OBE anchor movement) a'- loads induced in component supports due to response of civil structure of SSE (SSE anchor movement) h- loads resulting from occasional loads other than seismic (water hammer, steam hammer, safety relief valve opening or closing, etc.) as appropriate for plant condition al- loads induced in component supports due to the effects of LOCA 3.9-111 Rev. 30

Emergency Faulted Normal and Upset Conditions Conditions Conditio Primary Stress Intensity (Equation 9) PD o Do 2.25Sm but not 3.0Sm B 1 ---------- + B 2 ------ M 1 1.5S m 2t 2l greater than 1.8Sy Primary and Secondary Stress Range (Equation 10) Po Do Do 1 N/R N/R S n = C 1 --------------- + C 2 ------- M 1 + --------------------- E T 1 2t 2l 2(1 - v)

+ C 3 E ab a T a - b T b 3S m Peak Stress Range (Equation 11) Po Do Do 1 N/R N/R S p = K 1 C 1 --------------- + K 2 C 2 ------- M 1 + --------------------- K 3 E T 1 2t 2l 2(1 - v) 1

+ K 3 C 3 E ab a T a - b T b + ------------ E T 2 1-v Thermal Expansions Range (Equation 12) Do N/R N/R S e = C 2 ------- M 1 3S m 2l Primary and Secondary Membrane, and Bending Po Do Do 1 N/R N/R C 1 --------------- + C 2 ------- M 1 + C E ab a T a - b T b 3S m Stress (Equation 13) 2t 2l 2 Alternating Stress (Equation 14) S alt = 1 / 2 K e S p

N/R N/R Usage Factor Actual No. Cycles N/R N/R U = ---------------------------------------------------------- U 1.0 Allowable No. Cycles NOTES:

Nomenclature is as described in ASME Section III, NB-3600.

3.9-112 Rev

B1 = 0.5 may be used in lieu of 1.0 for branch connections, curved pipe/elbows, and tees.

C1 and K1 indices may be derived from NUREG CR-0778, June 1979.

LOAD COMBINATIONS FOR ASME III CLASS 1 PIPING1 Plant Operating Condition Equations Load (Moment Combination) 2 Normal/Upset 9 Pd + D + E+ H + V 10 Po + T + R + H + E + A + V + L 11 Po + T + R + H + E + A + V + L 12 T+R 13 Po + D + E+ H + V + L 14 Po + T + R+ H + E + A + V + L 12a S 9 P1 + D + H + Y Emergency 9 P1 + D + Y' + E' + H + A1 Faulted Pt Test Pt + PD 3.9-113 Rev

Plant Operating Condition Equations Allowable Stress rmal/Upset 9 1.5 Sm 10 -3.0 Sm 11 -

12 3.0 Sm 13 3.0 Sm 14 -

12a 3.0Sm ergency 9 2.25 Sm or 1.8Sy ulted 9 3.0 Sm st 0.9Sy 1.35Sy OTES:

This includes the following piping for Millstone 3:

Reactor Coolant Piping (including RC bypass, surge lines).

Portions of High and Low Pressure Safety Injection System Piping.

Portions of Residual Heat Removal System Piping.

Portion of Chemical and Volume Control Piping Portions of Pressurizer Safety, Relief and Spray Piping.

See Table 3.9B-12 for Definition of Loadings.

Class 1 piping does not experience loading designated by W.

Occasional dynamic loads such as waterhammer, steamhammer, safety valve discharge, etc. are combined with seismic inertia by SRSS method.

3.9-114 Rev. 30

PER NC3650 AND ND3650 Emergency Faulted Normal and Upset Conditions Conditions Conditions stained Loads PD o ( 0.75i )M A N/R N/R quation 8) ---------- + -------------------------- S h 4t Z casional Loads PD o ( 0.75i )M A ( 0.75i )M B 1.8Sh 2.4Sh quation 9) ---------- + -------------------------- + ------------------------- 1.2S h 4t Z Z ermal N/R N/R pansion Loads Mc i ------- S A quation 10) Z F ( 1.25S c + 0.25S h )

stained and PD o ( 0.75i )M A ( i )M C N/R N/R ermal ---------- + -------------------------- + --------------- S a + S h 4t Z Z pansion Loads xpansion 11)

TE:

All nomenclature are as defined in ASME Section III NC, ND 3600.

ABLE 3.9B-11 ASME III CLASS 2 AND 3* STRESS ANALYSIS REQUIREMENTS PER NC3650 AND ND3650 LOADING COMBINATIONS FOR QUENCH SPRAY, RECIRCULATION, AND SAFETY INJECTION ant Operating NC 3600 Condition Equations Load (Moment) Combinations ** Allowable Stress rmal/Upset 8 Pd + D Sh 9 Pp + D + E + H + V + W 1.2 Sh 10* T+R+A SA 10a S 3 SC 11* Pd + D + T + R + A Sh + SA ergency 9 Pp + D + H + Y + W 1.8 Sh ulted 9 Pp + D + E' + H + Y' + W + A1 1.8 Sh 10* T + R' + A' + X SA 3.9-115 Rev. 30

LOADING COMBINATIONS FOR QUENCH SPRAY, RECIRCULATION, AND SAFETY INJECTION ant Operating NC 3600 Condition Equations Load (Moment) Combinations ** Allowable Stress 11* Pd + D + T + R' + A' + X Sh + SA st 8 Pt Sc Pt + D 0.9 Sy TES:

Either the requirements of Eq. (10) or Eq. (11) must be satisfied.

See Table 3.9B-12 for definition of loading.

Class 1 portions of the Safety Injection are analyzed in accordance with Table 3.9B.10.

In break exclusion area, sum of stresses given by Equations 9 and 10 should not exceed 0.8 (SA + 1.2 Sh). In crack exclusion area, sum of stresses given by Equations 9 and 10 should not exceed 0.4 (SA + 1.2Sh)

Occasional dynamic loads such as waterhammer, steamhammer, safety valve discharge, etc. are combined with seismic inertia by SRSS method.

ABLE 3.9B-11 ASME III CLASS 2 AND 3* STRESS ANALYSIS REQUIREMENTS PER NC3650 AND ND3650 AD COMBINATIONS FOR ASME III CLASS 2 AND 3 PIPING EXCEPT QUENCH, RECIRCULATION, AND SI Plant Operating NC 3600 Condition Equations Load (Moment) Combinations* Allowable Stress rmal/Upset 8 Pd + D Sh 9 Pp + D + E + H + V + W 1.2 Sh 10(1) T+R+A SA 10a S 3 Sc 11(1) Pd + D + T + R + A Sh + SA ergency 9 Pp + D + H + Y + W 1.8 Sh ulted 9 Pp + D + E' + H + W + A1 + Y' 2.4 Sh st 8 Pt Sc 3.9-116 Rev. 30

AD COMBINATIONS FOR ASME III CLASS 2 AND 3 PIPING EXCEPT QUENCH, RECIRCULATION, AND SI Plant Operating NC 3600 Condition Equations Load (Moment) Combinations* Allowable Stress Pt + D 0.9 Sy TE:

See Table 3.9B-12 for definition of loading.

Either the requirements of Equation 10 or 11 must be satisfied.

In break exclusion area, sum of stresses given by Equations 9 and 10 should not exceed 0.8 (SA + 1.2 Sh). In crack exclusion area, sum of stresses given by Equations 9 and 10 should not exceed 0.4 (SA + 1.2Sh).

Occasional dynamic loads such as waterhammer, steamhammer, safety valve discharge, etc. are combined with seismic inertia by SRSS method.

ABLE 3.9B-11 ASME III CLASS 2 AND 3* STRESS ANALYSIS REQUIREMENTS PER NC3650 AND ND3650 LOAD COMBINATIONS FOR ANSI B31.1 PIPING Plant Operating 104.8 Condition Equations* Load (Moment) Combinations** Allowable Stress sign 11 Pd + D Sh rmal/Upset 12 Pp + D + H + W 1.2 Sh 13 T+R SA 14 Pd + D + T + R Sh + SA 10a S 3 Sc smic

  • st 11 Pt + D Sc Pt Sc TES:

Equation numbers are from 1973 Edition ANSI B31.1, dated June 30, 1973.

Refer to Table 3.9B-12 for identification of loadings.

3.9-117 Rev. 30

3.9-118 Rev. 30 TABLE 3.9B-12 LOADINGS APPLICABLE TO PIPING SYSTEMS Sustained mechanical loads, including deadweight of piping, components, contents, and insulation.

Internal pressure loads.

Loads due to thermal expansion of the system in response to average fluid temperature.

Loads induced in the piping due to the thermal growth of equipment and/or structures to which the piping is connected as a result of plant normal or upset plant conditions.

Loads induced in the piping due to thermal growth of equipment and/or structures to which the piping is connected as a result of plant faulted plant conditions.

Inertia effects of the OBE.

Inertia effects of the SSE.

Loads induced in the piping due to response of the connected equipment and/or civil structures to the OBE (commonly referred to as OBE anchor movements).

Loads induced in the piping due to response of the connected equipment and/or civil structures to the SSE (commonly referred to as SSE movements).

Loads induced due to building settlement effects.

Loads resulting from occasional loads other than seismic. Examples of these loads would be: water hammer, steam hammer, opening and closing of safety relief valves, etc.

Local stress effects in piping and/or piping components due to sudden changes in fluid temperature. These loads are commonly referred to as thermal transient effects.

Effects of piping striking pipe (pipe whip) or whip or effects of blowdown of an adjacent system (jet impingement loads), as defined for the emergency plant condition Effects of pipe striking pipe (pipe whip) or effects of blowdown of an adjacent system (jet impingement loads), as defined for the faulted plant condition.

Loads induced by operation of the system other than those loads described above (operational vibration).

TE: These loads are evaluated during preoperational and testing phase Loads on restraints induced by blowdown and subsequent pipe response of a ruptured system for emergency plant conditions.

Loads on restraints induced by blowdown and subsequent pipe response of a ruptured system for faulted plant conditions.

Loads induced in the piping due to pressure response (growth) of the containment during a faulted plant condition.

Internal pressure loads for emergency and faulted condition as applicable.

Internal pressure due to design pressure.

Internal pressure due to range of operating pressure.

Internal pressure due to peak pressure.

Internal pressure due to test pressure.

3.9-119 Rev. 30

Loads imposed by wind, snow, or ice.

3.9-120 Rev. 30

TABLE 3.9B-13 ACTIVE PUMPS AND VALVES s table has been deleted. Refer to the Plant Design Data System (PDDS), Seismic lification Tracking System (SQT) List for details.

3.9-121 Rev. 30

TABLE 3.9B-14 OMITTED 3.9-122 Rev. 30

eak No. Location of Break Type of Break Reactor Vessel Outlet Nozzle 1/4 Double-Ended Guillotine Reactor Vessel Inlet Nozzle 1/2 Double-Ended Guillotine Steam Generator Inlet Nozzle Double-Ended Guillotine Steam Generator Outlet Nozzle Double-Ended Guillotine Reactor Coolant Pump Inlet Nozzle Double-Ended Guillotine Reactor Coolant Pump Outlet Nozzle Double-Ended Guillotine 50-Degree Elbow on the Intrados Split Flow Entrance to the 90-degree Elbow (Steam Double-Ended Guillotine Generator Side) on the Crossover Leg Residual Heat Removal/Primary Loop Guillotine (viewed from the Connection residual heat removal line)

Safety Injection/Primary Coolant Loop Guillotine (viewed from the Connection Safety Injection line)

Pressurizer Surge/Primary Coolant Loop Guillotine (viewed from the Connection pressurizer surge line)

Loop Closure Weld in Crossover Leg Double-Ended Guillotine 3.9-123 Rev. 30

SNUBBER EMBEDMENT LOADS(1) (KIPS)

Component No. (2) Normal Operation (3) (4) Faulted Condition (4) (5)

Fx V M Fx V M am Generator per Snubbers 1 or 3 668 3 59 +1153 +59 +1144

-1443 -56 -1088 2 or 4 825 7 137 +1553 +59 +1144

-1443 -56 -1088 am Generator wer Snubbers 1 106 6 117 +1375 +63 +1226

-988 -73 -1423 2 74 9 413 +969 +92 +4523

-409 -78 -3607 3 118 6 117 +988 +63 +1226

-1375 -73 -1423 4 105 6 117 +988 +63 +1226

-1375 -73 -1423 actor Coolant mp Snubbers 1 83 6 297 +1168 73 3333

-1307 2 113 9 413 +841 +89 +4362

-930 -91 -4510 3 126 6 36 +970 +75 +3695

-1391 -86 -4395 3.9-124 Rev. 30

Maximum for all loops. All loads are +/- except where indicated. In those cases +

means outward, - means inward on embedment.

Refer to Figure 3.9B-5 Includes seismic (OBE) and deadweight.

Fx = axial, V = resultant shear, M = resultant moment Includes seismic (SSE), deadweight, and pipe rupture 3.9-125 Rev. 30

COLUMN EMBEDMENT LOADS(1)(KIPS)

Component No. (1) Normal Operation (1) (2) Faulted Condition (1) (2)

Fx V M Fx V M am Generator pport Columns 5 +74 Negligible Negligible +1,197 71 1200

-501 -1383 6 +74 Negligible Negligible +1479 72 1225

-501 -1411 7 +74 Negligible Negligible +1354 67 1146

-501 -1229 8 +74 Negligible Negligible +1455 71 1208

-446 -1282 actor Coolant mp Columns 4 +96 Negligible Negligible +1996 132 3039

-363 -1578 5 +96 Negligible Negligible +1145 187 4293

-363 -2228 6 +96 Negligible Negligible +1579 120 2767

-363 -1625 TES:

Refer to Table 3.9-2 Fx = axial resultant shear (horizontal) resultant moment 3.9-126 Rev. 30

COOLANT PUMP SUPPORTS Faulted Condition Support Section Material Material Stress (2) Allowable Stress(3) Factor of Members (1) Designation (ksi) Actual Stress (ksi) (ksi) Safety Steam Generator Upper Restraint Ring A543 C1.2 86.9 Stress Ratio = 0.74 1.36 Splice Plate A543 C1.2 86.9 Stress Ratio = 0.72 1.39 Connecting Rod Assy. A487 - Gr. 10Q 88.0 Stress Ratio = 0.77 1.3 Coupling A668-72 C1. L 76.8 Stress Ratio = 0.89 1.12 Cylinder Lug. Assy. A487 - Gr. 10Q 97.0 Stress Ratio = 0.2 5.0 A668-72 Gr. M 101.0 Stress Ratio = 0.89 1.12 Wall Clevis A487 - Gr. 10Q 97.0 28.9 109.0 3.8 Connecting Rod Bolt A668 C1. M 95.8 94.3 122.0 1.36 Studs SA-193 Gr. B-16 94.0 22.7 55.9 2.5 Yoke A-543-72 C1. 2 86.9 35 57 1.6 Lower Restraint Ring Clevis A487 - Gr. 10Q 83.5 Stress Ratio = 0.53 1.88 Snubber Lug Assy. A487 - Gr. 10Q 97.0 Stress Ratio = 0.258 3.88 Extension Tube A668 C1. L 76.8 Stress Ratio = 0.76 1.32 Snubber Clevis A471 C1.5 103.4 Stress Ratio = 0.24 4.1 Wall Mount A487 - Gr. 10Q 97.1 53.3 110.0 1.98 3.9-127 Rev

Faulted Condition Support Section Material Material Stress (2) Allowable Stress(3) Factor of Members (1) Designation (ksi) Actual Stress (ksi) (ksi) Safety Vertical Column Upper Column Clevis A487 - Gr. 10Q 83.5 Stress Ratio = 0.52 1.9 Upper Column Lug A487 - Gr. 10Q 88.0 Stress Ratio = 0.78 2.28 Column Tube A668-72 C1. L 76.8 Stress Ratio = 0.47 2.1 Floor Clevis A487 - Gr. 10Q 97.0 Stress Ratio = 0.4 2.5 Lower Column Lug A487 - Gr. 10Q 97.0 Stress Ratio = 0.42 2.38 S/G Cap Screws SA-540 - B23 C1. 2 87.0 Stress Ratio = 0.3 1.8 Reactor Coolant Pump Lateral Restraint Wall Mount A487 - Gr. 10Q 97.1 55.3 109.3 1.98 Snubber Clevis A487 - Gr. 10Q 97.0 Stress Ratio = 0.6 1.7 Extension Tube A668-72 C1. L 76.8 Stress Ratio = 0.68 1.47 Pump Link A668-72 C1. M 105.2 Stress Ratio = 0.817 1.22 Pump Lug A487 - Gr. 10Q 88.0 Stress Ratio = 0.6 1.7 Snubber Clevis Pin A668-72 C1. M 101.0 58.1 110.0 1.89 3.9-128 Rev

Faulted Condition Support Section Material Material Stress (2) Allowable Stress(3) Factor of Members (1) Designation (ksi) Actual Stress (ksi) (ksi) Safety Vertical Column Upper Clevis A487 - Gr. 10Q 88.0 Stress Ratio = 0.41 2.42 Upper Lug A487 - Gr. 10Q 93.2 Stress Ratio = 0.72 1.39 Column Tube A-668-72 Gr. L 76.8 Stress Ratio = 0.73 1.37 Lower Lug A487 - Gr. 10Q 93.2 Stress Ratio = 0.72 1.39 Floor Clevis A487 - Gr. 10Q 97.0 Stress Ratio = 0.37 2.7 Tie Down Pin A471-70 C1.5 102.6 Stress Ratio = 0.753 1.32 Pin A668-72 C1.M 115.0 84.2 98.0 1.16 NOTES:

(1) Refer to Figure 5.4-12.

(2) Minimum specified yield or ultimate at temperature.

(3) Even though the allowable stresses are lower for the normal operating conditions, the corresponding design loads were significan lower, thus making the faulted condition the critical design condition.

3.9-129 Rev

Faulted Material Stress Actual Stress Allowable (1) Material Designation (2) (ksi) (ksi) Stress (ksi) Factor of Safety Support Section Members Ring Girder SA-516 - Gr. 70 36.3 Stress Ratio = 0.48 20.5 Column Pipe SA-106-C 39.2 Stress Ratio = 0.84 1.2 Upper Vertical Clevis A487 - Gr. 10Q 100. Stress Ratio = 0.87 1.15 Lower Vertical Clevis A487 - Gr. 10Q 100. 20.8 41.08 1.97 Upper Vertical Lug A487 - Gr. 10Q 100. 19.8 39.5 2.02 Horizontal Ring and Wall A487 - Gr. 10Q 100. 11.6 27.5 2.36 Clevises Lower Horizontal Wall Lugs SA-533 - Gr. A C1. 1 49. 65.9 69.8 1.06 Spherical Plate SA-516 - Gr. 70 36.8 8 44.2 5.51 Adjusting Rod SA-540-B23 - C1.5 103.6 25.3 49.7 1.98 Upper Lug Rest. A-543-72 C1. 2 92. 55.2 77.3 1.4 Pin SA-540-B23 C1.2 137. 116.4 143.7 1.23 Embedment Plate SA-516 - Gr. 70 36.8 16.5 55.2 3.34 Skirt Anchor Bolts SA-540-B23 C1.2 137 84.8 106.1 1.15 NOTES:

1 Refer to Figure 5.4-13.

2 Minimum specified yield or ultimate at temperature.

3.9-130 Rev

PRESSURIZER SAFETY VALVE SUPPORT Faulted (1)

Material Support Material Stress (3) Actual Allowable Factor of Section(2) Designation (ksi) Stress (ksi) Stress (ksi) Safety lve Support SA-537-C1.2 52.2 25. 48. 1.9 nge Weld dial Arm SA-537-C1.2 52.2 9.1 55.7 6.1 ng Girder SA-537-C1.2 52.2 40. 55.7 1.39 lumn SA-537-C1.2 47.6 Stress Ratio = 0.73 1.37 SA-540 B24 C1.2 116.1 124 135 1.1 pport Bracket SA-537 C1.2 46. 44.3 55.2 1.2 TES:

A comparison of normal, upset, and faulted loads show that the faulted loads exceed the other loading conditions by a ratio greater than 2.

Refer to Figure 5.4-15 Minimum yield or ultimate at temperature 3.9-131 Rev. 30

Normal Operating Faulted Actual Allowable Material Allowable Support Section Material Stress Stress Factor of Stress(2) Actual Stress(3) Factor Member(1) Designation (ksi) (ksi) Safety (ksi) Stress (ksi) (ksi) of Safety R.V. Vertical Restraint A668-72 Gr. N small 44.3 large 110.7 28.7 95.3 3.3 Pad R.V. Support Gibkeys A668-72 Gr. N 19.0 71.1 3.7 118.5 66.3 101.3 1.5 R.V. Support SA-533-C1.2 6.0 39.96 6.6 66.6 39.8 59.9 1.56 Gibgussets Shell SA-537-C1.2 10.9 26.7 2.45 57.5 29.1 82.25 2.8 Base Flange SA-516-70 11.14 69.9 6.3 38. 27.0 68.4 2.53 Anchor Studs SA-193-B7 14.9 62.5 4.2 105 36.6 87.5 2.39 R.V. Bolts SA-540-C1.1 small 70.2 large 127.7 49.9 98.3 1.97 Leveling Device A668-72-N 2.2 73.7 33.5 123 87.5 105 1.2 NOTES:

(1) Refer to Figures 5.4-9 and 5.4-10.

(2) Minimum yield or ultimate at temperatures.

(3) Refer to Table 5.4-18.

3.9-132 Rev

Material Material Max. Actual Allowable Stress Primary Loop Bumper Section Designation Stress (1) (ksi) Stress (ksi) (ksi) Factor of Safety Cross Over Leg Bumper Intermediate Structure ASTM 514 Steam Generator Side 86.0 69.5 70.0 1.01 Pump Side 86.0 50.5 70.0 1.38 Saddle ASTM 514 Steam Generator Side 77.0 49.3 70.0 1.42 Pump Side 77.0 56.9 70.0 1.23 Hot Leg Bumper Vertical I-Beam ASTM 588 43.8 Stress ratio = 0.91 1.1 Saddle ASTM 514 70.4 65.2 70.4 1.08 Cold Leg Bumper Key Ring ASTM 514 77.0 15.0 33.3 2.2 Key Support ASTM 514 77.0 45.8 70.0 1.53 Key Support Ring ASTM 514 77.0 55.0 70.0 1.27 Attachment Bolts ASTM 193 GR B7 125 (ult. stress) 33.6 36.17 1.08 NOTE:

(1) Minimum specified yield at temperatures.

3.9-133 Rev

TRANSIENTS rmal Conditions Occurrences Heatup and cooldown at 100°F/hr (pressurizer cooldown 200°/hr) 200 (each)

Unit loading and unloading at 5% of full power/min 13,200 (each)

Step load increase and decrease of 10% of full power 2,000 (each)

Large step load decrease with steam dump 200 Steady state fluctuations

a. Initial fluctuations 1.5 x 105
b. Random fluctuations 3.0 x 106 Feedwater cycling at hot shutdown 2,000 Not Used Unit loading and unloading between 0 and 15% of full power 500 (each)

Boron concentration equalization 26,400 Refueling 80 Reduce temperature return to power 2,000 Reactor coolant pumps startup/shutdown 3,800 Turbine roll test 20 Primary side leak test 200 Secondary side leak test 80 Tube leakage test 800 Heaters out of service

a. One heater 120
b. One bank of heaters 120 set Conditions Occurrences Loss of load, without immediate reactor trip 80 Loss of power (blackout with natural circulation in the reactor 40 coolant system)

Partial loss of flow (loss of one pump) 80 Reactor trip from full power 3.9-134 Rev. 30

a. Without cooldown 230
b. With cooldown, without safety injection 160
c. With cooldown and safety injection 10 Inadvertent reactor coolant depressurization 20 Not Used Control rod drop 80 Inadvertent emergency core cooling system actuation 60 Operating basis earthquake (20 earthquakes of 20 cycles each) 400 Excessive feedwater flow 30 Reactor Coolant System (RCS) Cold Overpressurization 10 ergency Conditions (1)

Small loss of coolant accident 5 Small steam break 5 Complete loss of flow 5 ulted Conditions (1)

Main reactor coolant pipe break (large loss of coolant accident) 1 Large steam break 1 Feedwater line break 1 Reactor coolant pump locked rotor 1 Control rod ejection 1 Steam generator tube rupture 1 Safe shutdown earthquake 1 st Conditions Occurrences Primary side hydrostatic test 10 Secondary side hydrostatic test 10 TE:

In accordance with ASME III, emergency and faulted conditions are not included in fatigue evaluation.

3.9-135 Rev. 30

COMPONENTS Condition Classification Loading Combination sign Design pressure, Design temperature, Deadweight, Operating basis earthquake rmal Normal conditions transients, Deadweight set Upset condition transients, Deadweight, Operating basis earthquake ergency Emergency condition transients, Deadweight ulted Faulted condition transients, Deadweight, Safe shutdown earthquake, Pipe rupture loads 3.9-136 Rev. 30

COMPONENTS Operating Condition assification Vessels / Tanks Piping Pumps Valves rmal ASME Section III ASME Section III ASME Section III ASME Section III NB-3000 NB-3600 NB-3400 NB-3500 set ASME Section III ASME Section III ASME Section III ASME Section III NB-3000 NB-3600 NB-3400 NB-3500 ergency ASME Section III ASME Section III ASME Section III ASME Section III NB-3000 NB-3600 NB-3400 NB-3500 ulted ASME Section III ASME Section III ASME Section III Note 1 Appendix F Appendix F Appendix F NOTE 1 TO TABLE 3.9N-3:

CLASS 1 VALVE FAULTED CONDITION CRITERIA Active Inactive Calculate Pm from para. a) Calculate Pm from para.

NB3545.1 with Internal NB3545.1 with Internal Pressure Ps = 1.25Ps Pressure Ps = 1.50Ps Pm 1.5Sm Pm 2.4Sm or 0.7Su Calculate Sn from para. b) Calculate Sn from para.

NB3545.2 with NB3545.2 with Cp = 1.5 Cp = 1.5 Ps = 1.25Ps Ps = 1.50Ps Qt2 = 0 Qt2 = 0 Ped = 1.3X value of Ped Ped = 1.3X value of Ped from equations of from equations of 3545.2(b) (1) NB3545.2(b) (1) 3.9-137 Rev. 30

Sn 3Sm Sn 3Sm Pm, Pb, Qt, Cp, Sn, and Sm as defined by Section III ASME Code.

3.9-138 Rev. 30

AND 3 COMPONENTS AND SUPPORTS Loading Combinations (1) (2) Design/Service Level Requirements Design pressure, Design temperature, Design and Normal Deadweight, Nozzle loads Normal condition pressure, Normal Normal condition metal temperature, Deadweight, Nozzle loads Upset condition pressure, Upset condition Upset metal temperature, Deadweight, Nozzle loads, Operating basis earthquake Emergency condition pressure, Emergency Emergency condition metal temperature, Deadweight, Nozzle loads Faulted condition pressure, Faulted condition Faulted metal temperature, Deadweight, Nozzle loads, Safe shutdown earthquake TES:

Temperature is used to determine allowable stress only.

Nozzle loads, pressure, and temperatures are those associated with the respective plant operating conditions (i.e., normal, upset, emergency, and faulted) as noted, for the component under consideration.

3.9-139 Rev. 30

ABLE 3.9N-5 STRESS CRITERIA FOR SAFETY CODE CLASS 2 (1) AND CLASS 3 TANKS Design/Service Level Stress Limits sign and Normal m 1.0 S (m or L) + b 1.5 S set m 1.1 S (m or L) + b 1.65 S ergency m 1.5 S (m or b) + b 1.80 S ulted m 2.0 S (m or L) + b 2.4 S TE:

Applies for tanks designed in accordance with ASME III, NC-3300.

3.9-140 Rev. 30

TABLE 3.9N-6 STRESS CRITERIA FOR ASME CODE CLASS 2 TANKS(1)

Design/Service Level Stress Limits sign and Normal Pm 1.0 Sm PL 1.5 Sm (Pm or PL) + Pb 1.5 Sm set Pm 1.1 Sm PL 1.65 Sm (Pm or PL) + Pb 1.65 Sm ergency Pm 1.2 Sm PL 1.8 Sm (Pm or PL) + Pb 1.8 Sm ulted Pm 2.0 Sm PL 3.0 Sm (Pm or PL) + Pb 3.0 Sm TE:

Applies for tanks designed in accordance with ASME III, NC-3200.

3.9-141 Rev. 30

INACTIVE PUMPS Design/Service Level Stress Limits sign/Normal The pump shall conform to the requirements of ASME Section III, NC-3400 (or ND-3400) m 1.0S and (m or L) + b <1.5S set m 1.1S (m or L) + b 1.65S ergency 1.5S (m or L) + b 1.80S ulted m 2.0S (m or m) + b 2.4S 3.9-142 Rev. 30

AND CLASS 3 ACTIVE AND INACTIVE VALVES Condition Stress Limits* Pmax**

sign and Normal Valve bodies shall conform to ASME Section III set m 1.1S 1.1 (m or L) + b 1.65S ergency m 1.5S 1.2 (m or L) + b 1.80S ulted m 2.0S 1.5 (m or L) or b 2.4S TES:

Valve nozzle (piping load) stress analysis is not required when both the following conditions are satisfied: (1) the section modulus and area of every plane, normal to the flow, through the region defined as the valve body crotch are at least 110 percent of those for the piping connected (or joined) to the valve body inlet and outlet nozzles and, (2) code allowable stress for valve body material is equal to or greater than the code allowable stress of connected piping material. If the valve body material allowable stress is than that of the connected piping, the required acceptance criteria ratio shall be 110 percent multiplied by the ratio of the pipe allowable stress to the valve allowable stress. If unable to comply with this requirement, an analysis in accordance with the design procedure for Class I valves is an acceptable alternate method.

The maximum pressure resulting from upset, emergency, or faulted conditions shall not exceed the tabulated factors listed under Pmax times the design pressure. If these pressure limits are met, the stress limits in this table are considered to be satisfied.

3.9-143 Rev. 30

TABLE 3.9N-9 STRESS CRITERIA FOR ASME CODE CLASS 2 AND 3 PIPING Section 3.9B, Table 3.9B-11 3.9-144 Rev. 30

ading Combination Design Criteria Design ASME Section III Subsection NC-3400 and ND-3400.

Normal m 1.0S m + b 1.5S Upset m 1.0S m + b 1.5S Emergency m 1.2S m + b 1.65S Faulted m 1.2S m + b 1.8S 3.9-145 Rev. 30

Post Item ANS Safety Normal LOCA Pump No. System Class Mode Mode Basis Centrifugal charging Pump Number APCH CVCS 2 ON/OFF ON Required for high head safety 1, 2 and 3 injection (*); safe shutdown.

Residual removal Pump Number 1 APRH RHRS 2 OFF ON Required for safety injection.

and 2 Safety injection Pump Number 1 APSI SIS 2 OFF ON Required for safety injection.

and 2 Boric acid transfer Pumps 1 and 2 APBA CVCS 3 ON/OFF OFF Required for reactor shutdown.

NOTES:

(*) Only two charging pumps are required for proper safety injection operation, while the third pump is considered an installed spa 3.9-146 Rev

TABLE 3.9N-12 ACTIVE VALVES s table has been deleted. Refer to the Plant Design Data System (PDDS) Seismic Qualification cking System (SQT).

3.9-147 Rev. 30

SUPPORT STRUCTURES Allowable Deflections No-Loss-of Function Component (inches) Deflections (inches) per barrel, Radial inward 4.1 8.2 per barrel, Radial outward 1.0 1.0 per package 0.10 0.15 d cluster guide tubes 1.00 1.75 3.9-148 Rev. 30

s section presents information to demonstrate that instrumentation and electrical equipment sified as Seismic Category I is capable of performing designated safety related functions in event of an earthquake. The information presented includes identification of the Category IE rumentation and electrical equipment that are within the scope of Westinghouse nuclear steam ply system (NSSS), and balance of plant (BOP) scope. The qualification criteria employed for h item of equipment, the designated safety related requirements, definition of the applicable mic environment, and documentation of the qualification process employed to demonstrate required seismic capability are described in this section.

tions whose identification numbers include the letter B contain material within the nce-of-plant (BOP) scope, while sections whose identification numbers include the letter N tain material within the nuclear steam supply system (NSSS) scope.

3.10-1 Rev. 30

methods of meeting the general requirements for seismic qualification of Category I rumentation and electrical equipment as described by General Design Criteria (GDC) 1, 2, and re described in Section 3.1. The general methods of implementing the requirements of endix B to 10 CFR Part 50 are described in Chapter 17.

mic Category I instrumentation and electrical equipment are designed to maintain the ability to:

1. Initiate a protective action during a safe shutdown earthquake (SSE)
2. Withstand seismic disturbances during post-accident operation without loss of safety function ety related instrumentation and electrical equipment is seismically qualified in accordance h general instructions for earthquake requirements (Section 3.7B.3.1). In order to prevent any at of impacting damage to Class IE equipment during seismic events, nonsafety related rumentation and electrical equipment located adjacent to Class IE equipment is also mically qualified. The earthquake requirements and qualification methods conform to those ined in IEEE Standard 344-1975, IEEE Recommended Practices for Seismic Qualifications lass IE Equipment for Nuclear Power Generating Stations, (Section 1.8, R.G. 1.100) and are greement with the recommendations of Branch Technical Position EICSB 10. Instrumentation electrical equipment are tested as individual components, either as part of a simulated ctural section or as part of a completely assembled module or unit.

B.2 METHODS AND PROCEDURES FOR QUALIFYING ELECTRICAL EQUIPMENT AND INSTRUMENTATION hods and procedures for qualifying electrical equipment and instrumentation are contained in tion 3.7B.3.1 and 3.10B.4.

response of racks, panels, cabinets, and consoles is considered in assessing the seismic ability of instrumentation and electrical equipment. Mounted equipment is qualified, as a imum, to acceleration levels consistent with those transmitted by supporting structures. A gn objective is to minimize amplification of floor acceleration by supporting members on unted equipment.

lification of seismically related equipment is accomplished by the following methods:

1. Test the equipment under simulated operating conditions
2. Analyze the equipment for verification of proper function
3. Combinations of test and analysis 3.10B-2 Rev. 30

ipment.

B.3 METHODS AND PROCEDURES OF ANALYSIS OR TESTING OF SUPPORTS OF ELECTRICAL EQUIPMENT AND INSTRUMENTATION ports for Category I electrical equipment, instrumentation, and control systems are mically qualified by the analysis and testing procedures outlined in Section 3.7B.3.1. Supports designed to withstand the combined effects of normal operating loads acting simultaneously h two orthogonal horizontal and the vertical components of earthquake loading without loss of ctional capability or structural integrity as applicable. The stress levels due to the combined ing conditions do not exceed the maximum stress levels permitted under applicable codes. If e are no applicable codes, the stress level under combined loading for 1/2 SSE does not eed 75 percent of the minimum yield strength of the material at service temperature per the TM specification. Under safe shutdown earthquake, the stress level does not exceed the ller of 100 percent of the specified minimum yield strength or 70 percent of the specified mate strength of the material at temperature per ASTM specification.

esign objective is to provide supports for electrical equipment, instrumentation, and control ems that are seismically rigid, i.e., with fundamental natural frequencies above the cutoff uency, which separates the relatively flat rigid range of a response curve from the resonant ge of the relevant amplified response spectra curves. This assures that amplification of floor elerations through supporting members to mounted equipment is minimized.

dynamic analysis method is typically used to establish support spacing for cable trays and duit. Additionally, restraints are used as necessary to limit the horizontal loads to allowable gn values established on the basis of raceway loading and unsupported span lengths. Design visions for significant differential motions between buildings are made by breaks in raceways ese relative displacements would result in unacceptable equipment or support loadings.

eu of dynamic analysis, a provision is made for the use of static analysis in qualification of le tray systems and conduit support systems. For use in the static analysis, the peak resonant values from the ARS are amplified in accordance with Section 3.7. The damping values used qualification of cable tray systems are 4 percent for 1/2 SSE and 8 percent for SSE. The ping values used for the qualification of conduit support systems are 4 percent for 1/2 SSE 7 percent for SSE.

B.4 REPLACEMENT ITEMS hodologies for demonstrating equipment seismic qualification have been established for the lementation of GDC-2. Qualification methods have changed with time, resulting in various mic FSAR commitments.

E 934-1987 specifically provides seismic guidance for replacement items and states that parts l be selected and installed as to preserve the original seismic qualification of Class 1E 3.10B-3 Rev. 30

hese documents may be utilized for plant design changes and modifications. Replacement ipment originally qualified by Westinghouse to IEEE 344-71 meet the original criteria or er IEEE 344-75 or IEEE 344-87. The use of either method will ensure the original lification basis remain unchanged. The use of seismic experience data may be employed for curement of seismically insensitive and rugged components.

ases where a replacement item is not provided with documentation of seismic qualification ch is identical to the original item, an equivalency evaluation must be performed. The ivalency evaluation must document the methodology utilized for technically evaluating the acement item and provide justification that the item meets the seismic performance uirements necessary to maintain the seismic design basis.

B.5 OPERATING LICENSE REVIEW MP3 Plant Design Data System provides the information to demonstrate proper lementation of the criteria in Section 3.10B.1 and 3.10B.4 and demonstrate adequate seismic lification of Seismic Category I instrumentation and electrical equipment. Maintenance of this is in accordance with the Standard Review Plan Section 3.10.

3.10B-4 Rev. 30

s section presents information to demonstrate that instrumentation and electrical equipment sified as Seismic Category I is capable of performing designated safety related functions in event of an earthquake. The information presented includes identification of the Category IE rumentation and electrical equipment that are within the scope of Westinghouse nuclear steam ply system (NSSS), and balance of plant (BOP) scope. The qualification criteria employed for h item of equipment, the designated safety related requirements, definition of the applicable mic environment and documentation of the qualification process employed to demonstrate the uired seismic capability are described in this section.

tions whose identification numbers include the letter B contain material within the BOP scope, le sections whose identification numbers include the letter N contain material within the SS scope.

N.1 SEISMIC QUALIFICATION CRITERIA N.1.1 Qualification Standards methods of meeting the general requirements for seismic qualification of Category I rumentation and electrical equipment as described by General Design Criteria (GDC) 1, 2, and re described in Section 3.1. The general methods of implementing the requirements of endix B to 10 CFR Part 50 are described in Chapter 17.

Commissions recommendations concerning the methods to be employed for seismic lification of electrical equipment are contained in Regulatory Guide 1.100, which endorses E-344-1975. Westinghouse meets this standard, as modified by Regulatory Guide 1.100, by er type of test, analysis, or an appropriate combination of these methods. Westinghouse meets commitment employing the methodology described in the final staff approved version of AP-8587, Rev. 2 for all Seismic Category I instrumentation and electrical equipment.

N.1.2 Performance Requirements for Seismic Qualification ipment Qualification Data Packages (WCAP-8587) contains an equipment qualification data kage (EQDP) for every item of instrumentation and electrical equipment classified as Seismic egory I within the Westinghouse NSSS scope of supply. Table 3.10N-1 identifies the Category uipment supplied by Westinghouse for this application and references the applicable EQDP tained in Supplement 1 of WCAP-8587. Each EQDP in Supplement 1 contains a section tled Performance Specification. This specification establishes the safety related functional uirements of the equipment to be demonstrated during and after a seismic event. The required onse spectrum (RSS) employed by Westinghouse for generic seismic qualification is also tified in the specification, as applicable. The spectra employed have been selected to elope the plant specific spectra defined in Section 3.7.

3.10N-5 Rev. 30

mic qualification must demonstrate that Category I instrumentation and electrical equipment apable of performing designated safety related functions during and after an earthquake of nitude up to and including the safe shutdown earthquake (SSE). Any spurious actuation must result in consequences adverse to safety. The qualification must also demonstrate the ctural integrity of mechanical supports and structures at the operating basis earthquake (OBE)

l. Some permanent mechanical deformation of supports and structures is acceptable at the level provided that the ability to perform the designated safety related functions is not aired.

N.2 METHODS AND PROCEDURES FOR QUALIFYING ELECTRICAL EQUIPMENT AND INSTRUMENTATION ccordance with IEEE 344-1975, seismic qualification of safety related electrical equipment is onstrated by either type testing, analysis, or a combination of these methods. The choice of lification method employed by Westinghouse for a particular item of equipment is based upon y factors including practicability, complexity of equipment, economics, availability of vious seismic qualification to earlier standards, etc. The qualification method employed for a icular item of equipment is identified in the individual equipment qualification data packages DPs) of equipment qualification data packages (WCAP-8587).

N.2.1 Seismic Qualification by Type Test m 1969 to mid-1974 Westinghouse seismic test procedures employed single axis sine-beat uts in accordance with IEEE 344-1971 to seismically qualify equipment. The input form cted by Westinghouse was chosen following an investigation of building responses to seismic nts as reported in WCAP-7558 (1971). In addition, Westinghouse has conducted seismic sting of certain items of equipment as part of the Supplemental Qualification Program

-CE-692). This retesting was performed at the request of the NRC staff on agreed upon cted items of equipment employing multi-frequency, multi-axis test inputs (WCAP-8695,

5) to demonstrate the conservatism of the original sine-beat test method with respect to the dified methods of testing for complex equipment recommended by IEEE 344-1975.

original single axis sine-beat testing (WCAP-7821, 1971) and the additional retesting pleted under the Supplemental Test Program has been the subject of generic review by the

f. For equipment which has been previously qualified by the single axis sine-beat method and uded in the NRC seismic audit and, where required by the staff, the Supplemental lification Program, no additional qualification testing is required to demonstrate acceptability EEE 344-1975 since the Westinghouse aging evaluation program was implemented to rmine aging effects on complex electronic equipment located outside containment. This ipment is identified in WCAP-8587, Rev. 2 (1979), Table 7.1, and the test results in the licable EQDPs of WCAP-8587. Subprogram C of the Westinghouse aging evaluation program pendix B, WCAP-8587) has incorporated a representative sample of components from these ems which use complex electronic equipment. This program is completed and is reported in AP-8687, Supplement 2, Appendix A2. Subprogram C demonstrates that during the qualified 3.10N-6 Rev. 30

equipment tests after July 1974 (i.e., new designs, equipment not previously qualified, or viously qualified equipment that does not meet 1, 2, and 3 above), seismic qualification by test erformed in accordance with IEEE 344-1975. Where testing is utilized, multi-frequency, ti-axis inputs are developed by the general procedures outlined in WCAP-8695 (1975). The results contained in the individual EQDPs of WCAP-8587 demonstrate that the measured test onse spectrum envelopes the applicable required response spectrum (RSS) defined for generic ing as specified in Section 1 of the EQDP (WCAP-8587). Qualification for plant specific use stablished by verification that the generic RRS specified by Westinghouse envelopes the licable plant specific response spectrum. Alternative test methods, such as single frequency, le axis inputs, are used in selected cases as permitted by IEEE 344-1975 and Regulatory de 1.100.

N.2.2 Seismic Qualification by Analysis structural integrity of safety related motors (Table 3.10N-1 EQDP-AE-2 and AE-3) is onstrated by a static seismic analysis in accordance with IEEE 344-1975, with justification.

uld analysis fail to show the resonant frequency to be significantly greater than 33 Hz, a test is ormed to establish the motor resonant frequency. Motor operability during a seismic event is onstrated by calculating critical deflections, loads, and stresses under various combinations of mic, gravitational, and operational loads. The worst case (maximum) values calculated are lated against the allowable values. On combining these stresses, the most unfavorable sibilities are considered in the following areas:

1. Maximum rotor deflection
2. Maximum shaft stresses
3. Maximum bearing load and shaft slope at the bearings
4. Maximum stresses in the stator core welds
5. Maximum stresses in the stator core to frame welds
6. Maximum stresses in the motor mounting bolts
7. Maximum stress in the motor feet ere minor differences exist between items of equipment, analysis is employed to demonstrate the test results obtained for one piece of equipment are equally applicable to a similar piece of ipment.

analytical models employed and the results of the analysis are described in Section 4 of the ipment Qualification Data Packages (WCAP-8587).

3.10N-7 Rev. 30

ere supports for the electrical equipment and instrumentation are within the Westinghouse SS scope of supply, the seismic qualification tests and/or analyses are conducted including the plied supports. The EQDPs contained in WCAP-8587 identify the equipment mounting loyed for qualification purposes and establish interface requirements for the equipment to ure that subsequent inplant installation does not affect the qualification established by tinghouse.

N.4 OPERATING LICENSE REVIEW results of tests and analyses that ensure that the criteria established in Section 3.10N.1 have n satisfied employing the qualification methods described in Section 3.10N.2 and 3.10N.3 are uded in the individual EQDPs contained in WCAP-8587.

3.10N-8 Rev. 30

lacement items are procured and documented as shown in Section 3.10B.4.

.2 REFERENCES FOR SECTION 3.10 0-1 NS-CE-692. Letter of July 10, 1975 from C. Eicheldinger (Westinghouse) to D.B. Vasselo (NRC).

-2 WCAP-7536-L (Proprietary) and WCAP-7821 (Non-proprietary), 1971 plus Supplements 1-6, Seismic Testing of Electrical and Control Equipment (High Seismic Plants),

Povochnik, L.M. et al.

-3 WCAP-7558, 1971. Seismic Vibration Testing with Sine Beats. Morrone, A.

-4 WCAP-8587, Revision 2. 1979, Methodology for Qualifying Westinghouse WRD Supplied NSSS Safety Related Electrical Equipment, Butterworth, G. and Miller, R.B.

-5 WCAP-8587, 1978, Equipment Qualification Data Packages, Supplement 1 1978.

-6 WCAP-8634 (Proprietary) 1975 and WCAP-8695 (Non-proprietary), General Method of Developing Multi-Frequency Biaxial Test Inputs for Bistables, Jarecki, S.J.

-7 EPRI NP-7484: Guideline for Seismic Technical Evaluation of Replacement Items for Nuclear Power Plants.

-8 EPRI TR-104871: Generic Seismic Technical Evaluations of Replacement Items for Nuclear Power Plants.

3.10-9 Rev. 30

EQUIPMENT IN WESTINGHOUSE NSSS SCOPE OF SUPPLY Equipment EQD (1) fety-Related Valve Electric Motor Operators EQDP-HE-4 rrett Power-Operated Relief Valves (PORV) EQDP-HE-9 lenoid-Operated Isolation Valve EQDP-HE-10A ctronic Control Module EQDP-HE-10B dulating Valve EQDP-HE-10C rge Pump Motors (Outside Containment) EQDP-AE-2 nned Pump Motors (Outside Containment) EQDP-AE-3 ssure Transmitters EQDP-ESE-1B ferential Pressure Transmitters EQDP-ESE-3B sistance Temperature Detectors EQDP-ESE-5 and 6 core Neutron Detectors EQDP-ESE-8 and 9 clear Instrumentation System (NIS) EQDP-ESE-10 S Source Range Preamplifier EQDP-ESE-11 erator Interface Modules (OIMs) EQDP-ESE-12A cess Protection Sets EQDP-ESE-13 icators, Post-Accident Monitoring EQDP-ESE-14 corders, Post-Accident Monitoring EQDP-ESE-15 lid-State Protection System and Safeguard Test Cabinet (2 Train) EQDP-ESE-16 actor Trip Switchgear EQDP-ESE-20 op Stop Valve Cabinet EQDP-ESE-23 actor Coolant Pump Speed Sensor EQDP-ESE-24 ferential Pressure Indicating Switch (Group B) EQDP-ESE-40 ore Thermocouple EQDP-ESE-43A ermocouple Connectors and Thermocouple Splice WCAP-10919 ore Thermocouple Reference Junction Box Splice EQDP-ESE-43C ore Thermocouple Reference Junction Box EQDP-ESE-44A Equipment Qualification Data Package 3.10-10 Rev. 30

ctrical equipment qualification is an integral part in the design, construction, and operation of lstone Unit 3. The U.S. Nuclear Regulatory Commissions regulations in 10 CFR Part 50, mestic Licensing of Production and Utilization Facilities require that categories of ctures, systems, and components be designed to accommodate the effects of both normal and dent plant environmental conditions, and that control measures be employed to ensure the quacy of design. Specific requirements pertaining to environmental qualification of certain gories of electrical equipment are embodied in Section 50.49 of 10 CFR Part 50, vironmental Qualification of Electrical Equipment Important to Safety for Nuclear Power nts.

Millstone 3 Electrical Equipment Qualification (EEQ) Program complies with 10 CFR 50.49.

EEQ Program ensures the continued qualification of equipment that must function during and owing design conditions postulated for design basis accidents and the post-accident duration.

constituent parts of the EEQ Program include the program basis, verification of equipment rability during and following exposure to plant environmental conditions, and proper allation and maintenance of equipment in the plant. These elements are controlled through a of administrative documents consisting of a program description, implementing procedures, reference documents. The procedures are retrievable through the Stations electronic cedure management system.

gram output is generated by Equipment Qualification Records. These documents provide the itable bases and evidence, which demonstrate that Millstone 3 is compliant with CFR 50.49. EQRs utilize two main sources of design input - Test Report Assessments and an ironmental Specification.

t Report Assessments (TRA) are design calculations that evaluate and summarize the ironmental qualification test report(s). The TRA documents the process of assessing a lification test report, or analysis, as acceptable qualification documentation for use in the EEQ gram.

Environmental Specification provides the environmental conditions that are required to lify electrical equipment in performing its function while exposed to normal and accident rating conditions; and provides the environmental design conditions for use as input in elopment of Equipment Qualification Records.

mic qualification of safety-related mechanical and electrical equipment is presented in tions 3.9 and 3.10, respectively.

E INFORMATION PROVIDED IN SECTIONS 3.11, 3.11B, AND 3.11N BELOW PRESENTS HISTORICAL EEQ PROGRAM DETAILS, WHICH IS NOT SUBJECT TO TURE UPDATING.

3.11-1 Rev. 30

ormal, test, accident, and post-accident environmental conditions. The information presented udes the definition of the applicable environmental parameters, and description of the lification process employed to demonstrate the required environmental capability. The seismic lification of safety-related mechanical and electrical equipment is presented in Sections 3.9 3.10, respectively.

tions whose identification numbers include the letter B contain material within the nce-of-plant (BOP) scope, while sections whose identification numbers include the letter N tain material within the nuclear steam supply system (NSSS) scope.

3.11-2 Rev. 30

ety-related equipment and components are qualified to meet their performance requirements er normal, abnormal, and accident operating conditions. Evidence of qualification is presented he form of an equipment qualification record (EQR), located in Specification

-M3-EE-0353), that reflects the equipment functional capability during the described ditions and required time frame. The EQR also demonstrates that the equipment remains in a mode after its safety functions are performed.

ce environmental conditions vary for different areas of the plant, there are several ironmental zones. The environmental conditions to which the equipment is exposed depend on environmental zone in which the equipment is located. The safety-related equipment located hin each environmental zone can be determined from a sort of the information contained in the Q Master List (located in SP-M3-EE-0353).

ineering Specification SP-M3-EE-0333 provides a listing of the worst-case environmental ditions and profile figures for various areas in the plant and a listing of the environmental es for various areas in the plant. These conditions are determined by the criteria defined in tion 3.11B.1.1.

B.1.1 Environmental Conditions ironmental design conditions are specified in Engineering Specification SP-M3-EE-0333.

se environmental conditions are listed by a system of zones, each defining a specific area or s in the plant. The environmental design conditions in each zone are given for normal, ormal, and accident conditions as applicable. A further description of each of these conditions so included; for example, a one-time accident environment due to LOCA and MSLB.

ironmental parameters include temperature, pressure, relative humidity, chemicals, spray ntial, submergence potential, accident duration, and gamma/beta (where applicable) radiation

e. Where applicable, these parameters are given in terms of a time-based profile.

Equipment Qualification Program does not establish a fixed limitation on component service to account for the aging effects of non-seismic vibration. Rather than establish a fixed tation, the program relies on two complimentary elements to address this aging mechanism.

ially, the seismic qualification process provides a level of confidence that safety related ponents possess sufficient ruggedness to withstand a vibration environment. Thereafter, ular in-service surveillance and preventative maintenance diagnostics monitor vibration avior thereby providing early warning against premature service life expiration caused by

-seismic vibration environments.

mal operating environmental conditions are defined as conditions expected during routine t operations.

ormal operating conditions are any deviations from normal conditions, but do not include dent conditions. These environmental conditions include transients caused by a fault in a 3.11B-3 Rev. 30

ident environmental conditions result from design basis accidents (DBA), as described in tion 6.2 and Chapter 15. Safety-related equipment functions as required during all normal, ormal, and accident events.

B.1.2 Equipment Identification safety-related equipment and components for each Class 1E specification, located throughout plant, are listed in the EQRs which are in separate reports located in Specification (SP-M3-0353). Each device specified is labeled, in the EQRs, with its plant identification number, tion, and environmental zone.

B.2 QUALIFICATION TESTS AND ANALYSES EQRs, listed in separate reports, detail the Millstone 3 compliance with 10 CFR 50.49 and REG-0588, Revision 1, as endorsed by Regulatory Guide 1.89, Revision 1. The requirements eneral Design Criterion 1 (GDC 1) of 10 CFR 50, Appendix A are achieved by incorporating ormance, design, construction, and testing requirements into equipment specifications, and by establishment of a system of reviews to ensure conformance with the specified requirements.

ropriate auditable records are maintained in a permanent file. Chapter 17 provides a further nition of compliance to Criterion III of Appendix B, 10 CFR 50.

tection against earthquakes, as required by GDC 2, is provided by incorporating a description eismically induced vibrations in equipment specifications and by requiring qualification in ordance with Section 3.10.

environmental requirements of GDC 4 are addressed in Section 3.11B.1. The Class 1E ipment meets the requirements of GDC 4. The equipment is designed to operate satisfactorily o fail in a safe mode. Since components are procured for their ability to withstand the ironments resulting from both abnormal events and accidents, the Millstone 3 Class 1E ipment meets the requirements of GDC 23.

C 50 requirements are achieved by analysis and testing of pressure boundary components to ure containment integrity. Inservice inspection is performed to demonstrate leaktight integrity omponents, such as seals and seats.

B.2.1 Regulatory Guides recommendations provided in the following list of regulatory guides have been included in ropriate equipment specifications. A detailed discussion on compliance with the following ulatory guides is provided in Section 1.8:

1. Regulatory Guide 1.30 - Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment 3.11B-4 Rev. 30
3. Regulatory Guide 1.63 - Electrical Penetration Assemblies in Containment Structures for Water-Cooled Nuclear Power Plants
4. Regulatory Guide 1.73 - Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power Plants
5. Regulatory Guide 1.89 - Qualification of Class 1E Equipment for Nuclear Power Plants
6. Regulatory Guide 1.131 - Qualification Tests of Electric Cables, Field Splices, and Connections for Light-Water-Cooled Nuclear Power Plants B.2.2 Safety-Related (Class 1E) Equipment and Component Qualifications ety-related equipment located in a mild environment is environmentally qualified based on the owing:
1. equipment is purchased based on the normal and abnormal environmental conditions in which the equipment is required to function;
2. a periodic maintenance, inspection, and/or replacement program based on sound engineering practice and recommendations of the equipment manufacturer which is updated as required by the results of an equipment surveillance program;
3. a periodic testing program to verify operability of safety-related equipment within its performance specification requirements; and
4. An equipment surveillance program which includes periodic inspections, analysis of equipment and component failures, and a review of the results of preventive maintenance and periodic testing programs.

ild environment is an environment that would at no time be significantly more severe than the ironment that would occur during normal plant operation or during anticipated operational urrences. This type of environment is not in the scope of the EEQ program per CFR 50.49(c)(3).

ironmental qualification of all safety-related equipment meets the requirements of IEEE Std. -1974, Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations; intent of NUREG-0588, Interim Staff Position on Environmental Qualification ctrical equipment within the scope of the EEQ Program, has a designated qualified life derived ccordance with IEEE Std. 323-1974 by type test, or combination of type test/analysis and/or umented operating experience.

3.11B-5 Rev. 30

Qualification by Type Testing or Combination of Type Test/Analysis and/or Documented Operating Experience ting is the preferred method of qualification. Analysis has been used to verify or supplement results.

vice conditions simulated during the test were reviewed to ensure that they enveloped the dent environments. The test duration and environmental parameters utilized in the test were ewed to ensure that they equaled or exceeded specified values. When tests were found to be severe than specified, a determination of the adequacy of the test is made on a case-by-case s.

test specimen model, design, and construction material are reviewed against the equipment g qualified to verify applicability of test results.

selected test sequence is reviewed and, when determined not to be in accordance with the delines of IEEE Std. 323-1974, is reevaluated for adequacy.

ts which are successful using components that had not been pre-aged are considered eptable, provided the components do not contain materials known to be susceptible to ificant degradation due to aging effects.

rational modes tested are reviewed to ensure that they are representative of the actual lication requirements as defined in the procurement documents. The length of time that each of equipment is required to operate is reviewed against test data to ensure that the ipments designated life is greater than the length of time the equipment is required to operate.

failures identified during the environmental qualification effort are reviewed and evaluated tive to their effect on the ability of the component to perform its required function. If a ponent fails at any time during the test, the applicability of the test with regard to onstrating the ability of the component to function for the entire period prior to the failure is sidered on a case-by-case basis.

ere seals are included as part of the component, the test results and conclusions include the

s. Materials used for terminating cables and similar components are addressed separately in EQR.

lification by a combination of methods (test, evaluation, analysis) is identified in the EQR. A rmination of the adequacy of the qualification methods used is made on a case-by-case basis.

3.11B-6 Rev. 30

equipment with a designated life less than design plant life, a maintenance surveillance acement program based upon test data and analysis is utilized to lengthen the qualified life of quipment to that of the plant design life. When new test data becomes available, it is used to dify the program as necessary.

Aging siderations made for materials which are susceptible to thermal, vibration, electrical, hanical, and/or radiation aging are addressed in the applicable Environmental Qualification umentation. The order of application of the simulated aging conditions is reviewed to ensure the most severe sequence is applied. Known synergistic effects, or those found during testing, investigated on a case-by-case basis.

ipment, within the EQ Program is assessed for operability at specified time intervals by ematic application of the plant periodic test program. When this program detects any ipment degradation, the equipment or component is analyzed. Any equipment or component nd to have unsatisfactory aging characteristics is reviewed to determine necessary action.

ipment or component operability is defined as the capability to perform its specified function.

Margin lification type tests are reviewed to verify that adequate margin exists between the most ere specified service conditions of the plant and the conditions used in type testing. This ounts for normal variations in commercial production of equipment and reasonable errors in ning satisfactory performance. Increased levels of testing, number of test cycles, and test ation are considered as methods of ensuring adequate margin.

Submergence ipment has been evaluated for submergence as a result of Design Basis Accidents (e.g. LOCA ELB). Where submergence is possible, it is given as an environmental parameter in endix 3B. Equipment, within the EQ Program, located in areas with submergence potential, is er qualified to operate submerged, or it is electrically isolated, is administratively controlled, s location has been verified to be above the submergence level.

Documentation ing the review of submitted qualification documentation, when it is determined that actual test are not submitted (i.e., test summaries and/or certificates of conformance only are mitted), requests for actual test data are initiated. If such data are not submitted because the are considered proprietary by the manufacturer, an audit of these data are made. A certificate onformance by itself is not acceptable unless it is accompanied by verification of test data and rmation on the qualification program.

3.11B-7 Rev. 30

conclusions reached.

B.3 QUALIFICATION TEST RESULTS s section provides a discussion of the test results associated with environmental qualification quipment.

B.3.1 Nuclear Steam Supply System Equipment Qualification Program tion 3.11N describes the Nuclear Steam Supply System Equipment (NSSS) Qualification gram.

B.3.2 Qualification of Safety-Related Equipment (BOP) not covered by Section 3.11.

lification test results and supporting documentation for safety-related, non-NSSS equipment summarized in the EQR (Section 1.7).

B.4 LOSS OF VENTILATION following design features preclude the possibility of a total system failure for ventilation ems serving areas where equipment required to function during and following a DBA is ted.

1. All HVAC systems serving these equipment areas are designed to Seismic Category I requirements (Section 9.4).
2. Sufficient redundancy in equipment and power supplied is provided so that no single active component failure can result in loss of HVAC system function.
3. Redundant HVAC systems are connected to separate and independent onsite standby power supplies to ensure system operation upon loss of offsite power (Section 8.3).
4. Failure modes for isolation valves and dampers are described in Section 9.4.

Valves or dampers required for system operation after postulated accidents fail in the safe position.

5. Equipment outside the containment building required to operate following a LOCA or a high-energy pipe break is so located that it is not exposed to resultant post-accident ambient conditions or is designed to withstand these conditions.
6. Instrumentation and controls which incorporate audible and visual alarms enable the operator to monitor the HVAC systems performances. In the event of system 3.11B-8 Rev. 30

ed upon the above features and the detailed HVAC systems evaluations in Section 9.4, only ial loss of the ventilation or air-conditioning system could occur in areas where equipment uired to function during and/or following a DBA is located. This loss would not adversely ct the availability of the safety-related equipment to function during and following a DBA.

effects of any partial loss of HVAC is reflected in the environments of Engineering cification SP-M3-EE-0333.

B.5 CHEMICAL AND RADIATION ENVIRONMENT mponents of safety-related systems and their associated instrumentation and electrical ipment located inside the containment structure and elsewhere, which are required to function ng and subsequent to an accident, are designed to operate under normal, abnormal, accident, post-accident environmental conditions that may occur at their installation locations.

B.5.1 Radiation Environment ipment is qualified for the radiation doses found in Engineering Specification SP-M3-EE-

3. These doses were calculated using source terms given in Regulatory Guides 1.7 and 1.89, NUREG 0588.

umptions include:

1. For the design basis loss of coolant accident, which completely depressurizes the primary system, 100 percent of the core noble gases, 50 percent of the halogens, and 1 percent of the solid fission products are released into the containment atmosphere. The sump water is assumed to contain 50 percent of the core halogens and 1 percent of solid fission products. The airborne source is assumed to be 100 percent noble gases and 50 percent halogens.
2. For the gamma dose from airborne activity, no credit is taken for either internal shielding or radioactivity reduction by sprays or other removal mechanisms except radioactive decay.
3. The beta dose from airborne activity inside containment is based on the infinite medium dose. No beta dose has been considered for components that are sealed in shields which would preclude penetration of the beta particles.
4. LOCA dose values are derived from integration for a period of 1 year from the commencement of the accident. For HELB and fuel handling accidents, integration is over 1 month.

3.11B-9 Rev. 30

availability.

6. Equipment located within the containment building is qualified for radiation doses comprised of post-accident airborne nuclide concentrations, airborne nuclide concentration due to primary coolant leakage inside containment during normal operations, and location dependent doses from sources containing radioactive liquid, where applicable.
7. Equipment outside the containment building is qualified to integrated doses from post-accident radiation emanating from the containment structure, emergency fluid systems containing recirculating cooling water, and radiation sources resulting from normal operations.
8. Doses due to a fuel handling accident, in addition to doses resulting from normal operation sources, have been used for qualification of equipment in the fuel building.

B.5.2 Chemical Environment iscussion of the LOCA and the source terms is given in Section 15.6. The chemical position and resulting pH of the spray water in the containment atmosphere, the liquids in the tor core, and ESF sumps are identified in Section 6.1. The pH range (4.15 - approximately of the containment sump water following a LOCA is achieved from the trisodium phosphate ecahydrate baskets located in the containment sump structure. (See Section 6.2.2.)

concentration of chemicals in the containment sprays is equivalent to or more severe than that lting from the most limiting mode of operation, including a single failure in the spray system.

3.11B-10 Rev. 30

s section presents information to demonstrate that the safety-related electrical equipment of engineered safety features and the reactor protection systems are capable of performing their gnated safety-related functions while exposed to applicable normal, abnormal, test, accident, post-accident environmental conditions. The information presented includes identification of safety-related equipment that is within the scope of the Westinghouse Nuclear Steam Supply tem (NSSS) and for each item of equipment, definition of the applicable environmental meters, and description of the qualification process employed to demonstrate the required ironmental capability. The seismic qualification of safety-related mechanical and electrical ipment is presented in Sections 3.9 and 3.10, respectively.

N.1 EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL CONDITIONS omplete list of safety-related equipment within the Westinghouse NSSS scope of supply that is uired to function during and subsequent to an accident is presented in Table 3.11N-1. The plant cific environmental parameters are discussed in Section 3.11B for normal operating and for dent conditions together with the time each item of equipment is required to perform t-accident, when applicable.

N.2 QUALIFICATION TESTS AND ANALYSIS N.2.1 Environmental Qualification Criteria methods of meeting the general requirements for environmental design and qualification of ty-related equipment as described by GDC 1, 2, 4, and 23 are described in Section 3.1.

itional specific information concerning the implementation of GDC 23 is provided in tion 7.2.2.2. The general methods of implementing the requirements of Appendix B to CFR 50 are described in Chapter 17. Regulatory Guides 1.40, 1.73, and 1.89 concerning ironmental qualification are addressed in Section 1.8.

tinghouse meets the Institute of Electrical and Electronic Engineers (IEEE) Standard

-1974, IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating ions, including IEEE Standard 323a-1975, the Nuclear Power Engineering Committee EC) Position Statement of July 24, 1975, by either type test, operating experience, analysis, or ppropriate combination of these methods. Westinghouse satisfies this commitment by loying the methodology described in the final staff approved version of WCAP-8587.

N.2.2 Performance Requirements for Environmental Qualification esponse to the NRC staff request for additional detailed information on the qualification gram, Westinghouse submitted Supplement 1 to WCAP-8587. The latest revision of this plement, Supplement 1, WCAP-8587, contains an equipment qualification data package DP) for every item of safety-related electrical equipment supplied by Westinghouse within NSSS scope of supply. Table 3.11N-1 identifies the equipment supplied by Westinghouse for application and identifies the applicable EQDP contained in Supplement 1 of WCAP-8587.

3.11N-11 Rev. 30

onstrated under normal, abnormal, test, accident, and post-accident conditions. The ironmental qualification parameters, e.g., temperature, humidity, pressure, radiation, etc, loyed by Westinghouse for generic qualification purposes are also identified in the cification, as applicable. The parameters employed have been reviewed to ensure that the plant cific conditions as described in Section 3.11B have been enveloped.

N.2.3 Methods and Procedures for Environmental Qualification basic methodology employed by Westinghouse for qualification of safety-related electrical ipment is described in WCAP-8587. Each EQDP (Supplement 1, WCAP-8587) contains a cription of the qualification program plan for that piece of equipment. Qualification may be onstrated by either type test, operating experience, analysis, or a combination of these hods.

N.3 QUALIFICATION TEST RESULTS lification test results and supporting documentation for safety-related NSSS equipment are marized in the EQDPs and their associated test reports.

N.4 LOSS OF VENTILATION s of ventilation is discussed in Section 3.11B.4.

N.5 ESTIMATED CHEMICAL AND RADIATION ENVIRONMENT plant-specific estimates of the radiation dose incurred by equipment during normal operation discussed in Section 3.11B. The estimated doses and chemical conditions following an dent are defined in Section 3.11B. The radiation and chemical environments for which the SS scope equipment is qualified are defined in the performance specification of the applicable DP contained in Supplement 1, WCAP-8587.

3.11N-12 Rev. 30

-1 WCAP-8587, Revision 2, 1979, Butterworth, G. and Miller, R.B., Methodology for Qualifying Westinghouse WRD Supplied NSSS Safety-Related Electrical Equipment.

-2 WCAP-8587, 1978, Equipment Qualification Data Packages, Supplement 1.

3.11-13 Rev. 30

SCOPE OF SUPPLY (HISTORICAL, NOT SUBJECT TO FUTURE UPDATING)

Equipment EQD (1) ternally Mounted Limit Switches EQDP-H-3 and 6 fety-Related Valve Electric Motor Operators EQDP-HE-4 rrett Power-Operated Relief Valves (PORV) EQDP-HE-9 lenoid-Operated Isolation Valve EQDP-HE-10A ctronic Control Module EQDP-HE-10B dulating Valve EQDP-HE-10C rge Pump Motors (Outside Containment) EQDP-AE-2 nned Pump Motors (Outside Containment) EQDP-AE-3 arging Pump Motors EQDP-AE-5A ssure Transmitters EQDP-ESE-1A and 1B ferential Pressure Transmitters EQDP-ESE-3B and 4 sistance Temperature Detectors (well mounted) EQDP-ESE-6 WCAP-11587 core Neutron Detectors EQDP-ESE-8A clear Instrumentation System (NIS) EQDP-ESE-10 erator Interface Modules (OIMs) EQDP-ESE-12A cess Protection Sets EQDP-ESE-13 icators, Post-Accident Monitoring EQDP-ESE-14 corders, Post-Accident Monitoring EQDP-ESE-15 lid-State Protection System and Safeguard Test Cabinet (2 Train) EQDP-ESE-16 actor Trip Switchgear EQDP-ESE-20 ssure Sensor EQDP-ESE-21 op Stop Valve Cabinet EQDP-ESE-23A actor Coolant Pump Speed Sensor EQDP-ESE-24A ferential Pressure Indicating Switch (Group B) EQDP-ESE-40A sistance Temperature Detectors (surface mounted) EQDP-ESE-42A ore Thermocouple EQDP-ESE-43A C Series Connectors and Hardline Extension Cable EQDP-ESE-43F 3.11-14 Rev. 30

Equipment EQD (1) tting Adaptor/Cable Splice Assembly EQDP-ESE-43G ermocouple Connectors and Thermocouple Splice WCAP-10919 tting Adaptor/Splice Assembly WCAP-10920 ore Thermocouple Reference Junction Box EQDP-ESE-44A OTE:

Equipment Qualification Data Package 3.11-15 Rev. 30

PENDIX 3A - COMPUTER PROGRAMS FOR DYNAMIC AND STATIC ANALYSIS OF SEISMIC CATEGORY I STRUCTURES, EQUIPMENT, AND COMPONENTS PART 1 - STRUCTURES PART 2 - EQUIPMENT AND COMPONENTS PART 3 - PIPING SYSTEMS Rev. 30

1 STRUCTURES following are the major computer programs that are used in dynamic and static analysis of mic Category 1 structures. (Note: Minor programs such as post processors and simple grams are not listed.)

1. STRUDL II - Structural Analysis Program
2. SHELL 1 - Shell Analysis
3. STRUDLSW - Structural Analysis Program
4. ASAAS - Asymmetric Stress Analysis of Axisymmetric Solids
5. TAC2D - Heat Transfer Program
6. Time History Program - Dynamic Analysis
7. PLAXLY - Finite Element Soil-Structure Analysis for Plain Strain Problems
8. ANSYS
9. MAT5 - Axisymmetric Mat Analysis
10. MEMBRANE
11. SBMMI
12. GHOSH - WILSON 1.1 STRUDL II finite element method (Cheung and Ziankiewicz 1967) provides for the solution of a wide ge of solid mechanics problems. Its implementation within the context of the STRUDL lysis facilities expands these for the treatment of plane stress, plane strain, plate bending, low shell, and three-dimensional stress analysis problems.

UDL II (MIT 1968; 1971) has been designed as a modified subsystem of the Integrated Civil ineering System (ICES), which was designed and formulated at the Massachusetts Institute of hnology, Department of Civil Engineering.

UDL II also provides a dynamic analysis capability for linear elastic structures undergoing ll displacements. Either free or forced vibrational response may be obtained and, in the later

, the forcing functions may be in the form of time histories or response spectra.

3A.1-1 Rev. 30

lacements, velocities, accelerations, and forces.

UDL II is a recognized program in the public domain. Version 2- modification 2 (June 1972)

TRUDL is used. The software system is IBM-MVT-RELEASE 20.7. The hardware figuration is IBM-370 - Model 165. The Structural Design Language Engineering Users nual. Vol 1, Frame Analysis. Department of Civil Engineering.

1.2 SHELL 1 1.2.1 General Description s program is based upon the general numerical procedure proposed by B. Budiansky and P.P.

kowski (1963) and Greenbaum (1963) to analyze a shell of revolution subjected to arbitrary ings.

s is a finite difference stress analysis computer code. It can be used to determine the forces, ments, shears, displacements, rotations, and stress in a thin shell of revolution subject to trary loads expanded in a Fourier series of up to 150 terms. Single layer shells with up to 30 ply connected branches may be analyzed. Poissons ratio may change at discontinuity points, Youngs modulus and the thermal coefficient of expansion may be different at each point. The wed types of loading include elastic restraints, pressures in three orthogonal directions, perature changes which may have a gradient through the shell thickness, and simplified input weight of the shell or earthquake forces.

equilibrium equations for a thin shell are based on Sanders linear theory (Sanders 1959).

ders equations are expanded and modified slightly to handle a broader range of problems. All inent load, stress, and deformation variables are expanded into a Fourier series. The individual rier components of stress and deflection are found separately by solution of the finite erence forms of the appropriate differential equations. The algorithm used to solve these ations is a minor modification of the Gaussian elimination method.

1.2.2 Program Verification - Thin-Wall Cylinder ng thin-walled circular cylinder is subjected to a constant internal pressure distribution. A tion of this problem may be obtained (Roark 1965).

pertinent parameters of the cylinder are presented in Table 3A.1.2-1.

following solution can be verified (Roark 1965).

2 R = PR

(3A.1.2-1)

Et 3A.1-2 Rev. 30

t cylinder is idealized by 10 elements, as shown on Figure 3A.1.2-1. Computer results are ented in Table 3A.1.2-2 along with the results obtained from Equations 3A.1.2-1 and 1.2-2. As can be seen, the computer results compare very favorably. Therefore, this problem fies the accuracy of SHELL 1.

re:

s = hoop stress R = radial stress P = pressure R = radius t = thickness E = Youngs modulus 1.3 STRUDL-SW 1.3.1 General Description STRUDL-SW computer code uses the stiffness analysis method to analyze a wide range of ctural problems. It handles two and three-dimensional trusses and frames, having linear elastic mbers and statically applied loading.

RUDL-SW has been documented by bench marking procedures against the GTSTRUDL puter code. GTSTRUDL is a recognized program in the public domain.

1.4 ASAAS (Asymmetric Stress Analysis of Axisymmetric Solids) 1.4.1 General Description s is a finite element computer code (Crose 1971). It can be used to determine stresses and lacements in arbitrary axisymmetric solids, including problems involving asymmetric hanical and thermal loads and asymmetric temperature-dependent mechanical properties.

dependent variables, including the mechanical properties, are input by Fourier series ansions of the circumferential coordinate. The mechanical loads can be surface pressures, ace shears, and nodal point forces.

explicit stiffness relations for the axisymmetric solid ring elements of the triangular cross ion are based on the classical theorem of potential energy and the assumption that, within any ment, the displacement variation in the R-Z plane is linear. All dependent variables, including material properties, are expanded into the Fourier series. The harmonics are coupled and all 3A.1-3 Rev. 30

1.4.2 Sample Program Verification - Harmonic Axisymmetric Plane Strain infinitely long, solid, circular cylinder is subjected to cos and cos 2 pressure distributions.

losed-form solution of this problem may be obtained (Love 1944).

pertinent parameters of the cylinder are presented in Table 3A.1.4-1.

following solution can be verified (Love 1944):

r r = P o -- cos + cos 2 (3A.1.4-1) a 2 2 r 2r - a -

= P o 3 -- cos + --------------- cos 2 (3A.1.4-2) a a 2

2 2 r r -a -

r = P o -- sin - ------------ sin 2 (3A.1.4-3) a a 2

U r = U r1 + U r2 2

P o ( 1 - 4 ) ( 1 + )r U r1 = ------------------------------------------------- cos 2Ea (3A.1.4-4) 1+ 3 U r2 = P o ------------ r - 2r ----------- cos 2 E 2 3a U = U1 + U2 2

P o ( 5 - 4 ) ( 1 + )r sin U1 = ------------------------------------------------------------

2Ea (3A.1.4-5) 3 1+ 2 ---- r - - r sin 2 U 2 = P o ------------ 1 - ------

E 3 a2 cylinder is idealized by 16 elements, as shown on Figure 3A.1.4-1. Computer results are icted on Figure 3A.1.4-2, along with the exact results obtained from Equations 3A.1.4-4 and 1.4-5. As can be seen, the computer results are very close to the exact results. Therefore, this blem verifies the accuracy of ASAAS for mechanical loading problems where material perties are not variable.

3A.1-4 Rev. 30

1.5.1 General Description s is a finite difference computer code (Petersen 1969) which can be used to determine dy-state and transient temperatures in two-dimensional problems. The configuration of the y to be analyzed is described in the rectangular, cylindrical, or circular (polar) coordinate em by orthogonal lines of constant coordinate called grid lines. These grid lines specify an y of nodal elements. Nodal points are defined as lying midway between the bounding grid s of these elements. A finite difference equation is formulated for each nodal point in terms of apacitance, heat generation, and heat flow paths to neighboring nodal points. The equations all the nodal points are assembled and solved using an implicit alternating gradient algorithm.

1.5.2 Program Verification ample problem is presented to compare the results from TAC2D with an analytical solution.

objective is to show that the TAC2D program yields the correct solution.

problem is to determine the transient temperature distribution in a right circular cylinder ch is initially at temperature T. At time, t = 0, the temperature at the surface is instantaneously nged to T and maintained at that value.

hematically, the problem is defined by the following equations:

2 1--- ----

d- rdT T- = - dT

+ d-------- ------- (3A.1.5-0) r dr dr 2 dt dZ 0rR (3A.1.5-1)

T(r,z,0) = T1 (3A.1.5-2)

T(R,z,t) = T2 (3A.1.5-3)

L T(r, +/- ---, t) = T 2 (3A.1.5-4)

Z re:

t = is the time, r = the radius, z = the axial coordinate, R = the outside radius of the cylinder, L = the length of the cylinder, and

= the diffusivity.

3A.1-5 Rev. 30

k-

= ----- (3A.1.5-5) c re:

k = the thermal conductivity

= the density, c = the specific heat capacity.

the specific problem analyzed, the following numerical values were used:

R = 12.0 inches L = 48.0 inches K = 20.0 Btu/hr-ft-°F c = 40.0 Btu/cu ft-°F T1 = 0.0 °F 1.5.2.1 Analytical Solution ay be shown (Carlslaw and Jaeger 1959) that the solution is:

T - T - = 1 - f ( z, t ) g ( r, t )

(3A.1.5-6)

T2 - T1 n

4 ( -1 ) - 2 z f ( z, t ) = ----

( 2n + 1 )

e - t ------- ( 2n + 1 )

2L cos ------- ( 2n + 1 ) (3A.1.5-7) 2L n=0 2

2- J o ( rm ) - mt g ( r, t ) = --- R ----------------------------

m J i ( R m )

-e (3A.1.5-8) m=1 re the m are the roots of:

J o ( R ) = 0 (3A.1.5-9) roots m of Equation 3A.1.5-9 and the Bessell functions J and (Ji) are tabulated (Jahnke et al.,

5) and need not be computed.

3A.1-6 Rev. 30

ditions, Equations 3A.1.5-3 and 3A.1.5-4.

1.5.2.2 Numerical Solution With TAC2D ross section of the problem model for TAC2D is shown on Figure 3A.1.5-1. The model nds only to the axial midplane of the cylinder where an adiabatic boundary may be specified irtue of the symmetry condition described above. The solid material is represented by one erial block. The boundary conditions on the four external boundaries are described by lants 1 through 4 (specifically, Coolant Blocks 1 through 4). The material and coolant thermal meters, as specified by the input functions, are given in Table 3A.1.5-1. All coolants have the dard specific heat of 1.0 Btu per pound-°F (Btu/lb-°F). Coolants 1 and 2, which represent the batic external boundaries, have the standard heat transfer coefficient of 10 Btu/hr-sq ft-°F and standard flow rate of 10 pounds per hour.

1.5.2.3 Comparison of TAC2D Solution with the Analytical Solution omparison of the output from the code with the series solution is shown on Figure 3A.1.5-2.

temperature-versus-time function is plotted at three representative points within the cylinder.

an be seen that the results from TAC2D are almost identical to the series solution results. The imum difference between the two sets of results is about 2°F out of a mean magnitude of

°F.

1.6 Time History Program 1.6.1 General Description Time History Program computes time history response and amplified response spectra (ARS) ny mass location of a lumped mass system due to a synthetic earthquake input. The responses computed by integration of the modal equations of the system by exact methods (Nigam and nings 1968). The programs main application is the generation of ARS used in the design of mic Category I equipment and piping.

1.6.2 Sample Problem Time History Programs solution to a test problem is substantially identical to the solution ined using STRADYNE. STRADYNE is a recognized program in the public domain. The ple problem used consists of a structure subjected to an earthquake time history record. The cture is idealized by five lumped masses interconnected by five elastic beam elements, as wn on Figure 3A.1.6-1.

k acceleration and displacement as well as the horizontal ARS at the top mass point are pared in Tables 3A.1.6-1 and 3A.1.6-2, respectively.

3A.1-7 Rev. 30

1.7.1 General Description PLAXLY program provides a numerical solution for the dynamic analysis of plane systems er general dynamic loadings. This program works with a two-dimensional plane-strain finite ment idealization of the soil structure interaction problem.

original version of PLAXLY was developed at the University of California in Berkeley (Waas 2). It was later modified and extended at Stone & Webster (SWEC) to incorporate transient mic excitations, nonlinear soil behavior, and lumped mass representations of the structures.

1.7.2 Sample Problem PLAXLY programs solution to a test problem is substantially identical to the solution ined by using the FLUSH program. FLUSH is a recognized program in the public domain.

sample problem used consists of a structure represented by five lumped masses rconnected by four elastic beam elements. This structural model is connected to a finite ment representation of the soil, as shown on Figure 3A.1.7-1.

horizontal amplified response spectra (ARS) at the top mass point are compared on ure 3A.1.7-2.

1.8 MAT5 (Circular Mat with Axisymmetric Loading) 1.8.1 General Description s program is based upon the general numerical procedures proposed by Boris N. Zhemoshkin

62) to analyze a circular plate on an elastic foundation. It is used to determine moments, ars, vertical deflections, radial displacements, tangential and radial inplane forces, plus tions of the circular plate subjected to axisymmetric loadings.

soil subgrade may be modeled as either a Winkler (1867) or a Boussinesqu (1885) type tic foundation.

1.8.2 Program Verification results of this program have been reviewed in accordance with standard review procedures in ct at the time of use and, based on the users and reviewers knowledge of plate and shell ry, have been found to be satisfactory.

1.9 ANSYS ANSYS computer program is a large-scale general purpose computer program for the tion of several classes of engineering analysis problems. ANSYS is capable of analyzing 3A.1-8 Rev. 30

matrix displacement method of analysis based upon finite element idealization is employed ughout the program.

SYS is a recognized program in the public domain.

1.10 MEMBRANE (Membrane Stress Analysis) 1.10.1 General Description s program computes membrane stresses and strains in containment structures due to dead s, internal pressure, and temperature gradients across the wall. It analyzes cylinders, cones, spherical domes which consist of a fully cracked reinforced concrete section with a steel liner.

sses are computed by shell membrane theory (Billington 1965). The program automatically siders the effect of the uplift force acting on the roof of a cone or cylinder.

1.10.2 Program Verification an example, a cylindrical shell was analyzed by the program and by a hand calculation.

le 3A.1.10-1 presents the pertinent parameters of the cylinder. The comparison of results is n in Table 3A.1.10-2.

can be seen, the computer programs results compare very favorably. This problem verifies the uracy of MEMBRANE.

1.11 SBMMI (Single Barrier Mass Missile Impact) 1.11.1 General Description s program computes the elasto-plastic structural response of a barrier due to the following type oads: (a) static loads; (b) suddenly applied constant dynamic loads which remain permanently he structure; (c) suddenly applied constant dynamic loads representing missile impact with a te force and specific momentum; and (d) suddenly applied dynamic load of zero time duration specific momentum representing missile impact. The barrier is modelled as a single barrier s and a non-linear spring, with the above loads applied. The equation of motion is integrated me assuming constant acceleration in each time step (Billington 1965).

1.11.2 Program Verification an example, the program SBMMI was run for each load case separately. The computers result e compared to those obtained from a hand calculation. The hand calculation was based on the to-plastic response charts found in Billington (1965). Table 3A.1.11-1 presents the pertinent del and load parameters. The hand and computer program results are compared in Table 3A.1-9 Rev. 30

1.12 GHOSH-WILSON 1.12.1 General Description amic Stress Analysis of Axisymmetric Structures under Arbitrary Loadings, known as the OSH-WILSON computer code, is a finite-element based computer program developed by S.

sh and E. Wilson and modified by Stone & Webster as Code ST-200.

OSH-WILSON is capable of performing static and dynamic analysis of complex ymmetric structures subjected to any arbitrary static (mechanical and temperature) and amic loading.

method used to represent the three-dimensional continuum is either as an axisymmetric thin l, a solid of revolution, or a combination of both. The arbitrary loading in the circumferential ction is represented by a fourier series, and the analysis is carried out for each term and med up for the total response.

miltons variational principle is used to derive the equation of motion. This leads to a diagonal s matrix and a stiffness matrix and load vector which is consistent with the assumed lacement field. The equations of motion are solved numerically in the time-domain by direct gration using the Wilson method.

input required by GHOSH-WILSON is a description of geometry, materials, and boundary ditions. Loadings, damping factors, and time intervals for integration should be provided for h fourier term. Additional inertias can be added at joints during a dynamic analysis.

OSH-WILSON provides time history responses of the resultant forces, moments, shears, lacements, rotations, accelerations, and stresses at each node for the dynamic analysis.

ximum responses can also be obtained for each Fourier term.

1.12.2 Program Verification ic Case ylinder is subjected to a constant internal pressure. The cylinder is modeled using the shell ment, rectangular element, and the triangular element. The solutions for all three types of ments agree very well.

amic Case ylinder simply supported at both ends is subjected to a suddenly applied load at midspan. The tion of the equations of motion is obtained by the direct integration method. The cylinder is deled using rectangular elements. The GHOSH-WILSON solution (displacement under the 3A.1-10 Rev. 30

1.13 References for Appendix 3A.1 1-1 Bathe, Klaus-Jurgen and Wilson, E. Numerical Methods in Finite Element Analysis.

Prentice-Hall, Inc., Englewood Cliffs, NJ, 1976.

1-2 FBillington, D.P. 1965. Introduction to Structural Dynamics, First Edition, McGraw-Hill.

1-3 Boussinesqu, J. 1885. Application des Potentials a etude de l'equilibre et du Mouvement des Solides Elastiques. Paris, France.

1-4 Budiansky B. and Radkowski, P.P. 1963. Numerical Analysis of Unsymmetrical Bending of Shells of Revolution. AIAA Journal, 1, 1833.

1-5 Carlslaw, H.S. and Jaeger, J.C. 1959. Conduction of Heat in Solids. Oxford at the Clarendon Press, p 227.

1-6 Cheung, Y.K. and Zienkiewicz, O.C. 1967. The Finite Element Method. McGraw-Hill Book Company, Inc., New York, N.Y.

1-7 Crose, J.G. ASAAS Asymmetric Stress Analysis of Axisymmetric Solids with Orthotropic Temperature Dependent Material Properties that Can Vary Circumferentially. Air Force Report No. SAMSO-TR-71-197, Aerospace Report No.

TR-0172 (S2816-15)-1, December 29, 1971.

1-8 Ghosh, S. and Wilson, E. Dynamic Stress Analysis of Axisymmetric Structures Under Arbitrary Loading. Report EERC-69-10, University of California at Berkley, September 1969. Modified as Stone & Webster Engineering Corporation Computer Code ST-200, September 1973.

1-9 Greenbaum, G.A. 1963. Comments on Numerical Analysis of Symmetrical Bending of Shells of Revolution. AIAA Journal 2, 1833.

1-10 Janhke, E. and Emde, F. 1945. Table of Functions. Fourth Edition, Dover Publications, New York, N.Y.

1-11 Love, A.E.H. 1944. A Treatise on the Mathematical Theory of Elasticity. Dover Publications, New York, NY.

1-12 Massachusetts Institute of Technology (MIT) 1968. ICES STRUDL II, The Structural Design Language Engineering Users Manual. Vol 1, Frame Analysis. Department of Civil Engineering.

3A.1-11 Rev. 30

Design and Analysis Facilities. Department of Civil Engineering.

1-14 Nigam, N.C. and Jennings, P.C. 1968. Digital Calculation of Response Spectrum from Strong-Motion Earthquake Records. National Science Foundation.

1-15 Peterson, J.F., 1969. TAC2D - A General Purpose Two-Dimensional Heat Transfer Computer Code. USAEC Research and Development Report, Gulf General Atomic Inc.,

GA-9262.

1-16 Roark, R.J. 1965. Formulas for Stress and Strain. Fourth Edition, McGraw-Hill Book Company, Inc., New York, N.Y. p 298.

1-17 Sanders, J.L., Jr. 1959. An Improved First Approximation Theory for Thin Shells.

NASA Technical Report R-24.

1-18 Swanson Analysis Systems, Inc. 1979. Engineering Analysis System Users Manual (ANSYS). Volumes 1 and 2.

1-19 Timashenko, S. and Goodier, J.N. Theory of Elasticity. McGraw-Hill Book Company, Inc., New York, N.Y., 1951, pp 58-60.

1-20 Timoshenko, S. and Woinowsky-Krieger, S. 1959. Theory of Plates and Shells. Chapter 9, Plates of Various Shapes. McGraw-Hill Book Company, Inc., New York, N.Y.

1-21 Ulickii, I.I. 1972. Reinforced Concrete Structures (in Russian). Kiev, Russia.

1-22 Waas, G. 1972. Linear Two-Dimensional Analysis of Soil Dynamics Problems in Semi-Infinite Layered Media. Ph.D.Thesis, University of California, Berkeley.

1-23 Winkler, E. 1867. Die Lehre Vonder Elasticat and Festigkeit. Praha, Czechoslovakia.

1-24 Zhemochkin, B. N. and Sinitsin, A. P. 1962. Practical Method in Analysis of Beams and Plates on Elastic Foundations (in Russian). Gosstroiizdat, Russia.

3A.1-12 Rev. 30

TABLE 3.A.1.2-1 THIN-WALL CYLINDER, PERTINENT PARAMETERS Dimensions and Properties Loading and Boundary Conditions

= 25 inches At z = 0 inch; Fr = M = z = 0 20 inches At z = l = 20 inches; Fr = M = Fz = 0 0.5 inch P = 75 psi 28 x 106 psi issons ratio = 0.3

= Moment on free edge

= Radial force

= Force in z - direction

= Displacment in z - direction Page 1 of 1 Rev. 30

TABLE 3.A.1.2-2 EXACT AND COMPUTER STRESSES FOR THIN-WALL CYLINDER Variable Exact Shell 1 (inch) 3.348 x 10-3 3.348 x 10-3 (psi) 3,750 3,750 Page 1 of 1 Rev. 30

TABLE 3.A.1.4-1 INFINITELY LONG SOLID CYLINDER, PERTINENT PARAMETERS Dimensions and Properties Loading and Boundary Conditions

= a Pr = Po (cos + cos 2 )

=a ro = Po sin

= 10 x 106 psi Uz = 0

= 0.25 At r = 0, Ur = 0

= 1 inch Po = 10,000 psi Page 1 of 1 Rev. 30

TABLE 3.A.1.5-1 INPUT THERMAL PARAMETER FUNCTIONS FOR TAC2D SAMPLE PROBLEM Material Thermal Parameters Coolant Thermal Parameters EC1 (X)= 40.0 H3A (X) = 1.0 x 108 OH1 (X)= 20.0 FL03A (X) = 1.0 x 108 ON1 (X)= 20.0 TIN3A (X) = 1,460 H4A (X) = 1.0 x 108 FLO4A (X) = 1.0 x 108 TIN4A (X) = 1,460 Page 1 of 1 Rev. 30

TABLE 3.A.1.6-1 PEAK ACCELERATION AND DISPLACEMENT Time History Program Stardyne Program ak Acceleration 0.922 g 0.922 g ak Displacement 0.352 in 0.352 in Page 1 of 1 Rev. 30

TABLE 3.A.1.6-2 HORIZONTAL AMPLIFIED RESPONSE SPECTRA (Two Percent Oscillator Damping)

Period (Seconds) Time History Program (g) Stardyne Program (g) 2 0.929 0.928 4 0.999 0.999 6 1.040 1.039 8 1.301 1.301 0 1.253 1.252 2 1.694 1.697 4 2.419 2.419 6 3.158 3.153 8 5.962 5.961 0 11.104 11.101 2 6.777 6.802 4 4.694 5.025 6 3.441 3.451 8 2.576 2.576 0 2.417 2.428 4 1.613 1.603 8 1.720 1.731 2 1.493 1.491 6 1.487 1.507 0 1.201 1.201 0 0.653 0.660 Page 1 of 1 Rev. 30

TABLE 3.A.1.10-1 PERTINENT PARAMETERS OF A CYLINDRICAL SHELL Input Data Parameter dius of outer layer of rebars ro = 66.5 feet dius of inner layer of rebars ri = 63.67 feet dius of liner r1 = 63 feet dius of reference surface r = 65.25 feet ight of cylinder h = 122 feet ridional steel area/unit length, outer layer Ao = 4 in2 ridional steel area/unit length, inner layer Ai = 4 in2 er area/unit length Al = 4.5 in2 cumferential steel area, outer layer Ao = 8 in2 cumferential steel area, inner layer Ai = 8 in2 ernal pressure p = 9.72 ksi ll weight per unit surface q = 0.68 ksi tal load at top w = 9726.5 k mperature increment, outer rebars To = -12°F mperature increment, inner rebars i = 27°F mperature increment, liner l = 230°F Page 1 of 1 Rev. 30

TABLE 3.A.1.10-2

SUMMARY

OF RESULTS Hand Input Data Calculation Computer Run  % Difference er strain:

Meridional 0.001308 0.001308 0 Circumferential 0.001412 0.001412 0 mbrane stresses: (ksi)

Meridional rebars:

Outer layer 41.58 41.59 0.02 Inner layer 33.975 33.99 0.02 Liner -6.9857 -6.97 0.2 Hoop rebars:

Outer layer 42.47 42.47 0 Inner layer 36.65 35.65 0 Liner -4.586 -4.586 0 mbrane Forces: (k/ft)

Meridional 271.76 271.90 0.05 Circumferential 612.32 612.36 0 Page 1 of 1 Rev. 30

TABLE 3.A.1.11-1 TEST PROBLEM DATA Load Type Load Static load -15.5 k*

Suddenly applied constant load (remains on structure permanently) 62.9 k*

Missile Impact - finite force 264 k*

specific momentum 1.4 k*/sec Suddenly applied dynamic load w/zero time duration (applied impulse) 1.2 k*/sec TES:

Equivalent barrier weight = 16.66 k Barrier resistance function - From zero displacement to a displacement of 0.003 foot, the barrier resistance increases linearly from 0 to 87.2 k. For displacements greater than 0.0003 foot, the resistance remains at a constant value of 87.2 k.

Page 1 of 1 Rev. 30

TABLE 3.A.1.11-2 A COMPARISON OF HAND AND COMPUTER PROGRAM RESULTS Results from Hand Results from ad Number Load Case Calculation Computer Run 1 Barrier deflection -0.000533 feet -0.0005 feet 2 Barrier deflection 0.0054 feet 0.0054 feet Time of maximum deflection 0.01802 sec 0.018378 sec 3 Barrier deflection 0.0183 feet 0.0187 feet Time of maximum deflection 0.0180 sec 0.018391 sec 4 Barrier deflection 0.0171 feet 0.0172 feet Time of maximum deflection 0.01481 sec 0.014419 sec Page 1 of 1 Rev. 30

following computer programs were used for the analysis of Seismic Category I equipment components, as well as for pipe rupture design and analysis.

1. ASAAS - Asymmetric Stress Analysis of Axisymmetric Solids
2. LIMITA 25 - Nonlinear Static Analysis of Plane Frames
3. MISSILE - Turbine Missile Probability Program
4. SLOSH - Simplified Tank Sloshing Analysis
5. LION - Temperature Distribution for Arbitrary Shapes/Complicated Boundary Conditions
6. LIMITA 3 - Nonlinear Dynamic Analysis of Frame Structures
7. TAC2D - Two Dimensional Thermal Analysis
8. SHELL 1 - Thin Shell of Revolution Under Arbitrary Loading
9. NOZZLE (ST-147) - Vessel Penetration Analysis
10. DINASAW - Dynamic Inelastic Nonlinear Analysis by Stone & Webster Engineering Corporation
11. LIMITA 2 D - Nonlinear Transient Dynamic Analysis
12. STARDYNE - Linear and Nonlinear Elastic Structure Analysis
13. ASYMPR - Asymmetric Pressure Force - Time History
14. LIDOP - Local Inelastic Deformation of Piping
15. DLF - Dynamic Load Factors
16. STRUDL - Structural Analysis Program. STRUDL program descriptions and verification are presented in Sections 3A.3.8 and 3A.3.9 and are not duplicated here.

3A.2-1 Rev. 30

2.1.1 General Description s is a finite element computer code (Cross 1971). It can be used to determine stresses and lacements in arbitrary axisymmetric solids, including problems involving asymmetric hanical and thermal loads and asymmetric temperature-dependent mechanical properties. All endent variables, including the mechanical properties, are input by Fourier series expansions he circumferential coordinate. The mechanical loads can be surface pressures, surface shears, nodal point forces.

explicit stiffness relations for the axisymmetric solid ring elements of the triangular cross ion are based on the classical theorem of potential energy and the assumption that, within any ment, the displacement variation in the R-Z plane is linear. All dependent variables, including material properties, are expanded into the Fourier series. The harmonics are coupled and all equilibrium equations are solved simultaneously. The algorithm used to solve the equations is ock modified square root Cholesky method with iterative refinement (Cross 1971).

2.1.2 Program Verification - Harmonic Assymetric - Plane Strain infinitely long, solid, circular cylinder is subjected to cos and cos 2 pressure distributions. A ed-form solution of this problem may be obtained (Love 1944).

pertinent parameters of the cylinder are presented in Table 3A.2.1-1.

following solution can be verified (Love 1944):

= Po(r cos + cos2) (3A.2.1-1) 2 2 2r -

= P o 3r cos + ------------------ 2

- cos 2 (3A.2.1-2) 2 2 r -

r = P o r sin = --------------- sin 2 (3A.2.1-3) 2 r

2 3

( 1 - 4Y ) ( 1 + Y )r 1+Y U r = P o --------------------------------------------- sin + ------------- (r - 2Yr

) cos (3A.2.1-4) 2E E 3 2

2 3

( 5 - 4Y ) ( 1 + Y )r 1+Y U = Po --------------------------------------------- sin + ------------- (1 -- ) ---- r - - Y sin 2 (3A.2.1-5) 2E E 3 2 cylinder is idealized by 16 elements, as shown on Figure 3A.1.4-1. Computer results are icted on Figure 3A.1.4-2, along with the exact results obtained from Equations 3A.2.1-4 and 3A.2-2 Rev. 30

perties are not variable.

2.2 Limita 25 - 2D Nonlinear Transient Dynamic Analysis 2.2.1 General Description ita 25 (ST-224) is a computer code written and fully documented by SWEC, which predicts

-dimensional structures. A plane frame is represented mathematically as a discrete system of m members. Under loadings, the equilibrium at each joint is ensured by the system ilibrium equation (Martin 1966):

[K] {q} = {F} (3A.2.2-1) re:

[K] = System stiffness matrix

{q} = Global displacement vector

{F} = External force vector element of the stiffness matrix, Kij, situated in row i and column j, is the force in the i th ree of freedom required to produce a unit displacement in the j th degree of freedom when all er degrees of freedom are restrained from moving (Martin 1966; Przemieniecki 1968).

account for nonlinear effects, such as plasticity and large deflections, Equation 3A.2.2-1 is ed by an incremental method at any particular load step, Equation (1) may be written:

[K]i {q}i = {F}i (3A.2.2-2) re:

{q}i = {q}i - {q}i=l, {F}i = {F}i - {F}i-l

{q}0 = {q}0 , {F}0 = {F}0 stiffness matrix-[K], calculated based on the deformed structure at load step i (Martin, FDL IR66-80, 1966), is assumed constant through the load step. Displacements and member es are given at load step i by:

i

{q} = q } (3A.2.2-3)

=0 3A.2-3 Rev. 30

{Q} = [ K ] { q }

=0 re:

{Q} = member force vector

[k] = member stiffness matrix

{q} = member displacement vector equilibrium equations, Equation (2), are solved by a standard elimination technique.

ce no external loading is applied to a member between joints, the maximum value of the rnal force acting on a member occurs at its ends. The transition from the elastic to the fully tic state is disregarded, and the end sections are assumed to remain linearly elastic until a fully tic state is reached. The yield surface is defined by scalar function, of the internal member es, {Q}, having the form (Hodge 1959; Neal 1961; Stockey et al., 1966):

({Q}i) = 1 btained by integrating stress across the member section with the stress fully developed over section and satisfying the von Mises (or Tresca) yield criterion:

2 + 2T2 = y2 re:

= Normal stress T = Shear stress y = Yield stress in simple tension 2 = 3 (von Mises) or 4 (Tresca) s the function depends on the shape of the cross section and the force components being sidered. For a plane frame, the yielding normally occurs due to either a predominant bending ment or a predominant axial force. Therefore, two plastic models are used.

ding Yield Model:

ce a section is either elastic or fully plastic, there are four possible states:

1. Both ends A and B are elastic 3A.2-4 Rev. 30
3. End A is elastic; end B is plastic
4. Both ends A and B are plastic lastic hinge is introduced at any end section which is yielding. The force-displacement relation he plastic hinge follows an ideal bilinear curve (Clough 1965; Guberson 1967). In situations re force reversal occurs, the elastic stiffness of the hinged member is restored, providing tic unloading (isotropic strain hardening model).

al Yield Model:

re are only two possible states:

1. The entire member is elastic
2. The entire member is plastic en the member yields, the member elastic Youngs modulus is replaced by a plastic tangent dulus and the force-displacement relation follows a bilinear curve. If the member unloads, the tic modulus is restored.

2.2.2 Program Verification enter loaded beam, built in at one end and supported at the other end, is analyzed for parison to data obtained analytically. The analytical solution was obtained using limit lysis as explained in Hodge (1959). The displacement at the point of loading calculated using ita 25 and calculated analytically preshown in Figure 3A.2.2-1. Additional problems where lyzed to ensure that all program options were exercised and thus demonstrated the function adequacy of the program.

2.3 MISSILE 2.3.1 General Description Missile program calculates the impact probability (P2) of postulated turbine missiles on cified targets. The solid angle method is used to calculate P2:

1 P 2 = -------- d m

re:

= Solid angle subtended by the target 3A.2-5 Rev. 30

integral is evaluated by numerical integration, with consideration of the missile ejection city and the relative positions of the turbine and target (Figure 3A.2.3-1).

2.3.2 High Trajectory Verification tinghouse has derived a formula to predict the probability of impact for high trajectory siles. Some adjustments to the formula are necessary to enable direct comparison with the gram results. The formula has been derived on the basis that the initial velocity is random and ormly distributed between V1 and V2. The program uses a deterministic initial velocity. The mula may be specialized to this condition by setting V1 equal to V2 after applying LHopitals

e. Also, the formula has been derived assuming a missile fragment occurs in the quadrant of target, whereas the program assumes a missile fragment can occur in any of the four drants. These differing assumptions can be reconciled by using four fragments for program ut.

er making the above adjustments, the high trajectory formula becomes:

P = G2/(2V4) re:

P = Impact probability per square foot of target G = Acceleration of gravity (ft/sec2)

= Deflection angle range (radians) sile (MA-057) is a computer code written and fully documented by SWEC for inhouse use.

2.4 SLOSH 2.4.1 General Description purpose of this program is to compute the seismically-induced liquid pressures and the imum vertical displacement of the liquid surface in a container under horizontal acceleration.

mathematical procedures and formulas used in developing the program were taken from AEC ort TID-7024. The program uses data for intensity of ground motion taken from rage-acceleration-spectrum curves, as used in the analysis from the report.

program is used for circular or rectangular, shallow or slender, ground-supported tanks and ular or rectangular, shallow (not slender) tower-supported tanks.

3A.2-6 Rev. 30

omparison of results of computer program SLOSH and those given in the AEC Report (US C 1963) shows that the program yields correct results (Table 3A.2.4-1).

OSH (ME-111) is a computer code written and fully documented by SWEC for inhouse use.

2.5 LION (ME-112)

N is a digital computer program which is used to solve three-dimensional transient and steady e temperature distribution problems. The program may also consider subcooled nucleate ing and coolant heat transfer effects. The surface conditions may be forced convection, free vection, or radiation and heat may be externally or internally generated. Input to the program sists of structural geometry, physical properties, boundary conditions, internal heat generation s, coolant flow properties, and flow rates.

program solves the transient heat conduction equations for a three-dimensional field using a forward difference method. To ensure the temperature calculation stability, LION can rmine the suitable time increment, if the specified input time increment is too large.

ce the original program (Bray 1954) was developed, subsequent versions have evolved to e larger and more complex problems (Bray 1954, 1959; Personal Communication; Briggs 3; Lechliter 1963).

N is a recognized program in the public domain and has been used extensively.

2.6 LIMITA 3 2.6.1 General Description ITA 3 (ST-225) is a computer code written and fully documented by SWEC for inhouse use.

ormulation is identical to that of Limita 2 (Paragraph 3A.2.11) in that the equations are licable to a three-dimensional problem. For a space frame, yielding normally occurs due to er a predominant bending moment or a predominant torsion. Therefore, two plastic models are vided.

1. Bending Yield Model Since a beam section is either elastic or fully plastic, there are four possible states:
a. Both ends A and B are elastic
b. End A is plastic; end B is elastic
c. End A is elastic; end B is plastic 3A.2-7 Rev. 30

lastic hinge is introduced at any end section which is yielding. The force-displacement relation he plastic hinge follows an ideal bilinear curve (Clough et al., 1965; Giberson 1967). In ations where force reversal occurs, the elastic stiffness of the hinged member is restored, viding elastic unloading (isotropic strain hardening model).

2. Torsional Yield Model There are only two possible states:
a. The entire member is elastic
b. The entire member is plastic en the member yields, the member elastic modulus is replaced by a plastic tangent modulus the force-displacement relation follows a bilinear curve. If the member unloads, the elastic dulus is restored.

2.6.2 Program Verification 2.6.2.1 Elastic Example a checkout of LIMITA 3, a space frame is considered. All members are W14x500. A step load 0 kips is applied vertically at joint 6. This problem was analyzed by LIMITA 3 and ICES UDL II elastically. The results of displacements and moment Z at joint 6 were plotted against h other and found to be in excellent agreement.

2.6.2.2 Plastic Example s example is provided to illustrate this ability of the program to determine the inelastic sient response of a three-dimensional structure. The structures considered consist of tilevered steel tubes subjected to force transients causing bending and torsion in the structures.

results obtained from an analysis using the LIMITA 3 code are compared with data obtained erimentally.

experiment was a drop test in which the cantilever tubes were loaded by weights at each tube

. The results tabulated were the peak deflections and their corresponding times and the manent deflections. These results are compared to those obtained using the LIMITA 3 code in le 3A.2.6-1.

itional problems were analyzed (elastic and inelastic) to ensure that all program options were rcised and thus demonstrate the functions and adequacy of the program.

3A.2-8 Rev. 30

2.7.1 General Description er to Appendix 3A.1, Section 3A.1.5.1.

2.7.2 Program Verification er to Appendix 3A.1, Section 3A.1.5.2.

2.7.2.1 Analytical Solution er to Appendix 3A.1, Section 3A.1.5.2.1 2.7.2.2 Numerical Solution with TAC2D er to Appendix 3A.1, Section 3A.1.5.2.2.

2.7.2.3 Comparison of TAC2D Solution with the Analytical Solution er to Appendix 3A.1, Section 3A.1.5.2.3.

2.7.2.4 References for TAC2D er to Appendix 3A.1, Section 3A.1.12.

2.8 SHELL 1 2.8.1 General Description er to Appendix 3A.1, Section 3A.1.2.1.

2.8.2 Program Verification - Thin-Wall Cylinder er to Appendix 3A.1, Section 3A.1.2.2.

2.8.3 References for SHELL 1 er to Appendix 3A.1, Section 3A.1.12.

2.9 Vessel Penetration Analysis 2.9.1 General Description vessel penetration analysis computer code (ST-147) is written and fully documented by EC for inhouse use. The code performs various analyses on tanks and pressure vessels. All of 3A.2-9 Rev. 30

1. Applied load stresses at vessel-nozzle junction for:
a. Rigid attachment to cylinder
b. Rigid attachment to sphere
c. Hollow attachment to sphere
2. Pressure discontinuity analysis for thin shell interaction
3. Allowable load functions on nozzles for each case al stresses due to nozzle loads are found by the method prescribed by P.P. Bijlaard (Wichman l., 1965). The method prescribed by Johns and Orange (Johns and Orange 1961) is used for sure discontinuity stresses.

2.9.2 Program Verification ample problem of a thin-walled cylindrical vessel is subjected to applied loads from a rigid ndrical attachment. This problem may be solved using Johns and Orange's method (Johns and nge 1961).

ummary of manual calculations were then compared with the computer summary. Additional blems were considered to ensure that all program options were exercised and thus demonstrate function and adequacy of the program.

2.10 DINASAW (Dynamic Inelastic Nonlinear Analysis by Stone and Webster) 2.10.1 General Design ASAW may be used to predict the nonlinear, dynamic behavior of plane frames (pipes, rings, eams) including large displacements, plasticity, and impacts. Arbitrary force-time relations be applied at any station. DINASAW can also be applied to pipe whip and pipe impact blems.

analysis, as derived (Wu and Witmer 1972; Collins and Witmer 1973), employs the spatial te-element method in which the tangential and normal displacement fields are represented by ic interpolations. By applying the principle of virtual work in conjunction with DAlemberts ciple, the equations of motion may be derived in the form:

[ M ] { q** } = { F } - { P } - [ H ] { q }

3A.2-10 Rev. 30

and { q** } = the generalized displacements and generalized accelerations, respectively, for the complete assembled discretized structure defined, with respect to a global coordinate system

[M] = the lumped mass matrix for the complete assembled discretized structure

{F} = the assembled vector of externally-applied loading

{P} = an assembled internal force matrix (replaces conventional stiffness matrix)

[H] {q} = generalized loads arising from both large deflection and plastic behavior 2.10.2 Program Verification o examples are discussed here. The first (Wu and Witmer 1972), involves a ring subjected to al blast wave over a portion of its circumference. The resulting deformation severely distorts ring, flattening it considerably. Still, the computer code follows very closely not only to the lacement field, but also to the strain time history.

second case (Collins and Witmer 1973) involves the impact of a rotor segment onto a ring or ud. Again, the program, in conjunction with the CIVM (Collision Imparted Velocity Method),

ows experimental results very closely.

2.11 LIMITA 2 2.11.1 General Description ITA 2 (ST-223) predicts nonlinear, dynamic behavior of plane frames, including large lacements, plasticity, and impact. A plane frame is simulated as a lumped parameter system, sisting of an assembly of discrete lumped masses connected by beam members. Under any ing, the equilibrium at the rth mass point is ensured by the equation of motion:

i i m r q** r + C ri q i + K ri q i = fT (3A.2.11-1) re:

= a series with one term for each of the i displacements Cri = the damping coefficient, which applies to the ith velocity in the rth equation of motion 3A.2-11 Rev. 30

member from moving the rth degree of freedom when the ith degree of freedom is given a unit displacement when all other degrees of freedom are restrained from moving (Martin, 1966; Przemieniecki 1968).

fT = the external load factor ake account of nonlinear effects, such as plasticity and large deflections, Equation 3A.2.2-1 is ed by an incremental method (Clough and Wilson 1962). At any particular time, t, the lacement increment is obtained from t t t i t t m r q** r + C ri q i + K ri q i S i

- (3A.2.11-2) t S s

= fr - ( K ri q i )

t=i re:

Ctri = current damping coefficient Ktri = the member stiffness ch are calculated based on the current deformed structure (Martin, 1966) and assumed stant through the time step, t.

displacement and member forces are thus given by:

S S i t s S S qr = q r Qr = ( Ki qi ) (3A.2.11-3) t t second order differential system equations (Equation 3A.1.11-2) are solved by a linear eleration implicity method (Hildebrand 1956).

ce no external loading is applied to a member between nodes, the maximum value of the rnal force acting on a member occurs at its end sections. The transition from the elastic to the y plastic state is disregarded and the end sections are assumed to remain linearly elastic up to full plastic yield surface. The yield surface is defined by a scalar function of the internal mber forces, Q, of the form (Hodge 1959; Neal 1961; Stokey et al., 1966).

(Q) = l 3A.2-12 Rev. 30

2 + Y2T2 = y2 re:

= normal stress T = shear stress y = yield stress in simple tension Y2 = 3 (von Mises) or 4 (Tresca) s, the function depends on the shape of the cross section and the force components being sidered.

a frame structure, the yielding normally occurs due to either a predominant bending moment o a predominant tension or compression. Thus, two plastic models are provided:

1. Bending predominant members Since a section is either elastic or fully plastic, there are four possible states:
a. Both ends A and B are elastic
b. End A is yielding and B is elastic
c. End A is elastic and B is yielding
d. Both ends A and B are yielding lastic hinge is introduced at any end section which is yielding. The force-displacement relation he plastic hinge follows an ideal bilinear curve (Clough et al., 1965; Giberson 1967). In ations where force reversal occurs, the stiffness of the hinged member is restored, providing oading along the elastic line (isotropic strain hardening model).
2. Tension or compression of predominant members There are only two possible states:
a. The entire member is elastic
b. The entire member is plastic 3A.2-13 Rev. 30

curve. If the member unloads, the elastic modulus is restored.

2.11.2 Program Verification EC sponsored an experimental investigation performed by the Massachusetts Institute of hnology (Wilson 1968). The problem consisted of the cantilevered pipe (Figure 3A.2.11-1) jected to an impulsive load at its free end. The impulse is imparted by the detonation of a sheet igh explosive, separated from the pipe by a buffer material. A nearly uniform initial velocity is duced in the loaded region and is determined by high speed photography.

s problem was analyzed by LIMITA 2. The results were compared with experimental data and put from another computer program, DINASAW.

itional problems were considered to ensure that all program options were exercised and thus onstrate the function and adequacy of the program.

2.12 STARDYNE STARDYNE Structural Analysis System, written by Mechanics Research, Inc., of Los eles, California, is a fully warranted and documented computer program available at Control a Corporation. The latest version of this program became available August 1, 1973.

MRI STARDYNE Analysis System consists of a series of compatible digital computer grams designed to analyze linear and nonlinear elastic structural models. The system ompasses the full range of static and dynamic analyses.

static capability includes the computation of structural deformations and member loads and sses caused by an arbitrary set of thermal, modal applied loads, and prescribed displacements.

izing the normal mode technique, linear dynamic response analyses can be performed for a e range of loading conditions, including transient, steady state harmonic, random, and shock ctra excitation types. Dynamic response results can be presented as structural deformations internal member loads.

nonlinear dynamic analysis program is integrated in the rest of the STARDYNE system. The ations of motion for the linear portion of the structural model are generated and modified to ount for the nonlinear springs. The resulting nonlinear equations of motion are directly grated using either the Newmark or Wilson implicit integration operators. The user may r sets of structural loadings which vary with time, and specify time points at which the gram is to output the structural response.

3A.2-14 Rev. 30

2.13.1 General Description s computer program is written to calculate the time history of the resulting forces and ments at assigned nodes in the dynamic model of the RPV support system. These forces are duced due to the external asymmetric pressure in the reactor cavity, resulting from a LOCA r a hot leg or cold leg nozzle. The pressure forces may be acting at the reactor pressure vessel, primary shield wall, or the neutron shield tank.

program performs the following calculation:

P(t) = Pj ( t ) = Ai xPi ( t )

M( t) = Pj ( t ) x Rj re:

Pi(t) =A pressure time history for pressure area No. i Ai = An area vector corresponding to the pressure time history p (t)

Pj(t) = Force vector due to pressure acting on No. j area P(t) = Resultant force vector due to pressures acting on all areas Rj = Displacement vector from a force application point on an area to the point of rotation M(t) = Resultant moment vector due to all pressures calculate the force and moment time history at a node in a given structural model, the surface on which the pressure acts is divided into several regions, such that only a constant pressure on one region at any time.

projection of a surface area, multiplied by the pressure acting perpendicular to it, gives the sure force. This force is broken into three global components by defining the direction cosines he pressure force vector.

centroid coordinates of the projected area are then calculated, and when these are subtracted m the coordinates of the node, the displacement components are obtained, which are used in ulating the three moments at the node.

forces and moments for all projected areas, due to corresponding pressure forces, are thus ulated and summed for each time point to get the force/moment time history for the node.

3A.2-15 Rev. 30

ample problem was performed by the use of the computer code ME-171. The results were then pared to results obtained by hand calculations.

2.14 LIDOP (ME-184) 2.14.1 General Description OP (ME-184) generates crush rigidities and deformation energies for pressurized or ressurized piping in the following geometries:

1. Ring crush against flat rigid surface
2. Indent of straight pipe against rigid cylinder
3. 1.5D pipe elbow (extrados) against flat rigid surface
4. Pipe bend (extrados) against flat rigid surface
5. Indent of straight pipe against rectangular block h dynamic and material properties are considered in generation of the crush characteristics.

ressurized force-displacement and energy-displacement characteristics of pipe and elbows are erated from empirical equations which are based on experimental data. Pressurization effects, ed on fluid displacement during deformation, are superimposed on the unpressurized racteristics. The overall dimensions of the contact area, where applicable, are generated by irically corrected geometric relationships. Dynamic effects of elbows are empirically rmined from an experimental comparison of static and dynamic impact of spheres. Dynamic cts of all other geometries and elbows in certain cases are based on the results of finite ment computer simulations of rings impacting flat, rigid surfaces. The effects of material perties are determined from empirical relationships based on computer predictions TR-2-0, 1976; EMTG-33-0, 1977; EMTR-403-B, 1978; Standard Review Plan 3.6.2, REG-75/087).

2.14.2 Program Verification eral sample problems were performed by the use of the code ME-184 to ensure that all options e exercised and thus demonstrate the function and adequacy of the program.

3A.2-16 Rev. 30

2.15.1 General Description ME-185 program determines the dynamic load factor (DLF) for a single degree-of-freedom monic oscillator subject to an arbitrary force history. At time zero, the oscillator is assumed to n equilibrium and at rest. Its response to the force history, defined by a series of force-time s, is then computed. If the force is not specified at time zero, it is automatically set at zero and ps up linearly to the first specified force-time coordinate. If the force is specified at time zero, assumed to be suddenly applied.

ng the initial conditions at time zero, ME-185 solves the equation of motion and searches for ima during the interval up to the next specified force-time pair. It then determines the ndary conditions (position and velocity) at the end of the interval and uses these as initial ditions for the solution during the next time increment. The process is repeated until the last e-time pair is reached. Assuming this last force is applied as a continuing load, the steady e response is computed and maxima determined. The greatest maxima is then divided by the test applied load to determine the maximum load factor. Since the solution method searches the greatest absolute amplitude of the system response and applied force, the applied force be positive or negative and may arbitrarily change signs during the specified force history.

dynamic load factor depends on the natural frequency of the single degree of freedom llator. In order to provide the DLF at the frequency of the structural system being analyzed, as l as to show how the DLF changes to an error in the calculated frequency or in the duration of applied force, DLFs are computed for a range of frequencies. The above calculation method is ated for each of several discrete frequencies in the range. The frequency range extends roximately one order of magnitude to either side of the frequency corresponding to the period ation) of the applied force history.

a force which varies linearly between two specified force-time pairs, the equation of motion Mx** + kx = F o + a o t re:

M = mass of the oscillator k = stiffness of the oscillator F0 = applied force at the start of the interval a0 = rate of change in the applied force (F/ T) solution of this equation is:

3A.2-17 Rev. 30

re:

w= K/M , the natural frequency C1 = (Vo - ao/k)/w C2 = Xo - Fo/k C3 = Fo/k C4 = ao/k x0 = initial position V0 = initial velocity above relations are simplified and the magnitude of the spring force is made identical to lacement of K=1. This relation is used in ME-185.

en transferring from one time interval to the next, the position and velocity must be rmined before the new coefficients, C1...,C4 can be calculated. The position at the end of the vious time interval can be computed from the above equation for X and the velocity may be rmined from:

V = C1w cos(wt) - C2w sin(wt) + C4 ny interval, the maxima or minima may be computed by substituting the times of zero velocity he equation for X. These times may be computed from the relation:

0 = C1w cos(wt) - C2w sin(wt) + C4 If C2 = 0 then, wt = cos-1 (-C4/(C1w))

If C4 = tan-1 (C1/C2) wt = tan-1 (C1/C2)

If C1 = 0 then, wt = sin-1 (C4/C2w))

3A.2-18 Rev. 30

2 2 2 2

-1 - C 1 C 4 w +/- C 2 w ( C 1 + C 2 )w - C 4 wt = cos ---------------------------------------------------------------------------------------

2 2 2

( C 1 + C 2 )w 2.15.2 Program Verification eral problems were performed by the use of DLF (ME-184) to ensure that all options were d and thus demonstrate the function and adequacy of the program.

2.16 References for Appendix 3A.2 2-1 Bray, A.P. TIGER-Temperatures from Internal Generation Rates. KAPL, Schenectady, NY, May 1954.

2-2 Bray, A.P. and MacCracken, S.J. TIGER-II-Temperatures from Internal Generation Rates. KAPL-2044, May 29, 1959.

2-3 Briggs, D.L. TIGER-Temperatures from Internal Generation Rates. KAPL-M-EC-29, February 1963.

2-4 Clough, R.W. and Wilson, E.L. 1962. Dynamic Response by Step-by-Step Matrix Analysis. Symposium on Use of Computers in Civil Engineering, Lisbon, Portugal, 1962, p 45.1-45.14.

2-5 Clough, R.W.; Benuska, K.L.; and Wilson, E.L. Inelastic Earthquake Response of Tall Buildings. Proceedings of the Third World Conference on Earthquake Engineering, Vol.

II, Auckland and Wellington, New Zealand, January, 1965, p 68-69.

2-6 Collins, T. and Witmer, E. Application of the Collision Imparted Velocity Method for Analyzing the Responses of Containment and Reflector Structures to Engine Rotor Fragment Impact. Aeroelastic and Structures Research Laboratory, Department of Aeronautics and Astronautics, Massachusetts Institute of Technology, August 1973.

2-7 Crose, J.G. ASAAS Asymmetric Stress Analysis of Axisymmetric Solids with Orthotropic Temperature Dependent Material Properties that can Vary Circumferentially. Air Force Report No. SAMSO-TR-71-197, Aerospace Report No.

TR-0172 (S2816-15) - 1, December 29, 1971.

2-8 EMTG-33-0, Applicability of Pipe Crush Stiffness for Pipe Rupture Analysis, January 10, 1977.

2-9 EMTR-2-0, Extrapolation and Application of Experimental Static Pipe Crush Data, April 20, 1976.

3A.2-19 Rev. 30

2-11 Giberson, M.F. The Response of Nonlinear, Multi-Story Structures Subjected to Earthquake Excitation. Earthquake Engineering Research Lab, California Institute of Technology, Pasadena, California, June 1967.

2-12 Hildebrand, F.B. 1956. Introduction to Numerical Analysis. McGraw- Hill Book Company, Inc., New York, N.Y.

2-13 Hodge, P.G. 1959. Plastics Analysis of Structures. McGraw-Hill Book Company, New York, N.Y.

2-14 Johns, R.H. and Orange, T.W. 1961. Theoretical Elastic Stress Distributions Arising from Discontinuities and Edge Loads in Several Shell-Type Structures. NASA Technical Report R-103.

2-15 Larson, L.D. Inelastic Response of Pressurized Tubes Under Dynamic Bending and Torsional Loads. PhD Thesis, Mechanical Engineering Dept., Carnegie-Mellon University, 1973, University Microfilm Order No. 73-22872.

2-16 Lechliter, G.L.; Liedel, A.L.; and Schmid, J.R. Mathematics Programs Available on Philco 2000 Computer, Part II, Curve Plotting. KAPL-M-6416 (EC-40), October 1964.

2-17 Love, A.E.H. 1944. A Treatise on the Mathematical Theory of Elasticity. Dover Publications, New York, N.Y.

2-18 Martin, H.C. 1966. Introduction to Matrix Methods of Structural Analysis.

McGraw-Hill Book Company, Inc., New York, N.Y.

2-19 Martin, H.C. 1966. On the Derivation of Stiff Matrices for the Analysis of Large Deflection and Stability Problems. Proc. Conf. Matrix Methods Structure Mech, Wright-Patterson Air Force Base, Ohio, October 26-28, 1965, AFFBL TR 66-80.

2-20 Neal, B.G. The Effect of Shear and Normal Forces on the Fully Plastic Moment of a Beam of Rectangular Cross Section. Journal of Applied Mechanics, June, 1961, p 269-274.

2-21 Personal Communication with Mrs. M. Helme concerning TIGER-IV, Thermal Design Memorandum No. 88 (Date).

2-22 Przemieniecki, J.S. 1968. Theory of Matrix Structural Analysis, McGraw-Hill Book Company, Inc., New York, N.Y.

2-23 Schmid, J.R.; Lechliter, G.L.; and Fisher, W.W. LION-Temperature Distribution for Arbitrary Shapes and Complicated Boundary Condition. KAPL-M-6532 (EC-57),

Revision IV, December 1969.

3A.2-20 Rev. 30

2-25 Stokey, W.F.; Peterson, D.B.; and Wruder, R.A. 1966. Limit Load for Tubes Under Internal Pressure, Bending Moment, Axial Force and Torsion. Nuclear Engineering and Design, 4, North-Holland Publishing Company, Amsterdam, p 193-261.

2-26 U.S. Atomic Energy Commission. Nuclear Reactors and Earthquakes, Chapter 6, Dynamic Pressure on Fluid Containers. TID-7024, Division of Reactor Development, Washington, D.C., August 1963.

2-27 Wichman, K.R.; Hopper, A.G.; and Mershon, J.L. 1965. Local Stresses in Spherical and Cylindrical Shells Due to External Loading. Welding Research Council Bulletin, WRC-107.

2-28 Wilson, E.L. A Computer Program for the Dynamic Stress Analysis of Underground Structures. U.S. Army Corps of Engineers, Report No. 68-1, Structural Engineering Laboratory, University of California, Berkeley, California, January 1968.

2-29 Wu, R. and Witmer, E. Finite-Element Analysis of Large Transient Elastic-Plastic Deformations of Simple Structures with Application to the Engine Rotor Fragment Containment Deflection Problem. Aeroelastic and Structures Research Laboratory, Department of Aeronautics and Astronautics, Massachusetts Institute of Technology, January 1972.

3A.2-21 Rev. 30

TABLE 3.A.2.1-1 INFINITELY LONG SOLID CYLINDER, PERTINENT PARAMETERS Dimensions and Properties Loading and Boundary Conditions r0 =a r = P (cos + cos 2 )

l=a q = P0 sin E = 10 x 106 psi Uz = 0 Ur = 0 Y = 0.25 At r = 0, Po = 10,000 psi a = 1 inch Page 1 of 1 Rev. 30

TABLE 3.A.2.4-1 COMPARISON OF RESULTS OF SLOSH VS AEC ANALYSIS*

ample 1 Page 188* SLOSH

, Eq. Impulsive Force (kips) 298.5 299.7 Impulsive Force (kips) 105.4 105.8 (EBP), Impulsive Moment (kip-ft) 594 595 (IBP), Impulsive Moment (kip-ft) 1,120 1,118

, Convective Force (kips) 133 133 (EBP), Convective Moment (kip-ft) 212 214 (IBP), Convective Moment (kip-ft) 252 253 max, Maximum Moment (kip-ft) 1,372 1,371 ax, Maximum Shear (kips) 127.9 128.5 ample 2 Page 192* SLOSH

, Eq. Impulsive Force (kips) 458 458 Impulsive Force (kips) 277 277 (EBP), Impulsive Moment (kip-ft) 3,460 3,470 (IBP), Impulsive Moment (kip-ft) 4,070 4,074 Convective Force (kips) 139 137 (EBP), Convective Moment (kip-ft) 552 547 (IBP), Convective Moment (kip-ft) 560 552 max, Maximum Moment (kip-ft) 4,630 4,626 ax, Maximum Shear (kips) 301 301 ample 3 Page 197* SLOSH

, Eq. Impulsive Force (kips) 298.5 299.7 Page 1 of 2 Rev. 30

, Convective Force (kips) 133 133 de 1 quency (cps) 0.333 0.331 1, Seismic Force (kips) 23.46 23.80 1, Seismic Force (kips) 1.26 1.28 de 2 quency (cps) 1.486 1.482 2, Seismic Force (kips) -3.90 -3.91 2, Seismic Force (kips) 174.39 174.54 ax, Maximum Shear (kips) 195.21 195.72 rce:

U.S. Atomic Energy Commission. Nuclear Reactors and Earthquakes, Chapter 6, Dynamic Pressure on Fluid Containers. TID-7024, Division of Reactor Development, Washington, DC, August 1963.

Page 2 of 2 Rev. 30

BLE 3.A.2.6-1 COMPARISON OF EXPERIMENTAL DATA WITH ANALYTICAL DATA USING LIMITA 3 Experimental Limita 3 Percent Value Computer Results Difference ak Deflection Mode 6 0.297 0.310 4.2 ak Deflection Mode 9 Not Determined 0.870 -

me at Peak Deflection in Mode 6 ~0.004 ~0.0042 4.8 rmanent Deflection in Mode 6 0.144 ~0.140 2.8 rmanent Deflection in Mode 9 0.302 ~0.310 2.6 Page 1 of 1 Rev. 30

following computer programs are used for the analysis of Seismic Category I piping systems

1. NUPIPE II: Linear Elastic Analysis of 3-D Piping System Subjected to Thermal, Static and Dynamic Loads
2. PITRUST: Local Stress Analysis at Junction of 2 Cylindrical Vessels
3. PILUG: Local Stress Analysis at Junction of Lug with Pipe or a Cylindrical Vessel
4. SAVAL: Stress Analysis at Nozzle/Run-Pipe Junction Due to Atmospheric Safety Valve Discharge
5. STEHAM: Steamhammer Transient Analysis
6. WATHAM: Waterhammer Transient Analysis
7. PSPECTRA: Peak Spreads and Envelopes ARS Curves of Dynamic Events
8. STRUDL-SW: Structural Analysis Program
9. STRUDL-II (ICES): Structural Analysis Program
10. APEN: Anchor Penetration Analysis
11. PITRIFE: Finite Element Analysis of Integral Welded Attachment
12. PITAB: Piping Support Load Tabulation Program
13. CHPLOT: Piping Support Load Tabulation Program
14. LOADCOMB: Load Combination Generator for STRUDL
15. BEARST: Pipe Bearing Stress Analysis
16. ANCCOMB: Anchor Load Combination Program
17. BENDCORD: Bend Co-ordinate Program
18. NUPIPE-SWPC: Linear Elastic Analysis of 3-D Piping System Subjected to Thermal, Static and Dynamic Loads
19. PC-PREPS: Structural Analysis Computer Code
20. PILUG-PC: Local Stress Analysis at Junction of Lug with Pipe or a Cylindrical Vessel 3.1 NUPIPE II
1. General Description 3A.3-1 Rev. 30

utilizes the finite element method of analysis.

NUPIPE II handles all loading conditions required for complete nuclear piping analyses. A given piping configuration may be analyzed successively for a number of static and dynamic load conditions in a single computer run. Separate load cases, such as thermal expansion and anchor displacements, may be combined to form additional analysis cases. The piping deadload analysis considers both distributed weight properties of the piping and any added concentrated weights.

A lumped mass model of the system is used for all dynamic analysis; both translational and rotational degrees of freedom may be considered. Location of lumped masses and degrees of freedom at each mass point are preselected by the analyst. The program automatically computes values of translational lumped masses.

Program input consists basically of program control, piping configuration description, and load specification information. The output for each loading condition analyzed consists of support reactions, internal forces and moments, deflections and rotations, and member stresses. Output from seismic analysis includes system normal mode information. Several reports may be generated based on report specification. These reports include pipe stress summaries, pipe support tabulations, and piping isometric plots.

The NUPIPE II program performs analysis in accordance with ASME Section III, Nuclear Power Plant Components (Code). Features ensuring code conformance include use of accepted analysis methods, incorporation of specified stress indices and flexibility factors, proper combination of moment resultants, and provision to (automatically) generate results of combined loading cases. A program option is available to specify among Class 1 analysis in accordance with NC-3600 of the Code, analysis per ANSI B31.1.0 power piping code and combined Class 1 and Class 2 analysis per Articles NB-3600 and NC-3600 of the Code.

2. Program Verification The NUPIPE II program has been verified with ADLPIPE (A.D. Little Corp.) for thermal, weight, and response spectrum seismic analysis. The results from both programs are presented in Tables 3A.3-1 through 3A.3-7. The model used for this comparison is presented on Figure 3A.3-1.

The comparison is also made with ASME Benchmark solution (ASME 1972) for force time-history dynamic response. The model used for this comparison is shown on Figure 3A.3-2. The results for comparisons are presented in plots on Figure 3A.3-2. The natural frequencies are given in Table 3A.3-8.

3A.3-2 Rev. 30

3.2 PITRUST

1. General Description PITRUST is a computer program which calculates the local stress intensity at the junction of two cylindrical vessels. The calculated stresses, including those due to pressure, are determined for the run cylinder. The program has application where a trunnion is welded to a run-pipe or where a branch pipe exits from a vessel or run-pipe.

The method and theory of calculating stresses follows that promulgated by the Welding Research Council Bulletin No. 107 (Wichman et. al, 1965). The program is capable of complying with requirements of ASME Boiler and Pressure Vessel Code - Section III - Nuclear Power Plant Components and ANSI-B31.1 Power Piping. PITRUST input consists basically of program control options, run-pipe dimensions, internal operating pressure, trunnion outside diameter, and loading specification. If the design criteria for the stresses are exceeded, the program can incrementally increase the pad thickness and recalculate the stresses until the lug passes or until the pad reaches 1.5 times the pipe wall thickness.

Program output tabulates the applied loadings and the local stresses at the junction of the trunnion and run-pipe.

2. Program Verification PITRUST has been verified by comparing its solution of a test problem to the solution obtained by CYLNOZ, an independently written piping local stress program. CYLNOZ, written by Franklin Institute (Philadelphia, Pa.), is a recognized program in the public domain. The test problem is of a 72.375 inches O.D. x 0.375 inches thick run-pipe, reacting under an external loading of 1,000 lb force (normal and shear) and 1,000 in-lb moments transmitted by a 16 in O.D.

nozzle. A comparison of results is tabulated in Table 3A.3-11. PITRUST has also been verified by comparing its solution of a test problem to the experimental test results as outlined in Corum and Greenstreet (1971). A comparison of these results is tabulated in Table 3A.3-12.

3.3 PILUG

1. General Description PILUG is a computer program which calculates local stress intensity at the junction of a lug with a pipe or other cylindrical vessel. The stress intensity 3A.3-3 Rev. 30

The method and theory of calculating stresses follows that promulgated by the Welding Research Council Bulletin No. 107 (Wichman et al., 1965). The program is capable of complying with requirements of ASME Boiler and Pressure Vessel Code,Section III, Nuclear Power Plant Components and ANSI-B31.1 Power Piping.

PILUG input consists basically of program control, run-pipe dimensions, internal operating pressure, rectangular lug dimensions, and loading specification.

If the design criteria for the stresses are exceeded, the program incrementally increases the pad thickness and recalculates the stresses until the lug passes or until the pad reaches 1.5 times the pipe wall thickness.

Program output tabulates the applied loadings and the local stresses at the junction of the lug and run-pipe.

2. Program Verification PILUG has been verified by comparing its solution, of a test problem, to results obtained by hand calculations using the formulations specified in Wichman et al.

(1965). A comparison of results is tabulated in Table 3A.3-13.

3.4 SAVAL

1. General Description SAVAL is a computer program which calculates the stresses at the junction of a safety valve nozzle and run-pipe. Calculated moments due to the suddenly applied thrust load are premultiplied by a dynamic load factor (DLF) prior to computing the nozzle and localized piping stresses. The subroutine DLF is incorporated in the program which computes the dynamic load factor.

The stress intensity in the run-pipe is computed and compared to allowable stresses in accordance with Equation 9 of Subsection NC-3652 of the ASME Boiler and Pressure Vessel Code,Section III, Nuclear Power Plant Components.

Input for SAVAL includes discharge force vector, pipe, nozzle, and pad dimensions, valve weight, valve opening time, pressure, and code allowable stresses.

Computer output consists of resolved forces and moments at center line of run-pipe, dynamic loading factor, and calculated stresses.

2. Program Verification 3A.3-4 Rev. 30

Figure 3A.3-4. A comparison of results is tabulated in Tables 3A.3-14 and 3A.3-15.

3.5 STEHAM

1. General Description STEHAM is a computer program which is used to determine the steamhammer transients of piping systems. This program uses the method of characteristics with finite difference approximations both in space and in time (Jonsson et al.; Luk; Moody). It calculates the one-dimensional transient flow responses and the flow-induced forcing functions in a piping system caused by rapid operational changes of piping components, such as the actuation of a stop valve or safety/relief valve. Flow characteristics of piping components are mathematically formulated as boundary conditions in the program. These components include the flow control valve, the stop valve, the safety/relief valve, the steam manifold, and the steam reservoir. Frictional effects are taken into consideration in this program.

This program accepts the following as input: (1) the flow network representation of the piping system, (2) the initial flow conditions along the piping system, and (3) time-dependent flow characteristics of piping components. Output consists of time-histories of flow pressures, flow densities, flow velocities, inertia, and momentum functions.

2. Program Verification STEHAM is verified by comparing its solutions of a test problem (Figures 3A.3-5 and -6) to the results of the same problem obtained by an independent analytical approach, as well as an experimental measurement, as published in Progelhof and Owczarek (1963) and ASME Paper No. 63-WA-10). A comparison of results for time-history pressure responses is plotted on Figures 3A.3-7, -8, and -9. The forcing functions developed for nodal points of the piping system 3A.3-5 calculated from the relation F = (p + u2/g)A has also been checked by hand calculations as tabulated in Table 3A.3-16.

3.6 WATHAM

1. Description WATHAM is a computer program which is used to determine the flow-induced forcing functions acting on piping systems due to waterhammer. These forcing functions may then be used as input to a structural dynamic analysis such as a NUPIPE program run.

3A.3-5 Rev. 30

abnormal operational changes of piping components, such as the startup and trip of pumps or the rapid opening and closing of values.

The analysis is based upon the method of characteristics with finite-difference approximations both in time and space for the solution of one-dimensional liquid flows. Influences of piping components - including flow valves, pipe connections, reservoirs and pumps - have been considered in the analysis.

WATHAM input requires the geometry of the piping system, pipe properties, water properties, operational characteristics of pump and valve, flow frictional coefficients, and the initial water flow conditions. The output provides the; time history functions of piezometric heads, velocities, and nodal forces for all nodes and the inertial unbalanced force for each segment. It also gives the maximum value of all the above-mentioned functions and their occurring time in the process of flow-transient.

2. Program Verification For the verification of WATHAM, a problem from Reference SWEC 1977 is employed. Figure 3A.3-10 depicts the flow network with nine pipes, its geometrical properties and steady state flow conditions. The flow-transient mode analyzed is the sudden closure of a valve at the downstream end. Figure 3A.3-11 shows the hydraulic network for WATHAM. Table 3A.3-17 illustrates the input data needed for WATHAM run. Figures 3A.3-12 and 3A.3-13 show the comparison of head-time curves obtained from Streeter and Wylie (1967), Fabic (1967) and WATHAM. Table 3A.3-18 presents the comparison of nodal forces between hand calculation and WATHAM computation.

In general, WATHAM results are in good agreement with Streeters results (Streeter 1967). The small discrepancy is attributed to the modeling of reservoir boundary condition. In WATHAM, the energy equation between the reservoir is utilized rather than assuming that the head of pipe entrance is the same as that of the reservoir.

3.7 PSPECTRA

1. General Description The PSPECTRA program peak spreads and envelopes amplified response spectra curves of earthquakes or other dynamic events. The program reads ARS curves from tape, card, or disk files. The created curves are saved on a disk file, and optionally printed, plotted, and punched on a card file. Disk and card format is compatible with the NUPIPE program (ME-110). In fact, one important 3A.3-6 Rev. 30

There are two methods of peak spreading - either the sides of the spread peaks can be vertical or they can be parallel to the sides of the original peaks. There are three methods of enveloping - maximum value, absolute sum, and SRSS. Another program option allows enveloping E-W and N-S direction curves to form one horizontal curve for each curve set on a disk file. ARS curves can also be input from up to four disk files in one run for the purpose of enveloping or plotting curves with different damping values. The last option allows inputting curves with up to four different damping values and superimposing curves with different damping values on the same plot.

The sides of vertical peaks actually span 0.001 second. The sides of parallel peaks are perfectly parallel to the original peaks. That is, the spread up and down the peak is based on the period of the tip of the peak.

All the peaks on a curve are spread, no matter how small they are. The final curve is the maximum value envelope of the individual spread peaks.

The program envelopes as follows: At each 0.001 second interval it either sums, finds the maximum value, or takes the SRSS of the accelerations from the individual digitized curves. The envelopes, therefore, automatically become digitized at the same intervals as the input curves. This means that a maximum value envelope is slightly above the true envelope, wherever the input curves intersect at a nonmultiple of 0.001 second.

After peak spreading or enveloping is complete, the curves are processed to eliminate points on lines of constant slope. Many of the interpolated points are eliminated during this process.

PSPECTRA can also be directed to create plots of each curve before and after peak spreading, both superimposed on the same graph.

2. Program Verification For the verification of PSPECTRA, a problem was run three times, producing maximum, sum, and SRSS envelopes (Luong 1977). The results were verified by comparing its solution of the test problem to results obtained by hand calculations.

3.8 STRUDL-SW

1. General Description The STIFFNESS ANALYSIS Command of STRUDL-SW initiates the execution of a static, linear analysis of a structure considered as a lumped-parameter system.

3A.3-7 Rev. 30

program constructs the structural stiffness matrix and load vectors.

The computational procedure of the analysis is based on a network interpretation of the governing equations, the principal feature of which is the segmentation in processing of the geometrical, mechanical, and topological relationships of the structure. This allows a concise and systematic computational algorithm that is applicable for different structural types.

The basic steps in the procedure are as follows:

1. The stiffness matrix is determined for each member, considered as cantilevers (primitive stiffness matrix).
2. If there are any member releases specified, the local member stiffness matrix (Step 1) is modified.
3. The applied member loads, if any, are processed.
4. The structural stiffness matrix is assembled in the global coordinate system for free joints and released support joints.
5. The load vector is assembled and the global stiffness matrix and the load vector are modified to account for joint releases.
6. The governing joint-equilibrium equations are solved for the joint displacements.
7. The induced member distortions, member end forces and stresses are computed by back substitution.

Joint displacements are computed using a modified Gauss elimination procedure.

The modifications include partitioning of the stiffness matrix and load vectors into blocks that represent the equations of one or more joints, and a bookkeeping algorithm that takes advantage of the symmetry, banding, and sparseness of a typical stiffness matrix.

From the joint displacements, other results are calculated for joints and members.

For joints, member end forces are summed to yield reactions at support joints and to check the accuracy of analysis at free joints. For members, relative distortions at the end, and forces and moments at the beginning and end are calculated.

The above result types are not the only ones that may be listed following a stiffness analysis; they are, however, the only results that are calculated automatically 3A.3-8 Rev. 30

2. Program Verification STRUDL-SW has been verified by comparing the results of the STRUDL-SW runs with either hand calculations or results obtained by running the same test problems under a public domain version of STRUDL (Luong 1980). In this particular case, the public domain version of STRUDL used was GTICES STRUDL (February 1980 VIM7).

3.9 STRUDL-II (ICES)

1. General Description The present capabilities of STRUDL provide the engineer with the ability to specify characteristics of problems, perform analyses, reduce and combine results, and output any information stored within the system. Several different types of analysis procedures are currently available. This version of STRUDL also includes a variety of member design capabilities which provide both a set of useful techniques for aiding the designer in his design studies and a mechanism for adding new design procedures.

Analytic procedures in ICES STRUDL-II apply to both framed structures and continuous mechanics problems. Framed structures are two or three dimensional structures composed of slender, linear members, which can be represented by properties along a centroidal axis. Such a structure is composed of joints, including support joints, and members connecting the joints. A variety of force conditions on member ends and at support joints may be specified implicitly by means of structural type and orientation commands or explicitly for a member or joint.

Continuous mechanics problems are treated, in STRUDL, using the finite element method. In this method the domain of the problem is subdivided in one, two, or three dimensional elements, of different shapes, connected at a finite number of nodal points, or joints. This idealization is then entirely analogous to that of framed structures where members are a particular type of element. The present version of ICES STRUDL provides a variety of element types for the solution of plane stress/strain, plate bending, and shell analysis problems.

ICES STRUDL-II also contains the following types of design procedures applicable to the design of framed structures and their components: a frame optimization procedure which employs a discrete variable optimization procedure; a member selection capability with which various design procedures and codes can be used to design members from tables; designs subjected to arbitrary constraints; and reinforced concrete, beam, column, and slab adequacy checking and section proportioning, based on the ACI code.

3A.3-9 Rev. 30

Only partial verification has been performed on STRUDL-II (ICES). No unit testing has been performed on any of its commands (SWEC Computer Library 1977). All the commands, except for the CSM CALCULATIONS command, were qualified by using the comparison method. A system test was run on both STRUDL-II (V10L06) and GTICES (VIM6) using all commands except for the CSM CALCULATIONS command, and the results were compared. The closely-spaced modes feature of STRUDL-II was qualified by the manual method.

A system test on STRUDL-II (V10L06) with the CSM CALCULATIONS command included was run and the results were compared with hand calculations.

3.10 APEN

1. General Description APEN is a computer program that analyzes anchor penetrations. Anchor penetration consists of three phases, as follows:

ASE 1 Input In the input phase, the control cards are read and defaults are taken, where necessary. The loads from both sides of the anchor are read, identified, and stored. All earthquake type loads are stored as absolute values.

ASE 2 Design The loads from both sides are combined and summed according to the run condition.

Anchor dimensions input as zero (or blank) are selected from a table according to pipe size. Each dimension is then checked against the required size, which depends on the set of combined loads. Undersized dimensions are incremented in preset steps until they are equal to or greater than the required size, or until preset maximums are reached. These combined loads are not used in PHASE 3, Stress Analysis.

ASE 3 Stress Analysis The load combinations are made on each side of the anchor individually. For each loading combination, the maximum stress intensity at each side is computed. The stresses printed are for the worst side, for that loading combination. When all the loading combinations have been computed, the next problem is attempted. The end of the input signals the program stop. A detailed description of the analysis can be found in calculation 570.470.1.13-NP(B)-012 (Luong).

The restrictions on using the analysis procedure are:

a. Thickness of concrete wall must be 18 inches or greater.

3A.3-10 Rev. 30

c. Local stresses in the concrete wall which are induced by the anchor loads should be reviewed by the Structural group.
2. Program Verification APEN has been verified by using the manual method, found in the calculation titled, Verification of APEN Computer Program, Calculation No.

570.470.1.13-NP(B)-010-Z14 (Luong 1977). The results of the test problem from the APEN computer run were compared with that of the hand calculation.

3.11 PITRIFE

1. General Description PITRIFE is a post processor program that utilizes the results of a finite element model of two intersecting cylinders to determine the local discontinuity stresses at the interface of the two cylinders (pipe and trunnion).

The PITRIFE program provides an efficient and accurate means of going from the stresses generated in the finite element models to the stresses found in an actual pipe-trunnion problem subjected to actual loads. It accomplishes this translation from the finite element model to the actual pipe trunnion problem in stages. In the first stage, the program reads in the pipe and trunnion geometry, the type of weld securing the trunnion to the pipe, all the applied loads, and the Users choice of one of the four available load combination methods. In the second stage, based upon the problem geometry and the type of weld, the program generates a set of load independent stress coefficients by interpolating between the non-dimensional stress coefficients generated from the finite element models. In the third stage, it selects static and dynamic loads in accordance with the load combination method chosen, forms a total static and a total dynamic load vector, and uses these to generate 64 total load vectors. There are 64 total load vectors because the signs of the dynamic loads are arbitrary this results in 64 possible combinations of the components of the static and dynamic load vectors. In the fourth stage, it computes the principle stresses for each of these loads, determines the maximum stress intensity, and prints the results. The program repeats the third and fourth stages until stresses have been computed for all the desired combinations of static and dynamic loads.

2. Program Verification The PITRIFE computer program has been verified by demonstrating that the maximum stress intensities as given by PITRIFE equal the values given by the finite element analysis for specific size-on-size and 0.707 size-on-size models. A 3A.3-11 Rev. 30

coefficients and maximum stress intensities derived by hand calculation equal the coefficients used in the program to calculate maximum stress intensity. A comparison of these results is given in Table 3A.3-20.

3.12 PITAB

1. General Description PITAB provides tabulations of vertical and other restraint data consisting of general support data and final design loads. The final design loads for each support are calculated using Millstone procedure. This procedure specifies the method of combining individual support loading conditions for various types of supports to arrive at the final design loads.

The program accepts titles, notes, procedure selection, and support data for each support consisting of general support information, individual loading conditions for the support, and its thermal displacements. The final design load for each support is then calculated using the procedure selected and two copies of the vertical and other support tables are printed. Each table contains a listing of the supports in ascending order of support number with their general support information and final design loads.

2. Program Verification PITAB has been verified by using the manual method found in the calculation titled, Verification of PITAB Computer Program (Quan). The results of the test problem from the PITAB computer run were compared with that of the hand calculation.

3.13 CHPLOT

1. General Description CHPLOT is a program which plots any number of data values (variables) versus time. Although the plot input data file can be in the form of card data, the more appropriate application of this program is to be used in conjunction with a program that creates a plot data file (on disk or tape) having the format required for input to this program.

Plots are available in two sizes; one with axes of 5 inches (ordinate) by 8 inches (abscissa) that fits the standard 8-1/2 inch by 11 inch page, and the other is 8 inches by 12 inches for fitting an 11 inche by 15 inche page. Plots are normally one data value versus time per graph, although up to 14 data values (plots) can be plotted on one graph.

3A.3-12 Rev. 30

performed automatically to fit the size selected. Graphs can be grouped into a maximum of nine groups, within which each group of graphs will be scaled to the same scale factor.

An optional label is available for labelling all graphs at the bottom of each graph.

2. Program Verification All the output options such as plot size, scale groupings, curve labels, and data values versus time are verified by visual inspection. It is found that CHPLOT performs all the options defined in the users' manual.

Table 3A.3-22 shows a partial listing of the forces plotted on Figure 3A.3-15, which illustrates this visual verification of the plots.

3.14 LOADCOMB

1. General Description The purpose of LOADCOMB computer program is for the design and selection of pipe supports required to comply with subsection NF of ASME III, load conditions and combinations must be generated to be applied to the support frames. In order for the loads to be input to STRUDL and an NF code check performed, the final output from the program is generated in the form of STRUDL JOINT LOAD commands on a user specified data set.

The LOADCOMB program uses two different coordinate systems, the local run-pipe coordinate system (identical to the NUPIPE, ME-110, local run-pipe coordinate system) and the STRUDL global coordinate system.

The user of LOADCOMB must input the individual load components into LOADCOMB in the local run-pipe coordinate system. The local x axis for this system coincides with the longitudinal axis of the pipe. The y and z axes must coincide with the principal axes of the pipe. (See Attachments 6.1 and 6.2 of EMTG-31-0, 1977). The program internally forms the load combinations in the local run-pipe coordinate system.

The STRUDL JOINT LOAD commands are output in a STRUDL global coordinate system.

2. Program Verification LOADCOMB has been verified by using the manual method found in the calculation titled, Qualification Calculation to Verify Computer Program 3A.3-13 Rev. 30

3.15 BEARST

1. General Description BEARST is a computer program that calculates the maximum local stress in a pipe wall due to contact loads between the pipe and a structural element.

The program is capable of complying, as per SATM-17, Rev. 4 with the code requirements of ASME Section III (Class 1, 2, and 3) and ANSI B31.1.

Loads representing the forces that the run-pipe exerts on the frame are input according to a local 1-2-3 coordinate system whose origin is at the center of the run-pipe and where the 3 axis points along the centerline of the run-pipe, the 2 axis points in a general upward direction, and the 1 axis follows the right-hand rule.

The program is applicable to all types of 1-way, 2-way, and 3-way restraint problems. The program first assigns a set of signs to the loads in which each component can be + or - (earthquake loads); it then stores them along with the fixed - sign loads in an array of loads at each contact point. Loads at each restraining point are then combined to conform to the various load conditions, and within some load conditions there are a number of load combinations needed to assure for the appropriate multiple thermal loads and occasional loads. Local stresses are computed for each restraining point and stress intensity is listed for each load combination point by point. In general, stress intensities are calculated for each of the two elements taken from the outer and inner surface of run-pipe at the restraining point. Appropriate minimum normal pipe stresses are added to the largest one of the stress intensities for each load combination and the total compared to the appropriate allowables.

Provision is included for considering up to three multiple thermal loads for ASME Section III code compliance and also up to three multiple occasional loads within each operating condition.

2. Program Verification BEARST computer program has been verified by using the manual method found in the calculation titled, Verification of BEARST Computer Program, General

-NP(B)-011-Z14 (Luong 1976). The results of the test problem from the BEARST computer run were compared with that of the hand calculation.

3.16 ANCCOMB

1. General Description 3A.3-14 Rev. 30

type and to select the smaller of the two general stresses input for each side of the terminal anchor to obtain minimum normal stress (MNS). Terminal anchors are anchors common to two independently analyzed piping systems, one on each side of the anchor.

The load combination method is based on the EMTG-31-0, Section 2.16 rules.

a. For earthquake load types, OBEI; OBEA; SSEI; and EART, the total anchor load component is the square root of the sum of the squares of the two components from both sides of the anchor.

For example:

2 2 Mx OBEI = Mx + Mx Total OBEI 1 OBEI 2 Where:

MxOBEI = Operational basis inertia load component (Mx) from one side of the terminal anchor MxOBEI = Load component from the other side of the anchor.

The MNS value associated with these earthquake loads is zero.

b. Operational basis earthquake, combined inertia and anchor movement loads (OBET) are formed by adding together by absolute value summation the individual components of the OBEA and OBEI loads and then taking the +/- value of such sum. At terminal anchors, such summation is to occur only after the total OBEI and OBEA anchor loads have been formed.
c. For occasional dynamic loads associated with upset, emergency, and faulted conditions, i.e. OCC (UEF) the two OCCMAX, MIN sets from each side of the anchor are to be combined so as to yield one OCCMAX, MIN set by SRSS'ing individual components from each side (as in accordance with the SRSS operator of Appendix B, Section A) in such a fashion as to obtain the highest possible positive and negative magnitudes for the total OCCMAX, MIN set.
2. Program Verification ANCCOMB computer program has been verified by using the manual method found in the calculation titled, Verification of ANCCOMB Version 00 Level 3A.3-15 Rev. 30

3.17 BENDCORD

1. General Description BENDCORD is a Fortran IV Program which supplies, in printed and card form, data for coding segments of a circle for use in the NUPIPE piping program.

BENDCORD has the capacity of operating on an arc which lies in any one of three planes defined by the Cartesian Co-Ordinate System.

The piping system may be divided into equal or unequal segments so as to aide the user in placing supports.

For seismic piping systems, the user is provided the option of locating mass points at alternating nodes or, if desired, at every node.

Various elbow types may be inputted to reflect the requirements of Class 1 analysis.

BENDCORD divides an arc into tangent lines, the lengths of which are then calculated by subtracting the coordinates of the tangent point from the tangent intersection point of the tangent lines.

2. Program Verification The BENDCORD program is verified by calculating distances (or offsets) between nodal points by hand and comparing these values to those calculated by BENDCORD.

The problem consists of a 180 degree piping arc in the X-Z plane beginning with node 10 at phi = 45 degrees, and ending at node 36. There are 26 included angles in the arc, all equal. Mass points are at every other node. Refer to Figure 3A.3-14 for a graphic representation of the problem.

Partial results are shown in Table 3A.3-21. It can be seen that there are no significant differences between the results from BENDCORD and the hand calculations.

3.18 NUPIPE-SWPC

1. General Description 3A.3-16 Rev. 30

Code,Section III Nuclear Power Plant Components and the ANSI/ASME B31.1 Power Piping Code.

2. Program Verification Using an approved Quality Assurance Program, this computer program has been verified and validated and shown to be accurate and acceptable for use in evaluating piping systems in accordance with the ASME B&PV Code,Section III and ANSI/ASME B31.1 Power Piping Codes.

3.19 PC-PREPS

1. General Description PC-PREPS is a PC based computer program which performs a complete structural analysis, performing an AISC code check, weld qualification and baseplate/anchor bolt qualifications.
2. Program Verification Using an approved Quality Assurance Program, this computer program has been verified and validated and shown to be accurate and acceptable for use in evaluating ASME B&PV Code,Section III and ANSI/ASME B31.1 Power Piping Code components.

3.20 PILUG-PC

1. General Description PILUG-PC is a PC based stress analysis program used to calculate stress intensity at the junction of a rectangular attachment perpendicular to round pipe.
2. Program Verification Using an approved Quality Assurance Program, this computer program has been verified and validated and shown to be accurate and acceptable for use in evaluating ASME B&PV Code,Section III and ANSI/ASME B31.1 Power Piping Code components.

3A.3-17 Rev. 30

3-1 American Society of Mechanical Engineers. Pressure Vessel and Piping, 1972 Computer Programs Verification, Problem No. 5.

3-2 Arthur D. Little Corporation. ADLPIPE: Static, Dynamic, Thermal Pipe Stress Analysis.

3-3 Corum, J.M. and Greenstreet, W.L. 1971. Experimental Elastic Stress Analysis of Cylinder to Cylinder Shell Models and Comparison with Theoretical Predictions. First International Conference on Structural Mechanics in Reactor Technology, Berlin, Preprints, Vol. 3, Part G.

3-4 EMTG-31-0, Load Conditions, Combinations, and Allowable Stresses, April 11, 1977.

3-5 Fabic, S. 1967. Computer Program WHAM for calculation of Pressure, Velocity, and Force Transients in Liquid Filled Piping Networks, Report No. 67-49-R, Kaiser Engineers, November 1967.

3-6 Jonsson, V.K.; Matthews, L.; and Spalding, D.B. Numerical Solution Procedure for Calculating the Unsteady One-Dimensional Flow of Compressible Fluid, ASME Paper No. 73-FE-30.

3-7 Luk, C.H. Effects of the Steam Chest on Steamhammer Analysis for Nuclear Piping Systems, ASME Paper No. 75-PVP-61.

3-8 Luong, W.C. 1976. Verification of 'BEARST' Computer Program. Call. No. - General-NP(B)-011-Z14, December 15, 1976.

3-9 Luong, W.C. 1977a. Verification of 'PSPECTRA' Computer Program. SWEC Computer Library, March 16, 1977a.

3-10 Luong, W.C. 1977b. Verification of 'APEN' Computer Program. Call No. 570.470.1.13-NP(B)-010-Z14, October 28, 1977.

3-11 Luong, W.C. Call 570.470.1.13-NP(B)-012.

3-12 Luong, W.C. 1980. Verification of 'STRUDL-SW' Computer Program. SWEC Computer Library, February 1980.

3-13 Moody, F.J. Time - Dependent Pipe Forces Caused by Blowdown and Flow Stoppage, ASME Paper No. 73-FE-23.

3-14 Progelhof, R.C. and Owczarek, J.A. 1963. The Rapid Discharge of a Gas from a Cylindrical Vessel Through a Nozzle. AIAA Journal, Vol. 1, No. 9, September 1963, p 2182-2184.

3A.3-18 Rev. 30

3-16 Purohit, S.N. 1979. Verification of ANCCOMB Computer Program. Call. No.

570.47.01-NP(B)-017, November 6, 1979.

3-17 Quan, T.F. Verification of PITAB Computer Program. SWEC Computer Library.

3-18 Sexton, R.W. 1977. Qualification Call. to Verify Computer Program ME-163, LOADCOMB. SWEC Computer Library, November 23, 1977.

3-19 Streeter, V.L. and Wylie, E.G. 1967. Hydraulic Transients, McGraw-Hill Book Company, New York.

3-20 SWEC (Stone & Webster Engineering Corporation) Computer Library 1977. Part Verification of STRUDL-II (ICES) Computer Program. January 1977.

3-21 Wichman, K.R.; Hopper, A.G.; Mershon, J.L. 1965. Local Stresses in Spherical and Cylindrical Shells due to External Loading, Welding Research Council Bulletin, WRC-107.

3A.3-19 Rev. 30

ANCHOR MOVEMENT, AND EXTERNAL FORCE LOADING Forces (lb) Moments (in-lb) ode Program* FX FY FZ MX MY MZ 0 NUPIPE -9154 7541 4492 -5952 -823420 1241512 ADLPIPE -9178 7540 4492 -5529 -823420 1241512 8 NUPIPE 16650 ADLPIPE 16622 0 NUPIPE 34532 -33620 -31750 -486338 -1516811 573673 ADLPIPE 34511 -33608 -31736 -486386 -1519359 573438 0 NUPIPE 8631 ADLPIPE 8678 0 NUPIPE 1702 798 12553 -28147 164346 248852 ADLPIPE 1746 768 12541 -26917 166180 250956 TE:

See Figure 3A.3-1 for General NUPIPE II Model.

(All node points listed here are not shown.)

3A.3-20 Rev. 30

THERMAL, ANCHOR MOVEMENT, AND EXTERNAL FORCE LOADING Node Program Deflection (in) Rotation (rad)

DX DY DZ RX RY RZ 7 NUPIPE 0.348 -0.141 0.230 -0.0026 0.0025 -0.0084 ADLPIPE 0.348 -0.141 0.229 -0.0026 0.0025 -0.0084 2 NUPIPE 1.120 0.052 -0.023 -0.0092 -0.0051 -0.0115 ADLPIPE 1.120 0.052 -0.023 -0.0092 -0.0051 -0.0115 0 NUPIPE 1.276 -0.028 -0.548 -0.0066 -0.0044 0.0024 ADLPIPE 1.276 -0.027 -0.548 -0.0066 -0.0044 0.0024 0 NUPIPE 0.512 -0.001 -0.520 -0.0034 -0.0005 0.0035 ADLPIPE 0.512 -0.000 -0.520 -0.0035 -0.0005 0.0035 0 NUPIPE 0.066 -0.000 0.249 -0.0010 0.0026 -0.0020 ADLPIPE 0.067 -0.000 0.248 -0.0010 0.0026 -0.0020 0 NUPIPE -0.029 -0.079 0.011 -0.0002 -0.0002 -0.0007 ADLPIPE -0.029 -0.079 0.011 -0.0002 -0.0002 -0.0007 3A.3-21 Rev. 30

MOVEMENT, AND EXTERNAL FORCE LOADING Node NUPIPE (psi) ADLPIPE (psi) 180 18989 19013 199 17703 17731 214 23958 23955 236 14427 14416 265 6254 6251 305 12539 12532 344 11845 11838 370 6295 6296 395 3476 3473 430 3282 3308 3A.3-22 Rev. 30

ANALYSIS ode Program Forces (lb) Moments (in-lb)

FX FY FZ MX MY MZ 7 NUPIPE 295 2337 14 -35864 5218 51979 ADLPIPE 290 2341 15 -35108 5231 52081 2 NUPIPE 295 3306 14 59390 -5394 14010 ADLPIPE 299 3310 15 59735 -5500 14542 0 NUPIPE 330 2781 -29 30930 -22748 -84971 ADLPIPE 326 2783 -32 31920 -23105 -82784 0 NUPIPE 330 4933 -29 -255351 701 126476 ADLPIPE 336 4707 -32 -256444 916 126716 0 NUPIPE 330 -492 -29 -8972 27075 82202 ADLPIPE 336 -497 -32 -9181 27724 80676 3A.3-23 Rev. 30

DEADWEIGHT ANALYSIS ode Program Deflection (in) Rotation (rad)

DX DY DZ RX RY RZ 7 NUPIPE 0.007 -0.014 -0.004 0.0001 0.0001 0.0002 ADLPIPE 0.007 -0.014 -0.004 0.0001 0.0001 0.0002 2 NUPIPE -0.005 -0.013 0.013 0.0006 0.0001 0.0004 ADLPIPE -0.005 -0.013 0.013 0.0006 0.0001 0.0004 0 NUPIPE -0.008 -0.068 0.024 0.0004 -0.0000 -0.0004 ADLPIPE -0.009 -0.069 0.024 0.0004 0.0000 -0.0004 0 NUPIPE -0.015 -0.000 -0.003 0.0002 -0.0002 -0.0005 ADLPIPE -0.015 -0.000 -0.003 0.0002 -0.0002 -0.0005 0 NUPIPE -0.001 0.002 -0.001 -0.0000 -0.0001 -0.0002 ADLPIPE -0.001 0.002 -0.001 -0.0000 -0.0001 -0.0002 3A.3-24 Rev. 30

Node NUPIPE (psi) ADLPIPE (psi) 180 685 694 199 448 458 214 667 679 236 2472 2449 265 530 524 305 515 522 344 635 631 370 679 677 395 575 580 430 1101 1091 3A.3-25 Rev. 30

Mode 1st 2nd 3rd 4th 5th PIPE 7.109 9.328 12.297 14.681 18.043 LPIPE 7.118 9.329 12.492 12.427 17.714 3A.3-26 Rev. 30

Mode 1st 2nd PIPE 2.407 13.537 nchmark Pr. 2.3288 13.0808 3A.3-27 Rev. 30

Point No. 20 Hand Calculation NUPIPE n Wall Thickness 0.032 in 0.032 in mary Stress (Eq. 9) 3,713 psi 3,712 psi mary and Secondary Stress (Eq. 10) 16,041 psi 16,038 psi ernating Stress (Eq. 11 & 14) 13,468 psi 13,465 psi age Factor 0.0654 0.0631 int No. 30 n Wall Thickness 0.047 in 0.047 in mary Stress (Eq. 9) 8,748 psi 8,741 psi mary and Secondary Stress (Eq. 10) 117,655 psi 117,546 psi pansion Stress (Eq. 12) 99,884 psi 99,781 psi

. 13 18,252 psi 18,246 psi ernate Stress (Eq. 14) 218,258 psi 217,811 psi age Factor Out of Range 3A.3-28 Rev. 30

Hand Calculation NUPIPE r (1, 5) 0.183 0.1803 r (1, 8) 1.660 1.7361 r (1, 9) 0.0001 0.0001 r (1, 10) Not in Range r (5, 8) Not in Range r (5, 9) 0.221 0.2646 r (5, 10) 0.747 0.8051 r (8, 9) 0.857 0.8832 r (8, 10) 5.5518 5.8608 r (9, 10) 0.0001 0.0001 3A.3-29 Rev. 30

PROGRAM CYLNOZ AND HAND CALCULATION Franklin Institute Output from Hand Source of Stress Corrected Values PITRUST Calculation cumferential p (Normal) (lb) 395.00 399.00 399.99 p (Bending) (lb) 1875.00 1883.00 1877.30 M (Normal) (in-lb) 35.85 35.57 36.06 M (Bending) (in-lb) 364.70 366.60 354.30 M (Normal) (in-lb) 79.05 79.66 79.54 M (Bending) (in-lb) 90.52 80.57 79.42 ial p (Normal) (lb) 813.00 812.00 814.80 p (Bending) (lb) 812.30 827.00 810.60 M (Normal) (in-lb) 91.79 105.00 95.45 M (Bending) (in-lb) 158.80 160.00 158.80 M (Normal) (in-lb) 37.06 37.00 37.12 M (Bending) (in-lb) 117.90 105.00 103.85 Shear Stress 6.63 6.63 6.63 by M (psi)

Shear Stress 106.10 106.10 106.10 by V (psi)

Shear Stress 106.10 106.10 106.10 by V (psi) 3A.3-30 Rev. 30

Location and Cause PITRUST Results Exp. Results (Ref. 8) ment A ngt. Moment (Fig. 16, Ref. 8)

Circumf. Stress (psi) 20,438.9 20,000 Axial Stress (psi) 26,292.6 25,000 ment B cumf. Moment (Fig. 15, Ref. 8)

Circumf. Stress (psi) 22,016.2 24,000 Axial Stress (psi) 13,105.8 13,000 M

L ELEMENT M "A" C

0° ELEMENT "B"

270° 3A.3-31 Rev. 30

HAND CALCULATIONS st Problems: Run Pipe O.D. = 17 in; Run Pipe Thickness = 0.812 in Axial Length of LUG = 12 in; Width of LUG along Circumf = 3 in Loads: P = 3,399 lb; Vc = 1,788 lb; V = 2478 lb; Mc = 81,834 in-lb; M1 = 103,320 in-lb Mt = 76,284 in-lb ss in Circumferential Direction:

Stress From Hand ure Calculation Computer Output Remarks 0.5485 387 330 Membrane stress due to P 0.326 2,165 2,160 Bending stress due to P 0.294 671 629 Membrane stress due to Mc 0.388 18,976 19,904 Bending stress due to Mc 0.467 3,014 2,961 Membrane stress due to M1 0.416 6,143 5,969 Bending stress due to M1 ess in Axial Direction:

0.4447 683 690 Membrane stress due to P 0.4632 773 792 Bending stress due to P 0.294 1,897 1,864 Membrane stress due to Mc 0.550 6,357 5,942 Bending stress due to Mc 0.467 2,365 2,328 Membrane stress due to M1 0.582 4,989.7 4,842 Bending stress due to M1 3A.3-32 Rev. 30

ure Calculation Computer Output Remarks ear Stress:

1,304.8 1,304.8 Shear stress due to Mt

-366.99 -366.99 Shear stress due to Vc 127.15 127.16 Shear stress due to V1 3A.3-33 Rev. 30

WITH HAND CALCULATION AS-DESIGNED CONDITION Variable Hand Calculation SAVAL lve and Nozzle Weight (lb) 714.09 714.09 n Pipe Stiffness (in-lb/Rad) 37,950,000 38,083,056 zzle Stiffness (in-lb/Rad) 1,120,000,000 1,120,619,008 uivalent Stiffness (in-lb/Rad) 36,700,000 36,831,376

t. Rotational Frequency (cps) 22.0 22.19 me Ratio 1.11 1.11 namic Load Factor 1.22 1.22 cumferential Moment x DLF (in-lb) 314,760 314,760 t Vertical Force (lb) 14,423 14,436 zzle Stress (psi) 6,600 6,597*

TRUST Stress Intensity (psi) 62,782 62,789*

ess Intensification Factor 5.6 5.608 uation (9) Stress (psi) 36,651.9 36,681*

TE:

Allowable Stress = 1.2 Sh -1.2 (15,490) = 18,588 psi 3A.3-34 Rev. 30

WITH HAND CALCULATION REINFORCED CONDITION (1 1/4 PAD)

Variable Hand Calculation SAVAL lve and Nozzle Weight (lb) 714.09 714.09 n Pipe Stiffness (in-lb/Rad) 181,500,000 181,490,256 zzle Stiffness (in-lb/Rad) 1,120,000,000 1,120,619,008 uivalent Stiffness (in-lb/Rad) 156,200,000 156,193,808

t. Rotational Frequency (cps) 45 45.7 me Ratio 2.27 2.29 namic Load Factor 1.13 1.13 cumferential Moment x DLF (in-lb) 291,540 291,540 t Vertical Force (lb) 14,436 14,436 zzle Stress (psi) 6.144 6.144 TRUST Stress Intensity (psi) 17.046 17.046 ess Intensification Factor 2.59 2.59 uation (9) Stress (psi) 16,186 16,200 3A.3-35 Rev. 30

TABLE 3A.3-16 NODAL FORCE COMPARISON Diameter D = 0.25 ft Area A = D2/4 = 0.0490874 ft2 Nodal Force = (p + v2 /g) A - Patm A p = pressure lb/ft2

= density lb/ft3 v = velocity ft/sec g = gravitational constant 32.2 ft/sec2 Patm = 14.7 x 144 lb/ft2 At time t = 0.00650 sec de Pressure Velocity Density Force (lb) Force (lb)

o. (psia) (fps) (lb/ft3) (STEHAM) (Hand Calculation) 42.523 0.0 0.23954 186.57 196.67 42.785 5.7843 0.24076 198.43 198.53 44.231 31.219 0.24647 209.00 209.11 47.003 78.172 0.25737 230.62 230.73 50.214 129.89 0.26979 257.84 257.97 52.095 159.43 0.27697 274.93 275.06 52.209 161.97 0.27742 276.09 276.23 52.168 162.21 0.27731 275.83 275.97 3A.3-36 Rev. 30

Total Nodal pe Length Inside Dia Friction No. of Span Thickness Velocity

o. (ft) (ft) Factor Nodes (ft) (in) (fps) 2,000 3.0 0.030 7 333.33 0.30824 4.24413 3,000 2.5 0.028 9 375.00 0.44 2.92132 2,000 2.0 0.024 6 400.00 0.50026 4.98473 1,800 1.5 0.020 7 300.00 0.11108 3.59336 1,500 1.5 0.022 5 375.00 0.264 4.52142 1,600 1.5 0.025 6 320.00 0.13796 2.29183 2,200 2.5 0.040 8 314.29 0.21534 3.65878 1,500 2.0 0.030 6 300.00 0.14811 3.83245 2,000 3.0 0.024 7 333.33 0.30824 4.24413 TE:

initial heads of all nodes are calculated by using the Darcy-Weisbach equation.

3A.3-37 Rev. 30

AT TIME = 2.34 SECONDS Force, Kips Force Kips Pipe No. Node No. (WATHAM) (Hand Calculation) 1 1 276.34 276.48 1 2 300.46 300.62 1 3 317.78 317.94 1 4 329.59 329.76 1 5 341.39 341.56 1 6 355.31 355.49 1 7 369.52 369.71 al Force Calculation is based on the following equation:

F = A(pH + (/g)V2) re:

F = nodal force (lb)

= density (lb/ft3)

H = nodal head (ft) g = 32.2 ft/sec2 V = nodal velocity (fps)

A = pipe area (ft2) 3A.3-38 Rev. 30

WITH STRUDL-II OUTPUT 0.707 Test Problem: Size-on-Size Size-on-Size Average Pipe Radius (in) 3.00 3.00 Average Trunnion Radius (in) 3.00 2.12 Pipe Wall Thickness (in) 0.30 0.30 Y

X Trunnion Wall Thickness (in) 0.30 0.21 Z

SIZE-ON-SIZE MAXIMUM STRESS INTENSITY--PSI ( = 30°)

Load PITRIFE Output STRUDL-II Output FX = 10,000 lbs 5,763 5,768 FY = 10,000 lbs 7,844 7,846 FZ = 10,000 lbs 6,507 5,506 MX = 10,000 in-lb 1,329 1,329 MY = 10,000 in-lb 1,688 1,687 MZ = 10,000 in-lb 4,066 4,068 0.707 SIZE-ON-SIZE MAXIMUM STRESS INTENSITY--PSI ( = 30°)

Load PITRIFE Output STRUDL-II Output FX = 10,000 lbs 13,471 13,458 FY = 10,000 lbs 9,616 9,611 FZ = 10,000 lbs 20,105 20,030 MX = 10,000 in-lb 4,371 4,368 MY = 10,000 in-lb 2,467 2,467 MZ = 10,000 in-lb 6,178 6,176 3A.3-39 Rev. 30

WITH HAND CALCULATIONS Test Problem:

Average Pipe Radius = 1.5 in Average Trunnion Radius = 1.35 in Pipe Wall Thickness = 0.30 in Trunnion Wall Thickness = 0.27 in LOADS FOR EACH LOAD TYPE COMBINED (DL, OBEI, THER, OCCU, ETC.)

FX = FY = FZ = 10,000 lbs MX = MY = MZ = 10,000 in-lbs MNS Stress = 200 psi Internal Pressure = 100 psi STRESS COEFFICIENTS--0.9 SIZE-ON-SIZE--FX LOADING ( = 30°)

Coefficient By Coefficient From ess Type Hand Calculation PITRIFE ngitudinal--Inside Fiber -1.2652 -1.2652 cumferential--Inside Fiber -0.2764 -0.2764 ear--Inside Fiber 0.2041 0.2041 ngitudinal--Outside Fiber 0.7454 0.7454 cumferential--Outside Fiber 1.3509 1.3509 ear--Outside Fiber 0.2041 0.2041 MAXIMUM STRESS INTENSITY--0.9 SIZE-ON-SIZE ( = 30°)

Maximum Stress Intensity--psi ad Condition Hand Calculation PITRIFE DL + MNS1 28,181 28,182 DL + SRSS (OBEI, OCCU) + MNS2 73,220 73,220 DL + OBEA + THER + MNS3 88,216 88,216 DL + OCCE + MNS4 59,853 59,853 DL + SRSS (SSEI, OCCF) + MNS5 73,220 73,220 3A.3-40 Rev. 30

PROBLEM Offsets Between Nodal Points (feet)

Manual BENDCORD X1 4.27721 4.27715 Z1 4.27721 4.27713 X2 3.73046 3.73042 Z2 4.76158 4.76151 X3 3.73043 3.73041 Z3 4.76152 4.76151 X4 3.12932 3.12929 Z4 5.17652 5.17642 X5 3.12932 3.12929 Z3 5.17652 5.17642 X6 2.48254 2.48254 Z6 5.51598 5.51588 X7 2.48254 2.48254 Z7 5.51598 5.51587 X8 1.79956 1.77956 Z8 5.77500 5.77489 X9 1.79956 1.79959 Z9 5.77500 5.77490 3A.3-41 Rev. 30

S2807011 Time (sec) Seg. Force (lb) 0.0 0.0 0.004 0.0 0.008 4415.83 0.012 9626.32 (Maximum Force) 0.016 6712.86 0.020 2681.74 0.024 965.24 0.028 2081.88 0.032 4794.40 0.036 6089.20 0.040 5097.82 0.044 3360.32 0.048 2580.63 0.052 3234.01 0.056 4383.75 0.060 4857.52 0.064 4390.47 0.068 3631.02 0.072 3333.10 0.076 3652.30 0.080 4135.11 3A.3-42 Rev. 30

APPENDIX 3B - ENVIRONMENTAL DESIGN CONDITIONS FOR ENVIRONMENTAL DESIGN CONDITIONS SEE ENGINEERING SPECIFICATION SP-M3-EE-0333 Rev. 30

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